Nuclear Safety
NEA/CSNI/R(2021)5
February 2024
www.oecd-nea.org

Proceedings of the International

Workshop on Developments
in Safety Assessment Approaches
and Safety Management Practices
of Fuel Cycle Facilities

NEA/CSNI/R(2021)5  1
Nuclear Energy Agency

NEA/CSNI/R(2021)5
Unclassified

English text only
8 February 2024

NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

Proceedings of the International Workshop on Developments in Safety
Assessment Approaches and Safety Management Practices of Fuel Cycle
Facilities

Paris, France, 7-9 October 2019

This document is available in PDF format only.

JT03536870
OFDE

Unclassified
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to the delimitation of international frontiers and boundaries and to the name of any territory, city or area.

2  NEA/CSNI/R(2021)5
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NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consistsof 34 countries: Argentina, Australia, Austria, Belgium, Bulgaria, Canada, Czechia, Denmark, Finland, France,
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COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS (CSNI)The Committee on the Safety of Nuclear Installations (CSNI) addresses Nuclear Energy Agency (NEA)
programmes and activities that support maintaining and advancing the scientific and technical 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 abreastof 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 forin its activities. It initiates and conducts programmes identified by these reviews and assessments in order
to confirm safety, overcome discrepancies, develop improvements and reach consensus on technical issuesof common interest. It promotes the co-ordination of work in different member countries that serve tomaintain 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 participatingorganisations. The Committee ensures that valuable end-products of the technical reviews and analyses areprovided to members in a timely manner, and made publicly 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 and technicaldevelopments of future reactor technologies and designs. Further, the scope for the Committee includes
human and organisational research activities and technical developments that affect nuclear safety.

4  NEA/CSNI/R(2021)5

Acknowledgements
Gratitude is expressed to the workshop organising committee, the session chairpersons andthe workshop participants for their effort and co-operation.

Workshop organising committee
Name

Organisation, Country

Michel Guillard
Kenji Mori
Mitsuhiro Takanashi
Consuelo Alejano
Margie Kotzalas
Alayna Pearson
Diego Escrig Forano

IRSN, France
S/NRA/R, Japan
S/NRA/R, Japan
CSN, Spain
NRC, United States
NRC, United States
NEA

Chairpersons of technical sessions
M. Guillard (IRSN, France)
K. Mori (S/NRA/R, Japan)
C. Alejano (CSN, Spain)
M. Kotzalas (NRC, United States)

NEA/CSNI/R(2021)5  5Table of contents

Executive summary .............................................................................................................................. 6
List of abbreviations and acronyms ..................................................................................................... 8
1. Introduction ..................................................................................................................................... 10
2. Previous workshops ......................................................................................................................... 11
3. Workshop sessions and technical visit ........................................................................................... 13
4. Conclusion and recommendations ................................................................................................. 23
5. References ........................................................................................................................................ 24
Appendix 1 - List of participants ....................................................................................................... 25
Appendix 2 - Technical programme of the workshop ...................................................................... 27
Appendix 3 - Conference papers ........................................................................................................ 30

6  NEA/CSNI/R(2021)5

Executive summary
The Working Group on Fuel Cycle Safety (WGFCS), which is part of the Committee on
the Safety of Nuclear Installation (CSNI) of the Nuclear Energy Agency (NEA), held aninternational workshop on developments in safety assessment approaches and safety
management practices of fuel cycle facilities. The workshop was hosted by the NEA in cooperation with the French Institute for Radiation Protection and Nuclear Safety (IRSN) as
host of the technical visit, at the OECD Conference Center in Paris, France on 7-9 October
2019.
During the workshop, a total of 20 presentations were given in 4 sessions. In the frameworkof this workshop, a technical visit to the IRSN premises at Fontenay-aux-Roses, near Paris,
was organised on Wednesday 9 October. The workshop was attended by about 38
participants from 10 countries.
The event concentrated on the approaches of different national regulators and licensees inthe safety assessment of fuel cycle facilities (FCFs), focusing on the developments andimprovements adopted in the last years, as well as on recent international progress in safetyassessment methods, including feedback from the oversight practices of FCF operators.
Many of the safety assessment approaches discussed during the workshop had already been
under development and discussion in previous workshops, mainly in Toronto in 2011.
The Paris 2019 workshop discussions included the following conclusions andrecommendations:•

There is a need to carefully identify design extension conditions (DEC) specific toeach FCF, taking into account the lessons learnt from the Fukushima Daiichi
Nuclear Power Plant accident, and include them in the FCF safety assessment.

•

Although at the conceptual level, the graded approach in technical reviews can beseen as mature and widely shared, attention must be paid to defining clear ruleswith this approach. An agreed methodology including some quantitative criteriawith specific measures in place should be developed.

•

The development of a guidance document on the use of a graded approach is
identified as a future activity.

•

Reliable and updated statistical data, both for initiating events and component
failures, as well as deep knowledge of the installation processes are instrumental in
the FCFs’ accurate probabilistic analysis. A strong development work of specificmodels is required.

•

Complete and accurate analyses of credible plant conditions are important for theidentification and implementation of reliable safety controls (i.e. there is a need tosomehow validate the operator analysis methodologies).

•

When having to redo the FCF safety envelope to integrate new probabilistic
analyses, it is not enough to run additional analyses on top of the deterministic
existing ones. It is always necessary to conduct a thorough revision of the wholesafety assessments picture in the framework of the FCF safety management system.

•

Research and development (R&D) is considered a key tool to understand thestructures, systems and components (SSCs) ageing mechanisms and phenomena.
Identification of R&D activities to support ageing management programmes could
be a topic for future WGFCS work.

NEA/CSNI/R(2021)5  7•

FCF integrated safety analysis (ISA) is also an effective tool to identify criticalinfrastructure in need of ageing management. The effects of ageing are being
actively identified and managed through items relied on for safety (IROFS) or
safety controls and their critical safety infrastructure, i.e. the SSCs that support
them.

These workshop proceedings were approved by the Committee on the Safety of
Nuclear Installations (CSNI) at its 68th session on 2-3 December 2020 (NEA, 2021) andprepared for publication by the NEA Secretariat.

8  NEA/CSNI/R(2021)5

List of abbreviations and acronyms

AEGL

Acute exposure guideline level

ASN

Autorité de Sûreté Nucléaire (France)

BDB

Beyond design-basis

CAPS

Co-ordination action proposal sheet

CFD

Computational fluid dynamics

CNSC

Canadian Nuclear Safety Commission

CSF

Centralised waste storage facility

CSN

Consejo de Seguridad Nuclear (Spain)

CSNI

Committee on the Safety of Nuclear Installations (NEA)

DBA

Design-basis accident

DEC

Design extension conditions

DGR

Deep geological repository

DOE

Department of Energy (United States)

EDLCF

Evaporation to dryness due to the loss of cooling functions

FCF

Fuel cycle facility

FINAS

Fuel incident notification and analysis system

HLLW

High and low level waste

HLW

High-level waste

IAEA

International Atomic Energy Agency

IROFS

Items relied on for safety

IRSN

Institut de Radioprotection et de Sûreté Nucléaire (France)

ISA

Integrated safety analysis

JAEA

Japan Atomic Energy Agency

LILW

Low and intermediate level waste

NCS

Nuclear criticality safety

NEA

Nuclear Energy Agency

NRA

Nuclear Regulation Authority (Japan)

NRC

Nuclear Regulatory Commission (United States)

OECD

Organisation for Economic Co-operation and Development

PSA

Probabilistic safety assessment

PSR

Periodic safety review

NEA/CSNI/R(2021)5  9

QRA

Quantitative risk assessment

R&D

Research and development

RESEP

Resilience evaluation process

SAA

Severe accident analysis

SSCs

Structures, systems and components

SF&HLW

Spent fuel and high-level waste

S/NRA/R

Regulatory Standard and Research Department, Secretariat of Nuclear
Regulation Authority (Japan)

TSO

Technical safety organisation

VLLW

Very low-level waste

WGFCS

CSNI Working Group on Fuel Cycle Safety (NEA)10  NEA/CSNI/R(2021)5

1. Introduction
This report is a summary of the Committee on the Safety of Nuclear Installation (CSNI)
Working Group on Fuel Cycle Safety (WGFCS) international workshop on developments
in safety assessment approaches and safety management practices of fuel cycle facilities,
held at the OECD Conference Center in Paris, France, on 7-9 October 2019. It outlines thepresentations and conclusions drawn during the discussions and closing panel of theworkshop. During the workshop, a total of 20 presentations were given to more than 30
participants from 10 NEA member countries in 4 technical sessions:•safety assessment of fuel cycle facilities (FCFs), including analysis of external
events and design extension conditions;•experiences of the use of a graded approach in safety assessments in the design,
construction and licensing of FCFs;•

probabilistic approaches in safety assessments of FCFs, e.g. identification of eventsand integrated risk-informed decision-making, application of the ALARA
principle;•

FCF operating life and evaluation of life extension, ageing management as part ofthe management system.

In 2011, the WGFCS held a workshop on called “Safety assessment of fuel cycle facilities– regulatory approaches and industry perspectives” that was hosted by the Canadian
Nuclear Safety Commission (CNSC) in Toronto, Canada.
The workshop showed a high level of activity regarding safety assessments of FCFs inmany countries. Among the issues covered in more detail were the identification of
initiating events, risk-informed regulations and decision-making and regulatory oversight.
The WGFCS found it was worth revisiting the progress achieved in solving these issues
since the Toronto meeting almost one decade earlier. The definition of the scope of this
proposed activity takes into account the conclusions of that meeting.
The discussions on safety assessment approaches were based on work done in previous
workshops of the WGFCS, mainly the Toronto workshop in September 2011 (NEA, 2012),
but also the workshops in Albuquerque, United States in May 2015 (NEA, 2016), in
Aomori, Japan in November 2016 (NEA, 2017) and in Boulogne, France in April 2018
(NEA, 2019).
The main objective of the 2019 workshop was to bring together many specialists involvedin safety assessments and management practices to discuss new developments applied tonuclear fuel cycle installations.
FCFs include a wide diversity of installations. This activity will focus on facilities
dedicated to conversion, enrichment, fuel manufacturing, interim spent fuel storage, spent
fuel reprocessing, radioactive waste conditioning and on-site interim storage of radioactive
waste. Also addressed were the specificities related to range of ages of the facilities, whichspan 60 years or more, as well as to the design life of facilities currently in thedesign/construction phase.
The wide variety of installations and conditions presents different potential hazards, andhence requires a graded approach in the application of safety requirements based on afacility's associated risks. The challenge is to define the extent to which the application ofsafety requirements can be graded while assuring high levels of safety. This challenge isapplicable to the different topics addressed in the workshop.

NEA/CSNI/R(2021)5  11

2. Previous workshops

2.1. Toronto workshop in September 2011
In Toronto, Canada on 27-29 September 2011, the workshop on the “safety assessment offuel cycle facilities – regulatory approaches and industry perspectives” (NEA, 2012) drew
the following broad conclusions related to fuel cycle facility (FCF) safety assessment:•

It was recognised that the risk-informed decision-making process is an establishedand useful approach not only at the macro level, but also for complex, technicaltasks in FCFs.

•

It was also noted that it may be beneficial to benchmark with other areas both within
the nuclear industry (such as nuclear power plants) as well as outside (such as the
petrochemical industry) to identify cross-learning opportunities.

•

Both operators and regulators in member countries are engaged in continuous
improvement initiatives, including to improve evaluation methods and approaches,
and to recognise the importance of employing the ALARA (as low as reasonably
achievable) principle, in order to achieve better safety analysis and better protect
workers, the environment and the public.

•

Ageing FCFs face a challenge in that they don’t necessarily have modern design
features and equipment, making fully quantitative analysis challenging.

•

It was recommended that the WGFCS continue to support similar technicalworkshops and continue to foster similar information exchanges.

2.2. Albuquerque workshop in May 2015
The workshop on “operational and regulatory aspects of criticality safety” (NEA, 2016)was held in Albuquerque, United States on 19-21 May 2015. Presenters and participants
discussed various issues related to criticality safety at FCFs. Among the main topics
identified as themes for future consideration were the identification of the advantages anddisadvantages of deterministic and risk-informed approaches to nuclear criticality safety(NCS) assessments and the determination of the appropriate balance between the two
approaches.

2.3. Aomori workshop in November 2016
The workshop on “developments in fuel cycle facilities after the Fukushima Daiichi
Nuclear Power Station (NPS) accident” (NEA, 2017) was held in Aomori, Japan on 15-17
November 2016. The main objective was to review and discuss national activities, plans
and regulatory approaches in light of the lessons learnt from the accident at the Fukushima
Daiichi Nuclear Power Plant in terms of new safety requirements and operational issues of
FCFs.
It was noted that differences in national legislation and regulatory approaches don’t change
the responsibility of FCF operators to provide a full assessment of potential hazards and/or
any combinations thereof. This means that all hazards and their consequences are identified
and analysed in a comprehensive safety case for FCFs.

12  NEA/CSNI/R(2021)5
The Aomori workshop also highlighted that there are various characteristics andparticularities of FCFs that may differ from nuclear power plants, and it is therefore
important to use a graded approach in applying new requirements to all types and sizes of
FCFs. A graded approach means that the application of regulatory requirements, used
resources and other relevant measures are commensurate with the risks associated with theexposure situation, as well as with the likelihood and magnitude of exposures. The potential
hazards of an FCF could be assessed with deterministic or probabilistic tools, or with a
combination of them.

2.4. Boulogne workshop in April 2018
The workshop on “chemical hazards in fuel cycle facilities nuclear processing” (NEA,2019) was held in Boulogne, France on 17-19 April 2018. It focused on methods to ensurethat FCFs are designed and operated in a manner that ensures that the risks of hazardous
chemical exposure, corrosion, fire, explosion and contamination associated with nuclear
fuel processing are controlled and minimised.
It was noted during discussions that several FCFs fall within high chemical hazard
categories, and therefore require more detailed assessments of safety. FCFs use modern
methods and tools for quantifying chemical hazards, from quantitative risk assessments(QRAs) to detailed calculation codes, e.g. computational fluid dynamics (CFD). Some
codes are freely available, but adequate validation data may be missing for others.
Calculation codes and models for releases into the atmosphere include many uncertainties.
It was recognised that acute exposure guideline levels (AEGLs) are now used by some
operators for quantifying emergency exposure criteria from chemical accidents.

NEA/CSNI/R(2021)5  13

3. Workshop sessions and technical visit

3.1. Opening: Introduction and objectives of the workshop
The opening session was chaired by Mr Mitsuhiro Takanashi (Secretariat of the Nuclear
Regulation Authority [S/NRA/R], Japan), chair of the Nuclear Energy Agency (NEA)
Working Group on Fuel Cycle Safety (WGFCS). In the opening session Dr Escrig Forano
from the NEA gave a short overview of the objectives and goals for the workshop. Next,
Ms Veronique Rouyer, Head of the Division of Nuclear Safety Technology and Regulation
at the NEA, welcomed all participants and gave an introduction of the most important
challenges of safety assessment and safety management procedures for fuel cycle facilities(FCFs).

3.2. Session 1: Safety assessment of FCFs including analysis of external events anddesign extension conditions
Session 1 was chaired by Mr Kenji Mori (S/NRA/R, Japan) and featured six presentations:•

Overview of the Department of Energy (DOE) nuclear safety framework, T. Hiltz(US DOE, United States)•

Beyond-design-basis analysis in the Spanish SF & HLW centralised dry interim
storage facility, F.Lentijo (Spanish Radioactive Waste Management Organisation[ENRESA], Spain)•

Summary of guidelines for safety improvement evaluation of nuclear fuelfabrication facilities and spent fuel reprocessing facilities in Japan, A. Yamaguchi(S/NSR/R, Japan)•

Evaluating the resilience of nuclear facilities at Sellafield, A. Buchan (Sellafield
Ltd, United Kingdom)•

Overview of studies conducted at the Japan Atomic Energy Agency (JAEA) on
evaporation to dryness accident of high-level liquid waste in reprocessing plant, N.
Yoshida (JAEA, Japan)•

Assessment of chemical hazards for nuclear fuel cycle facilities, Y. Hemimou(French Nuclear Safety Authority [ASN], France) and J.P. Carreton(Radioprotection and Nuclear Safety Institute [IRSN], France)

1.1. Overview of the DOE nuclear safety framework: T. Hiltz
Mr T. Hiltz (US DOE) had gave a presentation on the nuclear safety framework of theDOE. The presentation explained regulation practice, oversight structure and safety
documents.
The DOE communicates about the safety conception with the US NRC. But the DOE’s
independence is maintained.

1.2. Beyond-design basis analysis in the Spanish centralised waste storage facility(CSF)
Mr F. Lentijo (ENRESA, Spain) gave a presentation about the CSF, a temporary centralised
storage facility in Spain.

14  NEA/CSNI/R(2021)5
The Spanish nuclear policy establishes the open cycle as the reference scenario. The
national strategy for the management of high-level waste (HLW), including of spent fuel,aims for future disposal in a deep geological repository (DGR). Meanwhile, the selected
option in Spain consists of a CSF to safely manage and store any radioactive waste thatcannot be disposed of (at the El Cabril low and intermediate level waste [LILW] and very
low-level waste [VLLW] disposal facility).
In January 2014, ENRESA submitted an application to obtain siting and construction
authorisations. During the licensing process, the regulatory body requested ENRESA
implement the beyond-design-basis (BDB) analysis (design extension scenarios and severe
conditions in the design and construction of the CSF) following lessons learnt from the
Fukushima Daiichi accident.
In this presentation, Mr Lentijo outlined the CSF design, the Spanish licensing process fora CSF and the results of the BDB analysis. The BDB analysis included seismic margin
analysis, extreme phenomena hazards, commercial aircraft impacts, station blackouts(SBOs) and the total loss of a heat sink.

1.3. Summary of guidelines for safety improvement evaluation of nuclear fuelfabrication facilities and spent fuel reprocessing facilities in Japan
Mr A. Yamaguchi (S/NRA/R, Japan) gave a presentation about the Japanese guidelines forsafety improvement evaluations of fuel cycle facilities.
After the Fukushima Daiichi accident, based on the requirement of the amended reactor
regulation, the Nuclear Regulation Authority Japan (NRA) established in November 2013
"The Guideline for Periodic Safety Assessment of Continuous Improvement of Fuel
Fabrication and Reprocessing Facilities” (hereafter referred to as the “guideline”).
Once all the licensees of uranium fuel fabrication or enrichment activities received
permission for changes to their business activities based on the new regulatory
requirements, the NRA in March 2019 established new guidelines for safety improvement
evaluations of uranium fabrication facilities (hereafter referred to as the “guidelines foruranium facilities”).
The presentation summarised the existing “guideline” and described changes made in the“guidelines for uranium facilities”.
It should be noted that when the “guidelines for uranium facilities” were revised, the NRA
decided that risk evaluations and safety margin evaluations (stress tests) were not requiredfor periodic safety assessments of continuous improvement because there are no structures,
systems and components (SSCs) important to safety in existing uranium fuel fabricationand enrichment facilities in Japan.

1.4. Evaluating the resilience of nuclear facilities at Sellafield
Mr A. Buchan (Sellafield Ltd, United Kingdom) gave a presentation about evaluating theresilience of nuclear facilities at Sellafield.
The Sellafield facilities used existing safety assessment processes to inform and prioritise
studies of beyond-design-basis events and resilience evaluations that were required after
the Fukushima Daiichi accident. These studies were used to inform the response to stresstests by UK regulators and industry bodies such as the European Nuclear Safety Regulators
Group (ENSREG).
In these stress tests, Sellafield developed two further processes, the severe accident analysis(SAA) and the resilience evaluation process (RESEP). These processes use theNEA/CSNI/R(2021)5  15

understanding already developed as a result of a comprehensive programme of safety cases
underpinned by extensive design-basis accident (DBA) and probabilistic safety analysis(PSA) studies. The SAA was developed to allow a comprehensive evaluation of highconsequence scenarios. The RESEP process was developed as a structured and consistent
approach to resilience assessment for the Sellafield site that satisfies the requirements ofthe ENSREG stress tests.
The output of severe accident analyses (SAA) has been further progressed to derive an
understanding of domino events and indicate the potential functionality of facilities andservices at a site level during a range of severe accident events, allowing further
development of the severe accident management strategies.

1.5. Overview of studies conducted at the JAEA on evaporation to dryness
accidents of high-level liquid waste in reprocessing plant
Mr Yoshida (JAEA, Japan) gave a presentation about experiments on the behaviour ofgaseous ruthenium (Ru) during a boiling accident in a reprocessing plant. According to thispresentation, in Japan, the “evaporation to dryness due to the loss of cooling functions(EDLCF)” is defined as a severe accident in reprocessing facilities and in recent years the
behaviour of ruthenium (Ru) has attracted much attention in the field of EDLCF study.
Existing studies have shown that Ru was released at a rate greater than other elements
because it formed volatile compounds such as ruthenium tetroxide (RuO4).
This presentation detailed the results of three kinds of studies on the behaviour of Ru in
EDLCF conducted by the JAEA:athe release characteristics of ruthenium from high-level liquid waste during
evaporation to dryness;b

the release of radioactive materials from high-level liquid waste in hot experiments
for EDLCF;cthe migration/deposition behaviour of volatile ruthenium.

1.6. Assessment of chemical hazards for nuclear fuel cycle facilities
Mr Y. Hemimou (ASN, France) and Mr J.P. Carreton (IRSN, France) gave a presentation
about assessments of chemical hazards for nuclear fuel cycle facilities in France.
The first half of the presentation summarised the main regulations involving dangerous
chemicals in nuclear fuel cycle facilities that have to be taken into account in the safety
demonstration.
At a facility that utilises dangerous chemicals, these may be present in quantities exceedinga certain threshold, which could lead to an accident such as a toxic emission, a fire or an
explosion resulting from uncontrolled developments. The safety demonstration concerning
non-radiological risks involving dangerous substances has to be implemented (seveso
directive). These facilities are considered to present major hazards due to the potential
accident risk associated with the presence of dangerous substances. Then the safety
demonstration is presented by deterministic approach and should include probabilistic
analyses of accidents and their consequences (for chemical substances, however,probabilistic analyses are not implemented).
Uranium hexafluoride (UF6) has low levels of radioactivity, but significant chemical
toxicity. The safety demonstration therefore has to provide for both effects.
The second half of the presentation described three examples of nuclear safety assessments
regarding chemical hazards: explosion hazards during the sintering stage of UO2 and MOX

16  NEA/CSNI/R(2021)5
fuel, on-site transportation of uranium hexafluoride, and corrosion of a fission productsolution concentrator.

3.3. Session 2: Experiences in the use of a graded approach in safety assessments inthe design, construction and licensing of FCFs
The second session was chaired by Ms Margie Kotzalas, NRC, United States, and included
five presentations:•

Graded approach of the DOE, T. Hiltz (US DOE, United States)•

Graded approach from the regulatory body/technical support organisation
perspective, F. Ledroit (IRSN, France)•“Fourniture Locale d’Entreposage d’Uranium de Retraitement” (FLEUR) Project:from safety requirements to optimised design principles, Y. Guegan (Orano,France)•Application of the graded approach at Sellafield Ltd., A. Buchan (Sellafield Ltd,
United Kingdom)•

Focus on the studies in support of fire safety analysis: IRSN fire modelling
approach for nuclear fuel facilities, R. Meyrand (IRSN, France)

This second session focused on experiences in the use of a graded approach in safetyassessments in design, construction and licensing of fuel cycle facilities. The IAEA defines
a graded approach as a process or method in which the stringency of the control measuresand conditions to be applied are commensurate with the likelihood and possible
consequences of and the level of risk associated with a loss of control (IAEA, 2018). Thissession had five presentations on the graded approach from the perspective of the regulator,
technical support organisation and operator.
1. The first presentation was from Mr Tom Hiltz, who discussed the use of the gradedapproach at the US Department of Energy.
2. The second presentation was from Mr Frédéric Ledroit from the IRSN. Hispresentation discussed the graded approach from the regulatory and TSO
perspectives.
3. Mr Yves Guegan from Orano gave the third presentation, which concerned the
FLEUR project and discussed how the graded approach is applied, from the safety
requirements to optimised design principles. Mr Guegan described how Oranoapplied the graded approach in the design and analysis of a new storage facility.
4. The fourth presentation was from Mr Andy Buchan from Sellafield Ltd, who
discussed the application of the graded approach at Sellafield in the United
Kingdom.
5. The final presentation was from Mr Raphael Meyrand from the IRSN. He discussed
the use of the graded approach in studies in support of fire safety analysis.

Graded approach summary
The US DOE, the IRSN and Sellafield Ltd all described very similar concepts andframeworks for use of the graded approach in technical reviews. This shows that there isagreement at a high level of what the graded approach is. At the conceptual level, the gradedapproach in technical reviews can be seen as mature and widely shared.

NEA/CSNI/R(2021)5  17

On the other hand, cultural challenges exist to the application of the graded approach as it
requires acceptance of some amount of risk. Challenges and differences exist in theimplementation of the graded approach by different countries. The extent to which thecultural aspect affects the graded approach is important.
Moreover, the question was raised of how the regulator and the operator agree with theapplication of the graded approach. The question is particularly significant regarding thedetermination of what hazards are most significant to safety and risk. It requires acceptanceof some amount of risk. Tolerance for risk can vary, so the DOE provided guidance for its
application to help staff understand the approach and feel comfortable applying it.
The following topics can be furthered explored either in additional workshops orbilateral/multilateral exchanges:•

Possibility of looking at a more agreed upon methodology for graded approaches
and the need to define clear rules with this approach.

•Possibility of making specific measurements, such as of the dose to public, andmaking some quantitative criteria.

•

Future activity: development of a guidance document on use of graded approaches.

•

New CAPS: Exchanges between different countries about implementation ofgraded approaches.

•

How to integrate the guidance into countries’ cultures and how to communicate
with the public.

•

Possibility of considering the wealth of safety criteria behind the graded approach.

•

Identifying barriers to implementing the graded approach (regulator and licensee)
and proposing strategies for overcoming these.

•

Comparison of how the graded approach is carried out in different countries (senda questionnaire to countries as a starting point; expand on the IAEA document on
the graded approach).

•First activity: case study of different countries to find common ground from whereto start.

Note: there is a good mix of regulators and operators in this WGFCS.

3.4. Session 3: Probabilistic approaches in safety assessments of FCFs
e.g. identification of events and integrated risk-informed decision-making,application of the ALARA principle, etc.
The third session was chaired by Ms Consuelo Alejano (Consejo de Seguridad Nuclear[CSN], Spain) and featured six presentations.
•

Intended IRSN probabilistic study developments for nuclear non-power facilities,
A. Luciani (IRSN, France)•

Expert system: a decision support tool for fire safety analysis in nuclear areas, W.
Plumecocq (IRSN, France)•

Integration of deterministic and probabilistic approaches to the safety assessmentof the Juzbado Fuel Fabrication Facility, C. Alejano (CSN, Spain)18  NEA/CSNI/R(2021)5•

Integration of deterministic and probabilistic safety assessments at the Juzbado Fuel
Fabrication Plant: operator's perspective, A. Romano (Empresa Nacional delUranio SA [ENUSA], Spain)•

An example of risk-informed methodologies to support nuclear criticality safety:
analysis of the pellet fabrication process at the Juzbado Plant, A. Romano (ENUSA,
Spain)•The effects of inaccurate analyses on the selection of safety controls, M. Kotzalas(US NRC, United States)

Session 3 promoted the exchange of information on the various probabilistic safety
approaches of national regulators and operators of fuel cycle facilities, with a focus on
recent development and assessments.
The six presentations – two from the IRSN, as technical safety organisation (TSO) of theFrench regulator, two from ENUSA as operator of the Juzbado Fuel Fabrication Plant, andtwo from regulators, the US NRC and Spain’s CSN – gave an overview of how probabilistic
approaches are understood, applied and integrated in the facility safety management
system.
There was a general consensus that the probabilistic assessment is a recognised tool for fuelcycle facility safety assessments, with very different application scopes, from use as a basic
and unique facility safety assessment to its supplemental use in an environment ofdeterministic basic philosophy. The IRSN, the US NRC, ENUSA and the CSN describedsimilar concepts but very different frameworks for the use of probabilistic approaches:•

The IRSN has recently started applying this technique to FCFs, starting from themethods, tools and long experience gained in PSA for nuclear power plants, with
the intention of developing successively probabilistic studies for three different
FCF installations in close contact with operators (share of plant information).

•In the framework of fire safety assessments, the IRSN has developed an expert
system approach for the dynamic use of large databases as part of this analysis tohelp identify configurations that could increase the risk in particular scenarios.
They consider it a new generation of computational tools in the field of probabilistic
fire simulation, highlighting the importance of using probabilistic techniques in thedetermination of safety items in fire analysis.

•

The Juzbado fuel fabrication facility has improved its deterministic safety analysis
with a full scope integrated safety analysis, as defined in US regulations, identifying
new items relied on for safety. It is currently working on the integration process ofthe new assessment results in the safety management system of the facility.
Regulator and operator perspectives were presented and a detailed example of theapplication of this risk-informed methodology to the pellet fabrication process,supporting nuclear criticality safety (NCS), was presented.

•

The US NRC described the effects of inaccurate analysis on the selection of safetycontrols, as a result of accumulated experience in the use of risk-informed
assessments. This analysis must be supported by three fundamental pillars:complete and accurate analysis of the installation conditions, re-validation of theassumptions, and verification of the accident analysis using operating data. The
presentation included recent real examples.

NEA/CSNI/R(2021)5  19

Among the main issues raised and discussed in the context of Session 3, the following
should be highlighted:
FCFs’ probabilistic safety analyses (PSAs) can be considered the base or a complement todeterministic safety analyses to make them more robust. However, following talks anddiscussions at the workshop, some challenges related to the use (methodologies
development, application) of probabilistic safety assessments could be highlighted:•

Need for strong development work of specific models for each installation andprocess (deep knowledge of the installation processes and data).

•

Need for statistical data that are reliable and updated, to allow parameter
estimations for initiating events and component failures. Some participants evenconsider data reliability as the main FCFs’ probabilistic methodologies problem.

•

Need to integrate a PSA to build the safety envelope of the installation when it
coexists with another safety analysis.

•

Very different degrees of development, application and implementation. For many
countries, the application is at the beginning for FCF safety assessments. There isinterest among some “newcomers” in the more advanced national approaches.

•

Importance of complete and accurate analyses of credible plant conditions to resultin the identification and implementation of reliable safety controls (need tosomehow validate the operator analysis methodologies).

Some of these items could be furthered discussed in the framework of WGFCS
activities.

3.5. Session 4: FCFs’ operating life and evaluation of life extension, ageing
management as part of the management system
The fourth session was chaired by Mr Michel Guillard (IRSN, France) and featured threepresentations on ageing management in participating countries.
•

Assessment of ageing management for nuclear fuel cycle facilities, B. Casoli(IRSN, France)•

Management of periodic safety reviews (PSRs), inspection, ageing and life
extension at Orano La Hague site, A. Genard and D. Scheehl (Orano, France)•

Using an integrated safety analysis to identify critical infrastructure in need ofageing management, M. Kotzalas (US NRC, United States)

The fourth session led to the following aspects being recognised:•

Ageing management is considered a growing issue, and is viewed much more
importantly than at the previous workshop in 2011.

•

Ageing management must be taken into account from the design of FCFs and mustcover the later stages of life.

•

PSRs are the tools to check ageing management during operating life, the key issue
being maintaining conformity.

•

Ageing management is a special concern for old facilities, especially whenconnected with loss of knowledge.

•

Research is another tool to understand the ageing phenomena and mechanisms.

20  NEA/CSNI/R(2021)5
Discussion on research activities necessary to support FCF ageing management
programmes should continue in the WGFCS.

3.6. Final panel discussions on sessions
The panel discussion was chaired by Mr Mitsuhiro Takanashi (S/NRA/R, Japan) andDr Diego Escrig Forano (OECD/NEA) and the four panellists were Mr K. Mori (S/NRA/R,
Japan), Ms M. Kotzalas (NRC, United States), Ms C. Alejano (CSN, Spain) and Mr M.
Guillard (IRSN, France).
The final panel discussion confirmed many of the topics recognised during the workshop
sessions. It was agreed that the workshop presentations pointed out some difficult topics
relating to the adaptation of nuclear power plant safety assessment techniques to nuclear
fuel facilities, severe accident analysis, beyond-design basis analysis, probabilistic safetyassessment analyses of accidents and their consequences and definitions of accepted risk
by countries/societies and cultural challenges for application of graded approach.
Safety assessment of FCFs including analysis of external events and design extension
conditions
During the presentations and discussions, the importance of having a clear high-level waste(HLW) disposal policy and a national strategy for its management was noted.
Participants discussed the guidelines for safety improvement evaluation of nuclear FCFs.
Regarding the US DOE nuclear safety framework, it was noted that the DOE has goodcommunication and shares the safety conception with the US NRC. But the independenceof the DOE is maintained.
In the beyond-design basis analysis in the Spanish centralised waste storage facility (CSF),
it was noted that:•

The Spanish nuclear policy establishes the open cycle as the reference scenario.

•

The national strategy for the management of HLW, including spent fuel, aims for
future disposal in a deep geological repository (DGR).

•

Meanwhile, the selected option in Spain consists of CSF for storage of any
radioactive waste that cannot be disposed of.

•During the licensing process of CSF, the regulatory body requested ENRESA
implement the beyond-design-basis (BDB) analysis (design extension scenariosand severe conditions in the design and construction of the CSF) following lessons
learnt from the Fukushima Daiichi accident.

In Japan, in November 2013, after the Fukushima Daiichi accident, the NRA established"The Guideline for Periodic Safety Assessment of Continuous Improvement of Fuel
Fabrication and Reprocessing Facilities”.
In March 2019, the NRA established new guidelines for safety improvement evaluations
for uranium fabrication facilities (“guidelines for uranium facilities”).
Changes from the existing “guideline” in the “guidelines for uranium facilities” were
described.
When the “guidelines for uranium facilities” were revised, the NRA decided that risk
evaluations and safety margin evaluations (stress tests) are not required for periodic safetyassessments of continuous improvement because there are no SSCs important to safety inexisting uranium fuel fabrication and enrichment facilities in Japan.

NEA/CSNI/R(2021)5  21

Use of graded approach:All the presentations described very similar concepts and frameworks for use of the gradedapproach in technical reviews. At the conceptual level, the graded approach in technicalreviews can be seen as mature and widely shared. On the other hand, cultural challenges
exist in the application of the graded approach, which requires acceptance of some amountof risk. Challenges and differences exist in the implementation of the graded approach bydifferent countries. Taking into consideration how the cultural aspect affects the gradedapproach is very important. Moreover, the question was raised of how the regulator and theoperator can agree on the application of the graded approach. The question is particularly
significant regarding the determination of what hazards are most safety and risk significant.
This topic can be furthered explored either in additional workshops or bilateral/multilateralexchanges. It was agreed that the topic can be furthered explored with a comparison ofgraded approach applications in different countries (first step to provide guidance and bestpractices).
Probabilistic approaches:
Probabilistic safety analysis (PSA) can be considered the basis or a complement to a
deterministic “classic” safety analysis to make it more robust. The key issue is thecollection of sufficient reliable data and updating those data. There is a need for strong
development work of specific models for each installation and process (deep knowledge ofthe installation processes and data) and a need for statistical data to allow parameter
estimations, both for initiating events and component failures.
During the panel discussions, it was noted that when a PSA coexists with other safety
analyses, there is a need to integrate it with those other safety analyses. Also,
correspondence with nuclear power plant PSAs may not be the right path to follow.
Complete and accurate analyses of credible plant conditions are important to identify andimplement reliable safety controls (there is a need to somehow validate the operator
analysis methodologies).
Regarding data sets, the emphasis was put on collecting data to produce a reliable product(not only failure data but also population data). Before looking for data, it’s important to
identify what they will improve with respect to risk knowledge. Some even consider
reliability of data to be the main problem of probabilistic methodologies for fuel cycle
facilities. If probabilistic assessment is integrated in the facility safety management system,
detailed information on probabilistic failures is then not necessary. The value of probability
can be enhanced with experience.
Regarding the use of safety cases, a balanced use of data and careful design consideration
were highlighted.
The integration of deterministic and probabilistic elements (not full scope PSA, not in linewith risks of these installations) is a good choice, with the goal being the understanding of
the likelihood of severe events and releases.
Due to their added value, it was agreed to continue exchanges on PSA approaches for FCFs
during further work of the WGFCS.
Ageing management:
Ageing management was considered a growing issue. It was highlighted that ageing
management must be taken into account from the design of FCFs and must cover the laterstages of life.

22  NEA/CSNI/R(2021)5
Ageing management is a special concern for old facilities, especially when connected withthe loss of knowledge.
Periodic safety reviews (PSRs) are the recommended tool for reviewing ageing
management during operating life, with the key issue being maintaining conformity.
Another tool is research to understand the ageing phenomena and mechanisms.
Discussion on research activities necessary to support ageing management should continue
in the WGFCS.
Furthermore, the importance of research and development (R&D) in support to safetyassessments was recalled. Indeed, it is crucial to identify R&D needs sufficiently inadvance (“longer-term anticipation”), taking into account the feedback from previous
safety assessments (gaps to be identified), the operating experience from FCFs, theexpected evolutions of the process or facilities, and technology and scientific monitoring.
It’s also important to periodically highlight the main topics for which state-of-the-art or
knowledge syntheses would be necessary to identify new research needs.

3.7. Technical visit to the IRSN premises at Fontenay-aux-Roses
The workshop ended on 9 October with a visit to the IRSN premises at Fontenay-aux-Roses
near Paris, France.
The group visited IRSN’s emergency response technical centre. The IRSN is responsible
for advising at the technical level the ASN, the French regulatory body, in case of nuclear
emergency. This centre monitors the status of France’s civil nuclear installations.
During the visit, IRSN experts presented the modelling software tools and experimental
platforms developed by the IRSN and used to support safety expertise of nuclear fuelfacilities (dispersion of radioactive substances, fire, explosion, criticality). At the end of thevisit, IRSN Director General Jean-Christophe Niel spoke to the participants.

NEA/CSNI/R(2021)5  23

4. Conclusions and recommendations

The discussions and papers presented during the four technical sessions of the 2019 NEA
workshop resulted in the following actions and recommendations:•

There is a need to carefully identify the design extension conditions (DEC) specificto each fuel cycle facility (FCF), taking into account the lessons learnt from the
Fukushima Daiichi accident, and include them in FCF safety assessments.

•

Although graded approaches for technical reviews are mature and widely applied,
it is important that clear rules be established for their application to FCFs, including
an agreed methodology that includes quantitative criteria with specific measures.

•

The NEA Working Group on Fuel Cycle Safety (WGFCS) identified thedevelopment of a guidance document on the use of a graded approach as a new
CAPS.

•

Reliable and updated statistical data, both for initiating events and component
failures, as well as a deep knowledge of the installation processes are instrumental
in the FCFs’ accurate probabilistic analyses. Strong development of specificmodels is required.

•

It is important to develop complete and accurate analyses of credible plantconditions to identify and implement reliable safety controls. There is a need tosomehow validate the operator analysis methodologies.

•

When having to redo the FCF safety envelope to integrate new probabilistic
analyses, it is not enough to put additional analyses on top of the deterministic
existing ones. It is always necessary to conduct a thorough revision of the wholesafety assessments picture in the framework of the FCF safety management system.

•

Research and development (R&D) is considered key to understanding the SSC
ageing mechanisms and phenomena. Identification of R&D activities to support
ageing management programmes could be a topic for future WGFCS work.

•

FCF integrated safety analysis (ISA) is also an effective tool to identify criticalinfrastructure in need of ageing management: the effects of ageing are being
actively identified and managed through items relied on for safety (IROFS) or
safety controls and their critical safety infrastructure, i.e. the SSCs that supportthem.

24  NEA/CSNI/R(2021)5

5. References

IAEA (2018), IAEA Safety Glossary 2018, International Atomic Energy Agency, Vienna,
www.iaea.org/publications/11098/iaea-safety-glossary-2018-edition.
NEA (2021), “Summary Record of the Sixty-Eighth (68th) Meeting of the Committee on the Safety of
Nuclear Installations”, NEA/SEN/SIN(2020)3, OECD, Paris (not publicly available).
NEA (2019), “Proceedings of the Nuclear Energy Agency International Workshop on Chemical Hazards
in Fuel Cycle Facilities Nuclear Processing”, 17-19 April 201, OECD Publishing, Paris, www.oecdnea.org/jcms/pl_19908.
NEA (2017), “Developments in Fuel Cycle Facilities after the Fukushima Daiichi Nuclear Power Station
Accident”, Workshop Proceedings, Aomori City, Japan, 15-17 November 2016, OECD Publishing,Paris, www.oecd-nea.org/jcms/pl_19802.
NEA (2016), “Operational and Regulatory Aspects of Criticality Safety”, Workshop Proceedings,
19-21 May 2015, Albuquerque, New Mexico, United States, OECD Publishing, Paris, www.oecdnea.org/jcms/pl_19698.
NEA (2012), “Safety Assessment of Fuel Cycle Facilities – Regulatory Approaches and Industry
Perspectives”, OECD/NEA Workshop Toronto, Canada, 27-29 September 2011, OECD Publishing,Paris, www.oecd-nea.org/jcms/pl_19118.

NEA/CSNI/R(2021)5  25

Appendix 1 - List of participants

Full Name

Affiliation

Country

1. Daniel MARLOYE

Belgonucleaire

Belgium

2. Aurelie BACHIRSN - Fontenay

France3. Jean-Pierre CARRETON

IRSN - Avignon

France4. Bernard CASOLIIRSN - Avignon

France5. Isabelle DUHAMELIRSN - Fontenay

France6. Yves GUEGAN

OranoFrance7. Michel GUILLARD

IRSN - Fontenay

France

8. Youcef HEMIMOUASNFrance

9. Frédéric LEDROIT

IRSN - Fontenay

France10. Anthony LUCIANIIRSN - Fontenay

France11. Christophe MEI

Orano - La Hague

France12. Raphael MEYRANDIRSN - Fontenay

France13. Mathieu MILIN

IRSN - Fontenay

France14. William PLUMECOCQ

IRSN - Fontenay

France15. Denis SCHEEHL

Orano - La Hague

France16. Eckhard WESTERMEIER

BfS

Germany

17. Kenji MORI

S/NRA/RJapan

18. Mitsuhiro TAKANASHI

S/NRA/RJapan

19. Kotaro TONOIKEJAEA - Tokai-muraJapan

20. Akinori YAMAGUCHIS/NRA/RJapan

21. Naoki YOSHIDA

JAEA - Tokai-muraJapan

22. Minjung SUNG

Korea Hydro & Nuclear

Korea

23. Joran DE JONG

ANVS

Netherlands26  NEA/CSNI/R(2021)5
Full Name

Affiliation

Country

24. Gaël MENARD

ANVS

Netherlands

25. Paul FITZSIMONSNational Nuclear Regulator

South Africa

26. Consuelo ALEJANO

CSN

Spain

27. Fernando LENTIJOENRESASpain

28. Antonino ROMANO

ENUSA

Spain

29. Andy BUCHAN

Sellafield Ltd

United Kingdom

30. Thomas G. HILTZUS DOE

United States

31. Andrea KOCKUS NRC

United States

32. Margie KOTZALASUS NRC

United States

33. Garrett SMITH

US DOE

United States

34. Diego ESCRIG ROMANONEANEA/CSNI/R(2021)5  27

Appendix 2 - Technical programme of the workshop
International workshop on developments in safety assessment approaches and safety
management practices of fuel cycle facilities
7-8 October 2019 – OECD Conference Center, room CC6
Day 1
10:00 – 10:20 Opening session - Chairperson: M. Takanashi (S/NRA/R, Japan)
Welcome- Veronique Rouyer, NEA SAF head of division (NEA)
NEA objectives and goals for the workshop – D. Escrig Forano (NEA)
10:20 – 10:50 Coffee break
10:50 – 13:00 Session 1: Safety assessment of FCFs including analysis of external eventsand design extension conditions, chair: K. Mori (S/NRA/R, Japan)
1.1 Overview of the DOE nuclear safety framework, T. Hiltz (US DOE,
United States)
1.2 Beyond-design-basis analysis in the Spanish SF & HLW centralised
dry interim storage facility, F.Lentijo (ENRESA, Spain)
1.3 Summary of guidelines for safety improvement evaluation of nuclear
fuel fabrication facilities and spent fuel reprocessing facilities in Japan,
A. Yamaguchi (S/NRA/R, Japan)
1.4 Evaluating the resilience of nuclear facilities at Sellafield, A. Buchan(Sellafield Ltd, United Kingdom)
1.5 Overview of studies conducted at JAEA on an evaporation to dryness
accident of high-level liquid waste in a reprocessing plant, N. Yoshida(JAEA, Japan)
1.6 Assessment of chemical hazards for nuclear fuel cycle facilities, Y.
Hemimou (ASN, France) and J.P. Carreton (IRSN, France)
13:00 – 14:00 Lunch break
14:00 – 15:10 Session 2: Experiences in the use of a graded approach in safetyassessments in design, construction and licensing of FCFs, chair: M.
Kotzalas (NRC, United States)
2.1 Graded approach of the DOE, T. Hiltz (US DOE, United States)
2.2 Graded approach from the regulatory body/technical support
organisation perspective, F. Ledroit (IRSN, France)
2.3 FLEUR Project: from safety requirements to optimised designprinciples, Y. Guegan (ORANO, France)
15:10 – 15:40 Coffee break
15:40 – 16:30 2.4 Application of the graded approach at Sellafield Ltd., A. Buchan(Sellafield Ltd, United Kingdom)
2.5 Focus on studies in support of fire safety analysis: IRSN fire modelling
approach for nuclear fuel facilities, R. Meyrand (IRSN, France)
16:30 – 17:00 Closing discussion of first day

28  NEA/CSNI/R(2021)5
Day 2
9:30 – 10:00

Morning coffee

10:00 – 12:20 Session 3: Probabilistic approaches in safety assessments of FCFs
e.g. identification of events and integrated risk-informed decision-making,application of the ALARA principle, etc. Chair: C. Alejano (CSN, Spain)
3.1 Intended IRSN probabilistic study developments for nuclear non-power
facilities, A. Luciani (IRSN, France)
3.2 Expert system: a decision support tool for fire safety analysis in the
nuclear area, W. Plumecocq (IRSN, France)
3.3 Integration of deterministic and probabilistic approaches to the safetyassessment of the Juzbado Fuel Fabrication Facility, C. Alejano (CSN,
Spain)
3.4 Integration of deterministic and probabilistic safety assessments at the
Juzbado Fuel Fabrication Plant: operator's perspective, A. Romano(ENUSA, Spain)
3.5 An example of risk-informed methodologies to support nuclear
criticality safety: analysis of the pellet fabrication process at the
Juzbado plant, A. Romano (ENUSA, Spain)
3.6 The effects of inaccurate analyses on the selection of safety controls,
M. Kotzalas (US NRC, United States)
12:20 – 13:20 Lunch break
13:20 – 14:40 Session 4: FCFs’ operating life and evaluation of life extension, ageing
management as a part of the management system, chair: M. Guillard(IRSN, France)
4.1 Assessment of ageing management for nuclear fuel cycle facilities, B.
Casoli (IRSN, France)
4.2 Management of PSRs, inspection, ageing and life extension at Orano
La Hague site, A. Genard and D. Scheehl (Orano, France)
4.3 Using an integrated safety analysis to identify critical infrastructure inneed of ageing management, A. Smith (US NRC, United States)
14:40 – 15:10 Coffee break
15:10 – 16:30 Final panel discussions, drawing of conclusions - chairs: all chairs,
D. Escrig Forano (NEA Secretariat)
Panel: Margie Kotzalas(US NRC, United States)
Consuelo Alejano (CSN, Spain)
Michel Guillard(IRSN, France)
Kenji Mori
(S/NRA/R, Japan)
Takanashi Mitsuhiro (S/NRA/R, Japan)
All the participants are expected to take part in the discussions and in the formulation of
conclusions during and after the final panel session.
16:30

Closing the workshop

NEA/CSNI/R(2021)5  29

9 October 2019 – Visit to IRSN premises at Fontenay-aux-Roses
Presentation of IRSN activities in relation to safety expertise of FCFs. The schedule was
as follows:
9:30 – 10:00

Reception of visitors

10:00 – 12:00 Guided tour of the new IRSN emergency response technical centre
12:00 – 13:30 Lunch buffet
13:30 –16:30

Presentation of modelling software tools and of experimental platforms
developed by the IRSN, used to support safety expertise of nuclear fuelcycle facilities (dispersion of radioactive substances, fire, explosion,
criticality).

In addition, the IRSN Director General, Mr Jean-Christophe NIEL, wishes to address
participants at the conclusion of the visit.

30  NEA/CSNI/R(2021)5

Appendix 3 - Conference papers

Prevention of chemical risks in Chemical Facilities in France
Philippe Bodenez, General directorate for Risks prevention, Deputy Head of Technological
Risks Department

The French legislation regarding the prevention of risks in the Chemical industry has been elaborated
since the beginning of the 19th century. It was based on the knowledge of hazards posed by atremendously evolving science at that time: Chemistry.
This legislation has fed the European directive on major-accident involving dangerous substances,like the need for operators to provide a safety report to the national competent authorities. The
mechanisms linked to the elaboration of safety reports were also used by the nuclear industry, andespecially by the nuclear fuel cycle facilities management, although the sector reclaimed specificresearches in fluorine chemistry. Other principles, like the defence-in-depth principle, were more, atleast at the beginning, used in the Nuclear industry. Nevertheless, there have been always
interactions between the sectors of Chemistry and Nuclear, as the matter of a comprehensive field
called “prevention of hazards”.

1°) Some examples of industrial accidents through the world :Since the beginning of its operation, the chemical sector has experienced accidents that led to, insome occasions, dramatic consequences. A few of them still remains in the memories of thespecialists of industrial hazards,-

The Flixborough accident in the United Kingdom in 1974

A huge explosion and fire resulted in 28 facilities, personal injury both on and off-site, and thecomplete destruction of the industrial site. It also had a domino effect on other industrial activity inthe area, causing the loss coolant at a nearby steel works which could have led to a further serious
accident.
-

The Bhopal Accident – India, 1984

At a pesticide manufacturing plant set up in 1969, a release of about 40 tonnes of methyl isocyanate(MIC) and other toxic gases continued for 2 hours during the night of 2nd to 3rd December 1984,
subsequent to a runaway reaction caused by the addition of water into a MIC tank. The
consequences of this release were uncontrolled due to many safety systems being defective or
shunted for purposes of lowering operating costs.
These toxic emanations led to a tremendous number of victims among the local population: between
3,000 and 25,000 deaths, and 170,000 to 800,000 intoxications according to sources. The injured
streamed into hospitals which were underequipped and not prepared for such a surge of victims andso were quickly saturated, requiring makeshift installations to be assembled. Due to the company’s
lack of communication and a denial regarding the potential presence of hydrogen cyanide in the
discharge, physicians faced with extreme symptoms (acute oedema in the lungs, respiratory distress,
persistent coughing, ocular lesions, etc.) were hard pressed to identify a suitable treatment.
-

The Seveso Accident – Italy, 1976

At about 12:37 p.m. local time on July 10, 1976, a runaway industrial chemical reaction ruptured apressure disk in a reaction vessel at the small ICMESA chemical manufacturing plant in the Lombardy
region of Italy--just north of Seveso, Italy, and about 16 miles (25 kilometers) north of Milan, Italy. It
released from one to fifteen pounds (0.45 to 6.8 kilograms) of the extremely toxic compound 2,3,7,8tetrachlorodibenzopara-dioxin (TCDD), or dioxin, into the air. A light breeze blew the dioxin cloudsouthward over the surrounding neighborhoods, while contaminating the area, killing many birds
and small animals, and injuring many people in its path. No people died from the accident, however,it resulted in what was considered, at the time, the highest known exposure to TCDD in a residential
area.
-

The AZF Accident in Toulouse – France - 2001

Between 20 and 120 tonnes of a stock containing more than 300 tonnes of ammonium nitrate refuse
detonated in the AZF fertiliser plant in Toulouse. The causes and circumstances surrounding theaccident would give rise to several investigations and expert appraisals during the following months.
Over a 250-m² floor area, the depot formed a band 25 m long, 8 to 10 m wide and 2 to 4 m high. The
explosion, corresponding to a 3.4 magnitude earthquake on the Richter scale, could be noticed up to
75 km away. Its intensity was evaluated as equivalent to 20-40 tonnes of TNT.
At the time of the explosion, 266 plant employees and 100 subcontractors were working onsite. 31
deaths are recorded:•21 deaths occurred at the AZF facility. Five of the casualties were working for subcontractors and 5
others were performing a range of onsite activities (delivery, elevator repair, etc.) or just passing by•1 was assigned to the SNPE neighbouring site
•9 were located outside the facility, all of whom were killed either immediately by the explosion orover the following days.
Moreover, hundreds of serious injuries were reported, of which 300 victims remained in hospital forover 6 days. On the whole, thousands of people in the vicinity would require hospitalisation. On
October 17th, 2001, the Haute-Garonne Prefecture tallied a total of 2,442 individuals affected by theaccident. Property damage to the plant was considerable: an oval crater 65 m long, 45 m wide and 7
m deep occupied the spot where the depot once stood; 80 ha of the plant were largely devastated.
This facility and 5 nearby chemical sites also affected by the accident were required to suspend
activities and ensure safety at their respective locations, by disposing over the course of a few
months their inventories of hazardous products; 1,300 other firms, combining industries, retailbusinesses and the building trades, affected to varying degrees (accounting for a total workforce of
21,000) were gradually inventoried over the subsequent weeks. Within a 3-km radius, 26,000
dwelling units were damaged, including 11,200 seriously, and more than 1,200 families had to be
relocated. Insurance companies assessed overall property damage at between €1.5 and €2.3 billion.
2°) A legal framework on the prevention of chemical accident in Europe
After almost three years of negotiations in Council and European Parliament, the Seveso I directive
was adopted in 1982. In the decade since the Directive’s adoption, its strict reporting requirements
resulted in the identification of some 130 major accidents European wide.
This directive contained a list of particular installations described by a number of activities. It had a
two-tier approach, which means that, for each substance of generic category of substances, twodifferent qualifying quantities (threshold levels) were mentioned. The directive imposes moreobligations on upper tier than on lower tier establishments.
On 9 December 1996, directive 96/82/EC on the control of major-accident hazards (so-called ‘Seveso
II Directive’) was adopted by the council of the European Union. It entered into force on 3 February
1997.
The aim of the ‘Seveso II Directive was twofold:-Firstly, the Directive aims at the prevention of major accident hazards involving dangerous
substances,
Secondly, as accidents do continue to occur, the Directive aims at the limitation of theconsequences of such accidents not only for man (safety and health aspects) but also for theenvironment (environmental aspects).

The Seveso-III-Directive (2012/18/EU) was adopted 4th of July 2012.
The legal framework established by the Directive creates a continuous improvement cycle ofprevention, preparedness and response to major accidents. The cycle is closed by provisions on
lesson learning.
The main obligations for the operators are:-Need to take all necessary measures to prevent major accidents and to limit their consequences
for human health and the environment. The requirements include:
Notification of all concerned establishments to the national authority1 ;Deploying a major accident prevention policy;
Producing a safety report for upper-tier establishments;
Producing internal emergency plans for upper tier establishments;
Providing information in case of accidents;

There are also obligations for Member State authorities
Member States need to ensure that a number of requirements are fulfilled, those include:-

1

Producing external emergency plans for upper tier establishments;
Deploying land-use planning for the siting of establishments;
Making relevant information publically available;
Ensuring that any necessary action is taken after an accident including emergency measures,
actions to ensure that the operator takes any necessary remedial measures and informing thepersons likely to the affected;
Reporting accidents to the Commission;
Prohibiting the unlawful use or operation of establishments;
Conducting inspections.

A “severity” scale was made official in February 1994 by the Committee of Competent Authorities (CCA) of EU
Member States as a means of applying the SEVESO Directive.
It is based on a series of 18 technical parameters intended to objectively characterise the effects orconsequences of accidents. Each of these parameters contains 6 levels, with the highest level determining theaccident’s index rating.

Member States may maintain or adopt stricter measures than those contained in the Seveso
Directive. The is the case in France where the main obligations of the directive, like the safety report
are required also for lower tier establishments. Another legal obligation that applies in France as aconsequence of the AZF accident in 2001 is the possibility to regulate the construction and the
inhabitation in the vicinity of upper-class Seveso establishments, through the Technological Risk
Prevention Plans (PPRT). The aim of the PPRT is to ensure the compatibility of the industrial activities
with the presence of population, the local development projects and the interests of local residents.
Finally, Citizens' rights are guaranteed.
-

The public concerned needs to be consulted and involved in the decision making for specificindividual projects;
Subject to the conditions outlined, Member State authorities need to make available any
information held pursuant to the Seveso Directive;
Access to justice needs to be granted.

All the principles that have been established for the prevention of major accidents in chemical plants
can be relied to principles used in legal prescriptions for nuclear facilities. Since the beginning of theindustrial developments of nuclear installations, cross safety assessments have been realized in bothfields.
The exhaustive assessments of actions generated by initiator events (earthquake, floodings, aircraft
crash…) in the safety report or the common use of alternative probabilistic approach anddeterministic approach, that are complementary to enhance the safety of chemical or nuclear
facilities shows that, in many countries, there is no tight border between the two fields.
Nevertheless, the legal obligations still remain in different regulations, more due to the fact that the
choice to rely on nuclear to provide energy is a political choice that remains in the hand of politicalleaders of European countries.

3°) International cooperation frameworks
The effects of major accidents do not stop at national borders. EU citizens can be seriously affectedby accidents in neighbouring countries. Therefore it is necessary to co-operate internationally and toaim at a high level of protection also in non-EU countries. The strengthening of civil societies andproviding a level playing field for industry are welcome side-effects of this cooperation.
In addition to bilateral activities and cooperation in the context of EU enlargement, the EU is also
active in all relevant international fora, some of which are listed below. Furthermore, actions on
major accident hazards also contribute to the international efforts to achieve Sustainable
Development Goals and to the Sendai Framework for Disaster Risk Reduction
UNECE: Convention on the Transboundary Effects of Industrial Accidents
The Convention on the Transboundary Effects of Industrial Accidents of the United Nations Economic
Commission for Europe is also known as 'Industrial Accident Convention', Helsinki Convention orunder the abbreviation 'TEIA'. It was signed in Helsinki (Finland) on 18 March 1992 and entered into
force on 19 April 2000. It aims at protecting human beings and the environment against industrial
accidents capable of causing transboundary effects and at promoting active international
cooperation between the contracting parties before, during and after such accidents.

The Seveso-III-Directive (2012/18/EU) is the main legislative instrument implementing the TEIAconvention in EU law.
UNEP: Flexible Framework on Accident Prevention and Preparedness guidance
The United Nations Environmental Programme (UNEP) is leading the international 'Flexible
Framework Initiative for Addressing Chemical Accident Prevention and Preparedness' which
promotes chemical accident prevention and preparedness (CAPP) primarily in fast growing
economies and developing countries.
OECD: Programme on chemical accidents
The Organisation for Economic Cooperation and Development (OECD) programme on chemical
accidents addresses three major areas:-

Development of common principles, procedures and policy guidance
Analysis of issues of concerns and recommendations for best practices
Sharing and communication of information and experience.

R&D PROGRAMME ON VOLATILIZATION AND TRANSPORTBEHAVIOUR OF RUTHENIUM UNDER A LOSS OF COOLING
ACCIDENT ON HIGH LEVEL LIQUID WASTE (HLLW)STORAGE TANKS IN REPROCESSING PLANTS AND
MITIGATION STRATEGIES
Lucie DE ALMEIDA1 , Philippe NERISSON2 , Marc BARRACHIN2 , Laurent CANT REL 2 ,Christian MUN2 , Laurent RICCIARDI 3 , Marc PHILIPPE 1
Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
1
BP 17, 92262 Fontenay-aux-Roses Cedex, France2
BP 3, 13115 Saint -Paul-Lez-Durance Cedex, France
3
BP 68, 91192 Gif-sur-Yvette Cedex, France
marc.philippe@irsn.fr
T he reprocessing of spent nuclear fuel produces high level liquid waste (HLLW ). Due to the
decay heat, the concentrated solutions containing fission products (FP) are stored in cooledtanks to prevent the solution from boiling, evaporating and drying out, which could lead tolarge releases of radioactive materials to the environmen t, especially ruthenium volatile
species. T he loss-of-cooling accident on HLLW storage tanks, considered as a beyonddesign-basis accident (BDBA) in the design stage of the La Hague reprocessing plants, is
one of the accident scenarios taken into account in defining the on-site emergency plan.
Within the scope of the current La Hague reprocessing plants safety reviews and post Fukushima complementary safety assessments (CSAs), a review of the initial studies carriedout on this accident scenario in the design stage confirmed the lack of reliable data on the
behaviour of ruthenium. It also highlighted the limits of the models used at the time toevaluate the source term of the accident.
On the basis of a comprehensive literature review, an IRSN research programme has been
defined firstly to re-examine the phenomenology of the accident scenario, to acquire missing
or updated data concerning the volatilization and transport behaviour of volatile ruthenium
and to reassess the associated source term. T he second o bjective of this programme is toinvestigate the possible implementation of additional ultimate mitigation measures to limit
radioactive releases, in case of a failure of the prevention and mitigation me asures adopted
by the licensee.
I.

INTRO DUCTIO N

In the La Hague reprocessing plants, after a period of radioactive decay in pools, the spent nuclear fuelassemblies undergo shearing and leaching in hot nitric acid. T he radionuclides present in the leachingsolution are separated by solvent extraction (PUREX1 process) in three fluxes: uranium, plutonium
and fission products with minor actinides (designated as FP solutions). FP solutions are concentrated
then transferred in dedicated storage tanks for an additional radioactive decay time, before being
incorporated in vitrified waste. T he concentrated FP solutions exhibit a very high radioactivity
associated with a considerable heat release (about 10 watts per lit re for a PWR fuel irradiated to
33 000 MWday per ton and cooled for three years before reprocessing), thus the storage tanks need tobe cooled permanently. Indeed, due to self-heating, a prolonged shutdown of the tank cooling system

1

PUREX: Plutonium Uranium Refining by Extraction1could cause the progressive evaporation of the FP solutions to dryness, and culminate in the formationof volatile species of ruthenium and their release in the tank venting circuit up to the environment.
T he cooling systems are designed and maintained to perform their function with a very high level ofreliability, so that a tank cooling system failure is extremely improbable. T he study of theconsequences of such an occurrence was nevertheless conducted in the design stage (1980s) as part ofthe safety demonstration of the La Hague plants, especially in order to determine the time interval after
which a substantial release of radioactive materials could take place and to evaluate the effectivenessof the measures planned in the on-site emergency plan to get such a situation under control [1].
Nevertheless, this study presents several limitations. First ly, it is based on old or incomplete data thatneed updating, as a result of the few studies that had been carried out on the behaviour of the different
forms of volatilized ruthenium, the reaction mechanisms in air/H 2 O/HNO3/NOx medium, the transport
behaviour of these materials and the ruthenium chemistry on the basis of local conditions andcompositions of gaseous phases. Secondly, there are uncertainties on modelling calculations as a resultof the assumptions adopted for transfer coefficients calculations (air flow, relative humidity, emissionrate, size and concentration of aerosols, flow mode, diffusion, sedimentation, etc.). T his confirms theneed of reassessing the source term and transport behaviour of volatile ruthenium under a total loss ofcooling of FP storage tanks accident .
T hese initial studies have been reassessed within the frame of the post-Fukushima complementary
safety assessments and the first periodic safety reviews of the La Hague plants, in light of the recent
knowledge acquired since then thanks to many experimental works that have been recently undertakenon this subject , especially in Japan [2].
II. ACCIDENT SCENARIO : O N-SITE EMERGENCY PLAN AND PO ST-FUKUSHIMACO MPLEMENTARY SAFETY ASSESSMENTS
T he principle of a concentrated FP solutions storage unit in La Hague plants is presented in Figure 1.

Figure 1: HLLW storage - Vents in normal operating conditions and in accident conditions2

In normal operation, the gases and vapours emanating from the storage tanks are extracted to thepurification system schematised on the right side of the figure: the inner diameter of the evacuation
duct is 100 mm. In case of boiling, this pipe is not sufficient to ensure the evacuation of the vapours
produced; the pressure inside the tanks increases until the tanks are connected with the bypass pipe ofa larger diameter (500 mm) linking the vents of the tanks to the main chimney stack (100 m high) of
the plant. T his bypass system, intended to limit the abnormal increases in pressure in the tanks, isprotected in normal operating conditions by two hydraulic seal pots (water) installed in series.
T he beyond-design-basis accident scenario considered in the on-site emergency plan is the following:- the initiating event is a total loss of the cooling function of the concentrated FP solutions storage
tanks;- due to self-heating of the highly radioactive FP solutions, their temperature increases progressively
in the storage tanks;- in case of non-intervention of the operators, the temperature of the FP solutions rises up to theboiling point: the time to reach the boiling point depends on the initial temperature of the solutions
and their thermal power;- during boiling, a vapour flow rate appears in the solution in the form of bubbles bursting at thesurface of the liquid: droplets are formed and carried along by the vapour out of the tanks, whichconstitute a radioactivity transfer vector of elements in aerosol form;- due to transport phenomena, part of the radioactive materials released from the storage tanks
deposit in the ventilation ducts and filters, which become irradiating, while the other part istransferred up to the chimney stack, causing radioactive releases to the environment ;- in addition, the evaporation of the solution causes an increase in nitric acid concentration and of theboiling temperature, which leads to the formation of volatile species of ruthenium by oxidation that
contribute to a large amount of the effective dose to the population.
T he measures presented in the on-site emergency plan, that could be implemented in order to manage
the accident situation until the initial failure is fixed, are the following:- implement means to extend the delay before reaching the boiling point, such as transferring thesolutions in order to divide them among all the available storage tanks, dilute them withdemineralised water to decrease their temperature, acidity and thermal power, sprinkle externallythe storage tanks to decrease their temperature, etc.;- use the fire extinction network in an open loop in order to supply the cooling system (internal orexternal circuits) of the storage tanks.
Additionally, after the Fukushima-Daiichi accident, the French nuclear licensees were required toperform Complementary Safety Assessments (CSAs) of their facilities, in order to check therobustness of the nuclear facilities and of the existing organizations to face e xtreme situations thatwere not taken into account at the design stage (extreme natural hazards - earthquake, flooding, natural
events - higher than required in the initial design standards and total loss of power supply and cooling
functions). T he CSAs carried out for La Hague plants identified several severe accidents or “dreaded
situations”, i.e. a degraded state of the facilities that could lead to a cliff -edge effect and has therefore
to be prevented or its consequences to be limited. Remediation means, mitigation means and crisis
management means were defined in order to enable the operator to be in total autonomy during thefirst 48 hours after a dreaded situation, whatever the site conditions, without any complementary
human or material means brought from outside the site.
T he loss of the cooling function of the concentrated FP solutions storage tanks leading to their boiling
was identified by the licensee as a dreaded situation. As such, in order to prevent the solutions fromreaching their boiling point, the remediation strategy is to supply cooling water, coming from the on 3

site storm-water tank and nearby dam, to the internal cooling coils of the tanks, in an open loop
configuration or with a recycling system. The water is supplied using flexible hoses that are connected
directly on the cooling coils of the tanks. In case of a failure of this remediation means, additional
mitigation means were proposed, in order to inject water directly into the tanks using existing
decontamination or instrumentation piping.
T hese new means complete the measures implemented within the frame of the on -site emergency planand could be also deployed to manage the accident situation described previously.
While this remediation strategy and the organizational and material means deployed were deemedsatisfactory on principle by the Nuclear Safety Authority in order to prevent the solutions fromreaching the boiling point , no mitigation means were formally identified by the licensee in order tolimit the radiological consequences if the boiling point is nevertheless reached and FP releases to theenvironment occur.
Consequently, in accordance with the defence-in-depth concept, the possible implementation of
additional mitigation measures to limit the releases of FP to the environment, and more particularly
ruthenium, if the measures proposed by the licensee fail for any reason, is under investigation. T o thispurpose, the phenomenology associated with the total loss of cooling of FP storage tanks accident
scenario needs to be re-examined, in order to reassess the source term and transport behaviour ofvolatile ruthenium species in the light of new available knowledge and simulation tools available, andto identify possible additional mitigation measures to reduce radioactive releases. T he R&D
programme described below attempts to answer these objectives.
III. LITERATURE REVIEW AND R&D PRO GRAMMET he behaviour of ruthenium in nitric acid solution has been studied for several decades, following theissues mentioned above. Research works mentioned in the literature can be classified in several
categories as listed below:- general information on ruthenium chemistry in a nitric medium, mechanisms for the formation ofruthenium tetroxide RuO4 ;- behaviour of volatile forms of ruthenium in presence of steam, nitric acid and nitrogen oxides(recombination, decomposition, etc.);- transfer phenomena of the different gaseous species containing ruthenium, solid or liquid species
formed, taking into account the conditions of the accidental scenario considered.
Subsequently, the efficiency and performance of various systems that can be used to trap and filter
ruthenium (gas/liquid absorbers, steel filters, etc.), or even prevent its volatilization (recombination,
addition of reducing agents in situ, etc.) has been investigated by research teams, taking into accountthe different physical and chemical forms expected of this element.
T hus, on these findings, the literature review performed by IRSN allows to identify several areas ofwork to try to improve knowledge on this topic, described thereafter:- ruthenium in nitric acid solution;- means of prevention to avoid ruthenium release;- ruthenium chemistry in gaseous phase;- transfer of ruthenium volatile compounds.

4

Ruthenium in nitric acid solution:
First, the chemical equilibrium of nitric acid is rather difficult to study, including many redox
reactions involving appearance or disappearance of several chemical species. Besides, if a nitric acidsolution is subjected to  radiation, new radical species may appear in solution under the effect ofwater and nitric acid radiolysis (H•, HO•, NO3 •). T hese compounds are likely to form secondary
harmful species, mainly H2 and H2 O2 [3].
Ruthenium in concentrated nitric acid solution (10 -2 M ≤ [HNO3 ] ≤ 10M during reprocessing process,2M in HLLW) presents a particular chemistry, with a potential coexistence of very many chemical
species, especially complex ions. It is mostly characterized by the formation of nitrosyl ruthenium ion(RuNO3+), relatively stable. It consists of a Ru2+ complex bound with a nitrosonium ion (NO+).
Moreover, RuNO3+ can form octahedral complexes nitrated (linker NO 3-), nitrous (linker NO 2-), andmixed nitrated-nitrous with general formula:[RuNO(NO3 )x (NO2 )y (OH)z(H2 O)5-x-y-z]3-x-y-z.
Among these chemical species, nitrated nitrosyl ruthenium complexes have been particularly studied
and analysed, as they are supposed to be responsible for the difficulties in decontaminating the flow ofuranium and plutonium in fission products [3]. T heir general formula is as follows:[RuNO(NO3 ) x (OH)y (H2 O) 5-x-y ]3-x-y .
T he important diversity of potentially existing nitrated nitrosyl ruthenium isomers (21 species) is
illustrated in Figure 2. Among these complexes, the predominance of one or more species depends
mainly on the concentration of nitric acid and nitrate ions.

Figure 2: Potentially existing isomers of nitrated nitrosyl ruthenium complexes [3]
In addition to these varied complex ions, ruthenium can also take other minor forms in nitric acid:- soluble species: soluble oxides, mainly Ru(VIII) in H2 RuO5 i.e. hydrated form of RuO 4 ; Ru(IV) in ion Ru4 O6 4+.xH2 O, from the reduction of Ru(VIII);5

- insoluble species: Ru(0) in metallic ruthenium; Ru(IV) in ruthenium dioxide RuO 2 .
R&D outlooks for characterization of ruthenium species:
In the R&D programme planned by IRSN, ruthenium in nitric acid solution will be studied, in order todetermine if particular chemical species mentioned above lead preferentially to the formation ofvolatile compounds containing Ru. T o perform this, the feasibility of identification of ruthenium
species by ion-exchange chromatography will be investigated in a simplified analytical nitric solution ,for several pH and temperature conditions. Indeed, literature mentions works using high performanceliquid chromatography to separate ruthenium species in nitric solution [4]. Information collected bythese analyses will be crossed with other analytical methods available in the laboratory (ultraviolet visible spectrometry, ICP -AES2 ), in order to refine the species identification .
Means of prevention to avoid ruthenium release:
Various methods are listed in the literature for ruthenium trapping in a nitric solution. T hey mainly
depend on the phase of the ruthenium species during trapping mechanism. In general, two types oftrapping are identified:- gas phase trapping of ruthenium (a priori in the RuO4 form) already volatilized;- in situ trapping in nitric solution, by addition of a specific reducing chemical compound, inhibiting
the volatilization of ruthenium species.
Gas phase trapping of ruthenium:
Igarashi et al. [5] have performed experimental tests of ruthenium absorption in water , in order toevaluate the influence of the presence of nitrogen oxides (NO and NO2 ) on the quantities absorbed.
When NO or NO2 concentration increases in the gas flow, quantities of gaseous Ru decrease sensibly(see Figure 3).

Figure 3: Effects of NO (left) and NO 2 (right) injection on RuO 4 absorption into water [5]2

ICP-AES: Induction Coupled Plasma - Atomic Emission Spectrometry6Analyses performed to determine the nature of ruthenium species trapped in the experimental device
suggest that this phenomenon may be due to RuO4 transformation into nitrated nitrosyl ruthenium
complex, less volatile and more absorbable into water. T hus, RuO4 trapping is better with NOx
injection in presence of water.
In the frame of the RSNR-MiRE3 project, IRSN has developed a test bench for characterization of
RuO4 retention by different devices. T he first step has been to characterize the PWR containment
depressurization line devices, namely the sand filter and the metallic prefilter. Figure 4 shows anoutline of this test bench, divided into three modules, corresponding to generation, filtration andtrapping for analysis of gaseous RuO4 .

Figure 4: Schematic outline of RuO 4 generation and filtration test bench [6]Figure 5 shows results obtained, respectively with sand filter and metallic prefilter.

Figure 5: Decontamination Factor obtained for RuO 4 with 10 cm thick sand bed (left) andmetallic prefilter (right) [6]T ests with a 10 cm thick sand bed in the filtration column show no significant retention of gaseous
RuO4 , for different operating conditions (1 ≤ DF 4 ≤ 2 after 1 h of test for all configurations tested,
DFmax < 16, see Figure 5 left [6]). More generally, the experimental results obtained with sand bed
show that sand does not trap RuO 4 . Similarly, retention of RuO4 is low on metallic prefilter
3

Research on Nuclear Safety and Radiation protection - Mitigation of Releases to the Environment,
https://gforge.irsn.fr/gf/project/mire/4

Decontamination Factor7(1 ≤ DF ≤ 3 after 1 h of test for the three configurations tested, DFmax < 15, see Figure 5 right [6]).
T hus, devices with design similar to the metallic prefilter of the containment depressurization line are
not suitable for trapping volatile ruthenium such as RuO 4 .
R&D outlooks in gaseous phase:
T he operational test bench presented on Figure 4 is currently used to characterize other materials likely
to trap gaseous RuO4 as a function of temperature and gas composition. It is planned to test thefollowing compounds:- zeolites;- MOFs (Metal Organic Framewok);- active charcoals.
T he objective is to evaluate the DF of each material tested for different configurations (temperature,gas flow rate, relative humidity). T hese tests started late 2017 and are planned throughout 2018.
In situ trapping in nitric solution
Sato and Motoki [4] focus on the trapping of ruthenium species in nitric solution by means of adecontamination column. The solution used is a real aqueous waste. According to the test considered,the decontamination column consists of different mixtures of zinc, graphite and active charcoals. Tests
carried out with decontamination columns, containing, o n the one hand a zinc-active charcoal mixture,on the other hand a zinc-graphite mixture followed by a second purification performed 6 months laterwith only activated carbon, show a good retention of ruthenium. Indeed, the DFs obtained arerespectively greater than 6500 and 36507300 (global DF). Conversely, columns with a zinc-graphite
mixture alone and active charcoal alone show less efficient purification of Ru (5 ≤ DF ≤ 10). Finally,
zinc alone seems to retain very few Ru (1 ≤ DF ≤ 2). T hese results show that flowing the solution in
the zinc-graphite mixture transforms the initial ruthenium species into more adsorbable compounds on
the activated carbon alone.
In the same way, Ishio et al. [7] and Sawada et al. [8] evaluated the effect of sucrose addition insolutions containing fission products, especially Ru. Ishio et al. [7] used a genuine HLLW sample,
called high active concentrate (HAC) derived from a liquid waste of the PUREX process. Results
show no decrease in the release of ruthenium if the molar ratio C12 H22 O11 /HNO3 = 1/48 (4% Ru
released), but an important decrease with a sufficient addition of sucrose: 0.16% Ru released if themolar ratio C12 H22 O11 /HNO3 = 1/24, and 0.03% Ru released if the molar ratio C12 H22O11/HNO3 = 1/12.
T his decrease in Ru release is probably due to the decomposition of HNO 3 by sucrose (denitration).
T he oxidation by HNO3 of Ru present in solution to form volatile compounds such as RuO 4 is theninhibited. Sawada et al. [8] also mention the addition of sucrose to limit the release of Ru, in a study
dealing with the influence of sodium nitrate NaNO 3 in a nitric solution of fission products. T hesolution used is a simulated HLLW with only inactive compounds. T ests with sucrose addition havebeen carried out at 180°C. As mentioned before in Ishio et al. [7], this addition appreciably decreases
the amounts of Ru released, from 1.33% to 0.11% without NaNO 3 and from 1.71% to 0.18% with
NaNO3 . T he same explanation as proposed by Ishio et al. is advanced, namely a significant
decomposition of HNO3 by sucrose.
Another way for trapping Ru is the separation by electro-volatilization, studied by Mousset [9]. T heindustrial implementation of this method would reduce the purification operations by eliminating Ru
from the dissolution liquors before separating operations of the PUREX process, and thus avoid the
addition of chemical species in the nitric solution.

8

T he experimental results of this study show that electro-volatilization yields obtained depend on many
parameters, such as the nature of the nitric solution (real fuel dissolution solution or RuNO3+
commercial solution) and the equipment used for electro-volatilization. For example, t he elimination
efficiency of Ru from a solution with [RuNO 3+] = 5.10 -3 mol.L-1 and [HNO3 ] = 3M is about 50% after
18 min.
R&D outlooks for in situ trapping in nitric solution:
T he objective of IRSN is to study the ruthenium release, coming from the heating of a nitric solution
containing dissolved ruthenium, with or without additional compound leading to inhibition ofvolatilization. To quantify a potential RuO 4 (g) release from a simplified analytical solution containing
ruthenium, representative in terms of acidity and temperature of real HLLW, the experimental setup
shown in Figure 6 is considered.

Figure 6: Schematic outline of the experimental test bench designed for characterization of Ru
in HNO 3 solution
T he reactor and the outlet line present a controlled heating device (liquid bath or heating cords
according to the configuration test), in order to set a fixed temperature or impose a heating ramp. T heacidity of the nitric solution will be fixed for each test in the range 1M ≤ [HNO3 ] ≤ 5M, its evolution
will be monitored by a pH electrode, as well as redox potential with a specific electrode. T he uppervapour phase of the reactor is continuously swept by a controlled argon flow, in order to optimize thegas outlet and the bubbling downstream the reactor, up to a potential drying out of the solution.
Bubblers with NaOH are used for the same function as on the test bench used for RuO4 filtration (seeFigure 4), i.e. trapping RuO4 for analysis. T he difference between quantities of RuO4 trapped with orwithout addition of a specific compound (solid or liquid) in the reactor will provide information about
the absorption/inhibition efficiency of this compound. T he potential trapping compounds to be tested
are reductive chemical species, porous materials, MOFs… Pre-tests are currently being carried out by
IRSN before elaborating a definitive test matrix for 2018 -2019.
Ruthenium chemistry in gaseous phase
T he form of ruthenium, in which it is supposed to distil from the nitric acid solutions, is RuO4 (g). T his
assumption is made according to the fact that ruthenium is known to mainly vaporize under this form
at low temperatures (below 200°C). Indeed, Nikol'skii [10] determined its vapour pressure between
0°C and boiling point to 140°C whereas the gas speciation was not known. In steam conditions, it was9recently shown by Miradji et al. [11] from equilibrium calculations that the RuO2 (OH) hydroxide
would be favoured at low temperature but with pressures well below RuO 4 (g) pressure. However thecomposition of the vapour phase above a prototypic waste solution in evaporating/boiling conditions
remains largely unknown for different reasons. T he release of ruthenium from nitric acid solutions,deeply studied, has been mainly investigated up to now by analysing the condensed distillate and suchan approach is not really appropriate to identify the speciation of the gas phase. Other reasons arelinked with gaseous ruthenium tetroxide itself. It is strongly instable regarding the RuO 2 +O2 (g)decomposition and can react with surfaces [12,13, 14, 15, 16]. It is consequently difficult to separate
the respective effects of the spontaneous decomposition of RuO 4 (g) and/or interaction with surfaces
from interaction with the other gases released from the solution.
T he most extensive study of ruthenium chemistry in the gas phase was performed by Brittain et al.
[17] who used a Knudsen cell coupled to a mass spectrometer in order to assess the impact of H 2 O,
HNO3 , N2 O, NO and NO2 on the RuO4 vapour pressure between 25°C and 130°C. T he main
conclusion is that it was never possible to detect other ruthenium gaseous species than RuO 4 in thepresence of the species mentioned above. More precisely, these experiments showed that in thepresence of NO or NO2 , the pressure of RuO4 tends to decrease whereas, for its part, the presence of
N2 O has no effect. T he addition of steam alone leads to a decrease in the RuO 4 vapour pressure (but toa lesser extent than in NO or NO 2 ) but Brittain indicated that the effect is not reproducible. T he
addition of HNO3 was also evaluated, in a NO 2 + H2 O atmosphere (equal proportions) with addition of
10% of HNO3 . T he experiment shows that the reduction of the RuO 4 signal is of the same amplitude asthat observed in the presence of NO2 alone. T he influence of the temperature is significant on thepressure of RuO4 since between 293 and 403 K, the signal decreases by 20% to disappear at 463 K.
Recent studies carried out in a Knudsen effusion cell coupled with a mass spectrometer have been
performed by Kato et al. [18]. They consisted in a sublimation of a sample prepared from a nitric acidsolution, containing Ru(NO)(NO3 )3 dried to 353 K. T he sample was then continuously heated up to753 K. T he spectrum shows the presence of RuO4 (g) between 390 K and 500 K with a maximum at
440 K. T he intensity ratios of ion current for Ru, RuO, RuO 2 and RuO3 to RuO4 (T able 2 in [18]) aresignificantly higher that those reported by Dillar (T able 1 in [19]) obtained from sublimation of a
RuO4 (c) sample for a similar ionization potential, likely indicating a possible interaction between NO x
(or other) molecules. In their interpretation, Kato assumed the existence of monatomic oxygen in thegas phase able to promote the formation of RuO 4 (g) from oxidation of Ru(NO)(NO3 )3 on the basis of
the observation of the O + ion pic in the experimental spectrum.
R&D outlooks on ruthenium chemistry in gaseous phase:
T he AST EC code developed by IRSN is the European reference code for severe accident simulationsin Pressurised Water Reactors. Among large number of functionalities, the AST EC lumped-parametercode can compute thermal-hydraulic conditions and fission product transport and chemistry in reactor
coolant systems (RCS) and in the containment. T he SOPHAEROS module of AST EC is based onmodelling using control volumes. In each control volume, it can compute vapour speciation assuminga thermodynamic equilibrium. In this case, the speciation is based on Gibbs energy of species involvedin vapour speciation and element inventory in the control volume. For each species, thermodynamic
data are the key factor to accurately compute the speciation at equilibrium. So the first step towards anadaptation of this module to the present issue for reprocessing plants would be an improvement of thethermodynamic properties of the gaseous species. T he different following items could be investigated:- consolidation of the thermodynamic properties of condensed RuO 4 by ab-initio calculations;- investigation of the interaction with steam (with different concentrations of H 2 and O2 ) toeventually put in evidence the formation of volatile acids (as H 2 RuO5 );10- detailed study of the gaseous phase speciation (RuO4 (g), nitrates, hydrated compounds...) above
dried samples prepared from nitric acid solutions by Knudsen effusion cell spectrometry;- tests in open-flow reactor to study chemical reactivity between RuO 4 and mixture of various
gaseous species, including NOx (start mid-2019).
Transfer of ruthenium volatile compounds:
Literature review did not provide much information about ruthenium transport, deposition (in RuO2
form for example) and its possible resuspension during accidental situation in nuclear facilities,
especially in ventilation ducts. Nevertheless, this point is very important for nuclear safety, notably interms of radiological protection (accumulation of radioactive Ru in specific areas) and to reassess thesource term released at the main stack of the plant. T hus, the R&D programme planned by IRSN willdeal with transport of ruthenium including experimental works on a thermal gradient tube with acontrolled injection of RuO 4 (g), see Figure 7. T he ruthenium deposits profile will be analysed in a firststep using a glass tube. MEB or XPS surface analyses will be performed to characterise ruthenium
deposits. T he thermal gradient and gas composition, including NOx and/or HNO3 , will be investigatedto study the thermal stability of ruthenium tetroxide. Later on, possible investigations of ruthenium
deposits remobilisation could be carried out.
Transport tube

controlled thermal profile

FURNACE

Argon + NOx

Aerosol filter

SG

Injection lines: steam + air + NOx

Alkaline liquid traps

Figure 7: START (Study of the TrAnsport of RuThenium ) experimental set-up
IV. CO NCLUSION
In nitric solution, literature review clearly shows that chemical properties of ruthenium nitrosyl ion
lead to the potential formation of a very large variety of nitrated, nitrous or mixed complexes, which
speciation mainly depends on the acidity and the temperature of the solution. T hat explains why t heaccurate inventory of ruthenium species in HLLW storage tanks is difficult to establish. T wo kinds ofmethods are identified in the literature for trapping volatile ruthenium from a nitric solution: gas phasetrapping for Ru already volatilized on the one hand, and in situ trapping in nitric solution by additionof a chemical compound inhibiting volatilization of ruthenium species on the other hand. Most ofthese methods have interesting efficiencies (in situ methods in particular). Following this literature
review, the objective of IRSN in 2018 is to initiate a R&D programme to improve knowledge on these
topics. It first concerns a better identification of ruthenium species in simplified analytical nitricsolution, to continue with a qualitative and quantitative characterization of tra pping devices, both in
the gas phase and in a nitric solution. T ransport and deposition of Ru in accidental scenario will be
studied in a second time, from 2019.
REFERENCES[1] J.P. Mercier & al., “An example of R&D on safety assessment: study of a prolonged loss ofcooling of HALW (beyond design accident)”, Institut de Protection et de Sûreté Nucléaire,11OECD/NEA/CSNI Specialist meeting on safety and risk assessment in fuel cycle facilities, T okyo,
Japan, October 1991
[2] N. Yoshida & al., “Experimental Evaluation of Release and T ransport Behavior of Gaseous
Ruthenium under Boiling Accident in Reprocessing Plant”, Japan Atomic Energy Agency, CSNI
Workshop on Developments in Fuel Cycle Facilities (FCFs) after the Fukushima Daiichi Nuclear
Power Station (NPS) Accident, Aomori City, Japan, 15 -18 November 2016
[3] Carron, V., “Étude du mécanisme d’oxydation des formes dissoutes du ruthénium dans lessolutions d’acide nitrique”, T hèse de doctorat de l’Université de Grenoble I, décembre 2000
[4] Sato, T ., Motoki, R., “Chemical species of ruthenium in radioactive aqueous waste anddecontamination mechanism of ruthenium with zinc-charcoal mixed column”, Radiochimica Acta 48,
pp 101-113, 1989
[5] Igarashi, H., Kato, K., T akahashi, T ., “Absorption behaviour of gaseous ruthenium into water” ,Radiochimica Acta 57, pp 51-55, 1992
[6] Nerisson, P., Vesin, C., “Génération par ozonation et filtration du RuO 4 . Projet RSNR MiRE –Synthèse des essais avec filtre à sable et filtre métallique”, Rapport IRSN PSN-RES/SEREX/201700631, 2017[7] Ishio, T ., Shibata, Y., Kodama, T ., Kato, T., Tsukada, T ., Serrano -Purroy, D., Glatz, J.-P., “Studyon radioactive material transport behavior from boiling/drying out high level liquid waste”,
Proceedings of Global 2015, September 20-24, 2015, Paris, paper 5164, pp 1069 -1075,2015[8] Sawada, K., Ueda, Y. and Enokida, Y., “Ruthenium release from thermally overheated nitric acidsolution containing ruthenium nitrosyl nitrate and sodium nitrate to solidify”, 5th International
AT ALANT E Conference on Nuclear Chemistry for Sustainable Fuel Cycles. Procedia Chemistry 21
pp 82-86, 2016[9] Mousset, F., “Électro-volatilisation du ruthénium en milieu nitrique – Influences de la nature desformes chimiques du ruthénium et de la composition des solutions modèles de dissolution”, T hèse dedoctorat de l’Université Pierre et Marie Curie - Paris VI, décembre 2004[10] Nikol'skii, A.B., Saturated Vapour Pressure of Ruthenium T etroxide, Russ. J. Inorg. Chem.,(Engl. T ransl.) 8(5) 541-543 (1963)[11] Miradji, F., Virot, F., Souvi, S., Cantrel, L., Louis, F., Vallet, V., T hermochemistry o f ruthenium
oxyhydroxyde species and their impact on volatile speciations in severe nuclear accident conditions,
T he Journal of Physical Chemistry A 210 606 (2016)[12] Mun, C., Cantrel, L. and Madic, C., Study of RuO4 Decomposition in Dry and Moist Air,Radiochimica Acta 95(11), pp 643-656 (2007)[13] Cains, P.W. and Barnes, S.J., Deposition of Volatilized Ruthenium on Stainless Steels, Journal of
Nuclear Materials 186 (1991) 83-86[14] Sakurai, T ., Hinatsu, Y., T akahashi, A. and Fujisawa, G., Adsorption o f Ruthenium T etroxide on
Metal Surface, Journal of Physical Chemistry 89 (1985) 1892 -1896[15] Sakurai, T ., T akahashi, A. and Fujisawa, G., Interaction of Ruthenium T etroxide with StainlessSteel, Journal of Nuclear Science and T echnology 20 (1983) 81 -83[16] Eichler, B., Zude, F., Fan, W., T rautmann, N. and Herrmann, G., Volatilization and Deposition of
Ruthenium Oxides in a T emperature Gradient, Radiochimica Acta 56 (1992) 133 -140
[17] Brittain, R.D. and Hildenbrand, D.L., Gas Phase Reactions of Ruthenium T etroxide with Nitrogen
Oxides, DOE, Report DOE/SR/00001-T 120 (1985)[18] Kato, T ., Usami, T ., T sukada, T ., Shibata, Y. and Kodama, T ., Study on volatilization mechanismof ruthenium tetroxide from nitrosyl ruthenium nitrate by using mass spectrometer, Jo urnal of Nuclear
Materials 479 123-129 (2016)[19] Dillard, J.G. and Kiser, R.W., Ionization and Dissociation of Ruthenium and Osmium T etroxides,
Journal of Physical Chemistry, vol 69, n°11 (1965), 3893-3897
12

REAS S ES S MENT OF CHEMICAL HAZARDS FOR THE FIRS T PERIODICS AFETY REVIEWS OF ORANO LA HAGUE REPROCES S ING PLANTS – THE
CORROS ION AGEING OF FIS S ION PRODUCTS EVAPORATORS ’ CAS E
Lucie DE ALMEIDA1 , Mamadou SOW 2 , Yohan LEBLOIS2 , François GENSDARMES2 ,T homas GELAIN2 , Nicolas LE ROUX2 , Jérôme RICHARD2 , Soleiman BOURROUS2 ,Alain BRUNISSO2 , Eloi BAUDUIN1 , Sarah FOURGEAUD2 , Florence GAUT HIER1 ,Marc PHILIPPE 1 , Igor LE BARS1Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
1
BP 17, 92262 Fontenay-aux-Roses Cedex, France
2
BP 68, 91192 Gif-sur-Yvette Cedex, France
marc.philippe@irsn.fr
T he reprocessing of spent nuclear fuel is carried out within two plants, UP3 -A and UP2-800,
located on the Orano La Hague site. In these plants, the spent fuel assemblies are dissolvedin nitric acid and the uranium and plutonium are then extracted by means of a chemical
process. T he remaining part consists of dissolved fission products with minor actinides (FPsolutions) and contains most of the radioactivity of the spent fuel. T hese FP solutions areconcentrated in evaporators before undergoing vitrification.
T he measures implemented in design phase to manage the chemical risks associated with thisprocess have been reassessed within the frame of the first periodic safety reviews of thesetwo plants. In particular, compliance and ageing management studies have been carried out
on the FP evaporators, revealing that the evaporators’ heated walls were more corroded than
anticipated in design phase, decreasing their residual “life expectancy”.
T he different compensatory safety measures proposed by Orano, to limit and monitor the
corrosion phenomenon and to limit t he radiological consequences in case of a leak, were
examined by IRSN. T o support its safety assessments regarding existing and new
evaporators, research programs have been carried out at IRSN, aiming at modelling theconsequences in case of a FP evaporator leak, in order to evaluate the sufficiency of Orano’s
prevention (materials grades, operating conditions…), surveillance and mit igation measures.
I.

INTRO DUCTIO N

T he reprocessing of irradiated nuclear fuel assemblies, carried out in France in the Orano La Hague
plants using the PUREX process (“Plutonium Uranium Refining by EXtraction”), aims at recovering
the plutonium and uranium contained in spent fuels and purifying them .
T he two reprocessing plants, UP3-A and UP2-800, are divided into units, where the different steps ofthe treatment process take place. After a period of radioactive decay in pools, the spent fuel undergoes
shearing and leaching in hot nitric acid in head-end units. T hen, in units R2 (UP2-800) and T 2(UP3-A), commissioned in the 1990s, the radionuclides present in the leaching solution are separated
by solvent extraction in three fluxes: uranium, plutonium and fission products with minor actinides(designated as FP solutions). T hese FP solutions are concentrated in evaporators (three per plant), thentransferred in dedicated storage tanks before being incorporated in vitrified waste, intended for deep
geological disposal.
T he separation-purification of radionuclides is carried out using several chemical reagents (acids,
bases, solvents, etc.), inducing a wide variety of chemical hazards linked with risks of disseminationof radioactive materials, irradiation, criticality, heat release, radiolysis, fire, explosion, corrosion, etc.
T he measures implemented in design phase to manage these risks have been reassessed by IRSN
1

within the frame of the first periodic safety reviews of these two plants, from 2010 to 2015 for UP3-A
and from 2015 to 2019 for UP2-800. In this context, compliance with the requirements defined for thedesign, operation, maintenance and monitoring of the process equipment were examined. Managementof the ageing of the process equipment and the results of the compliance review were analysed and,
where necessary, action plans were drawn up for the equipment concerned.
II. CO RRO SIO N AGEINGEVAPO RATO RS

OF

THE FISSIO N

PRO DUCTS

CO NCENTRATING

T he evaporators in the R2 and T 2 units consist of a boiler (in which the fission products solution israised to boiling temperature), topped by a plate column, where the vapours undergo initial
decontamination. T he boiler is heated with pressurized superheated water circulating in circuits
consisting of half-pipes welded to the outer surfaces of the boilers.

Figure 1: Scheme of a FP concentrating evaporator
T hese evaporators were designed in the 1980s using a grade of stainless steel, URANUS S1N, chosen
by the licensee for its corrosion endurance. Indeed, it is recommended for equipment containing
concentrated nitric acid solutions up to boiling temperature and functioning in oxidizing conditions. Its
use is however limited to a 5 N maximal acidity, in practice 2-3 N. T his material is neverthelessprohibited for solutions containing halides or depassivating elements and remains sensitive to under deposit corrosion. Conversely, it shows little sensitivity to intergranular corrosion [1,2]. In the design
stage, evaporator wall thickness margins were also adopted, to demonstrate their satisfactory seismic
performance and the ability of the heating coils to withstand pressure, even after corrosion during theirwhole life period planned. T he evaporators are also installed in individual reinforced concrete cell,which are inaccessible to the personnel owing to the ambient level of radioactivity. Owing to theseprovisions, the licensee only envisaged the possibility of moderate leaks and ruled out the possibilityof a clean break in this equipment and its heating circuit. T hus, the ventilation and air filtration
network in the cell containing this equipment was not designed for this accident scenario. Indeed, in2

this kind of chemical process cell with completely welded equipment and piping, the risk of occasional
contamination is considered really low and only one stage of very high efficiency filters is required on
the exhaust.
Between 2011 and 2015, in the frame of the compliance and ageing management examination of the
reprocessing plants, several wall thickness measurement campaigns have been carried out on the FP
evaporators, even though the possibility to perform such measurements was not planned in design
stage. Because of the high radioactivity of the solutions in the evaporators, the thickness
measurements are performed using articulated rods carrying an ultrasonic probe, operat ed through oneof the walls of the cells: only a limited part of the equipment is therefore accessible for measurement.
T hese measurements revealed a loss of wall thickness more important than anticipated in the design
margins in the heated parts of the boiler. T his observation call into question the residual “lifeexpectancy” of the FP evaporators, inasmuch as their walls have to retain a minimal thickness required
to withstand chemical, thermal and mechanical (pressure, earthquake) constraints.
T he licensee has hypothesised that the acceleration of the corrosion phenomena is linked to thepresence of deposits or solid residues in the FP evaporators, creating hot spots and making thechemistry locally more aggressive, increasing in fine the corrosion rate. Moreover, for some FP
evaporators, this increase could also be attributed to the treatment of effluents containing fluorides
with few complexing species of this halide (zirconium…). Based on these different hypotheses, thelicensee proposed several compensatory measures in order to limit and to mon itor this corrosion
phenomenon: use of additives to complex oxidizing ions, more frequent and effective rinsing to
remove deposits to prevent under-deposit corrosion, decrease of the temperature and pressure of thesuperheated water to decrease constraints, increased frequency of the thickness measurements. T herequired minimal thicknesses were re-evaluated with the new operating point.
Moreover, owing to the pressure of the superheated water circulating in the heating circuits, theevaporators are classified as “nuclear pressure equipment”. T herefore, a loss of integrity of theevaporators has to be studied and the licensee has to demonstrate its capability to bring back the
installation in a safe state. Four postulated failure scenarios were considered in these studies:- scenario 1: moderate leak of superheated water in the cell;- scenario 2: moderate leak of superheated water inside the evaporator;- scenario 3: major leak of the FP solution in the cell;- scenario 4: concurrent major leaks of superheated water and FP solution in the cell.
T he radiological consequences, inside the facility and to the environment, associated with thescenarios 3 and 4 were evaluated by the licensee.
After assessment of the failure scenarios studies and compensatory measures proposed by the licensee,
IRSN and ASN considered that , given the uncertainties inherent to the measurements and themodelling of the phenomena involved, a prudent approach should be adopted and the possibility of aloss of containment of a FP evaporator should be addressed. Indeed, the information available
presently is not sufficient to conclude definitively on the type of corrosion actually involved (uniform,
intergranular, under deposit…) and on the corrosion rates in chemical environments as complex as
those encountered in FP evaporators. T herefore, the corrosion monitoring programme has toencompass these uncertainties and both brutal mechanical breach and leak risks must be taken into
account in the safety studies. T hus, compensatory measures needed to be increased, in particular to
mitigate the consequences of major leaks. Afterward, Orano defined complementary measures related
to the corrosion monitoring program and the implementation of new measures to limit theconsequences of a potential FP evaporator leak. T hese complements were deemed satisfactory inprinciple by IRSN and to be implemented as soon as possible. Concurrently, Orano plans to replace3the FP evaporators within four years and submitted several related safety cases, which were assessedby IRSN [3-7].
T o support its ongoing safety assessments regarding existing and new evaporators, research
programmes have been carried out at IRSN. T hese programmes aim at modelling the consequences incase of a FP evaporator leak, in order to evaluate the sufficiency of Orano’s prevention (materials
grades, operating conditions…), surveillance and mitigation measures.
III. R&D PRO GRAMME
1. Modelling of the consequences of a FP evaporator failure
T o support its safety assessments regarding existing evaporators, IRSN carried out studies in order tobetter predict the consequences of a FP evaporator leak, in terms of resuspension of radioactive
particles inside the evaporator cells and transfers through the ventilation network and up to the static
containment barriers (filters). T hose studies have been performed considering, in a first step, the 3 rd
failure scenario (case of a major leak of the FP solution in the cell). Studies focused on the 4 th scenario(concurrent major leaks of superheated water and FP solution in the cell) are in progress. T he
behaviour of metallic filters submitted to an atmosphere saturated in water, situation representative of
the 4 th scenario, was also studied in order to evaluate their capacity to protect the HEPA (highefficiency particulate air) filters of the ventilation network.
a. Experimental study of particulate resuspension following leakage of fission product s
concentrates simulants
T he first study of this programme deals with aerosol particles resuspension following a hypotheticalscenario of fission products (FP) concentrates leakage due to a failed FP evaporator. A n airborne
release fraction (ARF) is determined experimentally using liquid simulants of FP impacting a surface
after a falling under conditions representative of an evaporator cell.
T he solution simulating FP concentrates was prepared in the laboratory with a mixture of soda
fluorescein and sodium chloride. Soda fluorescein is a particularly suitable tracer for detecting small
quantities, with a limit of detection by fluorescence measurement around 10 ng.L -1 of solution. Adding
sodium chloride helps to produce solution with a high surface tension to be in agreement with thevalues measured in FP solutions that are roughly between 70 mN.m -1 (surface tension of water) and
80 mN.m -1 . T he dynamic viscosity of the FP concentrates is mainly driven by the temperature with
reported values at boiling point near 0.30 mPa.s.
Measurements of the surface tension and the viscosity of the FP simulants were carried out by atensiometer (K10 Krüss model regulated in temperature) and by a micro -viscometer (model LOVIS
2000M regulated in temperature) respectively.
FP simulants leakage experiments were performed in a closed chamber of 2.30 m height and a floorarea of 7.42 m 2 . A leakage with a circular diameter of 2 mm is generated in a stainless steel tank with acapacity of 17 L perfectly sealed. T he tank is attached to the roof of the chamber at a height of 1.60 m
above a stainless steel drip tray with an area of 1 m 2 on which the liquid impacts at the end of its fall.
It is maintained at a pressure of 0.16 bar to reproduce a liquid height of 1.65 m and its temperature
regulated by a heating rod associated with a t emperature controller. T he spill and liquid column
heights are representative of actual conditions in a FP evaporator cell.
For each test, the tank is filled with 5 liters of 1 g.L -1 concentration solution in soda fluorescein. Oncethe desired temperature is reached the pressure controller is adapted to the correct setpoint and then thetank valve is opened to let the liquid flow.
4During the test, the ARF is collected by three high-volume samplers (HVS) at 1.13 m 3 .min -1 rate on
HEPA filters which are then dipped in water at slightly basic pH, the washing solution being analyzed
by a fluorimeter. T he aerosol concentration in the chamber is monitored in real time by an opticalcounter (Grimm 1.109) over a time step of 6 s.
Particle size analysis and electron microscopic observation revealed that the particles are fine dry
residues with a geometric median diameter around 5 μm.
T he airborne release fraction (ARF) is simply expressed as the ratio of the re-suspended aerosol mass
to the mass of potentially dispersible material. In this experiment, it is the ratio of the total mass of
soda fluorescein collected on the three HVS filters by the soda fluorescein mass initially dissolved in
the solution. T o assess the ARF, the experiment was repeated three times with blank measurements
taken from the filters without liquid leakage before each test. T he blank measurements showed anabsence of fluorescein particles in the chamber apart from those produced during the liquid leakage
tests.
T he results for a FP simulant solution at a temperature of 50 °C with a surface tension (γ) of 71 mN.m-1
and a dynamic viscosity (μ) of 0.55 mPa.s are exposed in T able 1.
Table 1: The ARF following the leakage of 5 liters of FP simulant of 1 g.L-1 concentration in soda
fluorescein
T est n°123

T (°C)
50
50
50

γ (mN/m)
71
71
71µ (mPa.s) Collected mass (g)0.55
2.04 10 -4
0.55
2.38 10 -4
0.55
1.99 10 -4

Initial mass (g)
5
5
5

ARF
4.08 10 -5
4.75 10 -5
3.97 10 -5

T he average aerosol fraction re-suspended during the three repetitions is 4.27 10 -5 with a coefficient ofvariation of 9.88 %. T he reproducibility of the measurements, despite the small amounts of aerosol (ofthe order of one hundred micrograms) collected on the filters, gives evidence of the very goodsensitivity of the experimental method.
T he ARF that is measured in this study is hardly comparable with experimental data from the literaturethat are obtained under less controlled conditions (ambient temperature), with relatively few
constraints on the nature of solution leakage. For instance, the ARF values obtained from Sutter et al.
(1981) investigations, for spilling liquids contained in a beaker, spread over a decade betwe en ~ 10 -5
and 10 -4 [8].
T his experimental study indicates that in a scenario of an evaporator loss of integrity with FP solution
concentrates leakage, part of the radioactivity will be re-suspended in particulate form. T he generated
particles could be in a size range that is fine enough to be easily transported by the ventilation ducts
and released in the environment .
Further experiments are currently scheduled on FP simulant solutions with lower viscosity and higher
surface tension to be more representative of an FP evaporator leakage.
b. Impact on the last level of HEPA filters of the facility ventilation network
T he aim of this study is to calculate the amount of radioactive particles which reaches the last level of
HEPA filters, taking into account the deposition inside the cell and the ventilation network. It is
assumed that the mass flow of particles released in the cell, 𝑞̇ 𝑖𝑛𝑗, and the aerodynamic diameter ofparticles, 𝑑𝑎𝑒 , are known. T hese data can be deduced from the previous study described in a).
It can be shown that the permanent mass flow of particles arriving on the filters, ṁp , is :𝑚̇𝑝 = 𝐾 𝑞̇ 𝑖𝑛𝑗
5

where K is a coefficient defined by the following relationship :𝑌𝐷𝑁𝐹 𝑇𝑖𝑛𝑡
𝐾 = 𝐶𝑇 ,𝑖𝑛𝑡
𝑉̇
𝑌𝑖𝑛𝑡 𝑇𝐷𝑁𝐹 𝑎𝑖𝑟
YDNF , TDNF and V̇air are respectively the mass fraction of particles, the flow temperature and thevolume flowrate just upstream the filters. Yint and Tint are respectively the mass fraction of particles
and the flow temperature in the red-colored section of the ventilation network in the Figure 2 (calledinterface section). T he transfer coefficient, 𝐶𝑇 ,𝑖𝑛𝑡, is defined by the ratio between the mean massconcentration of particles in the interface section and the mass flow of particles released in the cell.
T he study is therefore divided into two parts: the first deals with the calculation of CT ,int, realized witha CFD (Computational Fluid Dynamics) code called ANSYS CFX, the second aims at determining theratios

YDNF
Yint

and

Tint
TDNF

data with a 0D code called SYLVIA and developed by IRSN.

 ANSYS CFX calculations
ANSYS CFX software is a generalist commercial CFD code that is used for fluid flow modelling. T he
CFD modelling adopted by ANSYS CFX consists in the cutting of the computational domain into alarge number of little elements in which Navier-Stokes equations are solved. CFX calculations aim atevaluating aerosol dispersion and deposition in the evaporator cell and at the beginning of theventilation network. The calculation setup needs many steps: the geometry and the mesh build up, thedataset definition, the calculation execution and the result post -processing.
Geometry and mesh:
T his step aims at reproducing as close as possible the real geometry that has to be modelled and is
performed with ANSYS Design Modeler software by using the evaporator plans available. T he meshof this geometry has been realized with ANSYS Meshing and contains around 30 million tetrahedral
elements. T he evaporator geometry and mesh are presented in Figure 2.

Figure 2: Evaporator geometry and mesh
Results:
Simulations have been performed for four values of aerosol aerodynamic diameters: 0.13, 0.56, 2.2
and 6.7 microns. T he surface fields of deposition on the walls and of transfer coefficient in the cell andin the ventilation network for a diameter of 2.2 microns are presented in Figure 3.

6

Figure 3: Surface field of deposition and transfer coefficient (dp = 2.2 µm)
T ransfer coefficient in the ventilation network at the interface section between CFD and SYLVIA
calculations for all the studied diameters are presented in Figure 4.

Figure 4: Transfer coefficient at the interface section SYLVIA calculations
T he SYLVIA zonal code simulates airflows inside facilities equipped with ventilation networks innominal, damaged or accidental situations. T he SYLVIA modelling is based on representing the
network with nodes, for which mass, species and energy conservation equations are resolved, andbranches (ducts, filters, etc.), for which Bernoulli equation is resolved. T he SYLVIA code notably
enables to calculate the deposition of particles inside ducts considering sedimentation, thermophoresis,
turbulent impaction, Brownian and turbulent diffusion deposit models, and inside bends considering
impaction by centrifugal forces models.
Geometry of the ventilation network:
From the isometric plan of the ventilation network, the modelling has been done from the interface
section, shown in Figure 2, to the HEPA filters and contains 26 branches (11 horizontal ducts, 2vertical ducts, 3 inclined ducts, 5 horizontal bends and 5 vertical bends). Airflows coming from other7

branches of the ventilation network are modelled by considering volumetric flowrate boundary
conditions, representing dilution branches of the airflow studied. Note that temperature inside cells(60°C) and rooms (40°C or 25°C) are taken into account for dilution branches and heat exchanges
between ducts and rooms.
Results:SYLVIA simulations have been performed for the same aerodynamic diameters of particles as for
ANSYS CFX calculations (0.13, 0.56, 2.2 and 6.7 μm). Evolutions of the ratios between values ineach node and values in the interface node of the mass fractions o f particles for the 4 aerodynamic
diameters (left graph) and airflow temperatures (right graph) are presented in the Figure 5. T hesegraphs show that simulated values are mainly influenced by branches dilutions (notably on nodes 2
and 6) compared to deposit phenomena, explaining why results are quite similar for the 4 aerodynamic
diameters of particles.

Figure 5: Ratio between values in each node and values in the interface node of mass fractions(left graph) and airflow temperatures (right graph)
T he values of the coefficient K, defined above and calculated from values presented in Figure 4 for
ANSYS CFX calculations and in Figure 5 for SYLVIA simulations, are given in T able 2 for the 4
aerodynamic diameters of particles studied.
Table 2: Values of the coefficient K for the 4 aerodynamic diameters of particles studieddae (μm)
0.13
0.56
2.2
6.7

C T,int (s/m3) YDNF / Yint (-) Tint / TDNF (-)
1.85 10 -1
2.66 10 -1-1
1.83 102.67 10 -1
1.08-11.80 102.65 10 -1-1-11.40 10
2.46 10K (-)
9.75 10 -1
9.67 10 -1
9.46 10 -1
6.83 10 -1

T hus, knowing the aerodynamic diameter of the particles and the mass flow rate of the particles
injected into the cell, it is possible to deduce the mass flow rates of the radioactive particles that reachthe last level of HEPA filters.
T hese results corroborate the order of magnitude considered by the licensee for the radiologicalactivity trapped on the last level of HEPA filters and thus the associated dose rate.
However, concerning the deposits in the ventilation ducts, the licensee assumes a uniform surface
deposit of 10% of the re-suspended activity in the cell. T he results highlight that this deposit is nothomogeneous at all - due to the presence of radiation protection screw, bends, etc. - and can create
radiation hot spots that must be taken into account for the implementation of post -accidental measures.
8

2. Behaviour of metallic filters under humidity
T he HEPA filters used in the ventilation network of the evaporators’ cells to ensure the aerosol
containment are made from glass fibers medium. T his material ensures a very good efficiency vs.
pressure drop ratio. Hence, this filters design is widely used in the nuclear industry and their behaviour
is consequently well known.
In case of an important increase of the temperature and humidit y (as expected in case of evaporator or
heating loop breach), their mechanical resistance can decrease inducing a potential release of pollutant.
T o prevent this risk, IRSN recommended that Orano should implement measures to limit overpressure
and water quantities likely to reach the HEPA filters of the ventilation of FP evaporator’s cells, in caseof an accident (evaporator or heating loop breach). One of the potential solutions is to implement
metallic filters upstream the glass fiber HEPA filters. In case of water release, the pressure drop of themetallic filter should strongly increase and break up the flow rate inducing a static containment of thecells (to protect the HEPA filter disposed downstream the metallic filters).
T o evaluate the relevance of this solution, an experimental study has been carried out. T he main aimsare:- to check the feasibility of the proposed solution in terms of water transfer and mechanical integrity;- to evaluate the impact of the efficiency level required for the metallic filter (on the one hand,
metallic filter should be efficient enough to stop the ventilation before any release of water, on theother hand, its pressure drop should be as low as possible in order to avoid any overconsumption in
nominal operation).
T o do so, the first step of the study has been to characterize the microstructure and the behaviour of
two metallic fibrous media supplied by NOVINT EC with different initial efficiency. Initial spectral
efficiency, thickness, packing density and fibre diameters o f the two media are summarized in T able 3.
Table 3 : Characterization of the performances (for a velocity of 2 cm.s -1 ) andmicrostructures of the two metallic medium considered in this study
Cross section of themediumMicrostructure

70%

Overall thickness : 390 µm

60%Fiber diameters : 3 µm/ 12 µm

50%
40%

PorosityA

Spectral efficiency

Efficiency [%]

Ref.

30%
20%
10%
0%

0

200
Dp [nm]

300

0

400

100 200 300 400
Thickness [µm]Overall thickness : 490 µm

100,0%
99,8%
99,6%
99,4%
99,2%
99,0%
98,8%
98,6%
98,4%

Fiber diameters : 3 µm/ 12 µm
100%

Efficiency [%]80%
Porosity

B

100

100%
80%
60%
40%
20%
0%

60%
40%
20%
0%

0

200
Dp [nm]400

0 100 200 300 400 500

Thickness[µm]9One can notice that the medium A shows a very low efficiency ( minimal efficiency of ~20%) while
medium B reaches at least 98.5%. T he microstructure shows that this difference is mainly due to therepartition of the fibres diameters and thickness.
Filters have then been shortly exposed to a fluorescein soda aerosol (t o simulate the contamination).
After this step, they have been exposed to a humid airflow, pressure drop has been monitored andwater has been sampled downstream to evaluate the pollutant transfer. T he results are presented in
Figure 6.
2000

A

Pressure drop [Pa]Pressure drop [Pa]

150
100
50

B

1500

1000
500

0

0
0

50

100
Time [min]

150

200

0

50

100
Time [min]

150

200

Figure 6: Pressure drop as a function of the time for two different metallic media (A and B)
exposed to a saturated air at a filtration velocity of 2 cm .s -1 (the blue points represent thedetection of pollutant downstream )

6000

250

800

200

5000

200

190

600

180
400

A

200
0
00:00

170
160

02:24

04:48
Time

150
07:12

4000
3000

150

B

100

2000

50

1000
0
00:00

01:12

Flowrate [m3 .h-1 ]210
Pressure drop [Pa]1000

Flowrate [m3 .h-1 ]Pressure drop [Pa]T hose results show that when exposed to a saturated air, the filters pressure drop increases in threesteps. In the first step, pressure drop increases slowly. During this step, no contamination transfer can
be noticed. During the second step, pressure drop increases quickly and the contaminant transfer canbe identified at the transition between this phase and the last one where the pressure drop reaches alimit value which depends on the efficiency of the filter. T he same experiment has been carried out
with full scale (cylindrical) pleated filters (Figure 7).

0
02:24

Time

Figure 7 : Pressure drop and flowrate measured for two full scale metallic filters (A and B)
exposed to a humid airflow
T he experiment performed at the full scale confirms the behaviour identified at the media scale. It isalso to be noticed that when the limit pressure drop of the filter is higher than the capacity of the fan,
the flowrates tends to zero before any transfer of pollutant, ensuring the containment of the cells andthe protection of the HEPA filters.

10

T he main conclusion of this study is that the use of metallic filters to protect the HEPA glass fibers
filters from high humidity is a reliable solution. However, the efficiency level of those filters needs tobe high enough to lead to a maximal pressure drop higher than the capacity of the fan used in theventilation network.
IV. CO NCLUSION
Evaporators are used in the La Hague reprocessing plants to concentrate FP solutions. Despite theprovisions taken in design phase, the evaporators’ heated walls display a loss of thickness due tocorrosion that is higher than expected, reducing their life expectancy and rendering them more
susceptible to the risk of leakage. Consequently, worst -case accident situations, concerning failures of
the nuclear pressurised equipment, were studied, such as a major leak of FP solution or superheatedwater in the cell. T his would lead to the release of radioactive materials in the evaporator’s cell, withpressure and humidity conditions likely to degrade the performance of the HEPA filters installed
upstream and downstream in the nuclear ventilation systems.
In order to make a realistic estimate of potential releases into the environment in such a situation andto support its safety assessments of the operating and monitoring measures deployed by the licensee
for the evaporators currently in operation or under fabrication, research have been conducted by IRSN
on the following themes:- the resuspension of radioactive aerosols caused by the FP solution dropping into the drip tray underthe evaporator;- the transfer of radioactive materials into the ventilation system, and their deposition orcondensation in the evaporator cell and the ventilation ducts;- the behaviour of the ventilation system HEPA filters, given the temperature and humidity
conditions.
T he results obtained through these studies will be integrated in IRSN’s current assessment of theperiodic safety review file for the UP2-800 spent fuel reprocessing plant at Orano La Hague site.
REFERENCES[1] Nuclear corrosion Science and engineering, ed. Damien Féron, Woodhead publishing, chapter 19:“Corrosion issues in nuclear fuel reprocessing plants – P. Fauvet”, 2012
[2] ArcelorMittal - UR S1 Rev 3 « Acier URANUS® S1 », February 11 2012[3] IRSN assessment n°2015-00376, June 27 2015[4] IRSN assessment n°2016-00158, May 17 2016[5] IRSN assessment n°2016-00345, November 04 2016[6] IRSN assessment n°2016-00270, August 04 2016[7] ASN decision n° 2016-DC-0559, June 23 2016[8] Sutter, S. L., Johnston, J. W., & Mishima, J. (1981) Pacific Northwest National Laboratory(PNNL), Richland, WA (United States)

11

The non-radiological risks involving dangerous
chemicals in nuclear fuel cycle facilities- French framework regulation Youcef HEMIMOU, Brice DELIME, Raffaello AMOROSI, Josquin VERNON
French Nuclear Safety Authority
15, rue Louis Lejeune CS70013 – 92541 Montrouge Cedex
youcef.hemimou@asn.fr
Accidents involving hazardous chemicals pose a significant threat to the population and the environment.
Among nuclear facilities, this threat applies in particular to the fuel cycle facilities. As a consequence, theactivities related to hazardous chemicals are covered by a legal framework that, depending on the nature of
the activity and the associated risks, aims to guarantee that, they will not be likely to be detrimental to safety,
public health and environment [1].
The aim of this article is to present a summary of the main regulations involving dangerous chemicals in
nuclear fuel cycle facilities that have to be taken into account in the safety demonstration. The latter must
prove that the risks of an accident - radiological or not - and the scale of its consequences, given the current
state of knowledge, practices and the vulnerability of the installation environment, are as low as possible
under acceptable economic conditions.
Key words : nuclear fuel cycle facilities, safety demonstration, principle of defence in depth, deterministic
approach, probabilistic approach, dangerous chemicals, non-radiological risks, Seveso Directive, domino
effects.
1. THE DANGEROUS SUBSTANCES IN NUCLEAR FUEL CYCLE FACILITIES
1.1. List of main fuel cycle facilities
In France, the nuclear fuel cycle includes uranium conversion, uranium enrichment, fuel manufacturing
and fuel reprocessing facilities [2]. These facilities are different from each other, each concerning a different
stage of the nuclear fuel cycle. The list of theses facilities (both manufacturing and reprocessing facilities)
and their locations are shown in Figure 1.

Figure 1 : location of main fuel cycle facilities

In general, these are subject to the order of the 7 february 2012 setting the general rules relative to basic
nuclear installations (BNI) [3]. It provides the general rules that can be applied to the design, construction,1

operation, final shutdown, dismantling, maintenance and monitoring of BNI, to protect the interests
mentioned in article L. 593-1 of the Environmental Code (public security, health and safety, protection ofnature and of the environment).
The licensee has to provide a safety demonstration which must prove that the risks of an accident radiological or not - and the scale of its consequences, given the current state of knowledge, practices andthe vulnerability of the installation environment, are as low as possible under acceptable economic
conditions.
According to the article 3.1 of the order of the 7 february 2012 [3], the safety demonstration is based on theprinciple of defence in depth (Figure 2) “which consists in deploying successive and sufficiently independent levels ofdefence aiming, with regard to the licensee, at :- preventing incidents ;- detecting incidents and applying measures that will firstly prevent them from leading to an accident, and secondly restore anormal situation or, in case of failure, place and maintain the installation in a safe condition ;- controlling accidents that cannot be avoided or, in case of failure, limit their severity by regaining control of the installationin order to return it to and maintain it in a safe condition ;- managing accident situations that could not be controlled so as to mitigate the consequences, especially for humans and theenvironment”.

Figure 2 : the defence in depth concept (International Nuclear Safety Advisory Group)

Other internal hazards mentioned in article 3.5 of the order of the 7 february 2012 [3] to be considered in
the safety demonstration include : fires, explosions, hazardous substance emissions. The BNI where
dangerous chemicals may be present in quantities exceeding a certain threshold can lead to an accident
such as a toxic emission, a fire, or an explosion resulting from uncontrolled developments during an
operation of any BNI, and consequently may give rise to the need for specific safety requirements in addition
to nuclear safety requirements.
The implementation of the safety demonstration concerning non-radiological risks involving dangerous
substances has to be in line with the Seveso Directive1 whose purpose is to prevent and control such
accidents. The Directive was transposed into the French regulation.
Depending on the quantities of dangerous substances listed in the first appendix (Part 1 - Dangerous
substances or Part 2 - Named substances) to Seveso Directive, the BNI or the establishments including
several BNI in the whole location under the control of the same licensee, are categorized in :1 The Seveso Directive (Directive 82/501/EEC) was amended in view of the lessons learned from later accidents suchas Bhopal, Toulouse or Enschede resulting into Seveso-II (Directive 96/82/EC). In 2012 Seveso-III (Directive
2012/18/EU) was adopted taking into account, amongst others, the changes in the Union legislation on theclassification of chemicals and increased rights for citizens to access information and justice.

2-no Seveso ;lower tier Seveso ;upper tier Seveso.

1.2. Part 1 of the first appendix to Seveso Directive : categories of dangerous substances
Dangerous substances covered by the hazard categories are listed in Column 1 of Part 1 of the appendix toSeveso Directive. These are subject to the qualifying quantities set out in Columns 2 (Low tier) and 3 (Uppertier) of Part 1. An excerpt is shown on Figure 3.

Figure 3 : excerpt of the Annex I (Part 1)

The dangerous substances referred to in the first appendix (Part 1) of the Seveso Directive are from theEuropean regulation EC no 1272/2008 on classification, labelling and packaging (CLP) of substances andmixtures, called the “CLP Regulation”. The latter aligns the previous EU legislation with the GHS (Globally
Harmonised System of classification and labelling of chemicals), a United Nations system used to identify
hazardous chemicals and inform users about these hazards. It also has links to the European regulation
EC no 1907/2006 on Registration, Evaluation, Authorization and Restriction of Chemicals, called “REACH
legislation”.
The hazards of chemicals are communicated through a standard hazard statement. Hazard statement meansa statement assigned to a hazard class and category that describes the nature of the hazards of a hazardous
substance or mixture, including, where appropriate, the degree of hazard.
A code, starting with the letter H and followed by 3 digits is designated for each hazard statement :- H2xx : physical hazards ;- H3xx : health hazards ;- H4xx : environmental hazards.
The hazard statement linked to Part 1 of Annex 1 is on Figure 4.
HazardH statement

Physical200, 201, 202,203, 204, 205 (Explosives)
220, 221, 222, 223, 224, 225, 226 (Flammable)
240, 241, 242, 250, 260, 270, 271, 272 (Self reactive/pyrophoric/in contact with water, emit flammablegases/Oxidizing)
014, 029 (Reacts violently with water., contact with water liberates toxic gas)

Health300, 301, 310, 330, 331, 370

Environment

400, 410, 411

Figure 4 : hazard statement linked to Seveso Directive3

Many substances stored, used or produced in the fuel cycle facilities are subject to the Seveso Directive. For
example, hydrofluoric acid present in the BNI is shown in Figure 5.
Hydrofluoric acid (UF6)
H300

Acute Tox. 2

H310

Acute Tox 1

H314

Skin Corr 1A

H330

Acute Tox 2

Figure 5 : hazard statement linked to Seveso Directive

The qualifying quantity of these substance as referred to in the Seveso Directive is for the application of :-lower-tier requirements : 5 tonnes ;upper tier requirements : 20 tonnes.

1.3. Part 2 of the appendix to Seveso Directive : named dangerous substances
48 dangerous substances are included in Part 2 of the appendix to Seveso Directive. Where a dangerous
substance is covered by Part 1 of this appendix and is also listed in Part 2, the qualifying quantities set out
in Columns 2 and 3 of Part 2 have to be applied. Among these 48 substances, some substances of those areused in the BNI are directly subject to Part 2. For example : hydrazine and petroleum products andalternative fuels.
The qualifying quantity of these substances as referred to in the Seveso Directive is :-Hydrazine (lower-tier : 0,5 tonnes / upper tier : 2 tonnes)
Petroleum products (lower-tier : 2 500 tonnes / upper tier : 25 000 tonnes)

1.4. Categorisation of establisments as regards the Seveso Directive
4 establishments (3 upper tier and 1 low tier) including several BNI are considered to present major hazards
due to the potential accident risk associated with the presence of dangerous substances as defined by theSeveso Directive (Figure 6).
Seveso upper tierSeveso lower tierFigure 6 : categorisation of establisments

In the safety demonstration, the requirements are the same for these 3 categories (no Seveso, lower tierSeveso, upper tier Seveso). Information has to be given with regard to the quantity and nature of dangerous
substances produced, used and stored , the circumstances under which an accident might occur and themeasures to prevent such accidents and to mitigate their consequences.
The main difference is the update of the elements needed for the safety demonstration linked to nonradiological risks :4

-for the upper tier category : at least once every five years ;for the no Seveso and lower tier categories : at least once every ten years.

2. THE ELEMENTS TO PROVIDE FOR THE SAFETY DEMONSTRATION LINKED
TO NON-RADIOLOGICAL RISKS
2.1. General principle of the safety demonstration involving dangerous chemicals
For the contemplated incidents and accidents involving dangerous chemicals, the safety demonstration
includes an assessment of the non-radiological potential consequences as shown on Figure 7. For eachscenario, this assessment includes at least the information referred to in the second appendix of the Seveso
Directive and in the order of the 7 February 2012 [3] to provide the safety demonstration.
Domino effects
Non-radiological risks involving
dangerous chemicals

Domino effects

Fire

Explosion

Environment and health humanconsequences

Radiological risks

Toxic emission

Damages on the EIP*/AIP** fromthe safety demonstration

Pollution, death, injuries, damages on
structures
Figure 7 : consequences linked to non-radiological risks involving dangerous chemicals

EIP* : Important Element for the Protection of public security, health and safety, natureand environment. That is to say a structure, an equipment, a system (programmed or not),hardware, a component or software present in a basic nuclear installation or placed under
the responsibility of the licensee, fulfilling a function necessary for the demonstration
mentioned.
AIP**: Important Activity for the Protection of public security, health and safety, natureand environment. That is to say activities participating in the technical or organizational
provisions.
2.2. Minimum data and information to be considered in the safety demonstration
According to article 3.2 of the order of the 7 february 2012 [3], the safety demonstration is demonstrated
by a prudent deterministic approach. This procedure integrates the technical, organisational and humandimensions, and takes into account all the possible statuses of the installation, whether permanent ortransient. In addition to the postulated single initiating events, the demonstration of safety addresses
plausible situations of combined initiating events.
According to article 3.3 of the order of the 7 february 2012 [3], the safety demonstration shall also include
probabilistic analyses of accidents and their consequences.
According to Article 3.7 of the order of the 7 february 2012 [3], for each scenario, this assessment includes :-a presentation of the assumptions used in the release calculations and exposure scenarios ; therelease calculation assumptions must be reasonably pessimistic and the exposure scenarios
must be based on realistic parameters, but without considering any population protection
actions that could be implemented by the public authorities ;

5

-

-

an estimation of the effective doses and the intensity of the non-radiological phenomena to which
persons and the environment could be exposed in the short, medium and long term, distinguishingthe different age classes where necessary, and considering the different hazardous substancetransfer pathways ; the estimation includes equivalent doses to the thyroid in the event ofradioactive substance releases that justify this ;an estimation of the extent of the areas likely to be affected ;for incidents or accidents having consequences outside the site, the kinetics of the development of
the hazardous phenomena and the propagation of their effects.

The intensity of the non-radiological hazardous phenomena is defined with respect to reference values
expressed as toxic effects, overpressure effects, thermal effects, and effects associated with the impact of a
projectile on humans and structures. The reference values to use are those figuring in appendix II of theorder of 29 September 2005 [4] (Figure 8).
The effect thresholds

On human

Thermal radiation

8 kW/m2 or heat load of 1800 (kW/m2)4/3.s : firstsignificant deaths

On structures
200 kW/m2 : significant glass broken

5 kW/m2 or heat load of 1000 (kW/m2)4/3.s : first
deaths20 kW/m2 : concrete's performance in fire
during several hours and most severe on
concrete structures

3 kW/m2 or heat load of 600 (kW/m2)4/3.s : firstirreversible effects

16 kW/m2 : most severe on structures,
excluding the concrete structure
8 kW/m2 : domino effect threshold and most
severe on structures
5 kW/m2 : significant glass broken

Overpressure

200 mbar : first significant deaths (5%)

300 mbar : most severe on structures

140 mbar : first deaths (1%)

200 mbar : domino effect threshold

50 mbar : first irreversible effects

140 mbar : severe on structures

20 mbar injuries caused by broken glass

50 mbar : moderate damage
20 mbar significant glass broken

Based on Lethal Concentration 5% and exposure time
(passage of the cloud) - first significant deaths (5%)

Toxic dose

Based on Lethal Concentration 1% and exposure time
(passage of the cloud) - first deaths (1%)
Based on irreversible effects (first injuries) and
exposure - first irreversible effects
Figure 8 : the reference values

Some substances can have a double effect : radiological and non-radiological like uranium hexafluoride(UF6) to and from enrichment plants. Uranium is in the form of UF6, which has low levels of radioactivity,
but significant chemical toxicity, especially because it will be hydrolyzed in hydrofloryc acid (HF) in contact
with the humidity of the air. So, the safety demonstration has to provide for the both effects.
Finally, the assessment of the extent and severity of the consequences of identified major accidents have toinclude maps as shown on Figure 9.

140 mbar : first deaths

50 mbar : first irreversible
effects

20 mbar : injuries caused bybroken glass

6
Figure 9 : distance of overpressure effects

3. REFERENCES[1] http://ec.europa.eu/environment/seveso[2] Chapter 13 – Annual report ASN - 2016[3] Order of 7 February 2012 setting the general rules relative to basic nuclear installations[4] Order of 29 September 2005 relative to the evaluation and integration of the probability ofoccurrence, the kinetics, the intensity of the effects and the severity of the consequences of potential
accidents in the hazard studies of classified installations subject to authorization7

Chemical Effects and their Evaluation Methods for
Accidental Release of Uranium Hexafluoride
K. Mori
Regulatory Standard and Research Department,
Secretariat of Nuclear Regulation Authority (S/NRA/R), Japan
1-9-9, Roppongi, Minato-ku, Tokyo, Japan 106-8450
kenji_mori@nsr.go.jp
This paper is mainly a summary of English translation of chapters 2 to 4 of the Japanese report,
JNES-RE-2013-20211), concerning chemical effects of UF6 for uranium processing facilities. In addition,
the author of this paper described the utilization of the result of this report in the part of chapter 1 andchapter 5 of this paper.
Abstract
After the accident at the Fukushima Daiichi Nuclear Power Station occurred in March 2011, the new
regulatory requirements for nuclear fuel fabrication facilities were enforced in December 2013, which
require licensees of UF6 enrichment and reconversion activity to evaluate chemical effects on the work
environment after a UF6 leakage in an accident that could lead to severe accidents.
Preparatory to enforcement of this new regulatory requirements, the Incorporated Administrative
Agency, Japan Nuclear Energy Safety Organization (JNES), has, as one of the efforts to consider theevaluation methods of UF6 chemical effects and criteria for UF6 leakage required for the regulation,
investigated and studied UF6 chemical effects and the regulatory statuses in the U.S., France and U.K..
As the result, it was clarified that it is important to evaluate the consequences of UF6 leakage accidents
focused on the acute chemical exposure effects. Concerning the evaluation methods of chemical effects,
views, methods and items to be considered about acute chemical exposure effects were investigated. The
results were issued as the JNES report JNES-RE-2013-20211).
In order to share the information on the safety assessment of UF6 chemical effects to the members of
OECD/NEA/WGFCS (the Organisation for Economic Co-operation and Development/the Nuclear
Energy Agency/the Working Group on Fuel Cycle Safety) as a good practice of the safety assessment of
UF6 chemical effects in Japan, outlines of the original JNES report and some of the latest information
were shown in this paper.
1. Introduction
IAEA (International Atomic Energy Agency) Safety Standards Series No. NS-R-5, "Safety of Nuclear
Fuel Cycle Facilities2)" and GS-R-2, "Preparedness and Response for a Nuclear or Radiological
Emergency3)" require specific responses against the chemical toxicity of UF6.
In Japan, the handling method has been regulated to prevent health hazards due to UF6. Moreover, HF
produced by the hydrolysis reaction of UF6 and the moisture in the air has been regulated for the controlof venting and work environments.
After the accident at the Fukushima Daiichi Nuclear Power Station occurred in March 2011, the new
regulatory requirements for nuclear fuel fabrication facilities were enforced in December 2013, which
require licensees of UF6 enrichment and reconversion activity to evaluate chemical effects on the work
environment (inside and outside UF6 processing buildings) after a UF6 leakage in an accident that could1lead to severe accidents. Moreover, in the same month, the NRA requested licensees to evaluate chemical
effects of UF6 on the public and to report the results to the NRA by the time when the verification
examination for conformity to the new regulatory requirements is completed4).
Preparatory to enforcement of this new regulatory requirements, JNES1 has, as one of the efforts toconsider the evaluation methods of UF6 chemical effects and criteria for UF6 leakage required for the
regulation, investigated and studied actual UF6 chemical effects and the current regulatory statuses in the
U.S., France and U.K.. The investigation and study results have been referred to during theabove-mentioned activities for "verification of chemical effects on the public" directed to the UF6
processing plants by the NRA. Moreover, the investigation and study results were issued as the JNES
report JNES-RE-2013-2021.
2. Summary of the UF6 leakage accident(1) The cause of death during UF6 handling at UF6 leakage accidents is inhalation of HF produced by thereaction of UF6 with the moisture in the air 5).
(2) Acute or chronic exposure to UF6 causes chemical and radiological effects. Since uranium
enrichment and reconversion facilities in Japan process the uranium within confined equipment, thechronic exposure of workers taking a small amount of uranium continuously during normal operationis hard to consider. Therefore, this report outlines the acute exposure due to UF6 leakage accidents
only. Serious acute chemical exposure consequences caused by UF6 leakage accidents are hazardous
to people due to the heavy metal toxicity (deterioration of kidney function) of accumulated uranium
in the kidney and HF produced by the hydrolysis addressed in item (1). The classification of healtheffects due to exposure to substances, including uranium, is investigated and shown in Table 2.1.
Late chemical effects in Table 2.1 are shown in parentheses because clear evidence of late effects,
including carcinogenicity, due to uranium exposure was not found and a zero rate of the person's
carcinogenicity was not confirmed yet.
(3) The HF generated by a UF6 leakage is very corrosive and stimulative. Chemical effects of HF
inhalation include stimulation and damage to the nasal cavity mucosa, bronchus and alveolar. The
increase of the exposure concentration worsens the symptoms, leading to bronchitis, pulmonaryedema or injury, bleeding and other symptomatic states of the pancreas, which may cause pneumonia.
Further increase of exposure concentration may cause dyspnea and result in death.
Table 2.1 Classification of health effects due to chemical exposure
Acute effects
Chronic effects

Chemical effects
Radiation effects
Chemical effects
Radiation effects

Various injuries, death (late effects)
Cataracts, Epilation, death (+ stochastic effects)
Malignant symptoms of workers (late effects)
Increase of the carcinogenic rate

3. Characteristics of UF6 (property and behavior)
3.1. Radiation and chemical effects due to UF6 intake
The specific activity of UF6 enriched to 5% or less used for Japanese light water reactors is 121
1

JNES was unified into the NRA in March 2014.

2

Bq/mg-U, and the amount of uranium corresponding to the committed effective dose of 1 mSv for 50
years is 13 mg and 370 mg for inhalation and ingestion, respectively2.The effect of inhalation is moresignificant than that of ingestion.
Figure 3.1 shows the inhalation quantity of soluble uranium and its acute effect (classified into three
endpoints depending upon the inhalation quantity) and the chronic exposure limit (obtained from Subpart
Z of U.S. OSHA (Occupational Safety and Health Administration) 29 of CFR 1910.10006). This figure
shows that, for example, the chemical effect of inhalation of 13 mg-U, which is equivalent to thecommitted effective dose of 1 mSv for 50 years, is level 2, and 40 mg-U which is about 3 times 13 mg-U
is level 3, leading to a potentially fatal adverse effect to health, which may cause a criticalincurable renal
impairment. On the other hand, the committed effective dose for 50 years of 40mg-U is approx. 3 mSv,
which is a value that not only acute radiation effect can not be caused but also significant carcinogenic
effects are not seen.
From the above result, the evaluation of the acute effect of inhalation focused on the chemical effect of aUF6 leakage accident is considered necessary3.
4. Evaluation method of UF6 diffusion and migration in the air at the UF6 leakage accident and thecriteria of chemical effects
4.1 Evaluation method of UF6 diffusion and migration in the air
Concerning the evaluation method of chemical accident consequences, the situation of each country
was investigated, and the U.S. evaluation method was referred to.
4.1.1. Evaluation method of UF6 diffusion and migration in the air in U.S.
NUREG-1520 Rev. 110) shows requirements concerning the evaluation method of chemical accident
consequences. The requirements are to prove that the model to be used provides conservative evaluation
results and that the diffusion model for the calculation of the UF6 concentration and its reaction products
conform to the guidance for the model applicability specified in NUREG/CR-648111). Moreover, it isrequired that the analysis of the degree of effect must follow the guide, NUREG/CR-641012), concerning
the atmospheric dispersion and consequence modeling.

2

Fell T.P. et al; Assessment of internal doses of workers potentially exposed to enriched uranyl fluoride and uranium
tetrafluoride, NRPB-W56, 2004. Obtained from the value of U.K. HSE publication. In the case of natural uranium, the radiation
effect becomes smaller.
3 Since the UO F is soluble like UF , the acute effect by the UO F and UF inhalation is, if the uranium inhalation quantity is
2 2
6
2 2
6
the same, considered to be the same. As mentioned later, U.S. regulation specifies similarly.

3

Soluble uranium inhalation quantity (mg-U)

Figure 3.1 Relationship between the inhalation quantity
and the chemical effect of soluble uranium(Acute thresholds and limits of normal operation)*1

3

The lower limits of the acute effect level 1 to 3 due to UF6 were calculated based on the adult's respiration rate (1.2 m /h) and other7)

factors referring to the UF6 threshold concentration of the AEGL level 1 to 3, which is the acute exposure guideline of U.S. EPA to
be mentioned later.
8)*2 Concerning LD50 of uranium for persons, various values have been reported. Although NUREG-1391 defines the value to be 230
9)

mg for persons of 70 Kg, another report defines it to be 1.0 g. For this report, the value of NUREG-1391 is used.

As a specific application, the 5-factor-method data of NUREG/CR-6410 or the former DOE (U.S.
Department of Energy) DTL (Dangerous Toxic Load) handbook13) are used for the source term evaluation.
Moreover, the evaluation method according to the atmospheric dispersion models for safety analysis of
nuclear power reactor facilities of the U.S., R.G. 1.145 14) (hereinafter referred to as "R.G. 1.145 model"),and the evaluation method using a calculation code, "HGSYSTEM” and “RASCAL3.0.5 and ARCON96"
are approved by NRC (U. S. Nuclear Regulatory Commission).
Where, the evaluation method according to the atmospheric dispersion models means the method to
calculate the substance concentration in the air by multiplying the release rate of radioactive materials to
the relative concentration obtained assuming that the spacial substance concentration is a horizontally andvertically normal distribution based on the specified meteorological conditions, such as the atmosphericstability and wind speed, during the accident period. However, the building effect is adjusted, if any.
The HGSYSTEM/UF6 code evaluates the diffusion after the UF6 is released to the atmosphere. The
chemical and thermodynamic model of UF6 is incorporated into the code based on the HGSYSTEM
Version 3.0 code developed by Shell Research Ltd. The code can treat high-density gas, drying andwetting deposition, involvement effect of the building, plume floating, and effects of complex
topography15) ~ 17). NUREG/CR-6481 "Review of Models Used for Determining Consequencesof UF6
Release"11) specifies criteria of the source term evaluation and criteria of atmospheric dispersion
evaluation, and shows a pairwise applicability comparison of two or more calculation codes, suggestingthat the HGSYSTEM/UF6 code is excellent as of 1997.
The RASCAL code was developed by the US NRC, and the NRC emergency activity center uses thecode to evaluate the exposure dose by radiological materials released to the atmosphere in an accident atpresent.
4.1.2. Results of JNES study on the evaluation of UF6 diffusion and migration in the air
4

JNES started, using simple calculation methods 4 according to the R.G. 1.145 model and theatmospheric dispersion model which is used in Japan 5 , 18) (hereinafter referred to as "Japanesemeteorological guide model") and the method of RASCAL4.2 and HGSYSTEM/UF6 codes, evaluation
and analysis of diffusion and migration of UF6 in the air for the following two typical cases;(1) Liquid UF6 leakage accident due to 30B cylinder damage;(2) UO2F2 + HF leakage accident.
Examples of the results are shown in Figure 4 .1 and Figure 4 .2.
Consequently, the following results were obtained;The simple calculation method according to the Japanese meteorological guide model (in the case of
no involvement effect by buildings) gives the most significant inhalation value.
RASCAL 4.2
HGSYSTEM/UF6
Japanese meteorological guide model without building effect

1.0E+6

1.0E+7RASCAL 4.2 average concentration
RASCAL 4.2 1hr equivalent concentration
HGSYSTEM/UF6
Japanese meteorological guide model without building effect

1.0E+6

1.0E+5

1.0E+5

U Inhalation Intake (mg)

1.0E+4

HF concentration (ppm)

1.0E+4

1.0E+3

1.0E+3

1.0E+2

1.0E+2

1.0E+1

1.0E+1

1.0E+01.0E+010100

distance (m)

1000

10

10000

Uranium inhalation from the release point100

distance (m)

1000

10000

HF concentration from the release point

Figure 4.1 Liquid UF6 leakage accident due to 30B cylinder damage
1.0E+5

1.0E+4RASCAL 4.2 average concentration

1.0E+6RASCAL 4.2
HGSYSTEM/UF6
Japanese meteorological guide model without building effectJapanese meteorological guide model with building effect
RG-1.145 Eq.(3)RASCAL 4.2 1hr equivalent concentration
HGSYSTEM/UF6

1.0E+5

Japanese meteorological guide model without building effectJapanese meteorological guide model with building effect

1.0E+4RG-1.145 Eq.(3)1.0E+3

HF Conentration (ppm)

U Inhalation Intake (mg)

1.0E+3

1.0E+2

1.0E+2

1.0E+1

1.0E+1

1.0E+01.0E+01.0E-1

1.0E-1

10

100

distance (m)

1000

10000

10

Uranium inhalation from the release point100

distance (m)

1000

10000

HF concentration from the release point

Figure 4.2 UO2F2 + HF leakage accident
4

The simple calculation method means the method to obtain the relative concentration supposing the severe simple
meteorological condition (atmospheric stability: F type, wind speed: 1 m/s, etc.), avoiding statistical processing of meteorologicalobservation data.
5 Although R.G. 1.145 model and Japanese meteorological guide model use the same evaluation formula when the involvement
effect of buildings is not considered, those models use different evaluation methods when the involvement effect of buildings is
considered.

5





In the case of including the involvement effect of buildings in the short distance where theinvolvement effect is significant, the simple calculation method according to the R.G. 1.145 model
gives a relatively larger inhalation value than the Japanese atmospheric dispersion model.
Concerning the liquid UF6 release accident, since the HGSYSTEM/UF6 code incorporates thevolatile phenomena from a failed part, it provides detailed analysis and is considered more suitable
than the RASCAL4.2 code.

4.2. Criteria of chemical effect4.2.1. Criteria of U.S.
As shown in Table 4.2, NUREG-1520 classifies significance (degree of incidence) into three levels andconsiders the levels of ERPG (Emergency Response Planning Guidelines) of the AIHA (American
Industrial Hygienist Association) and AEGL (Acute Exposure Guideline Level) of the EPA (U.S.
Environmental Protection Agency) so that definitions of the amount of chemical dose (CD) of each effectrate of Codes of Federal Regulation and each level of ERPG6 and AEGL7) are consistent.
AEGL has the minimum number of authorized compounds. It is due to strict scientific examination byexperts, which is considered more reliable. TEEL (Temporary Emergency Exposure Limits) data containchemical substances handled by DOE facilities (more than one thousand and several hundred), whichhave no threshold in AEGL and ERPG data and are temporarily specified to evaluate the toxicity, andsubstances and the values are reviewed each year, and the substances open to AEGL or ERPG are deleted.
In the U.S., there are three threshold data types of chemical substance exposure, as compared in Table
4.3. The DOE specifies on the webpage that the priority for adopting the thresholds is AEGL > ERPG >TEEL7.
Table 4.2 Classification of significance by the Codes of Federal Regulation 10 CFR 70.61 and
NUREG-1520
Workers

Offsite Public* RD > 1 Sv

Highsignificance** CD = endanger life.
(Exceeding ERPG-3 or
AEGL-3)250 mSv < RD ≤ 1 Sv

MediumsignificanceLowsignificanceCD = irreversible or otherserious and long-lasting
health effects(Exceeding ERPG-2 or
AEGL-2)
Accidents with lower
radiological and chemical
exposures than those abovein this column* RD: Radiological Dose (effective dose)

RD > 250 mSv
30 mg soluble uranium
intake, CD = irreversible orother serious and long-lasting health effects(Exceeding ERPG-2 or
AEGL-2)
50 mSv < RD ≤ 250 mSv
CD = mild transient healtheffects(Exceeding ERPG-1 or
AEGL-1)
Accidents with lower
radiological and chemical
exposures than those abovein this column

Environment

Radioactive release > 5000
x threshold limits specified
in Table 2 of the
supplementary provision Bof 10 CFR Part 20
Radioactive releases
producing lower effects
than those referenced abovein this column* *CD: Chemical Dose, (XX) show the data of NUREG-1520.

6

http://response.restoration.noaa.gov/oil-and-chemical-spills/chemical-spills/resources/
emergency-response-planning-guidelines-erpgs.html
7 https://orise.orau.gov/emi/scapa/chem-pacs-teels/default.htm

6Furthermore, the standard review plan, NUREG-171819), of MOX fuel fabrication facilities states that
ERPG or AEGL may be adopted. It also describes that, when other data are to be used, materials thatprove the validity must be shown. Table 4.4 shows the AEGL and ERPG thresholds of UF6 and HF.
Moreover, NIOSH (National Institute for Occupational Safety and Health) has released the IDLH
(Immediately Dangerous to Life or Health) values of the case of exposure time of 30 minutes of workers8.
4.2.2. Criteria of France
The threshold criteria of France are shown in Appendix 2 of the arrêté (Ministry Order) of September
29, 2005. The classification of thresholds of acute toxicity effects, which is described in Appendix 2, isshown in Table 4.5.
The threshold of each chemical substance is specified for the land use plan based on Table 4.5 by theexpert group of toxicology9. Table 4.6 shows a part of the France thresholds of UF6 shown in the report20)
and of HF shown in the report21), which are research reports published by Institut national del'environnement industriel et des risques (INERIS) (National Institute for Industrial Environment and
Risks) as the France thresholds of acute toxicity effects. Moreover, Reference 22, which incorporates
some of the results of European project ACUTEX23), describes that persons of high susceptibility are notincluded in the population exposed to chemical substances.
Table 4.3 Comparison of toxic threshold data on chemical substances in U.S.
AEGL
ERPG
TEEL
Responsible organization
EPA
AIHA
DOE
Evaluation of the
Planning of
Implementation of theacute exposure health emergency response
impact evaluation of
Purpose of the data
effects, planning of
measures, evaluation
chemical substances
emergency responseof the acute exposure
not contained in themeasureshealth effects
AEGL or ERPG data
All persons, including Excluding individuals All persons, including
Assumed exposed
individuals of highof high susceptibility
individuals of highpersons
susceptibility (Note)(Note)
susceptibility (Note)
Level of health effect
1, 2, 3
1, 2, 30, 1, 2, 3
10 minutes, 30
1 hour15 minutes including
minutes, 1 hour, 4
the time of thehours, 8 hours
maximum
concentration; in the
Exposure time
future, 1 Schedule
unified into a time
To be unified in 1
hour in the future
Number of open
130 (as of the middle
145 (as of the middle> 3000compoundsof 2013)of 2013)
Note: Individuals of high susceptibility mean infants, children, and persons whose resistance capability of the lung functions,such as pneumoconiosis, pulmonary emphysema, and liver function, cardiac function and immune function is weak, who are
expected to show health effects often.
8

http://www.cdc.gov/niosh/idlh/intridl4.html. IDLH values are specified assuming that the workers can maintain the capability
to evacuate within 30 minutes not resulting in death or irreversible health effects.
9 Organized by Ministère de l'Écologie, du Développement durable et de l'Énergie Institut national de l'environnement industriel
et des risques (INERIS) Institut national de la recherche scientifique (INRS), Institut de radioprotection et de sûreté nucléaire
(IRSN), university hospital centre, industry representative, and toxicologist.

7

INERIS showed the guideline for licensees to decide the not-described (ND) values in the research
report24) in 2009. For the "land use plan," the first priority is to refer to the references used to specify U.S.
AEGL, ERPG and other criteria, and to decide the values according to the France method described in the
INERIS research report, the second priority is to use values of ERPG, IDLH (SEI values only) and others,
and the third priority is to use the values of AEGL10. The priority of the values for the "emergency
response" is ERPG, AEGL, IDLH (SEI values only), and TEEL. Since SER of Table 4.6 for UF6 specifies
the not-described (ND) values, the specified values are adopted based on the base data of U.S. AEGL-1 or
ERPG-1 according to the way of thinking mentioned above.
Table 4.4 AEGL and ERPG thresholds of UF6 and HF
UF6 (mg-UF6/m3)

AEGL

ERPG

10
min.

30
min.

1
hr.

4
hr.

8
hr.

AEGL-1

3.6

3.6

3.6

NR

NR

ERPG-1

5

AEGL-2

28

19

9.6

2.4

1.2

ERPG-215

AEGL-3

216

72

36

9.0

4.5

ERPG-330

Exposure timeLevel

HF (ppm)

Exposure time
Level

AEGL

1
hr.

ERPG

10
min.

30
min.

1
hr.

4
hr.

8
hr.

AEGL-1

1.0

1.0

1.0

1.0

1.0

ERPG-1

2

AEGL-2

95

34

24

12

12

ERPG-220

AEGL-3

170

62

44

22

22

ERPG-350

Exposure time
Level

Exposure time
Level

1
hr.

http://www.epa.gov/oppt/aegl/pubs/final.htm

Table 4.5 Classification of thresholds of acute toxicity effect in France24)
Threshold
SER(Seuil des Effets Réversibles)
SEI(Seuil des Effets Irréversibles)
SPEL (SEL) *1(Seuil des Premiers Effets Létaux
(Seuils des Effets Létaux))
SELS(Seuil des Effets Létaux
Significatifs)

Description
Reversible Effects threshold
Concentration of substances in the air that may cause reversible
effects when exceeding the specified exposure time
Irreversible Effects threshold
Concentration of substances in the air that may cause irreversible
effects when exceeding the specified exposure timeFirst Lethal Effects threshold or Lethal Effects Threshold
Concentration of substances in the air that may cause 1% deathwhen exceeding the specified exposure time
Significant Lethal Effects threshold
Concentration of substances in the air that may cause 5% deathwhen exceeding the specified exposure time

*1: Appendix 2 describes SPEL as SEL.
10

Reference 25 describes that although AEGL is scientifically reliable, the guideline is too conservative as the criteria used forthe purpose of land use plan since individuals of high susceptibility are also taken into consideration.

8

Table 4.6 France thresholds of UF6 and HF acute toxic effects
UF6 (mg-UF6/m )Source: INERIS-DRC-10-1009947-04672A
ND: Not Described
3

1-min

10-min20-min30-min

1-hr

2-hr

4-hr

8-hr

SER

ND

ND

ND

ND

ND

ND

ND

ND

SEI

ND

123

ND

41

21

ND

ND

ND

SPEL

7,790

3,614

1,800

1,210

605

302

144

72

SELS

12,614

5,861

2,923

1,958

979

490

245

115

HF (ppm)

Source: INERIS-DRC-08-94398-12729A

ND: Not Described

1-min

10-min20-min30-min

1-hr

SER

60

12

7

5

1

SEI

ND

600

ND

200

100

SPEL

11,100

1,123

563

377

189

SELS

17,147

1,705

851

567

283

4.2.3. Criteria of U.K.
The criteria of U.K. include SLOT (Specified Level of Toxicity) and SLOD (Significant Likelihood of
Death) for HF specified by the British Health and Safety Executive (HSE). SLOT is specified for the land
use plan as the following level: all persons in the area suffer from intense pain, considerable persons
require a doctor's medical examination, some seriously injured persons require treatment over a long
period, and highly susceptibile persons die, and, on the other hand, SLOD is specified as a level of 50%
lethality26). Both criteria are indicated as the amount of toxic loading (= concentration (ppm) x exposure
time (min)), In the case of HF, it is 12,000 for SLOT and 21,000 for SLOD.
4.2.4. Criteria of JapanThe criteria of Japan concerning UF6 and HF are as follows: there is no quantitative limit for public
safety since the criteria of UF6 is for registration and notification. The standard of HF is for venting andwork-environment control, not for an accident. Therefore, the criteria of Japan cannot be used as criteria
for accidents that could lead to severe accidents(1) For UF6 (soluble uranium, including UO2F2)
 The duty of registration and notification concerning sale, manufacturing, handling and otheractivities according to the Poisons and Deleterious Substances Control Law.
(2) For HF
 Emission standard (minimum) of Enforcement Regulations of Air Pollution Control Act: 1
mg/Nm3 (1.5 ppm)
 Ministry Order of Effluent Standard for the Water Pollution Control Act (Fluorine and its
Compounds): 8 mg/liter (except offshore areas), and 15 mg/liter (offshore areas) Control standard of the workspace concentration according to the Prevention of Hazards due toSpecified Chemical Substances of Industrial Safety and Health Act: 0.5 ppm
9

4.3. Discussion
4.3.1. The UF6 chemical effect evaluation method for accidents that could lead to severe accidents
The way of thinking, methods, and items to be considered to identify to perform the UF6 chemical
effect evaluation for accidents that could lead to severe accidents are as follows:(1) Identification of accidents that could lead to severe accidents(2) Evaluation of the source term of each identified accident that could lead to severe accidents
 To evaluate the chemical effects for each of the worker of the processing room, on-site staff, andthe public in the vicinity of the site, the source term of severe accidents must be obtained for boththe building inside (processing room) and building outside.
 Identify the UF6 leaked and the pathway of the reaction products.
(3) Exposure concentration evaluation of hazardous chemical substances The exposure concentration should be evaluated for each of UF6, UO2F2 and HF.
 For the evaluation considering the diffusion outside the building, the following method i results inconservative effects. However, if the effect is too conservative, the following method ii may be
applied to the evaluation considering the chemical and thermodynamic effects of UF6.
i. Evaluation method according to the Japanese meteorological guide modelii. Evaluation method using the proved and verified calculation codes (HGSYSTEM/UF6 code,
RASCAL4.2 code etc.)(4) Evaluation of exposure time The exposure time assumed for each of the worker, the staff and the public should be specified
appropriately based on the condition of the facilities considered, and the basis should be shown.
(5) UF6 chemical effect evaluation for accidents that could lead to severe accidents
 The UF6 chemical effect is evaluated based on the exposure concentration and time.
 AEGL is desirable to use as the criteria of chemical effect evaluation because of the following
reasons. When other criteria are used, the appropriateness of the criteria should be explained.
(a) AEGL is considered most reliable since many references are available, strict examination byexperts has been performed, and it is utilized for evaluation of most UF6 related facilities in
the U.S.
(b) Individuals of high susceptibility are also included in the study, and the thresholds areconservative compared with other data.
4.3.2. Validation of measures against UF6 leakage in accidents that could lead to severe accidents
The following items should be considered for the validation:(1) Acceptability of chemical effect evaluation results of the above item (5) is judged for the public in
the vicinity and workers, with due consideration of the following two endpoints.
(a) For the defenseless public in the vicinity, the effect does not cause any “mild transient healtheffects”.
(b) For the workers and staff to whom defenses and exercises are available, the effect does notcause any “irreversible or other serious and long health-lasting effects”.
(2) Based on these items to be considered, AEGL is used to determine the acceptability of the effect. The
measures are appropriate when acceptable, but further measures should be taken when notacceptable.
105. Confirmation of current status of licensees with enrichment and reconversion activites
Learning from the lessons of the accident of the Fukushima Daiichi Nuclear Power Station, the "Act for
the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors (Nuclear RegulationAct)" was revised in June 2012. Moreover, new regulatory requirements for nuclear fuel and otherfacilities were enforced in December 2013. For the enforcement, it was decided that the new regulatory
requirements should not interfere with the fuel processing for five years after the enforcement. But, the
NRA requested, by its direction4), licensees who manage the equipment of UF6 treating processes underpositive pressure: the licensees of enrichment activity (JNFL: Japan Nuclear Fuel Limited who owns thehomogenizing and blending equipment for the enriching process) and the licensees of reconversion
activity (MNF: Mitsubishi Nuclear Fuel Co. LTD, who owns the UF6 evaporation and hydrolysis
equipment for the reconversion process) to confirm that there is no possibility of the risk or remarkablechemical effects due to excessive radiation exposure on the public and to report the results to the NRA.
The contents addressed in this paper were reflected in the evaluation by licensees for such confirmation,and it was recommended to use the AEGL values as the criteria and appropriate evaluation methods likethe Japanese meteorological guide model or HGSYSTEM/UF6 code, etc. for the analysis.
The secretariat of NRA evaluated the reports27), 28) submitted by JNFL and MNF and the evaluation
results were accepted in the NRA meetings in April 2017 and September 2017, respectively.
6. Conclusion
The new regulatory requirements of the nuclear fuel cycle facilities in Japan incorporated measuresagainst the UF6 chemical effect in accidents that could lead to severe accidents. The current statuses of the
U.S., France and U.K. were investigated in the study of the method to evaluate UF6 chemical effects
necessary to enforce the related regulations, and the results were summarized in the report,
JNES-RE-2013-20211). Moreover, the results, such as using AEGL values as the criteria and using theJapanese meteorological guide model or HGSYSTEM/UF6 code for UF6 exposure concentration
evaluation, were applied to the confirmation conducted by licensees of enrichment and reconversion
activity to verify that UF6 leakages did not result in a remarkable radiation and/or remarkable chemical
effects to the public based on the direction of the NRA4).
Reference
1)
2)
3)
4)

5)
6)
7)

JNES, “Chemical Effects and their Evaluation Methods for Accidental Release of Uranium
Hexafluoride”, JNES-RE-2013-2021, (Dec. 2013), [in Japanese]
IAEA, “Safety of Nuclear Fuel Cycle Facilities”, NS-R-5, (2008).
IAEA, “Preparedness and Response for a Nuclear or Radiological Emergency”, GS-R-2, (2002).
Nuclear Regulation Authority, ”Submission of a report on the chemical effects of uranium
hexaflouride handed in the uranium fuel processing facility to the public (NRA’s instruction)”,
NRA-11d-13-023, (Dec. 11, 2013), [in Japanese].
U.S. Department of Health and Human Service, Agency for Toxic Substances and Disease
Registry, “Toxicological Profile for Uranium”, (2013).
U.S. Code of Federal regulations, “29 CFR 1910.1000, Subpart Z”, (2006).
EPA, “Acute Exposure Guideline Level”,
http//www.epa.gov/opt/aegl/process.htm.
118） NRC, “Chemical Toxicity of Uranium Hexafluoride Compared to Acute Effects of Radiation”,
NUREG-1391, (Feb. 1991).
9)
R. L. Kathren, et. al., “Acute chemical toxicity of uranium”, Health Physics 94(2), 170-179,(2008).
10) NRC, “Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility”,
NUREG-1520 Rev.1, (May 2010).
11) NRC, “Review of Models Used for Determining Consequences of UF6 Release”,
NUREG/CR-6481, (Nov. 1997).
12) NRC, “Nuclear Fuel Cycle Facility Accident Analysis Handbook,” NUREG/CR-6410,(Mar.1998).
13) DOE, “Airborne release fractions/rates and respirable fractions for nonreactor nuclear facilities”,
DOE-HDBK-3010-94, (Dec.1994).
14) NRC, “Technical Basis for Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential
Accident Consequence Assessments at Nuclear Power Plants”, NUREG/CR-2260, (Jul. 1981).
15) DOE, et al., “Technical Documentation of HGSYSTEM/UF6 Model”, K/SUB/93-XJ947/1, (Jan.
1996).
16) S. R. Hanna, et al., “HGSYSTEM/UF6 Model Enhancements for Plume Rise and Dispersion
around Buildings, Lift-Off of Buoyant Plumes, and Robustness of Numerical Solver”,
K/SUB/93-XJ947/2R1, (Jan. 1997).
17) M. W. Yambert, et al., “A Summary of Recent Refinements to the WAKE Dispersion Model, aComponent of the HGSYSTEM/UF6 Model Suite”, ORNL/TM-13666, (Aug. 1998).
18) Nuclear Safety Commission of Japan, “The meteorological guide for safety analysis of nuclear
power reactor facilities”, （Mar. 2001）, [in Japanese].
19) NRC, “Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel
Fabrication Facility”, NUREG-1718, (Aug. 2000).
20) INERIS, “Méthodologie de détermination des valeurs seuils de toxicité aiguë françaises en casd’émission accidentelle de substances chimiques dans l’atmosphère”, DRC-07-82347-07520A,(2007).
21) INERIS, “Seuils de Toxicité aiguë Hexafluorure d’uranium”, DRC-10-103128-02257A, (2010).
22) NRC, “Final Safety Evaluation Report on the Construction Authorization Request for the Mixed
Oxide Fuel Fabrication Facility at the Savannah River Site, South Carolina”, NUREG-1821, (Mar.
2005).
23) ACUTEX project, “Methodology to develop AETLs (Acute Exposure Threshold Levels)”,
ACUTEX, (Jan.2006).
24) INERIS, “Technical Guidance document for selecting acute toxicity threshold values in theabsence of official French values”, DRC-09-103128-07577A”, (2009).
25） DOE, “O 151.1C FAQ Program Elements: Consequence Assessment; Hazards Survey/Hazards
Assessment (Technical Planning Basis); Protective Actions and Reentry Rev. 0”, (Jul. 2007).
26) HSE, “Toxicity levels of chemicals”, http://www.hse.gov.uk/chemicals/haztox.htm.
27) JNFL, “Report on chemical effects that handling of uranium hexafluoride in Rokkasho uranium
enrichment plant effect on the public”, Reported as of Aug. 19, 2016, partly amended as of Mar.
23, 2017 and Apr. 14, 2017.
28) MNF, “Report on chemical effects that handling of uranium hexafluoride in the reconversion plant
effect on the public”, Reported as of Mar. 9, 2017, partly amended as of Sept. 6, 2017.
12

13

Comparison of Operating Experience from Nuclear Fuel Cycle Facilities
and Industrial Chemical Plants
Margie Kotzalas
U.S. Nuclear Regulatory Commission
Abstract: While nuclear fuel cycle facilities and commercial industrial plants have different
regulatory schemes and process requirements, both use hazardous chemicals to manufacture
their products. This paper discusses operating experience feedback from nuclear fuel cycle
facilities and from commercial industrial plants in order to identify lessons in the safe
management of chemical hazards.
Introduction
This paper compares lessons learned from industrial chemical incidents with those from nuclear
fuel cycle facilities. While nuclear fuel cycle facilities are often more regulated and held tostricter safety standards and inspections by regulatory authorities, lessons can be learned fromoperating experience gained from industrial incidents.
Incidents that occurred during maintenance or non-routine activities at nuclear fuel cycle andindustrial chemical facilities were selected for comparison. In most cases, the causes of theincidents involve: (1) inadequate procedures or procedural adherence, (2) inadequate process
hazards analyses or improper assumptions, (3) inadequate management of change control
process, and (4) improper configuration management and equipment lineup.
Lessons learned and actions taken to address causes apply to many situations. While each
incident is different, a robust safety culture that includes a wide variety of operating experienceis helpful in preventing the occurrence or lessening the severity of future incidents.
Background
The U.S. Chemical Safety Board is an independent U.S. Federal agency that investigates
industrial chemical incidents, determines the root causes for the incidents, identifies lessons
learned, and develops recommendations to address the identified weaknesses. Their reports
are publicly available on the internet 1. Information about nuclear fuel cycle incidents, causes,
and lessons learned are maintained in the joint Nuclear Energy Agency/International Atomic
Energy Agency Fuel Incident Notification and Analysis System 2 (FINAS) database of operating
experience. Information about the incidents analyzed in this paper was gathered from thesesources.
Periods of non-routine operations, such as maintenance, have greater risk significance because
they are not frequently performed; equipment may be out of service; equipment configuration
may be different; and procedural knowledge and accuracy may not be adequate. Fourincidents, two industrial chemical and two nuclear fuel cycle, were selected for comparison. The
incidents, causes, and lessons learned are described in the Discussion section.

1
2

U.S. Chemical Safety Board investigation reports are available at www.csb.gov.
FINAS database is available at www.iaea.org.

Discussion
La Hague Plutonium Oxide (PuO2) Spill 3
Process Description
Spent nuclear fuel assemblies are processed in various workshops at La Hague. One of theseworkshops is used to purify, convert, and package plutonium oxide recovered from spent fuel.
The PuO2 powder is poured into a homogenizer. The powder is then directed by a pipe, which
is monitored by neutron counters, to a sieve. The latter is watched by a camera and supplies afilling head. The downstream part of the filling head is located in a glove box. A can is bound to
the downstream part of the filling head to be filled with PuO2 powder. The full can is weighed,sealed and transferred into another glove box to be put into a case by means of a bracket
before being stored in another workshop.
Safety controls are in place for filling the cans with PuO2 powder and for packaging the filledcans into cases. In the filling process, a can is bound to the filling head, and its mass ismonitored by means of spring scales. The spring scales automatically stop the filling process
when a set point for the PuO2 mass in the can is reached. In addition, a sensor, located in thefilling head, stops the filling process when the can is full.
When the can is full, a valve closes the filling head to prevent the loss of PuO2 powder into theglove box. In the workshop, the operating employees monitor the process with several process
software controllers (PSC). At this point, an operator undocks the full can from the filling head,transfers it to another glove box, and it is then packaged into a case.
Description of Incident
On September 18, 2008, the workshop was not in operation, and the operator decided to modify
one of the PSCs in order to reduce the packaging time of the cans (by reducing the height of thetransfer by the bracket from 2 m to 0.5 m). The change was governed by a modification
authorization document. Only post-modification testing of the semi-automatic mode of operation
was performed, because the automatic control mode can only be tested when the workshop isin operation.
On September 23, 2008, a new campaign to fill the cans with PuO2 powder and to packagethese into cases was started. This enabled post-modification testing of the automatic control
mode. During that testing phase, a malfunction in the automatic control mode concerning thebracket occurred. Subsequently, the bracket was then used in the semi-automatic control
mode.
On September 24, 2008, in the morning a maintenance team began the repair of the deficient
controller. During the maintenance, no movement of the cans was permitted in the associatedglove box. So, the supply of PuO2 powder to the filling head was stopped. At that time an
empty can was bound to the filling head and a full can remained in the glove box to betransferred.
In the beginning of the afternoon, an operator, who was not aware of the ongoing maintenanceactivities, restarted filling of cans and continued for an hour and a half. The operator then put
3

FINAS incident #142

the glove box back to the same condition as reported 2 hours earlier, except now a full can wasconnected to the filling head.
Shortly thereafter, the maintenance employees came back to perform their work. They tested
that operation in the semi-automatic mode was correct.
The facility was then put back in operation to validate the modification to the automatic mode,and the operator restarted the filling process. He monitored the task from the control room.
Within 20 minutes, he observed that the increase of the mass of the can was unusual so hestopped the filling operation.
Since the sensor in the filling head had not been triggered, the manufacturing manager thoughtthat the spring scales were deficient and decided to undock the can from the filling head toconfirm his hypothesis (not realizing that the can and the filling pipe were full of PuO2 powder).
During the undocking, PuO2 powder spilled into the glove box and the full can was re-docked to
the filling head.
Around 15 kilograms (kg) of PuO2 powder had accumulated in the pipe of the filling head andapproximately 600 grams of PuO2 powder spilled into the glove box. This did not present a
criticality risk and, due to the design of the glove box, the incident had no consequence to thestaff or environment.
Causes of Incident
1. The modification procedure did not consider that the validation tests after completion of themodifications resulting from maintenance were being performed while the facility was inoperation. The procedure was not revised to account for the PuO2 powder in the can docked to
the filling head.
2. The sensor in the filling head failed. Therefore, it did not perform its safety function to stop
the filling of the docked can.
3. The process hazards analysis regarding the sensor was incomplete. The geometry of the
glovebox ensured criticality safety. So, the sensor in the filling head was mistakenly notconsidered in the process hazards analysis. Because of this, it was not surveilled or
maintained.
Lessons Learned
1. Adequate procedures for controlling modifications are critical. They should require process
hazards analyses to consider the risk caused by new features of a modification.
2. Thoroughness of process hazards analyses is important such that equipment like the sensor
will be identified as a safety control and will undergo periodic surveillance.

Nuclear Fuel Services (NFS) Generation of Nitrogen Oxides (NOx) 4
Process Description
The Uranium (U)-Aluminum bowl cleaning station system is designed to remove uranium from
centrifuge bowls by circulating nitric acid through the bowls. The system has historically
produced NOx during the nitric acid dissolution process. Safety controls include a NOx
detection system with sensors located at the nitric acid knockout column's siphon break(potential NOx release point if process ventilation fails) and at the employee working level.
Description of Incident
On October 13, 2009, NFS began using the bowl cleaning station system to dissolve UAluminum fines rather than adding them to the normal dissolver column. Laboratory testing
indicated that dissolving the fines with caustic in the normal dissolver column was notrecommended due to potential for a vigorous reaction. The fines were loaded into strainers andplaced directly into the bowls to be dissolved with nitric acid. After the dissolution process
began, the operator noticed that the temperature of the system was increasing and that NOx (inthe form of a brown cloud) was beginning to form inside the bowl cleaning station containment
vessels. The system was shutdown. The NOx detector alarmed and the facility was evacuated.
The safety assessment determined that NFS experienced an unexpected exothermic reaction.
The elevated temperatures from the reaction created nitrogen compound gases within theassociated process off-gas piping. An instrument located near the ceiling of the facility detected
these gases and generated an alarm which resulted in the evacuation of employees from theaffected area. Additionally, the elevated temperature of these gases caused portions of the offgas piping to deform and sag in the nearby area. NFS personnel took action to shut down thesystem. As a result, no personnel were injured and offsite environmental releases during theperiod associated with the event were within regulatory limits.
Laboratory analysis of similar U-Aluminum fines material was conducted and a NOx generation
rate specific for the fines material was estimated. Based on engineering calculations, it wasdetermined that the NOx generation for the fines was significantly higher than the previously
analyzed NOx generation for the U-Aluminum ingots.
Causes of Incident
The incident was caused by the failure to recognize the possibility of increased NOx generation
due to the greater surface area and aluminum content of the material to be dissolved. Although
the safety basis assumed that NOx would be generated during processing and controls were inplace to alert workers to evacuate, the consequence calculations did not include a NOx
generation rate that specifically addressed the differences between the U-Aluminum finesmaterial and the material previously processed in the bowl cleaning station. No equipment
failed; however, the steam evolved during the chemical reaction forced solution and slurry into
ventilation piping which resulted in deformation of the polyvinylchloride off-gas piping.

4

FINAS incident #148

Specifically, the causes included:
1. Inadequate management oversight of facility process changes; and2. Inadequate process safety hazard evaluation with inaccurate assumptions.
Lessons Learned
1. Robust process hazards analyses with technically justifiable assumptions are necessary toensure that appropriate safety controls are identified and implemented to provide adequate
protection and to ensure that management controls for those controls are sufficient, such thatthey are available and reliable to perform their intended safety function when needed.
2. Management expectations for strong safety culture set the example for all the employees.
Management presence and engagement on the process floor enables open and timely
communication of potential safety concerns.
3. Effective change management process should include a requirement that changes toprocedures associated with processes, process parameters, and process inputs be clearly
defined prior to implementation.
ExxonMobil Refinery Explosion5
Process Description
On February 18, 2015, an explosion occurred in the ExxonMobil refinery’s electrostatic
precipitator (ESP) in Torrance, California. The incident occurred when the operator was
attempting to isolate the equipment for maintenance while the unit was in an idled mode ofoperations. Preparations for maintenance caused a pressure deviation that allowedhydrocarbons to backflow through the process and ignite the ESP. Debris from the explosion
was thrown into other units of the refinery directly surrounding the ESP. One of these pieces ofdebris hit scaffolding in the refinery’s alkylation unit, narrowly missing a tank containing tens of
thousands of pounds of hydrofluoric acid (HF). Had the debris struck the tank, a rupture could
have been possible, resulting in a potentially catastrophic release of toxic HF into theneighboring community.
Description of Incident
The sequence of events that eventually led to the explosion at the refinery began on February
12, 2015, when problems with a piece of equipment called an expander caused the refinery’s
fluid catalytic cracking (FCC) unit to be put into an idled condition.
With the FCC unit shut down, steam was forced into a reactor to prevent hydrocarbons fromflowing back from the main distillation column. On the morning of the incident, this steam wasescaping through an open flange on the expander, preventing operators from continuing theirmaintenance work. It had traveled through a leaking slide valve connected to the reactor.

5

U.S. Chemical Safety Board ExxonMobil Torrance Refinery Electrostatic Precipitator Explosion Torrance California
Investigation Report, Published May 2017, No. 2015-02-I-CA

An outside supervisor then reduced the amount of steam being forced into the reactor so thatwork could continue. However, at the time, workers were unaware that hydrocarbons wereleaking into the main distillation column from interconnected equipment. As the pressure of thesteam dropped, the hydrocarbons flowed back into the reactor, out through the leaking slide
valve, and eventually into the ESP. There, the hydrocarbons found an ignition source andexploded.
Causes of Incident
Multiple process safety management deficiencies helped contribute to the incident.
1. In order to perform work to bring the FCC unit back online, the operators determined theyneeded to deviate from several existing operating procedures. So, they used a written
temporary deviation from normal operating procedures. This deviation was created in 2012 toaddress problems with the expander. However, ExxonMobil did not conduct a management ofchange review before implementing this outdated deviation, even though conditions within the
FCC unit had changed over the previous three years.
2. ExxonMobil performed inadequate process hazards analyses, which could have identified
more effective safeguards against the flow of hydrocarbons, such as installation of
instrumentation to detect flammable hydrocarbons flowing through the equipment into the ESP,or establishment of safe operating limits for the FCC unit.
Lessons Learned
1. It is essential to identify and define safe operating limits for all modes of operation andmeasure process conditions and parameters that can verify the operation of the process relative
to those safe operating limits. When a facility relies on operating parameters that only indirectly
provide information on critical process parameters it can lead to the inability to identify when aprocess is in an unsafe mode.
2. When implementing a deviation from an existing procedure, it is critical that the company
conduct a management of change review to verify and authorize the technical basis and theimplementation time period and to identify any new or affected hazards and associatedmitigation strategies.
3. It is essential to schedule and perform maintenance of safety critical equipment, such that theequipment is available to perform its safety critical function.
4. It is important to consider all modes of operation including non-routine operations when
performing process hazards analyses. Incident scenarios could be possible during non-routinemodes of operation that may not have been considered when analyzing process hazards fornormal continuous operation.
5. Companies should develop procedures for all modes of operations that detail safe operating
limits, consequences of deviating from those limits, and specific actions to implement in theevent that the process deviates outside of its safe operating limit.
6. Robust management of change practices are needed when making changes to procedures.
Similar to process hazards analyses, conducting management of change as a multidisciplinarygroup composed of individuals with different areas of expertise can assist in identifying hazards
introduced by the procedure change.
Williams Geismar Olefins Plant – Reboiler Rupture and Fire6
Description of Incident
On June 13, 2013, a catastrophic equipment rupture, explosion, and fire occurred at theWilliams Geismar Olefins Plant in Geismar, Louisiana, in which two employees were killed. The
incident occurred during non-routine operational activities that introduced heat to a reboiler
which was offline, creating an overpressure event while the vessel was isolated from its
pressure relief device. The introduced heat increased the temperature of the liquid propane
mixture confined within the reboiler shell resulting in a dramatic pressure rise within the vessel.
The reboiler shell catastrophically ruptured causing a boiling liquid expanding vapor explosion
and fire.
Causes of Incident
Process safety management program weaknesses at the facility during the 12 years leading up
to the incident caused the reboiler to be unprotected from overpressure. These weaknesses
include deficiencies in:
1. Implementation of management of change reviews or conduction of pre-start up safety
reviews. As a result, the company did not evaluate and control all hazards introduced to theprocess by those changes.
2. Performance of adequate process hazards analyses such that effective over-pressure
protection by either a pressure relief valve or administrative controls would be applied.
3. Development of procedures for the non-routine operation performed the day of the incident.
Lessons Learned
1. Robust management of change practices are needed to ensure the review analyzes hazards
affected by the change in the entire process. Similar to process hazards analyses, conducting
management of change reviews as a multidisciplinary group can assist in identifying hazards
introduced by a process change.
2. Pre-startup safety reviews are key opportunities to verity effective implementation of design
intent, accuracy of process safety information, and proper installation and configuration of fieldequipment.
3. Ensure that the selected safety control adequately addresses the original safety concern
identified by the process hazards analysis team.
4. Operating procedures need sufficient detail to ensure effective performance of critical steps
including performing steps in the correct order. Affected employees should receive training on

6

U.S. Chemical Safety Board Williams Geismar Olefins Plant Reboiler Rupture and Fire Geismar, Louisiana Case
Study, published October 2016, No. 2013-03-I-LA

the procedures. Management should establish expectations to maintain and follow accurate
procedures.
5. Process hazards analyses and management of controls teams should effectively use thehierarchy of controls to the greatest extent feasible when evaluating safeguards.
Conclusions
While the incidents described in this paper have different causes and consequences, thelessons learned from the operating experience are applicable to industrial chemical and nuclear
fuel cycle facilities. These lessons can be summarized as the importance of conducting robust
process hazards analyses and diligent management of the change process. Process hazards
analyses should be conducted by a multidisciplinary team and ensure that assumptions aretechnically justifiable. The safety controls and operating limits identified by the analyses should
include the implementation of management controls to ensure that the established controls will
be available and reliable when called upon to perform their intended safety function. The
management of change process should ensure that procedures are adequate and expectations
that they must be followed are established. Additionally, the management of change process
should ensure that a process hazards analyses is conducted for the maintenance evolution andthe modified operation.
Through the sharing of operating experience from a variety of sources and the consideration oflessons learned, improvements will continue to be realized in chemical process safety.

Chemical Safety Requirements in the United States for Commercial
Fuel Cycle Facilities
M.Diaz and J.Hammelman
U.S. Nuclear Regulatory Commission
Mail Stop TWFN-4A60
Washington, DC 20555
ABSTRACT
Chemical safety regulatory requirements and practices vary somewhat between countries. This paper
presents the regulatory requirements and regulatory practices for commercial fuel cycle facilities in the
United States (U.S.). Regulations regarding chemical safety for commercial U.S. fuel cycle facilities arefound in Title 10 of the Code of Federal Regulations (CFR) Part 70, Domestic Licensing of Special
Nuclear Material. The U.S. Nuclear Regulatory Commission (NRC) regulations for fuel cycle facilities
that process fissile material have performance assessment requirements for acute chemical exposures anda requirement that licensees or applicants conduct a safety analysis to assess compliance with theperformance requirements. This paper presents an overview of these chemical safety performancerequirements and discusses the NRC’s approach to its review of an applicant or licensee or safetyassessment. Specifically, this paper addresses the analysis of toxic and reactive chemicals used at U.S.
commercial fuel cycle facilities, quantitative standards used to assess consequences and acute chemical
exposures to workers via different pathways.

I. OVERVIEW OF NRC REGULATIONS AND REGULATORY GUIDANCEPart 70 of Title 10 of the U.S Code of Federal Regulations provides the regulatory framework forlicensing of special nuclear material (SNM). SNM is defined as plutonium, uranium 233, anduranium enriched in 233U or 235U. NRC regulations, 10 CFR 70 Subpart H, provides specificrequirements addressing chemical safety requirements and requires applicants and licensees toperform an integrated safety analysis (ISA). The ISA is a systematic analysis that identifies: (a)facility and external hazards; (b) the potential of these hazards to initiate accident sequences; (c)what these potential accident sequences are, including their likelihood and consequences; and (d)items relied on for safety (IROFS).
The U.S. also has a Memorandum of Understanding (MOU) between the U.S. Occupational
Safety and Health Administration (OSHA) and the NRC, assigning the responsibility for the
management of chemical hazards related to processing radioactive material to the NRC. As
specified in 10 CFR Part 70, a fuel cycle facility applicant or licensee must develop and maintainan ISA. They must identify the chemical hazards associated with NRC-licensed material as wellas the hazardous chemicals produced from licensed material. The term hazardous chemicals
produced from licensed materials includes substances having licensed material as precursor
compound or substances that physically or chemically interact with licensed materials; and thatare toxic, explosive, flammable, corrosive, or reactive to the extent that they can endanger life or
health if not adequately controlled. These include substances commingled with radioactive
material, and include substances such as hydrogen fluoride that is produced by the reaction of
uranium hexafluoride and water, but do not include substances prior to process addition tolicensed material or after process separation from licensed material.

Additionally, a fuel cycle facility applicants or licensees must identify in its ISA, facility hazards
that could affect the safety of licensed materials and thus present an increased radiological risk.
For chemical hazards, applicants and licensees must limit the risk of accidents involving suchchemicals at their facilities. The performance requirements for acute chemical exposures statethat accident sequences with high consequences must be highly unlikely. High consequence
events are those that could endanger the life of a worker or lead to irreversible or other serious,
long lasting health effect to an individual located outside of the controlled area (e.g. site’s
property boundary). The performance requirements also state that accident sequences that could
lead to irreversible or other serious, long lasting health effect to a worker or cause mild transient
health effects to any individual located outside of the controlled area are considered events withintermediate consequences and must be unlikely. Highly unlikely and unlikely are terms definedby the applicant or licensee which identifies the frequency of occurrence of an event.
The regulations require an applicant or licensee to conduct an ISA that identifies credible chemicalhazards, accident sequences that could lead to the chemical hazards, and estimate the likelihoodand consequences of the accident sequences. The results are then compared to the performancerequirements for acute chemical exposures. The analysis is also used to determine if controls are
required to ensure compliance with the performance requirements. The regulations define suchcontrols as, “items relied on for safety” or IROFS.
NRC Regulations requires an applicant or licensee to describe in its ISA Summary proposed
quantitative standards, “used to assess the consequences to an individual from acute chemical
exposure to licensed material or chemicals produced from licensed materials.” The purpose of
these standards is to help classify the consequences into high or intermediate in the ISA.
NRC documents available to provide guidance on how to analyze chemical hazards andconsequences include: NUREG1513, “Integrated Safety Analysis Guidance Document;”
NUREG/CR-6410, “Nuclear Fuel Cycle Facility Accident Analysis Handbook;” Information
Notice 2007-022, “Recent Hydrogen Fluoride Exposures at Fuel Cycle Facilities;” and NUREG1520, “Standard Review Plan for Fuel Cycle Facilities License Application.” These documents
provides guidance on how to consider chemical hazards in an ISA and address relevant exposure
pathways (e.g., dermal and ocular pathways).
The following sections will provide details regarding the NRC’s evaluation of proposed
quantitative standards.
II. NRC REVIEW OF CHEMICAL SAFETY ANALYSISThe NRC examines the methodology and information used by the applicant or licensee to
identify hazards and accident sequences that could result in an acute chemical exposure toworkers and individuals outside the controlled area. The methodology should be systematic anduse information about the applicant or licensee’s material quantities, process, process equipment,
and operations. The NRC encourages the use of historical experience at similar facilities andoperations.

When evaluating chemical hazards, the NRC considers typical material properties such as
toxicity, flammability, and reactivity. The NRC evaluates the applicant or licensee’s
identification of credible chemical hazards that could potentially produce “high” or“intermediate” acute chemical exposure events. Table 1 provides information on the toxic orhazardous characteristics of some common chemicals used at fuel cycle facilities. Table 1
references sources of publically-available information on toxic or hazardous characteristics ofchemicals commonly present at U.S. fuel cycle facilities. However, this is not an exhaustive listof chemicals that may be used at such facilities.
The NRC recognizes that accidents often occur: (1) during non-routine operations including
maintenance where the hazards and controls are different from those of normal operation, (2) asa result of unanalyzed plant modifications where new hazards might be introduced, and (3) as aresult of operations being conducted outside of conditions examined in previous safety analyses.
Any locations where hazardous licensed material, including fissile material, could inadvertently
be located should also be considered. A review of previous accidents related to similar
operations is useful when evaluating the hazards present at a fuel cycle facility.
Review of Chemical Accident Consequences
NRC generally uses a multistep process to examine how the applicant or licensee has evaluatedthese potential chemical exposures consequences. When reviewing chemical exposure
consequences, it is important to understand whether the receptor is a worker inside the controlledarea, or is an individual outside the controlled area. This is important because the highconsequence and intermediate-consequence events described in the performance requirements
are different.
The first step involves assessing the material’s form and its concentration as it moves from therelease point to the receptor location, and the major physical processes involved in the initial
release and subsequent transport. Estimating and classifying chemical exposure consequences
further involve an assessment of multiple parameters such as vessel size and pressure, hole size,building ventilation characteristics, building dimensions, and local meteorology. Methods for
conducting these types of analyses are discussed in NUREG/CR-6410, “Nuclear Fuel Cycle
Facility Accident Analysis Handbook,” and the Center for Chemical Process Safety’s,“Guidelines for Chemical Process Quantitative Risk Analysis,” published in 1999. The NRC
determines if the methods for estimating release rate and release conditions are reasonable (i.e.,
the results are not clearly biased in a way that underestimates consequences) for the physicalproperties of the material being released.
The second step involves determining the nature (e.g., exposure pathways) and the approximate
duration of the chemical exposure. This estimate requires an understanding of the properties of
the transported material (developed by the first step), an estimate of the effectiveness of any
protective equipment, and an awareness of any actions of the exposed individual that would
influence exposure (e.g., exposure time). The NRC then determines if the methods the applicant
or licensee uses for estimating exposures are reasonable given factors such as the layout of theplant and the qualifications and training of the workers. In general, for chemical exposures via
the inhalation pathway, it is reasonable to expect that both workers and individuals outside thecontrolled area may be subject to such exposures. For non-inhalation chemical exposures, it isreasonable to expect that only workers would be subject to such exposures.
The third step involves the assessment of the consequences from the exposure event. Thisevaluation requires an understanding of the estimated exposure (developed during the secondstep) and information on the toxic characteristics of the released material or its anticipatedreaction products. The same information on chemical toxicity characteristics that is used toestimate consequences is generally used to identify proposed standards.
Estimation of dermal and ocular exposure consequences for workers is generally morechallenging than estimating inhalation exposure consequences. Dermal and ocular exposure
often involve liquids or aerosols (gas-liquid mixtures), and the estimation of exposure
parameters, such as exposure location on the receptor (e.g., hand vs. chest), the percent of
impacted body surface area, and the duration of exposure may be difficult. Effects of dermal andocular exposure often correlate to the concentration of the material involved in the exposure(e.g., severe skin burns are associated with short exposure to nitric acid in concentrations greaterthan 20 percent). So, in many cases, it may be more practical to estimate whether an exposure is
likely and, if it is, correlate exposure effects to the concentration of the material involved in the
exposure.
The NRC examines the method(s) the applicant or licensee used to estimate exposure of theworker or the individual outside the controlled area. The NRC evaluates the reasonableness ofany model used for the analysis and the specific parameters used in the analysis.
Review of Chemical Accident Likelihood
The NRC reviews the methods the applicant or licensee used to estimate the likelihood of anacute chemical exposure event. The regulation in 10 CFR 70.61 also specifies the permissible
likelihood of occurrence of accident sequences of different consequences. High-consequence
accident sequences must be “highly unlikely” and intermediate-consequence accident sequencesmust be “unlikely.” Implicitly, accidents in the low-consequence category can have a likelihoodof occurrence less than “unlikely” or simply “not unlikely.”
The NRC review consists of examining the overall estimate of the scenario frequency to makesure there is reasonable assurance that the overall likelihood is not unreasonably low. Historicalinformation for similar facilities and processes is examined when it is available to support the
NRC’s review. Additional information on NRC’s review of likelihood can be found in NUREG
1520, “Standard Review Plan for License Applications for Fuel Cycle Facilities.”
Review of Quantitative Standards
An applicant or licensee must describe in their ISA Summary proposed quantitative standards toassess the consequences to an individual from an acute chemical exposure to licensed material orchemicals produced from licensed materials. The proposed quantitative standards serve to
identify the event consequence categories for the ISA’s chemical safety discussions. The NRC

verifies that the proposed quantitative standards used to assess consequences to an individualfrom acute chemical exposures are appropriate.
When evaluating the proposed quantitative standards, the NRC considers specific accident
sequences described in the ISA Summary. If the accident sequence is determined to be highly
unlikely, no proposed quantitative standards are required. Similarly, if the accident sequence isdetermined to result in consequences that are less than the intermediate events, no proposed
quantitative standards are required.
The proposed quantitative standards should be based on generally available information fromindependent sources (e.g., government agencies or organizations, well-recognized professionalorganizations) describing the chemical’s toxicity and hazardous properties. The applicant or
licensee should provide a basis for the use and applicability of any proposed quantitative
standard. It is not expected that they will need to conduct their own experimental testing or
toxicity tests to generate data supporting their proposed standards.
Standards may have many forms. For inhalation exposures, the standard may be based on air
concentration for a given exposure time. For non-inhalation exposures, the standard may bebased on body surface area (BSA) exposure for a given time. The NRC reviewer ensures that the
proposed standard is consistent with available toxicological information, and that the use of the
proposed quantitative standard provides a reasonable estimate of event consequence (i.e. does
not result in an underestimate of the event’s severity).
The following paragraphs provide specific examples of information sources that are acceptable
to the NRC when evaluating an applicant or licensee’s proposed quantitative standards forclassifying acute chemical exposure events as of high or intermediate consequence.
When reviewing exposure scenarios, the NRC has identified several useful information sources
to evaluate a applicant or licensee’s proposed quantitative standards. Acceptable exposure
standards include, but are not limited to, those based on the Emergency Response Planning
Guidelines (ERPGs), the Acute Exposure Guideline Levels (AEGLs), Temporary Emergency
Exposure Levels (TEELs), the Globally Harmonized System of Classification and Labelling ofChemicals (GHS), and the exposure limits established by OSHA or other Federal agencies andscientific organizations.
The two most common data sources for the NRC to use when reviewing proposed quantitative
standards are the AEGLs and ERPGs. The AEGLs 1 are intended to describe the risk to humans
resulting from once-in-a-lifetime, or rare, typically accidental exposure to airborne chemicals.
The American Industrial Hygiene Association (AIHA) establishes the ERPGs 2. While thesestandards were developed for other purposes, such as emergency guidelines for once in a lifetime
exposures, the NRC accepts the ERPG values to define “high” and “intermediate” consequencesin ISAs. These are inhalation exposure limits that the NRC has previously accepted as meetingthe quantitative standards requirement.
1 The history and nature of AEGLs is discussed on an Environmental Protection Agency Web site: http://www.epa.gov/oppt/aegl/index.htm
2 ERPGs are discussed on an American Industrial Hygiene Association Web site: https://www.aiha.org/get-involved/AIHAGuidelineFoundation/EmergencyResponsePlanningGuidelines/Documents/ERPGIntroText.pdf

Another common data source that the NRC uses when reviewing proposed quantitative standards
are the TEELs, which were developed by the U.S. Department of Energy (DOE). TEELs aretemporary and alternative guidelines used for chemicals that do not have established ERPGs and
AEGLs values 3.
The NRC may also use the information available in the GHS database when evaluating a
proposed standard. The GHS is an internationally standardized system for characterizing andlabeling chemical hazards to help protect consumers, transportation workers, and emergency
responders. The GHS defines different types of hazards (physical, health, and environmental)
and establishes methods for assigning standardized GHS hazard statements used to communicate
information about the severity of the hazard for specific exposure routes 4. OSHA’s Hazard
Communication Standard has been aligned with the GHS to improve the quality and consistencyof hazard information in the workplace. 5 Table 1 provides GHS hazard statements for common
chemicals in the fuel cycle process.
Table 2 presents the descriptions from the various information sources (i.e. AEGLs, ERPGs, andGHS) and compares it to the descriptions of “high” and “intermediate” consequence chemical
exposure events specified in NRC regulations. The hazard statements in the GHS database areconsidered useful when reviewing proposed standards, particularly when AEGLs, ERPGs, orTEELs are not available.
The NRC uses the National Institute for Occupational Safety and Health (NIOSH) Skin Notations as
another useful data source when reviewing proposed quantitative standards. The NIOSH Skin

Notations involve the assignment of multiple skin notations for distinguishing systemic (SYS),
direct (DIR), and sensitizing (SEN) effects caused by exposure of skin (SK) to chemicals. These
notations are useful for understanding the health effects and consequence of a specific chemical.
For example, in the past, NRC has used the skin notation profile for hydrofluoric acid to assess alicensee’s quantitative standard for HF.
In summary, the NRC verifies that an applicant or licensee’s proposed quantitative standards are
consistent with available and technically sound information or well-established data sources. If
the applicant or licensee proposes other sources of information as the basis for a proposed
standard, the NRC assesses the adequacy of the information the applicant or licensee isreferencing.

3

TEELs are discussed in a DOE Handbook DOE-HDBK-1046: http://orise.orau.gov/emi/scapa/files/doe-hdbk-1046-2008_ac.pdf

4 Globally Harmonized System of Classification and Labelling of Chemicals (GHS), fifth revised edition, 2013, Part 3 Health Hazards
5 Occupational Safety and Health Administration, Final Rule on Hazard Communication, Federal Register, Vol. 77, No. 58, March 26,

2012, pp. 17574-17896.

Table 1 – Acute Exposure Hazard Information for Common Fuel Cycle Process Chemicals 6
Chemical

NIOSH skin notation 8

ammonium hydroxide(NH4OH)GHS Hazard Statement in GHS database 7(Inhalation, dermal, ocular, ingestion exposure)H314 1B (causes severe skin burns and eye damage)H335 (may cause respiratory irritation): C≥ 5 %Noammonium fluoride(NH4F)

H301 (toxic if swallowed)
H311 (toxic in contact with skin)
H331 (toxic if inhaled)
H314 1B (causes severe skin burns and eye damage): C ≥25% for 1 hour exposure
H335 (may cause respiratory irritation): C ≥ 10 %H300 (fatal if swallowed)
H310 (fatal in contact with skin)H314 (causes severe skin burns and eye damage): C ≥ 7%
for 3 minute exposure; 1% ≤ C < 7% for 1 hour exposure
H330 (fatal if inhaled)
H302 (harmful if swallowed)
H314 (causes severe skin burns and eye damage): C ≥ 70%
for 3 minute exposure; 50% ≤ C < 70% for 1 hour exposure
H322 (harmful of inhaled)
H335 (may cause respiratory irritation) C ≥ 35%
H314 (causes severe skin burns and eye damage): C ≥ 20%
for 3 minute exposure; 5% ≤ C < 20% for 1 hour exposure
H315 (causes skin irritation)hydrochloric acid (HCl)hydrofluoric acid (HF)

hydrogen peroxide (H2O2)

nitric acid (HNO3)
perchloroethylene(C2CL4)
sodium hydroxide (NaOH)

Noted by OSHA list
for skin adsorption 9

AEGL; ERPG 10;TEEL(Inhalation exposure)

No

No

No

No

No

Yes

SK: SYS (FATAL)-DIR (COR): may
be potentially lethal or lifethreatening following exposure of
the skin; corrosive following exposureof the skin
No

Yes

Yes

No

Yes

No

No

Yes

No

No

Yes

No

Yes

H314 (causes severe skin burns and eye damage): C ≥SK: DIR (COR). corrosive followingNoYes
5% for 3 minute exposure; 2% ≤ C < 5% for 1 hour exp.
exposure of the skinsulfuric acid (H2SO4)
H314 (causes severe skin burns and eye damage): C ≥ 15%
No
No
Yes
for 3 minute exposure
Tributyl phosphate
H302 (harmful if swallowed)
No
No
Yes
H315 (causes skin irritation)((CH₃CH₂CH₂CH₂O)₃PO)
uranyl nitrate (UO2(NO3)2) H300 (fatal if swallowed)
No
No
Yes
H330 (fatal if inhaled)
Note: Exposure to chemicals with hazard or skin notation statements in bold would generally be considered a high consequence event in the context of an ISA. Exposure to chemicals with ahazard or skin notation statement that is underlined would generally be considered an intermediate consequence even in the context of an ISA. Skin Corr 1A is for exposure less than 3minutes. Skin Corr 1B is for exposure less than 1 hour.

6
7
8
9
10

The user should verify that current information is used because the sources identified in Table 1 are occasionally revised.
GHS information source: http://www.dguv.de/ifa/Gefahrstoffdatenbanken/GESTIS-Stoffdatenbank/index-2.jsp
NIOSH skin notation profiles: http://www.cdc.gov/niosh/topics/skin/skin-notation_profiles.html
Table A-1 of OSHA Technical Manual Section II, Chapter 2: https://www.osha.gov/dts/osta/otm/otm_ii/otm_ii_2.html
The AEGL and ERPG levels were established considering the more vulnerable receptors in the exposed public (elderly, children).

Table 2 – Chemical Exposure descriptions and statements related to the performance requirements in 70.61
High
Consequences

Description in 70.61

Description in AEGL 11

Description in ERPG 12

Could endanger the lifeof a worker

AEGL-3 is the airborne concentration of a
substance above which it is predicted that thegeneral population, including susceptible
individuals, could experience life-threatening
health effects or death.
AEGL-2 is the airborne concentration of a
substance above which it is predicted that thegeneral population, including susceptible
individuals, could experience irreversible orother serious, long-lasting adverse healtheffects or an impaired ability to escape.

ERPG-3 is the maximum airborne
concentration below which it is believed thatnearly all individuals could be exposed for up
to 1 hour without experiencing or developing
life-threatening health effects.
ERPG-2 is the maximum airborne
concentration below which it is believed thatnearly all individuals could be exposed for up
to 1 hr without experiencing or developing
irreversible or other serious health effects or
symptoms which could impair an individual'sability to take protective action.
ERPG-2 is the maximum airborne
concentration below which it is believed thatnearly all individuals could be exposed for up
to 1 hr without experiencing or developing
irreversible or other serious health effects or
symptoms which could impair an individual'sability to take protective action.
ERPG-1 is the maximum airborne
concentration below which it is believed thatnearly all individuals could be exposed for up
to 1 hr without experiencing other than mildtransient adverse health effects or perceivinga clearly defined, objectionable odor.

Could lead to irreversible
or other serious, longlasting health effects to
any individual locatedoutside the controlled
areaIntermediate
Consequences

11

12

Could lead to irreversible
or other serious, longlasting health effects to aworker

AEGL-2 is the airborne concentration of a
substance above which it is predicted that thegeneral population, including susceptible
individuals, could experience irreversible orother serious, long-lasting adverse healtheffects or an impaired ability to escape.

Could cause mildtransient health effects to
any individual locatedoutside the controlled
area

AEGL-1 is the airborne concentration of a
substance above which it is predicted that thegeneral population, including susceptible
individuals, could experience notablediscomfort, irritation, or certain asymptomatic
non-sensory effects. However, the effects are
not disabling and are transient and reversible
upon cessation of exposure.

Description in GHS Hazard
Statements
H330 Fatal if inhaled
H310 Fatal in contact with skin

H331 Toxic if inhaled

H331 Toxic if inhaled
H311 Toxic in contact with skinH314 Causes severe skinburns and eye damage
H318 Causes serious eye
damageH332 Harmful if inhaled

The Acute Exposure Level Guidelines have been developed primarily to provide guidance in situations where there can be a rare, typically accidental exposure to aparticular chemical that can involve the general public. They are based primarily on acute toxicology data and not subchronic or chronic data. They are designed to protect
the general population including the elderly and children, groups that are generally not considered in the development of workplace exposure levels.
The Emergency Response Planning Guideline (ERPG) values are intended to provide estimates of concentration ranges where one reasonably might anticipate observing
adverse effects as described. The ERPG values should not be expected to protect everyone but should be applicable to most individuals in the general public. Since thesevalues have been derived as planning and emergency response guidelines, not exposure guidelines, they do not contain the safety factors normally incorporated into
exposure guidelines. They are estimates, by the committee, of the thresholds above which there would be unacceptable likelihood of observing the defined effects. The
estimates are based on the available data that are summarized in the documentation. In some cases where the data are limited, the uncertainty of these estimates is large.
Users of the ERPG values are encouraged strongly to review carefully the documentation before applying these values.

III. CONCLUSION
The NRC regulations for SNM use risk informed performance standards. Assessment ofcompliance with the performance standards requires a thorough identification of credible
hazards, including chemical hazards, potential accident scenarios, and an evaluation of thesescenarios in terms of likelihood and consequences. New hazards may be introduced as the resultof the introduction of new tech and /or new processes. Therefore, it is important,… The NRC
recognizes that new technology and processes may introduce new hazards and it is important that
regulators and operators maintain the safety mindset and conduct thorough analysis and reviewof the new hazards. Additionally, it is important that guidance and reports be routinely updated
as new information comes out (e.g. toxicity information, new safety datasheets and safety
practices).
IV. REFERENCES
U.S. Code of Federal Regulations, “Domestic Licensing of Special Nuclear Material,” Part 70,
Chapter I, Title 10, “Energy.”
U.S. Code of Federal Regulations, “Occupational Safety and Health Administration:
Subpart Z—Toxic and Hazardous Substances, Tables Z-1 and Z-2,” Chapter XVII, Title 29,
Section 1910.100, “Labor.”
U.S. Nuclear Regulatory Commission, NUREG-1520, “Standard Review Plan for License
Applications for Fuel Cycle Facilities.” Revision 2, June 2015.
U.S. Nuclear Regulatory Commission, NUREG-1513, “Integrated Safety Analysis GuidanceDocument.” May 2001.
U.S. Nuclear Regulatory Commission, NUREG/CR-6410, “Nuclear Fuel Cycle Facility Accident
Analysis Handbook.” March 1998.
U.S. Nuclear Regulatory Commission, Information Notice 2007-22, “Recent Hydrogen Fluoride
Exposures at Fuel Cycle Facilities.” June 2007.
U.S. Nuclear Regulatory Commission/Occupational Safety and Health Administration,“Memorandum of Understanding between the Nuclear Regulatory Commission and the
Occupational Safety and Health Administration, ‘Worker Protection at NRC-Licensed
Facilities,’” September 6, 2013.
Trevor Kletz, IChemE/Butterworth-Heinemann, Oxford, England “What Went Wrong, Case
Histories of Process Plant Disasters and How They Could Have Been Avoided.” 2009.

R&D TO IMPROVE ASSESSMENT OF EXPLOSION RISKS ASSOCIATED WITH
RED OIL PHENOMENA IN ORANO LA HAGUE REPROCESSING PLANTSAngelica ORTEGA NICAISE1, François VIROT2, Lucie DE ALMEIDA1, Christian MUN2,
Marc PHILIPPE1, Patrick POCHON3, Marie-Christine CHARBONNEL3, Catherine EYSSERIC3,
Julia BISSON4, Frédéric LEPRETRE4, Stéphanie DE SIO5, Gaëtan BECKER6, Nicolas VIGIER5
1
Institut de Radioprotection et de Sûreté Nucléaire (IRSN), BP 17,92262 Fontenay-aux-Roses Cedex, France
2
Institut de Radioprotection et de Sûreté Nucléaire (IRSN), 13115 Saint-Paul-Lez-Durance, France
3
CEA/DEN/DMRC Marcoule, BP 17171, F-30207 Bagnols-sur-Cèze Cedex, France
4
ORANO Cycle, Établissement de la Hague, Beaumont-Hague, 50444 La Hague Cedex, France
5
ORANO Cycle, 1 Place Jean Millier, 92 084 Paris La Défense, France
6
ORANO Projets, 1 rue des Hérons, 78182 Montigny Le Bretonneux, France
marc.philippe@irsn.fr
The PUREX process is the main industrial process applied to recover uranium andplutonium from spent nuclear fuel. Linked to such processes, the “red oil” phenomena comefrom the possible reaction between an organic phase (the extractant TBP and its degradation
products) and an acidic phase (concentrated nitric acid potentially containing extractableheavy metal nitrates). These complex reactions can lead to a thermal runaway in the plant’s
equipment followed by an explosion. They caused several significant accidents in the
nuclear industry, especially in the USA and Russia.
In design phase of La Hague plants (1980-1985), the explosion risks due to red oilcompounds formation were considered: the risk control measures adopted then were basedon lessons learned from previous accidents and scientific knowledge of the time. However,Tomsk accident (1993) feedback and the substantial research works carried out to better
understand the phenomena have confirmed the need to reassess the criteria and safetyparameters chosen initially to control these explosion risks.
In 2013, an important experimental research programme started as part of an ORANO, CEA
and IRSN collaboration. It aims to a better understanding of the phenomena and the main
influential parameters, as well as acquiring thermokinetic and physicochemical data to
support the modelling studies that are being developed simultaneously by ORANO and
IRSN. The results already obtained have allowed IRSN to formulate recommendations
regarding the enhancement of red oil risks control, as part of the safety review of the plants,
and ORANO to suggest concrete actions to limit risks of overpressure and to improve
explosion risks prevention in future equipment.
I. INTRODUCTIONIn France, the reprocessing of spent nuclear fuel using the PUREX (“Plutonium Uranium Refining by
Extraction”) chemical process is carried out within two plants, UP3-A and UP2-800, located on the
ORANO (ex AREVA) La Hague site. In these plants, after the spent fuel assemblies are dissolved innitric acid, TBP (tributyl phosphate) diluted in HTP (hydrogenated tetrapropylene – TPH in French) isused to extract uranium and plutonium from fission products and minor actinides.
Under the process operating conditions in La Hague plants, reactions between TBP and its
degradation products on the one hand, and nitric acid, its derivatives and extractable heavy metal
nitrates (uranyl nitrate or plutonium nitrate) on the other hand, can lead to the formation of unstable,1nitrous organic compounds known as “red oil”. These can violently decompose and generatepotentially explosive gases.
The formation of these compounds has already led to several accidents, notably in the United States in1953 and 1975 and, more recently, in Russia in 1993, generally resulting in the destruction of theequipment and rooms involved and in significant radioactive releases to the environment [1]. The
feedback and lessons learned from these accidents were the subject of an earlier report in which IRSN
presented its state of knowledge at the time [2].
II. REX before 1990
In 1953, the first known red oil incident took place during the evaporation of a solution of uranyl
nitrate hexahydrate in nitric acid at the Savannah River plant. The presence of TBP in the evaporator
coupled with the absence of sufficient control methods led to an over-evaporation of the solution
reaching its thermal decomposition point and causing a violent thermal runaway reaction between
TBP and nitrates, followed by an explosion. The equipment was totally destroyed and the building
was severely damaged.
The same year a similar event happened at Hanford site, caused by an undetected failure in the pumpfeeding the uranyl nitrate to the evaporator, by the presence of TBP in the equipment and by theabsence of safety control. The consequences were less important thanks to the presence of a rupture
disk in the evaporator.
An incident involving plutonium took place at Oak Ridge laboratory in 1959 at a stripping tower used
for preparation of plutonium nitrate concentrate. The presence of TBP in the equipment led to an
explosion, the destruction of the evaporator and the building and the radioactive contamination of thesurroundings.
In 1975, a new explosion occurred at the Savannah River plant. Over 100 L of TBP were
unexpectedly carried over with the concentrated uranyl nitrate solution even though it had alreadypassed two evaporators. While the solution was undergoing denitration, in a boiler-type batch
evaporator used for conversion into uranium trioxide, and although the established procedures were
undertaken, temperature rose from 200°C to 450°C, flammable gases were massively produced andfinally, despite the start of the safety ventilation, an explosion occurred. The accident’s cause was thephase inversion of concentrated TBP due to reduced uranium concentration during rinsing operationthat led ultimately to the inadvertent transfer of bulk solvent to a uranium denitrator and to theaccident [[3]]. Prior to this explosion, the possibility of a phase inversion was not properly recognized
and all tanks were only periodically skimmed to remove floating organic solution (TBP-HTP). No
agitation was planned. Over 200 L of TBP were recovered from various hold tanks and process
vessels following the accident investigation. Such a quantity was not expected to accumulate within
the system, nor was it expected that TBP could be pumped from one process tank to another. Indeed,
TBP organic phase was assumed to float over the aqueous phase and should have remained in the firsttank, since the solution is pumped from the bottom of the tank but not entirely in order to avoid
carrying over the floating organic phase. Following the accident, agitation was implemented to ensurethat organic compounds do not build-up in process vessels [1] [3] [4] [5].
These incidents showed that the amount of TBP, the solution temperature, the nitrates’ concentrationin the solution and the adequate equipment venting play a significant role in the red oil runaway
reaction.

2III. ORANO La Hague UP2-800 & UP3-A plants’ design
During the design of the UP2-800 and the UP3-A plants, in the 1980s, the risks of explosion
associated with red oil compounds were taken into account in the safety analysis of the evaporators
used to concentrate acidic aqueous streams, coming from different extraction or purification cycles aswell as from effluent management units, containing nitrates (either nitric acid or uranyl and plutonium
nitrate) and residual amounts of organic solvent. The units identified as concerned by this risk werethe uranyl nitrate concentration units, the intercycle concentration units, the high-level effluents
concentration units, the fission products concentration units, the acid recovery units and the oxalic
liquor treatment units. Several measures were implemented based on the state of knowledge andavailable operating experience feedback available at that time. These included maintaining theevaporator temperature below 135°C (to avoid any thermal runaway reaction below this temperaturethreshold), and very low concentrations of solvent (TBP) in the solutions fed to the evaporators [3].
These two functional requirements were adopted in the safety demonstrations and incorporated in thegeneral operating rules of the facilities concerned.
The organic effluents treatment units are also concerned by red oil risks. In these units, the feeding
stream is mainly composed of organic solvent and therefore the risk prevention measures consist in
minimizing the amount of nitrates that it contains.
To guarantee a temperature below 135°C in the evaporators under normal operating conditions, theheating loop is equipped with temperature and pressure controllers as well as a high temperature andhigh pressure alarms which, once breached, will shut off the heat fluid supply. Moreover, the circuitalso has two relief valves that limit its pressure rise, and therefore, the heating fluid’s temperature.
The organic effluents treatment units are equipped with liquid falling film evaporators, with pressure
and temperature controllers located on the heating system as well as high pressure and temperature
warnings. At the upstream of the evaporator, the solvent temperature is also controlled.
Regarding the TBP amount limitation in the evaporators feeding solutions, a permanent diluent wash
is implemented for the aqueous feeds arriving from the extraction cycles: this wash is carried out in abattery of mixer-settlers or in a pulsed column. A solvent flushing operation also takes place,
periodically, in the tanks located upstream the evaporators, allowing to separate TBP from the feeding
stream by decantation. These measures allow the aqueous stream to limit its TBP content to a few
dozen milligrams per litre.
As not all streams feeding the evaporators can be washed, a systematic control of the TBP
concentration is realized before transfer to an evaporator. Moreover, the evaporator’s feeding tank is
not completely emptied to avoid the transfer of any eventually floating TBP.
For the specific case of the organic effluents treatment units, before arriving to the units, the feeding
stream undergoes a carbonate and soda treatments as well as water wash to reduce the nitrateconcentration.
IV. Tomsk accident - REX
In 1993 at Tomsk plant, an accident happened in a full-scale tank equipped with a cooling/heating
jacket and a bubbler-mixer at the bottom. Upon discharge from an evaporator in TOMSK-7 PUREX
first extraction cycle, an aqueous solution carried over a “substantial volume” of TBP and its
degradation products into a 35 m3 pressurized process tank containing 25 m3 of uranyl nitrate andplutonium at around 40°C. The volume carried over is reported to have sat for at least several months.
In addition, the uranyl nitrate solution contained various other nitrates and residual fission products.
After this long period, 1.5 m3 of concentrated nitric acid (14 N) was added to the tank, which had no
mixing and was only cooled from the bottom. This addition led to the formation of three phases: aheavy phase of aqueous uranyl nitrate, a second one of used solvent and a third floating one of3

concentrated nitric acid. The uranyl nitrate phase heated to around 80°C leading to the progressive
heating of the second phase until its self-heating reaction with nitric acid caused the mixture to boil
for over two hours, producing a gaseous mixture at an accelerated rate (possibly butene, butane,butanol, carbon monoxide, and other explosive vapours, along with steam). All the produced gases
generated a pressure increase, causing two explosions in the evaporator buffer tank and releasingaround 20 TBq of alpha and beta emitters to the environment.
The feedback acquired from Tomsk accident showed an incapacity to manage the heat duty and the
produced gases at temperatures higher than 70/80°C during the initiation TBP degradation step. It has
led to an increase of pressure and temperature reaching the threshold temperature of red oil runaway,
where red oils reaction occurred. Such phenomena could be observed in unheated equipment, such as
buffer tanks, relay or storage tanks and transfer pots. The explosion was the result of the combinationof several factors: sufficient amount of organics, lack of agitation, reduction of the blowdown section,
increase of nitric acid concentration, insufficient temperature and pressure monitoring and presence ofeasily oxidizing degradation products in the organic phase. Separately, these factors would not havebeen able to lead to an explosion.
Since that accident, considerable international R&D work has been carried out to better identify
reaction mechanisms and key parameters that may lead to red oil runaway reactions and to learn more
about the factors or conditions that could help control this risk.
The R&D and feedback led to review the control measures to prevent red oil explosions in nuclear
facilities. In 2003, an assessment of the potential for red oil explosion in the Department of Energy’s(DOE) defence nuclear facilities stated that four controls could be used to prevent a red oil event: atemperature below 130°C, a sufficient vent for the process, an optimal organic removal from theprocess and a nitric acid concentration below 10 M [5]. Nevertheless, more research was needed since
Tomsk accident started below 130°C.
V. R&D
After an extensive literature review carried out by ORANO and IRSN in 2012, an experimental
programme, led jointly by ORANO, IRSN and CEA, was launched in 2013 in order to acquire newscientific knowledge on red oil thermal runaway phenomena and the parameters that influence andgovern them (nitric acid concentration, presence of uranium and of organic impurities, phase ratio,pressure, etc.) with the objective of improving safety measures implemented to control such risks.
This R&D experimental programme is still underway at the CEA Marcoule laboratory.
More precisely, the aim of the programme was to improve thermochemical knowledge of biphasic
systems (TBP/HTP – HNO3) at temperatures higher than the boiling point under the operating
conditions prevailing in the evaporation-concentration units of the UP3-A and UP2-800 plants at La
Hague. In particular, this involves reviewing the conditions under which thermal runaway is initiated(Tonset values), the maximum permissible variation ranges of control parameters (temperature, acidity,
pressure, etc.), the possible impact on equipment and facilities, and prevention and mitigation
measures aimed at controlling these risks when the units are in operation.
This experimental programme was performed using closed cells which provide penalising results, andwas divided in two different tasks. Firstly, an extensive campaign was performed in order to identify
the most influential parameters for the TBP-HNO3 thermal runaway reaction. The investigated species
were: nitric acid (present in every solution), nitrous acid, butanol, butyl nitrate, DBP (di-butyl
phosphate), uranium, HTP, nitrates and fission products (as metallic ions). This campaign took place
in a non-adiabatic calorimeter TSU (thermal screening unit – Figure 1) [6], equipment designed forfast hazard screenings and therefore perfectly suitable for heterogeneous liquid systems like the twophase TBP and nitric acid one.

4

Figure 1: Thermal screening unit (TSU) calorimeter
In addition, the steel cell selected can withstand the very high pressure variation associated with thehighly exothermic reactions. Nevertheless, its volume was small (8 cm3) leading to a phi factor1 larger
than 1, typically around 2. Given this fact, these first experimental results shall be considered as
qualitative data from the thermochemical point of view.
Experiments were conducted with an aqueous-organic phase (A/O) volume ratio of 4 using a 2.5 mLsolution where the organic phase was always saturated with nitric acid beforehand. The heating ramp
was settled at 1°C.min-1.
From these experiments, the following conclusions can be drawn. The nitric acid concentration is themain factor affecting the thermal runaway: as the acidity rises up to 13.8 M, the onset temperature(the threshold of the self-sustained thermal runaway) decreases and the rate at which temperature andpressure rise increases. At low nitric acid concentration (1 M), the presence of uranyl nitrate leads thesolution towards a runaway behaviour, certainly due to the increase of nitrate concentration in theorganic phase. However, its effect at higher acidity seems to be negligible. Like the organic solvent,the diluent HTP similarly reacts with nitric acid leading to thermal runaway but this substance may beless stable due to a decrease of the onset temperature compared to a system with undiluted TBP.
Secondly, the identified parameters were tested with a targeted set of experiments in order toaccurately investigate temperature and pressure behaviour and to determine thermo-kinetic data likeheat release and pseudo rate constant. For this task, a pseudo-adiabatic calorimeter Phi-Tec II model,
as shown in Figure 2, was used [7].

Figure 2: Phi-Tec II calorimeter and its external reactor (left), reaction cell (right)

1

Phi-factor, or adiabatic factor: ratio of the thermal mass of sample and container over the thermal mass ofsample; the closer it is to 1, the more adiabatic the experiment.

5This calorimeter is often considered as a reliable instrument to generate quantitative kinetic andthermodynamic data allowing the scale-up to the industrial level. The data are extracted from recorded
thermograms by applying the methodology described by McIntosh & Waldram [8]. In addition, thelarge volume of the cell, about 100 cm3, allows for more representative studies. To avoid some
repeatability issues regarding the reactivity of the diphasic solution, the linear ramp rate mode waschosen instead of the heat-wait-search one.
1.

Development of the experimental protocol

Several input parameters have been optimised in order to define the most reliable protocol toinvestigate the phenomena related to the red oil risks. The Phi-Tec II calorimeter is operated with aclosed cell, therefore the quantity of solution (or filling ratio) as well as the phase ratio affect thedevelopment of the thermal reaction. These parameters are inter-dependent, the variation of one ofthem modifying the optimal value of the other one. Therefore, the filling ratio value was constant(30% and 55%, representative ratio in the evaporators of the PUREX process). The aqueous toorganic (A/O) phase volume ratio modifies the heat release, which is also a function of nitric acidconcentration. Thus, for each acidity, the A/O ratio selected reflected the maximum of heat release.
An example of such evolution of the heat release (𝛥𝐻), expressed by mole of organic solvent, isshown in Figure 3. This phenomenon, already mentioned by different authors [9] [10], indicates thatthe chemical runaway may not only be driven by the degradation of TBP or that a bias may comefrom the evaluation of the heat capacity of the solution. The intensity of the exothermic process in
two-phase systems remains an open issue. In these experiments, the nitric acid concentration varied
from 2.5 M to 13.8 M and the corresponding optimal A/O ratios always evolved between 20 and 30.
To obtain accurate thermochemical data, the value of phi-factor was carefully checked for each run.
That implies the mandatory use of Hastelloy cells, instead of glass ones. For the experiments
performed at a 30% filling ratio (the most unfavourable value), the adiabatic factor was close to 1.1.

Figure 3: Example of classical effect of the aqueous-organic phase volume ratio on the heat ofreaction - conditions: organic phase TBP-aqueous phase HNO3 8.5 M, filling ratio 30 %2.

Overview of main results

The main accomplishments of this study are its completeness and its realistic conditions, excepted theuse of penalising closed cells, as the evaporator’s working conditions in terms of acidity as well as
filling ratio were taken into account.
The trend about the onset temperature previously highlighted during the first task has been confirmed
and evaluated. The nitric acid concentration has an important effect, indeed onset temperature6

gradually decreases when the acidity rise occurs to tend towards 105°C for the highest investigatedvalue (13.8 M). It is to be noted that experimental onset temperatures cannot be directly extrapolated
to industrial equipment (open system, heat transfer…). The runaway rate as well as the maximal
temperature increase when the amount of organic solvent increases or when the aqueous volume
decreases. Nevertheless, for these criteria, the acidity has a fairly low effect especially in [5-13.8 M]
range. A similar trend was noticed regarding the pressure data. The uranyl nitrate modifies the thermal
runaway only at low acidity (1 M of nitric acid). For this set of experiments, the increases of maximal
temperature and of pressure and temperature rise rates were dependent on the uranyl nitrateconcentration. That is certainly due to a highest extracted amount of nitrate compounds in the organic
solvent. The experiments showed no influence of uranyl nitrate on onset temperature which is inagreement with the ascertainment of Chandran et al. [11], whereas former study made by Glossop [9]predicted the opposite behaviour.
Regarding the thermochemical results, all experiments presented a runaway behaviour which follows
a pseudo first order kinetic rate law. It agrees with most of the published data (see reviews [12][13]and references therein) and corroborates the fact that the degradation process is driven by thehydrolysis of TBP as the principal reaction. The obtained activation energies always decrease if theacidity increases, confirming the catalytic effect of the nitric acid on the chemical processes. For the
8.5 M acidity, the following Arrhenius parameters are obtained at optimal A/O ratio: the activationenergy (Ea) and the pre-factor logarithm (ln(A), A in s-1) equal 108 kJ.mol-1 and 23.4, respectively.
These parameters of the overall decomposition rate are in good agreement with previous ones
obtained with single organic phase saturated with nitric acid [11] [14]. Indeed, Chandran et al. [11]reported (for both acidity 8 M and 15.6 M) the following datasets: Ea between 108.8-103.5 kJ.mol-1
and ln(A) varying from 24.8 to 22.5, while Nichols’ kinetic parameters were 112 kJ.mol-1 and 24.5.
These data seem to show similar rates in one-phase organic and in two-phase TBP-nitric acid systems;
nevertheless, as the ascertainment is not substantiated, further investigations in one-phase systemusing the same aqueous acidity range as studied (from 2.5 M to 13.8 M) will be carried out. However,it confirms that the exothermic processes could occur at the interface or inside the organic phase.
An increase in nitric acid aqueous concentration (from 2.5 M to 13.8 M) does not have a significant
effect on heat release: from 2000 kJ.mol-1 to 2300 kJ.mol-1. The comparison of these experimental
data with theoretical calculations, based on heats of formation coupled to postulated degradation
paths, is difficult due to the lack of accurate chemical speciation of liquid end-products (to estimate
the oxidation extend of TBP by-products) as well as the existence of parallel processes which
participate to the overall heat release. In addition, corresponding estimation does not take into accountthe enthalpies of solvation of each species and the temperature effect, inducing a bias on the enthalpyof reaction.
3.

Ongoing R&D

In order to assess the quantity of solvent that is likely to accumulate in the evaporators and to simulate
their dynamic behaviour during normal operation and accident condition, a simulation tool is being
developed at IRSN and will be validated by the experimental data obtained as part of the ongoing
research programme. This tool is built from some ASTEC [15] modules dealing with the thermal andthe chemical phenomena.
Once the tool has been validated, computation and resolution of the kinetic equations system for the
known reaction scheme of TBP (see reference [12] for the detailed scheme therein) will take place.
Thermal and chemical phenomena are taken into consideration by two different modules: a thermal
and a kinetic one; the values of heat and gas generation determined by the chemistry module are used
to feed the thermal one.

7

To feed the tool, a theoretical study based on quantum chemistry has been carried out in order toimprove the knowledge of some physical properties for the key species [16]. For example, thethermodynamic properties are mandatory to calculate the heat generated by chemical reactions.
Similar approach is ongoing regarding the uranyl nitrate-TBP complexes. Additional data will beobtained experimentally in order to establish a more precise reaction mechanism of red oil reactions(kinetics, speciation, etc.).
In the near future, the current simulation efforts will contribute to the expertise analysis regarding theenhancement of red oil risks control.
VI. ORANO’s new control parameters
Even though the programme was not completely finished, the results allowed ORANO to strengthen
safety demonstrations for the UP3-A plant as part of its safety review. As a matter of fact, ORANO
has modified its red oil risk analysis in line with the safety principles adopted at the design stage [3].
The new phenomenological approach taken by ORANO indicates that the simultaneous conditions are
required for a red oil runaway reaction to occur:
 an organic phase containing TBP and/or its degradation products;
 a concentrated nitrate ion phase (nitric acid, nitrates coming from salts);
 a temperature higher than the runaway initiation temperature; a ventilation system that lacks the capacity to remove the gases generated.
This new approach confirms that control of the red oil risk must be demonstrated not on the basis oftemperature alone, but simultaneous control of the quantity of solvent in the equipment, thetemperature and acidity of the medium, and adequate removal of process gases. Thus, assuming thatthe evaporators contain a significant quantity of solvent (whether due to gradual accumulation orsudden inlet), the risk of thermal runaway in units concerned by the red oil risks cannot be ruled out
simply by limiting the evaporator temperature.
ORANO identified the heated units and equipment concerned by these risks in the UP3-A plant. In
addition, unheated equipment such as storage, receiving, and feeder tanks or buffer tanks were studied
as, based on feedback from Tomsk accident, they could also be concerned by red oil risks. The
analysis of the associated risks in tanks is currently being performed by ORANO.
For future fission products concentration units, ORANO has suggested concrete actions to improve
explosion risks prevention and to limit overpressure in the identified equipment in case of a thermal
runaway, particularly regarding generated gases extraction.
VII. CONCLUSION
As part of the La Hague plants periodic safety review, studies were carried out in order to better
understand TBP organic phase behaviour when in contact with acidic aqueous solutions at different
temperatures. Datasets of interest were determined by means of a TSU calorimeter to confirm the
major influential parameters. Among these datasets, the thermokinetic parameters were measured
using a pseudo-adiabatic calorimeter (in closed system) under chemical conditions similar to those of
the equipment identified as “at risk” in terms of red oil reactions at La Hague plants.
Consequently, qualitative and quantitative research made by an experimental programme has allowed
IRSN and ORANO to determine specific values of thermokinetic data for the TBP-HNO3 reaction as
well as the influence of other species, like diluent or uranyl nitrate. IRSN will use the results of thisresearch in a computational tool to simulate specific scenarios based on industrial conditions, in orderto support its ongoing assessment of the periodic safety review file for the UP2-800 spent fuelreprocessing plant at ORANO La Hague site.
References:

8

[1]
[2]

[3]

[4]
[5]
[6]
[7]
[8]
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9Canadian Nuclear Fuel Cycle Facilities: The Regulation of Chemical Hazards in Canada
Sarah EATON, Julian AMALRAJ, Kavita MURTHY
Canadian Nuclear Safety Commission
280 Slater Street, Ottawa, ON, Canada K1P-5S9
sarah.eaton@canada.ca
Radiation hazards at nuclear facilities are typically the focus of attention of nuclear regulators and the
public alike, but at nuclear fuel cycle facilities (FCFs), chemical hazards may pose a greater risk to theworkers and the public if not managed properly. In Canada, the Nuclear Safety and Control Act (NSCA),
authorizes the Canadian Nuclear Safety Commission (CNSC) to regulate all of the activities in the nuclear
fuel cycle, from mining and milling to spent fuel management. Operating commercial FCFs in Canada
include: uranium mines and mills, front-end uranium processing facilities, interim spent fuel storage
facilities and several isotope processing and research and development facilities. With a wide range offront-end fuel cycle facilities comes a wide range of chemical hazards. FCF operators are required to takeall reasonable precautions to ensure control of hazards from radioactive and hazardous substances that areassociated with licensed activities.
The CNSC uses a regulatory framework with 14 safety and control areas that ensure requirements andguidance on various aspects of safety associated with the operation of FCFs. The CNSC regulates several
aspects of use of chemicals in FCFs which include facility management, personnel training, process
design, assessment and control of hazards, occupational exposure, personal protective equipment,
environmental releases, environmental monitoring, transportation, emergency management anddissemination of information including operational experiences from events and incidents associated withchemical hazards and safety.
This paper provides an overview of two events involving chemical hazards (hydrofluoric acid andberyllium) at Canadian FCFs with the goal of sharing operational experiences with the global FCF
community. Sharing operation experiences are important in the nuclear industry to share lessons learned
and helping reduce the possibility of similar events occurring elsewhere. This paper will also provide an
overview of the CNSC’s approach to conducting compliance.
I.

INTRODUCTION

The Canadian Nuclear Safety Commission (CNSC) is the regulatory body created under the Nuclear
Safety and Control Act (NSCA) by the Parliament of Canada, to regulate the use of nuclear energy
and materials to protect the health, safety and security of persons and the environment; to implement
Canada's international commitments on the peaceful use of nuclear energy; and to disseminate
objective scientific, technical and regulatory information to the public.
The NSCA established the CNSC as the regulatory body for nuclear activities in Canada and among
other things, gave the CNSC the power to make regulations and the authority to grant licences. There
are 13 regulations made under the NSCA, several that apply to all activities regulated by the CNSC
such as the Radiation Protection Regulations and some specific sector-focused regulations such as the
Uranium Mines and Mills Regulations. The Canadian nuclear regulatory framework consists of the
NSCA, associated Regulations, licences, licence conditions handbooks and regulatory documents(Figure 1).

Figure 1: The CNSC's regulatory framework is made up of the Nuclear Safety and Control Act, the Regulations made under the Act,
licences accorded by the Commission and regulatory documents.

Under the NSCA, one of the CNSC’s key responsibilities is protection of the environment. To meetthis responsibility, the CNSC requires the potential environmental effects of all nuclear facilities or
activities to be considered and evaluated when licensing decisions are made. The CNSC conducts
environmental assessments (EAs) under the Canadian Environmental Assessment Act (CEAA) or
under the NSCA. An EA under CEAA is a planning and decision-making tool, its objectives are tominimize or avoid adverse environmental effects before they occur, and incorporate environmental
factors into decision making for designated projects. An EA under CEAA is carried early in thelicensing process (before any licence is granted) and considers the entire proposed lifecycle of a
project. It includes information prepared by the applicant and CNSC staff, as well as comments
received from Aboriginal groups and the public. An EA under the NSCA is conducted for projects notlisted in the Regulations Designating Physical Activities or for projects previously assessed under
CEAA. The nuclear industry is also subject to the provisions of the Canada Labour Code and theassociated Regulations which in part deal with regulating occupational exposure to hazardous
substances and management of workplace safety.
The decision making body with respect to major nuclear facilities in Canada is the independent
quasi-judicial administrative tribunal of the CNSC (the Commission), made up of up to sevenpermanent members. Licensing decisions related to major facilities are made by the Commission
transparently in public hearings. The Commission provides public and other interested parties
opportunities to participate when major decisions are under contemplation. Roughly 800 scientific,
technical and professional staff support the Commission. Staff review applications and makerecommendations to the Commission and are also responsible for regulatory compliance verificationand enforcement.
II.

CANADIAN NUCLEAR FUEL CYCLE FACILITIESAt the present time, Canada has four operating nuclear power plants (NPPs) and one plant in safe
shutdown. All of Canada’s power reactors are CANDU (Canadian Deuterium-Uranium) reactors.
These pressurized heavy water reactors use natural uranium as fuel and heavy water as a coolant andmoderator. Operating commercial FCFs in Canada include: Uranium mines and mills, front-end
uranium processing facilities, interim spent fuel storage facilities and one FCF research &development facilities (Chalk River Laboratories). Spent fuel is also stored at the respective reactor
sites. There are no enrichment facilities or re-processing facilities in Canada.
Radiation hazards are often the focus of nuclear regulators and the public at nuclear FCFs, while chemicalhazards if not controlled effectively, may have a higher risk to the workers and the public. With a wide
range of facilities comes a wide range of chemical hazards and FCF operators are required to take allreasonable precautions to ensure control of hazards from radioactive and hazardous substances that areassociated with licensed activity.

This paper provides an overview of two events involving chemical hazards (hydrofluoric acid andberyllium) at Canadian FCF’s with the goal of sharing operational experiences with the global FCF
community. Sharing operation experiences are important in the nuclear industry to share lessons learned,
helping reduce the possibility of similar events occurring elsewhere. This paper will also provide an
overview of the CNSC’s approach to conducting compliance. A brief description of Canadian uranium
processing facilities is provided below and in Figure 2.

Figure 2: Location of uranium processing facilities in Ontario, Canada

The Blind River Refinery (BRR) facility is located near Blind River, Ontario. The facility refines uranium
concentrates (yellowcake) received from uranium mines worldwide to produce uranium trioxide, anintermediate product of the nuclear fuel cycle.
The Port Hope Conversion Facility (PHCF) is located in Port Hope, Ontario. The facility primarily
converts uranium trioxide (UO3) powder to uranium dioxide (UO2) and uranium hexafluoride (UF6). UO2
is used in the manufacture of CANDU reactor fuel, whereas UF6 is exported for further processing into
fuel for light water reactors.
Cameco Fuel Manufacturing Facility (CFM) is also situated in Port Hope, Ontario. It receives natural anddepleted uranium dioxide powder (UO2) from the PHCF and fabricates ceramic fuel pellets and thereafter
manufactures finished fuel bundles for CANDU reactors for use in Canadian nuclear power reactors andresearch reactors.
BWXT Nuclear Energy Canada (BWXT), whose operations are similar to CFM, operates two uraniumprocessing facilities in Ontario. The Toronto facility processes natural UO2 powder from the PHCF into
ceramic pellets. The majority of these pellets are shipped to BWXT’s Peterborough facility where they
are assembled into CANDU reactor fuel bundles.
III.

CNSC COMPLIANCE PROGRAM

The CNSC compliance monitoring program consists of verification, enforcement and reporting. These
processes are in place to ensure that CNSC licensees operate their facilities safely and exhibit a highlevel of compliance with the regulatory framework. These activities enable the CNSC to provide
assurance to Canadians of the continuing compliance and safety performance of licensees.

Compliance verification activities consist of on-site inspections of the various safety and control areas
and assessments of periodic reports submitted by licensees of their own monitoring of licensed activities.
Through inspections of nuclear facilities, activities and processes, as well as through the review oflicensee reports, CNSC staff continuously monitor the nuclear industry to ensure that licensees areoperating safely, securely, and in compliance with the requirements set out in the NSCA, its associatedRegulations, and licences. In cases of non-compliance, CNSC staff will use various enforcement actions,in a graded approach, to ensure that licensees take effective and timely corrective actions. Enforcement
actions can range from issuing a written notice for corrective actions for minor infractions to issuingorders, issuance of Administrative Monetary Penalties (AMP), recommending licensing actions(including suspension and revocation) withdrawal of certification or prosecution for more serious
violations. In certain situations, licensees may be called to appear before the Commission.
The CNSC has implemented its Independent Environmental Monitoring Program (IEMP) to verify
that the public and the environment around licensed nuclear facilities are safe. It is separate from but
complementary to, the CNSC's ongoing compliance verification program. The IEMP involves taking
samples from public areas around the facilities, measuring and analyzing the amount of radiological(nuclear) and hazardous substances in those samples using the CNSC's laboratory. The IEMP process
consists of developing site-specific sampling plans for each nuclear facility, and then processing andanalyzing the samples collected. The sampling plans focus on measuring concentrations ofcontaminants in the environment at publicly accessible locations such as parks, residential
communities and beaches, and in areas of interest identified in environmental risk assessments(ERAs). Samples may be taken for air, water, soil, sediment, vegetation (e.g., grass and weeds) andsome food (e.g., meat and produce).
Samples are analyzed at the CNSC's state-of-the art laboratory by highly qualified scientists using bestindustry practices. Samples are measured for both radiological and non-radiological contaminants related
to the activities of the nuclear facility and as identified in the site-specific ERA. Contaminant levels arecompared to those in applicable guidelines and/or natural background levels to confirm there is no impact
on health or the environment. Conclusions and data are published on the CNSC website, illustrated in auser-friendly map [Ref 1].
CNSC staff regularly communicates the status of the nuclear industry to Canadians through reports
published on our website and through social media [Ref 1]. In addition, the CNSC requires licensees todevelop and implement public information and disclosure programs for all phases of the FCF’s lifecycle.
The licensee is required to proactively post information on general site operations, events, environmental
sampling results and other pertinent information on its website [Ref 2].
IV.

SAFETY AND CONTROL FRAMEWORKTo ensure that licensees in Canada meet all of their regulatory requirements and expectations, CNSC staff
assesses, evaluates, reviews, and verifies how well licensees are complying with these requirements.
CNSC staff base their evaluation on safety and control areas (SCAs). Each SCA includes technical areas
and topics, which are selected based on the risks of the specific licensed activity. Consistently using thesame terms when referring to the same SCA, facilitates communication internally as well as externally,
with licensees, the public and the Commission. In total, there are 14 SCAs that can be broadly sorted into
three functional areas: management, facility and equipment, and core controls and processes.
Management-specific SCAs address the organizational and human elements of safety in the Canadian
nuclear facilities. Specifically, they cover management systems, human performance management, andoperating performance. CNSC staff ensure that licensee staff are adequately trained, knowledgeable, andequipped to handle all safety duties. Performance, safety culture, organizational contingency plans, andmany other specific measures are all covered under this functional area.

Facility and equipment SCAs include safety analysis, physical design, and fitness for service. These areas
assess the potential hazards and risks of operating (as well as the preventative measures taken to minimizerisk), the integrity of facility infrastructure design, and the overall long-term performance of equipment
and systems. When CNSC staff analyze safety on the basis of facility and equipment, they are looking toverify a wide breadth of measures, ranging from the facility’s safety as a structure down to themaintenance of components.
Core controls and processes form the largest functional area. These SCAs include radiation protection,
conventional health and safety, environmental protection, emergency management and fire protection,
waste management, security, safeguards and non-proliferation, and packaging and transport. What these
SCAs have in common is that they all cover how a facility operates, they measure actions and plans thatare in place, all against the unique and specific nature of each facility. The different core controls andprocesses, as well as the associated regulatory requirements are all outlined in facility licensing
agreements.
SCAs are a template for confirming regulatory compliance and facility safety. All 14 SCAs are evaluatedat the Canadian FCFs using a long term compliance program. The licence and licence conditions
handbooks for all Canadian FCFs identify requirements for all SCAs. CNSC staff use these requirements
as compliance verification criteria during inspections and reviews.
V.

2017 HF EVENT AT CAMECO’S PORT HOPE CONVERSION FACILITY

Cameco’s PHCF is a uranium conversion facility operating in Ontario where nuclear materials have beenprocessed at the site since the 1930’s. In February 2017, Cameco was issued a 10 year operating licenceby the CNSC following a public hearing in Port Hope, Ontario in November 2016.
On May 5, 2017 Cameco reported a small release of hydrogen fluoride (HF) at its UF6 plant. HF is ahighly corrosive acid which can cause significant health impacts if inhaled, absorbed into the eyes ormakes contact with the skin. HF differs from other highly corrosive acids in that the fluoride ion readily
penetrates the skin causing destruction of deep tissue layers and if left untreated this can continue for days
[Ref 3]. During the night shift a junior instrumentation technician was independently performing
maintenance work on the HF transfer tank system. The employee was in the process of calibrating a
differential pressure transmitter mounted to the wall next to the tank. When the junior technician opened
the connection to the impulse line, HF gas was released creating a mist of gas above the transmitter. At
the time of the event, the HF transfer tank was depressurized with an internal pressure slightly above
atmospheric pressure. An isolation valve was not activated and was left in the open position. The
emergency ventilation system was activated by a local HF detector. Upon arriving to the affected area,
Cameco’s emergency response team secured the connection to the impulse line. The junior technician was
directed to Cameco’s medical department where he received precautionary medical attention due toexposure to HF. The worker was not injured and there were no environmental impacts as a result of thisevent.
According to Cameco’s management system, the maintenance plan for this task requires that prior tocommencing work the necessary permits and clearances shall be obtained from the production supervisor.
These clearances and permits are important to protect all workers and verify that all hazardous energy
sources have been isolated, the necessary permits obtained and controls are in place. Work clearances are
required to ensure that all hazardous energies, including in this case HF, are isolated removing any risk to
the workers and the public. The maintenance plan also stipulates that the valve for the impulse line shallbe closed.

As per the CNSC regulatory requirement, the licensee was required to do a detailed investigation of thisreported event and submit a final report for review by CNSC staff. Cameco conducted an investigation
into the event and determined that the impulse line of the pressure transmitter was not properly closed
from the HF transfer tank system. The required work clearance and permits were not obtained prior to thestart of the maintenance work at the HF transfer tank. Cameco determined that the necessary work
clearances and permits were completed after the leak occurred and not before the job, as required by
Cameco’s procedures. The junior instrument technician did not request a job clearance from the UF6
production supervisor and failed to inform the senior instrument technician of the planned work activity
as required by Cameco’s procedures.
Cameco determined that the junior and senior instrument technicians were performing maintenanceactivities without the necessary work clearances and permits for an unspecified period of time. This was
not an isolated incident and this practice was known to the UF6 production supervisor. As a result of theevent, Cameco’s management team issued a bulletin notice to all staff stressing the importance ofcompleting work clearances. Cameco terminated the junior instrument technician and the UF6 production
supervisor and the senior instrument technician was suspended from work. Cameco concluded from theirinvestigation that they had interviewed other UF6 supervisors and reviewed job safety clearances of otherteams and no other issues were identified. As a result, Cameco treated this incident as an isolated event.
Cameco posted this event on its website in compliance with the CNSC requirements.
Cameco initiated three corrective actions:a) Improve maintenance work management process for assigning work tasks to shift maintenanceemployees. The improved work process should include more visibility and accountability for thework assigned to shift maintenance employees to the production supervisors allowing improved
communication with the tradesperson.
b) Evaluate and assess current work clearances and permits to ensure they provide clear andunambiguous direction and eliminate any needless requirements.
c) Perform an assessment of the site audit program to determine what is required to enhance theaudit program to better monitor programs, especially during shift work.
Following the event CNSC inspectors conducted an unplanned, reactive inspection [Ref 4]. CNSC staff
determined that Cameco had not been conducting verification activities related to this maintenanceactivity. Cameco’s licence conditions handbook provides clarity, compliance verification criteria andguidance on meeting the CNSC requirements and expectations. Under the management system SCA therequirement states “licensees shall implement and maintain a management system”. CNSC staff assessed
the May 5, 2017 event and the compliance history of procedural non-adherence and determined that
Cameco failed to verify whether work is being performed correctly and according to approved
procedures, as required by its management system.
An AMP was issued to Cameco on September 6, 2017 in accordance with section 6(1)(b) of the CNSC’s
Administrative Monetary Penalties Regulations, for the violation of failure to comply with a condition ofa licence, in accordance with paragraph 48(c) of the NSCA [REF 5]. The purpose of the AMP was topromote compliance with licence conditions and deter future violations. The value of the AMP was $17,
830. Cameco was given a period of 35 days to review the AMP and pay the penalty or request a review of
the AMP by the Commission. On October 11, 2017 Cameco requested a review of the AMP which isscheduled for March 2018 [REF 6].
The CNSC will submit this event to the International Atomic Energy Association (IAEA) and Nuclear
Energy Agency (NEA) Fuel Incident Notification and Analysis System (FINAS) to share operational
experience with other FCF operators and regulators.

VI.

BERYLLIUM EXPSOURE LEVEL EXCEEDANCE AT BWXT’S PETERBOROUGHFACILITYBWXT manufactures nuclear fuel bundles using the fuel pellets from the Toronto facility together withzircaloy tubes manufactured in-house. The zircaloy tubes, spacers and bearing pads that form the
CANDU® nuclear fuel bundle are assembled through a process called brazing. Brazing is a metal joining
process used to join two metallic components using a filler metal as a brazing agent. BWXT uses
beryllium as the brazing agent and the spacers and bearing pads are coated with beryllium before brazing.
The beryllium coating process of the spacers and bearing pads create airborne particulates of beryllium
that can be an occupational hazard for workers at the facility.
Beryllium particulates in air are a chemical hazard and due to this, occupational exposure of beryllium toworkers in Canada is regulated by establishing Occupational Exposure Limits (OEL). OELs areapplicable to workers within a facility and not to be confused with release limits for emissions andeffluents from a facility which have an impact on the public and the environment. The Canada
Occupational Health and Safety Regulations [REF 7] specifies that employers shall use the American
Conference of Governmental Industrial Hygienists® published Threshold Limit Value® of 0.05µg/m3 as
the OEL for beryllium particulates in air - the time weighted average exposure of 8 hours that a worker
can be exposed to, above which there may be adverse health effects. With prolonged exposure, beryllium
is known to produce adverse health effects including scarring of lung tissue, beryllium sensitization and acondition called chronic beryllium disease. BWXT’s licence stipulates a value of 0.05µg/m3 as the OEL
for workers to control hazards associated with beryllium particulates present in the air. Occupational
Exposure to beryllium for workers is controlled by engineered controls (room ventilation), safe work
procedures, protective clothing and the use of full face respirators as personal protective equipment(PPE). The workplace is monitored through area monitors and personal air monitors to ensure safety.
For non-routine work or high risk tasks such as cleaning, maintenance and spillage handling (whenairborne beryllium levels are expected to be above the OEL), the approach to safety is to perform thework under a work permit system that includes identification of hazards present, confirm appropriate
training and the use of PPE like Powered Air Purifying Respirators (PAPRs) with cartridges/filters
specific for the type of hazard present.
The use of PAPRs with appropriate cartridges/filters specific to the type of hazard present provides aprotection factor of 1000 to the hazard to which the worker is exposed. This allows workers to performthese operations in a safe manner. Workers performing these tasks rely on the PAPRs to ensure beryllium
exposure is below the OEL. Workers performing these tasks, accessing beryllium process areas prone forsignificant airborne beryllium particulates are specifically trained in beryllium hazards and are monitored
regularly through blood tests to detect any beryllium sensitization.
On August 22, 2017 BWXT discovered that incorrect respirator cartridges/filters were being used with
PAPRs as PPE while performing non-routine work. As per the CNSC regulatory requirement, the licensee
was required to do a detailed investigation of this reported event and submit a final report for review by
CNSC staff. BWXT’s investigation report was provided to CNSC staff. CNSC staff issued a request
under subsection 12(2) of the General Nuclear Safety Control Regulations as the report did not provide
sufficient information about the measures BWXT has taken to minimize beryllium air concentrations in
the affected area in the future. BWXT provided a more detailed response as requested. BWXT determinedthat incorrect filters for PAPRs were procured in March 2013. Based on available general area air
sampling results from the affected areas between December 4, 2015 and August 23, 2017 BWXT
determined the average and maximum air concentration values were 0.12 µg/m3 and 0.29 µg/m3
respectively. General room air samples cannot be correlated to OELs and as a result, BWXT transitioned
from conducting routine general room air samples to routine personal air samples in late 2015. As aresult, most of the air sampling data for the period is from personal air sampling.

Based on available personal air sampling results from the affected areas between December 4, 2015 andAugust 23, 2017 the average air concentration (8-hour time weighted average) was 0.39 µg/m3 and the
maximum was 4.63 µg/m3. Based on available air sampling data for that time period, a total of 15
instances were identified where exposure to airborne beryllium by the two workers was likely to haveoccurred above the OEL. Fourteen of the identified instances impacted one worker and the other single
instance impacted a second worker. Both workers were referred to the plant occupational health nurse andphysician for follow up which includes screen tests for beryllium sensitivity. Both workers havesubsequently returned to normal duties. No health effects have been noted since the discovery of the error.
Both workers are under increased monitoring for any potential long term effects of exposure to beryllium.
BWXTs investigation determined a number of causal factors leading to the use of incorrect filters
including inadequate verification of the newly procured filters when received from supplier and thepurchase orders issued did not contain an adequate description of the item. The critical to safety list used
for enhanced purchasing controls included PAPRs but did not identify filters as critical to safety
individually by part numbers. Finally, BWXT personnel and workers did not recognize that incorrect
cartridge/filters were issued by the facility stores for use with PAPRs.
BWXT initiated several corrective actions to prevent a future occurrence of a similar event. The critical tosafety list was updated to include respirator filters individually by part number. BWXT implemented adocumented process for new or changed critical to safety items, including an approval process to ensure
workers with permissions are trained accordingly and items received are properly verified. In addition,
BWXT implemented training to include correct filter/cartridge type identification for work performed that
require respirator wearer training and provided refresher training to personnel conducting pre-job briefs inrelation to non-routine work conducted under the environmental health and safety work permit systems,
to ensure identified PPE is adequately explained.
CNSC inspectors conducted an unplanned, reactive inspection in response to the event [REF 8]. The
inspection confirmed that BWXT has responded to the event as per requirements and the root causes andassociated corrective actions were appropriate. CNSC staff also presented the event to the Commission as
an Early Incident Report, where the Commission was able to ask both BWXT and Cameco questions on
the event and the corrective actions taken. BWXT posted this event on its website in compliance with the
CNSC requirements.
The CNSC will submit this event to FINAS to share operational experience with other FCF operators andregulators.
VII.

CONCLUSIONSThe CNSC’s regulatory framework provides many mechanisms for ensuring that nuclear activities in
Canada are conducted safely, ensuring the health and safety of workers, the public and the environment.
The NSCA is especially powerful, as it provides the CNSC with the regulatory oversight of hazardous
substances at nuclear facilities.
The CNSC compliance program is multifaceted with many tools available to CNSC staff and inspectors
in ensuring compliance with regulatory requirements. This paper outlined two specific examples of events
that occurred at Canadian FCFs and highlighted the different approaches used by CNSC staff in ensuring
the health and safety of workers.
Sharing operational experiences provides the opportunity for both regulators and operators to learn fromevents at other FCFs. From the regulators perspective there were a number of lessons learned from these
events that have been applied to the compliance program for all FCFs in a graded approach. This includes
supply chain validation, confirming the licensee’s audit program covers all aspects of the work activity
and a closer examination of licensee self-assessments. The CNSC strongly encourages licensees to sharelessons learned amongst the industry. In addition the CNSC utilizes FINAS to share operational
experiences with Canadian FCF operators.

VIII.

REFERENCES

REF 1 – www.nuclearsafety.gc.ca
REF 2 – RD/GD 99.3 Public Information and Disclosure March 2012
REF 3 – Cameco Site Information Handbook April 2017
REF 4 – CNSC Staff Inspection Report PHCF May 11 and 12, 2017
REF 5 – Administrative Monetary Penalty Notice of Violation Cameco PHCF
REF 6 – Request for Review of Administrative Monetary Penalty: Cameco Corporation–2017-AMP-06
REF 7 – Canada Occupational Health and Safety Regulations 2017
REF 8 – CNSC Staff Inspection Report BWXT October 5, 2017