Science and Technology Series

Jean Couturier, Michel Schwarz
Current State of Research
on Pressurized Water Reactor Safety
Jean Couturier, Michel Schwarz

IRSN is a public industrial and commercial body undertaking research and consultancy
in the field of nuclear safety and radiation protection. It provides the public authorities
with technical support. It also carries out various public service missions entrusted toit under national regulations. In particular, these include radiological monitoring of
the environment and workers in France, managing emergency situations, and keeping
the public informed. IRSN makes its expertise available to partners and customers
both in France and worldwide.

Head office31, avenue de la Division Leclerc
92260 Fontenay-aux-Roses
RCS Nanterre B 440 546 018
Telephone +33 (0)1 58 35 88 88
Postal Address
B.P. 17 - 92262 Fontenay-aux-Roses Cedex
FranceWebsite www.irsn.fr

ISBN : 978-2-7598-2164-8

9 782759 821648
Price : 35 euros

Current State of Research on Pressurized Water Reactor Safety

This publication reflects IRSN’s desire to keep an enduring record of its results andto share its knowledge.

Enhancing nuclear safety

Science and Technology Series

For more than 40 years, IPSN then IRSN has conducted research and developmenton nuclear safety, specifically concerning pressurized water reactors, which are
the reactor type used in France. This publication reports on the progress of thisresearch and development in each area of study – loss-of-coolant accidents, coremelt accidents, fires and external hazards, component aging, etc. –, the remaining
uncertainties and, in some cases, new measures that should be developed toconsolidate the safety of today’s reactors and also those of tomorrow. A chapterof this report is also devoted to research into human and organizational factors,
and the human and social sciences more generally. All of the work is reviewed
in the light of the safety issues raised by feedback from major accidents such as
Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments
conducted, for example, as part of the ten-year reviews of safety at French nuclear
reactors. Finally, through the subjects it discusses, this report illustrates the many
partnerships and exchanges forged by IRSN with public, industrial and academic
bodies both within Europe and internationally.

Current State of Research
on Pressurized Water
Reactor Safety

Science and Technology Series

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Page 1

Current State of Research
on Pressurized Water Reactor Safety

Jean Couturier, Michel Schwarz

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Page 2

Cover illustrations: photographs showing research and development in pressurized water
reactor safety at IRSN. From left to right: the hydrogen risk in nuclear reactors © IRSN Institut
Icare; image from a 3D simulation by the MC3D code of corium dispersion in the water in areactor pit in the event of a central reactor vessel failure © IRSN; states of degradation of fuelassemblies from the Phébus-PF program © IRSN; a prototype of the conformable transducer –IRSN/CEA patent © IRSN/CEA (bottom); Phébus-LOCA – section view (post-mortem) of a test
fuel cluster following a temperature transient typical of a LOCA © IRSN (top); an experiment
involving a ﬁre in a stack of cable raceways © Florent–Frédéric Vigroux/IRSN (bottom); 2D
model of the Nice basin; below, ampliﬁcation of seismic waves calculated by a numericalsimulation of seismic wave propagation in this model © Fabien Peyrusse/Inria; ﬁre test on
electrical cables © Florent–Frédéric Vigroux/IRSN (bottom).

Printed in FranceISBN (print): 978-2-7598-2164-8 – ISBN (ebook): 978-2-7598-2165-5
DOI: 10.1051/978-2-7598-2164-8

All rights relative to translation, adaptation and reproduction by any means whatsoever are
reserved, worldwide. In accordance with the terms of paragraphs 2 and 3 of Article 41 of theFrench Act dated March 11, 1957, “copies or reproductions reserved strictly for private use andnot intended for collective use” and, on the other hand, analyses and short quotations forexample or illustrative purposes, are allowed. Otherwise, “any representation or reproduction- whether in full or in part – without the consent of the author or of his successors or assigns, isunlawful” (Article 40, paragraph 1). Any representation or reproduction, by any means
whatsoever, will therefore be deemed an infringement of copyright punishable under Articles
425 and following of the French Penal Code.
 IRSN 2018

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Page 3

PrefaceThis publication sets out to provide an overview of much of the research anddevelopment work carried out over the past forty years in the ﬁeld of pressurized water
reactor safety, in particular by the Institut de Protection et de Sûreté Nucléaire (IPSN),
later becoming the Institut de Radioprotection et de Sûreté Nucléaire (IRSN), whether
alone or in collaboration with other organizations. This work – and the lessons that can belearned from it – is set in front of the safety issues raised during assessments of French
nuclear power plants, or following incidents and accidents that have occurred, such as
those affecting reactor 2 at the Three Mile Island nuclear power plant in 1979, reactor 4 at
the Chernobyl nuclear power plant in 1986 and, more recently, at the Fukushima Daiichi
nuclear power plant in 2011.
The safety of nuclear facilities is part of a continuous improvement process that isbased on:– national and international operating experience feedback,– and knowledge acquired through research and development, with assessment andresearch activities fueling each other.
Since the 1980s, safety reviews have been carried out at nuclear power plants
operated by Électricité de France (EDF), the national power utility, as part of a ten-yearly
outage program. These provide an opportunity to ensure that lessons learned fromoperating experience feedback and R&D work ﬂow down to the operational level.
The research and development themes relating to pressurized water reactor safety inwhich IRSN has been – and for the most part still is – particularly involved are part of thegeneral initiative to prevent and mitigate the impact of postulated events. These include
internal events, such as the loss of coolant or an uncontrolled increase of reactivity in thecore until it melts, and "hazards" such as earthquakes, external ﬂooding, and ﬁres insideor outside facilities. The knowledge acquired through research on core melt accidents isAnglais_BookIV

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Page 4Current state of research on pressurized water reactor safety

fed into ASTEC, an IRSN computer code used for simulation purposes, and internationallyrecognized as a reference tool.
The authors also wished to provide a panorama of studies and research work relating
to human and organizational factors, a theme that has been of concern to theinternational community since the accidents at Three Mile Island and Chernobyl. The
safety issues raised by the growing tendency of licensees to work with subcontractors
have become major themes of studies and research today. Furthermore, since theaccident at the Fukushima Daiichi nuclear power plant, and the conclusions drawn fromthe event – in particular those of the National Diet of Japan Fukushima Nuclear Accident
Independent Investigation Commission, according to which various social factors contributed to the accident – studies and research initiatives have covered topics in thehuman and social sciences.
For each of the research and development themes addressed, the remaining
uncertainties and the new knowledge required are highlighted. Research in new areas,such as topics relating to new reactor designs (passive safety, the possibility of retaining
the corium in the reactor vessel in the event of a core melt accident, etc.) is also
discussed.
The research and development work discussed in this publication illustrates IRSN’s
policy of opening up to the international community and the many ties it has developedwith it. Examples include direct ties with similar organizations (such as the United States
NRC or GRS in Germany) research operators (in particular the Commissariat à l’énergie
atomique et aux énergies alternatives [CEA] and universities) as well as involvement inknowledge sharing organizations (such as the OECD Nuclear Energy Agency) and themany projects funded in France by the Agence nationale de la recherche (ANR) and at theeuropean level by the European Commission through its multi-year Framework Programs(FP) for research and development. In addition, IRSN is closely involved in deﬁning
European research strategies within the European Sustainable Nuclear Energy Technology Platform (SNETP) and, since 2012, as a member of the NUGENIA association.
I would especially like to thank the two main contributors, Jean Couturier and Michel
Schwarz, for this impressive summary (representing more than three years’ work) and,more generally, all the IRSN experts for their invaluable assistance in this work.
Jean-Christophe Niel
IRSN Director-GeneralAnglais_Book

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Page 5The authors

Jean Couturier works at the ofﬁce of IRSN’s Director General, supporting the roll-out
of a knowledge management program. He is also a senior expert (safety policies, risk
analyses). He began his career working on the design of fast reactors. From 1982, his work
focused on nuclear safety not only for this type of reactor (PHENIX, SUPERPHENIX) but
also for research reactors, pressurized water reactors, etc. He is a member of the Standing
Group of experts for Reactors and the Standing Group of experts on Nuclear Pressurized
Equipment.
Michel Schwarz retired from IRSN in 2012. He spent his career in nuclear safety
research. In particular, he ran the Phébus-PF international research program on core melt
accidents in light water reactors. He was director of major accident prevention and then
IRSN’s scientiﬁc director. He is a member of ASN’s Scientiﬁc Committee.

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List of abbreviations

Glossary of institutions
AEC: Atomic Energy Commission, USA (forerunner of the U.S.NRC)
AECL: Atomic Energy of Canada Limited, a nuclear science and technology research
institute
AFPS: Association française du génie parasismique (French Earthquake Engineering
Association)
ANDRA: Agence nationale pour la gestion des déchets radioactifs (French National
Radioactive Waste Management Agency)
ANL: Argonne National Laboratory, USA
ANR: Agence nationale pour la recherche (French National Research Agency)
AREVA: French nuclear operator
ARMINES-SPIN: Institut Carnot M.I.N.E.S – Centre des Sciences des processus industriels
et naturels de l’École des Mines de Saint-Étienne, France (Carnot M.I.N.E.S Institute –Center for Industrial and Natural Processes of the École des Mines in Saint-Etienne)
ASME: American Society of Mechanical Engineers, USA (commonly used to refer to thedesign and construction rules drawn up by this American society and used by American
designers [Westinghouse, etc.])
ASN: Autorité de sûreté nucléaire (French Nuclear Safety Authority)
BelV: Belgian Federal Agency for Nuclear and Radiological Inspections of nuclear
installations
BETCGB: Bureau d’études techniques et de contrôle des grands barrages (French
Technical and Inspection Ofﬁce for Large Dams)
BRGM: Bureau de recherche géologique et minière (French Geological Research Mining
Bureau)Anglais_Book

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C3R: Laboratoire de cinétique chimique, combustion et réactivité (Laboratory of Chemical Kinetics, Combustion and Reactivity is a French joint research laboratory establishedby IRSN, CNRS and the University of Lille 1 Science and Technology)
CEA: Commissariat à l’énergie atomique et aux énergies alternatives (French Alternative
Energies and Atomic Energy Commission)
CEBTP: Centre d’essais du bâtiment et des travaux publics (the French Construction andPublic Works Test Center is a center of expertise in soil engineering, materials andstructural engineering, and construction)
CEGB: Central Electricity Generating Board, UK
CEMAGREF: Centre national du machinisme agricole, du génie rural, des eaux et desforêts (French National Center for Agricultural Machinery, Rural Engineering, Water andForestry, research institute)
CERIB: Centre d’études et de recherches de l’industrie du béton (French Concrete
Industry Study and Research Center)
CETMEF: Centre d’études techniques maritimes et ﬂuviales (French Technical Maritime
and River Study Center)
CNL (formerly AECL): Canadian Nuclear Laboratories
CNPP: Centre national de prévention et de protection (French National Center for
Prevention and Protection)
CNR: Compagnie nationale du Rhône (French National Company of the Rhone)
CNRS: Centre national de la recherche scientiﬁque (French National Center for ScientiﬁcResearch)
CNSC: Canadian Nuclear Safety Commission
CORIA: Complexe de recherche interprofessionnel en aérothermochimie (French Aerothermochemistry Research Complex)
CRL: Chalk River Laboratories, Canada
CSNI: Committee on the Safety of Nuclear Installations, OECD
CSN: Consejo de Seguridad Nuclear (Spanish Nuclear Safety Council)
CSO: Centre de sociologie des organisations (French Center for Sociology of Organizations, a joint research unit of Sciences Po and CNRS)CSTB: Centre scientiﬁque et technique du bâtiment (French Construction Science andTechnology Center)
CTICM: Centre technique industriel de la construction métallique (French Industrial
Technology Center for Construction in Metal)
DCNS: French hi-tech company specializing in defense naval systems
DGA: Direction générale de l’armement (Ministère de la Défense) (French General
Directorate for Armament, [French Ministry of Defence])
DGPR: Direction générale de la prévention des risques (French General Directorate for
Risk Prevention, a government department reporting to the Ministry of Ecology,
Sustainable Development and Energy)
DMT: Département de mécanique et thermique (Mechanical and Thermal Engineering
Department at CEA’s research platform in Saclay, France)Anglais_Book

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List of abbreviations

Page 9IX

EDF: Électricité de France (French power utility)
ENEL: Ente Nazionale per l’Energia Elettrica, Italian power utility
ENS Cachan: École nationale supérieure de Cachan (a prestigious French higher National
School)ENSI: Eidgenössisches Nuklearsicherheitsinspektorat (Swiss Federal Nuclear Safety
Inspectorate)
ENSREG: European Nuclear Safety Regulators Group
EPRI: Electric Power Research Institute, USA
ETIC: Laboratoire d’étude des incendies en milieux conﬁnés (Laboratory for the Study ofFire in a Contained Environment, IRSN, France)
Euratom: European Atomic Energy Community
FzK: Forschungszentrum Karlsruhe (Karlsruhe Institute of Technology, Germany)
Framatome: nuclear steam supply system manufacturer, France
GDF: Gaz de France (French gas utility)
GeM: Institut de recherche en génie civil et mécanique, unité mixte CNRS/École centralede Nantes/Université de Nantes (French Civil and Mechanical Engineering ResearchInstitute, a joint CNRS/École Centrale de Nantes/University of Nantes unit)
GFH: Groupe facteurs humains (Human Factors Group, EDF, France)
GrDF: Gaz, réseau, distribution France (French Gas, Network management, Distribution
Company)
GRS: Gesellschaft für Anlagen – und Reaktorsicherheit (reactor safety organization,
Germany)
HAMMLAB: Halden Man-Machine Laboratory (Norwegian laboratory researching manmachine interactions and control processes)
HSE: Health and Safety Executive, UK
IAEA: International Atomic Energy Agency, Vienna, Austria
IBRAE: Nuclear Safety Institute of the Russian Academy of Sciences, Russia
ICARE: Institut de combustion aérothermique réactivité et environnement (Institute for
Combustion, Aerothermal Engineering, Reactivity and Environment, CNRS, Orléans,France)IFE: Institutt for energiteknikk (Institute for Energy Technology, Norway)
IFREMER: Institut français de recherche pour l’exploitation de la mer (French ResearchInstitute for Exploitation of the Sea)
IFSTTAR: Institut français des sciences et technologies des transports, de l’aménagement
et des réseaux (French Institute of Science and Technology for Transport, Development
and Networks)
INERIS: Institut national de l’environnement industriel et des risques (French National
Institute for the Industrial Environment and Risks)
INL: Idaho National Laboratory, USA
Inria: Institut national de recherche dédié au numérique (French National ResearchInstitute for the Computational Sciences)Anglais_BookX

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Current state of research on pressurized water reactor safety

INQUA: International Union for Quaternary Research (organization responsible forpromoting international collaboration in the earth sciences)
INSA: Institut national des sciences appliquées (French National Institute of Applied
Sciences)
IPSN: Institut de protection et de sûreté nucléaire (French Institute for Nuclear Safety
and Protection)
IREX: Institut pour la recherche appliquée et l’expérimentation en génie civil (FrenchInstitute for Applied Research and Experimentation in Civil Engineering)
IRSN (formerly IPSN): Institut de radioprotection et de sûreté nucléaire (French Institutefor Radiological Protection and Nuclear Safety)
ISTC: International Science and Technology Center
IUSTI: Institut universitaire des systèmes thermiques industriels (French University
Institute for Industrial Thermal Systems, a joint research unit of CNRS/Universities ofProvence and the Mediterranean)
JAEA: Japan Atomic Energy Agency
JAERI (formerly JAEA): Japan Atomic Energy Research Institute
KAERI: Korea Atomic Energy Research Institute, South Korea
KEPCO: Kansai Electric Power Company, Japan
KfJ: Forschungszentrum Jülich (Jülich Research Center, Germany)
KfK: Kernforchungszentrum Karlsruhe (Karlsruhe Institute of Technology, Germany)
KIT (formerly FzK and KfK): Karlsruhe Institut für Technology (Karlsruhe Institute ofTechnology, Germany)
KWU: Kraftwerk Union (German nuclear power plant construction company)
Labra: Laboratoire des rayonnements appliqués (Laboratory for Applied Radiation, CEA,France)
LaMCoS: Laboratoire de mécanique des contacts et des structures (Laboratory for the
Mechanics of Contacts and Structures, a joint research unit of INSA in Lyon and CNRS,France)
LCPC: Laboratoire central des ponts et chaussées (French Central Laboratory for Bridges
and Roads)
LEFH: Laboratoire d’étude des facteurs humains (Laboratory for the Study of Human
Factors, IPSN, France)
LEMTA: Laboratoire d’énergétique et de mécanique théorique et appliquée (Laboratory ofEnergy and Theoretical and Applied Mechanics, jointly run by the University of Lorraine
and CNRS, France)
LEPMI: Laboratoire d’électrochimie et de physico-chimie des matériaux et des interfaces(Laboratory of Electrochemistry and Physico-chemistry of Materials and Interfaces in
Grenoble, France)
LMA: Laboratoire de mécanique et d’acoustique, (Mechanics and Acoustics Laboratory,
France)
LMDC: Laboratoire matériaux et durabilité des constructions (Construction Materials andDurability Laboratory, France)Anglais_Book

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List of abbreviations

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LNE: Laboratoire national de métrologie et d’essais (French National Laboratory for
Metrology and Testing)
LSHS: Laboratoire des sciences humaines et sociales (Human and Social Sciences
Laboratory, IRSN, France)
LVEEM: Laboratoire Vellave sur l’élaboration et l’étude des matériaux (Vellave Laboratory on the Development and Study of Materials in Le Puy-en-Velay, France)
MAI: Materials Ageing Institute, France (an international research and development
institute set up by EDF, specializing in research into the ageing of materials used in power
plants)
MIST: Laboratoire de micromécanique et intégrité des structures, laboratoire “ sans mur ”
commun au CNRS et à l’IRSN (Micromechanics and Structural Integrity Laboratory,
jointly funded by the CNRS and IRSN, could be described as a laboratory “without walls”,
France)
NEA: Nuclear Energy Agency, OECD
NEI: Nuclear Energy Institute, USA
NIIAR: Scientiﬁc Research Institute of Atomic Reactors, Russia
NIST: National Institute of Standard and Technology, USA
NITI: Aleksandrov Scientiﬁc Research Technological Institute, Saint-Petersburg, Russia
NRA: Nuclear Regulation Authority, JapanNUGENIA: NUclear GENeration II & III Association (international association dedicatedto the safety of generation II and III reactors)
NUPEC: Nuclear Power Engineering Center, Japan
OECD: Organization for Economic Co-operation and Development
NGO: Non-Governmental Organization
ORNL: Oak Ridge National Laboratory, USA
PC2A: Laboratoire de physico-chimie des processus de combustion et de l’atmosphère(Laboratory for the Physical Chemistry of Combustion Processes and the Atmosphere,
Lille, France)
PNNL: Paciﬁc Northwest National Laboratory, USA
PROMES: Procédés, matériaux et énergie solaire (Processes, Materials and Solar Energy,
laboratory in Perpignan, France)
PSI: Paul Scherrer Institute, Switzerland
RESIF: Réseau sismologique & géodésique français (French Seismologic and Geodesic
Network)
ROSATOM: Federal Agency on Atomic Energy, Russia
SHOM: Service hydrographique et océanographique de la marine (French Naval Hydrographic and Oceanographic Service)
SNCF: Société nationale des chemins de fer (French rail operator)
SNL: Sandia National Laboratory, USA
STUK: Radiation and Nuclear Safety Authority, Finland
U.S.NRC: United States Nuclear Regulatory Commission, USAAnglais_Book

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Page 12Current state of research on pressurized water reactor safety

VTT: Technical Research Center, Finland
WGAMA: Working Group on Analysis and Management of Accidents, OECD/NEA/CSNIWGELE: Working Group on Electrical Power, OECD/NEA/CSNIWGEV: Working Group on External Events, OECD/NEA/CSNIWGFCS: Working Group on Fuel Cycle Safety, OECD/NEA/CSNI
WGFS: Working Group on Fuel Safety, OECD/NEA/CSNIWGHOF: Working Group on Human and Organizational Factors, OECD/NEA/CSNIWGIAGE: Working Group on Integrity and Ageing of Components and Structures, OECD/
NEA/CSNIWGRISK: Working Group on Risk Assessment, OECD/NEA/CSNI
Technical glossary
AAR: Alkali-Aggregate Reactions
ABI: name of a program of tests conducted as part of research into reactor vessel failure
and basemat erosion due to corium in a reactor core melt accident
ACHILLE: name of an experimental program to study the behavior of fuel rods
ACRR: Annular Core Research Reactor (experimental reactor run by SNL to study coretemperature rise and core melt in a reactor vessel)
AGORAS: Amélioration de la gouvernance des organisations et des réseaux d’acteurs pourla sûreté nucléaire (Improving Governance of Organizations and Networks involved in
Nuclear Safety, research project)
AIC: Argent-Indium-Cadmium (Silver-Indium-Cadmium)
ALPHA: name of an experimental facility run by JAERI to study steam explosions in areactor
ALPS: Advanced Light water reactor Performance and Safety (international research
program)
APHRODITE: name of a series of analytical tests to study two-phase thermal hydraulics
ARTEMIS: name of a test program conducted as part of research into reactor vessel failure
and basemat erosion due to corium in a reactor core melt accident
ARTIST: Aerosol Trapping in a Steam Generator (experimental facility run by PSI – studyof aerosol retention in a steam generator)
ASTEC: Accident Source Term Evaluation Code (system of simulation codes for evaluating the physical phenomena that occur during a core melt accident in a pressurized water
reactor)
ASTRID: Advanced Sodium Technological Reactor for Industrial Demonstration (demonstration sodium-cooled fast reactor project)
AZALEE: name of a CEA shake table
BALI: name of a series of tests conducted as part of research into reactor core melt
accidents
BALISE: name of a test program conducted as part of research into reactor vessel failure
and basemat erosion due to corium in a reactor core melt accident

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List of abbreviations

Page 13XIII

BETA: name of a KIT experimental facility used to study corium/concrete interaction aspart of research into reactor core melt accidents
BETHSY: name of a CEA experimental facility – tests used to check the ability of the
CATHARE code to predict satisfactorily the behavior of a nuclear steam supply system inan accident situation
BFC: Bottom of Fissile Column
BILLEAU: name of tests to study steam explosion in a reactor (study of the dispersion of ajet of solid spheres in a test section ﬁlled with water)
BIP: Behavior of Iodine Project (international program to study iodine behavior in areactor containment)
BK: fuel building
BWR: Boiling Water Reactor
CABRI: name of a CEA test reactor used to study accident situations in reactors (PWR,
FNR)
CAD: Computer-Aided Design
CADUCEE: name of an experimental apparatus on the GALAXIE platform for research
into controlling ﬁre risks in nuclear facilities
CALIST: Characterization and Application of Large and Industrial Spray Transfer (IRSN
experimental facility for studying spray mechanisms)
CANON: name of a series of analytical tests to study two-phase thermal hydraulics
CARAIDAS: name of an IRSN experimental facility – study of the behavior of radioactive
products in a reactor containment
CAIMAN: name of a CEA experimental facility – study of iodine behavior in a reactor
containment
CARINEA: name of an experimental facility on the GALAXIE platform used for research
into controlling ﬁre risks in nuclear facilities
Cast3M: name of a ﬁnite-element computer code used for the mechanics of structuresand ﬂuids
CATHARE: Code avancé de thermohydraulique pour les accidents de réacteurs à eau(Advanced Thermohydraulics Code for Water Reactor Accidents, a simulation code used
for safety analyses)
CATHODE: name of some analytical test programs conducted as part of research into
controlling ﬁre risks in nuclear facilities
CAV: Cumulative Absolute Velocity
CCI: Corium Concrete Interaction
CCM: Cold Crucible Melting (test program to study the fragmentation of molten
mixtures)
CCWS: Component Cooling Water System
CEOS.fr: Comportement et évaluation des ouvrages spéciaux (Behavior and Evaluation ofSpecial Structures, a French research project)
CESAR: name of a module of the ASTEC codeAnglais_Book

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Current state of research on pressurized water reactor safety

CIP: CABRI International Program (international program to study the behavior of nuclear
fuel rods and their cladding in a reactivity injection accident in PWR)
CIR: Cooperative Irradiation-assisted stress corrosion cracking Research (an international
research and development program on the corrosion of stainless steels under stress andunder irradiation)
CIVA: name of a numerical simulation platform for non-destructive testing
CFD: Computational Fluid Dynamics
CFR: Code of Federal Regulations
CHIP: Chimie de l’iode dans le circuit primaire (Iodine Chemistry in the Reactor Coolant
System, an IRSN facility or research program contributing to better evaluation of thequantity of iodine that can be released in a reactor core melt accident)
CHRS: Containment Heat Removal System
CIRCE: name of a research program on the corrosion of nickel-based alloys under stress
CLARA: name of a test program conducted as part of research into reactor vessel failure
and basemat erosion due to corium in a reactor core melt accident
CODAZIR: research program to study the fuel rod behavior in a LOCACOMET: name of an experimental facility and a concept developed by KIT to research
reactor core melt accidents
COPAT: Centre opérationnel de pilotage des arrêts de tranche (EDF), see OCC
COPO: Corium Pool Facility (Finnish experimental facility – research into the potential for
retaining the corium in the reactor vessel in a reactor core melt accident)
CORA: name of a research program on core temperature rise and core melt in the reactor
vesselCORDEB: Corium-Debris (experimental program of research into reactor core melt
accidents)
CROCO: name of a code for simulating corium spreading
CSD: Combustible sévèrement dégradé (Severely Degraded Fuel)CSS: Containment Spray System
CVCS: Chemical and Volume Control System
DA: diagnostic automatique (automatic diagnostics)
DANAIDES: Dispositif analytique pour l’étude, en cas d’incendie, du dysfonctionnementélectrique par les suies (Analytical Equipment for Studying Electrical Malfunctions caused
by Soot during a Fire, an experimental facility on the GALAXIE platform)
DBE: Design Basis Earthquake (used during the design of nuclear facilities)
DCH: Direct Containment Heating
DDF: durée de fonctionnement (Operating Life [EDF project])
DDT: Deﬂagration–Detonation Transition
DEBORA: name of a series of analytical tests for research into two-phase thermal
hydraulics
DEF: Delayed Ettringite Formation

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DELTA: name of some analytical test programs run as part of research into controlling ﬁrerisks in nuclear facilities
DENOPI: Dénoyage piscines (Spent Fuel Pool Water Uncovery, a research program on
accidental water uncovery of a nuclear fuel storage pool)
DEVAP: Dépôt en phase vapeur des produits de ﬁssion volatils sur les surfaces des circuits(Volatile Fission Product Deposits in the Vapour Phase on Reactor System Surfaces, ananalytical test program to study the transfer of radioactive products in reactor systems
during a core melt accident)
DF: Damage Fuel (a research program to study core temperature rise and core melt in areactor vessel)
DISCO: DISpersion of simulated COrium (KIT experimental facility to study airborne
contamination/corium dispersion, using inert powders)
DIVA: Dispositif incendie ventilation et aérocontamination (Fire, Ventilation and Airborne
Contamination Device, IRSN experimental facility for conducting ﬁre tests in laboratories
and factories or in a pressurized water reactor)
DNB: Departure from Nucleate Boiling
DRACCAR: Déformation et renoyage d’un assemblage de crayons de combustible pendantun accident de refroidissement (Deformation and Reﬂooding of a Fuel Rod Assembly
during a Loss-Of-Coolant Accident, simulation code)
DRIVER: name of a KIT experimental facility for studying hydrogen risk
ECO: Experiments on energy COnversion during a steam explosion
ECOBA: Étude du conﬁnement des ouvrages en béton armé (Study of Reinforced ConcreteContainment Structures, a research project to improve the assessment of containment
integrity in pressurized water reactors during a core melt accident)
ECS: Complementary Safety Evaluation
EDGAR: name of a CEA experimental facility – study of fuel rod behavior
ELISA: name of an experimental loop for research into cooling water recirculation under
accident conditions
EMAIC: Émission de l’argent, de l’indium et du cadmium (Silver, Indium and Cadmium
Emission, a series of tests to study the release of products from Ag-In-Cd control rods
during accident transients)
EMIS: Émission de produits de ﬁssion (Fission Product Emission, a code for simulating therelease of ﬁssion products [a forerunner of ELSA])ENACCEF: Enceinte accélération de ﬂamme (Flame Acceleration Enclosure, an experimental facility run by CNRS/ICARE, Orléans, France)
ENISTAT: Experimental and Numerical Investigation of Shear wall reinforced concretebuildings under Torsional effects using Advanced Techniques (a European program ofexperimental and numerical studies)
EPICUR: Études physico-chimiques de l’iode conﬁné sous rayonnement (Physical andChemical Studies of Contained Iodine under Radiation, an IRSN facility or research
program to validate models of iodine chemistry in the containment of a pressurized water
reactor in an accident situation)

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Current state of research on pressurized water reactor safety

EPR: European Pressurized Water Reactor
ERCOSAM-SAMARA: Containment thermal-hydraulics of current and future LWRs for
Severe Accident Management (a research program on hydrogen risk)
ESSOR: Essai orgel (ORGEL Reactor Test, research reactor at the Ispra Joint ResearchCentre, Italy)
EVA: étude du vieillissement des aciers (Study of the Ageing of Steel)
FABIME: name of a series of tests on thermal fatigue and their test equipment
FALCON: name of a program of analytical tests to study the transfer of radioactive
products in reactor systems during a core melt accident
FARO: name of a test program to study interactions between fuel and coolant (steam
explosions)
FAT3D: name of a series of tests on thermal fatigue and their test equipment
FDS: Fire Dynamics Simulator
FEBA: Flooding Experiments with Blocked Arrays (experimental program to study fuel rod
behavior in a loss-of-coolant accident)
FGD: Fission Gas Dynamics (analytical tests to study reactivity accidents)
FITS: Fully Instrumented Test Series (Sandia National Laboratories experimental facility
used to study steam explosion in a reactor)
FLECHT-SEASET: Full-Length Emergency Core Cooling Heat Transfer–Separate Effects
tests And System-Effects Tests (experimental program – study of fuel rod behavior in aloss-of-coolant accident)
FLHT: Full Length High Temperature (research program on core temperature rise and coremelt in a reactor vessel)
FLIP: Feux de liquide en interaction avec une paroi (Interaction of Liquid Fires with a Wall,research program on solvent ﬁres for fuel reprocessing plant safety)
FLUENT: name of a numerical ﬂuid mechanics code
FNR: Fast Neutron Reactor
FP: Fission Products
FP: Framework Program, European Commission
FPCPS: Fuel Pool Cooling and Puriﬁcation System
FPT: Fission Product Test (acronym associated with the tests run as part of the Phébus-PF
program)
GALAXIE: name of an IRSN platform of experimental facilities for research into controlling ﬁre risks in nuclear facilities
GONDOLE: name of a research program – swelling of steel under irradiation
GMR: Giant Magnetoresistance
GPS: Global Positioning System
H2-PAR: Hydrogen Passive Autocatalytic Recombiners (test program)
HEPA: High Efﬁciency Particulate Air
HEVA: Hélium, vapeur (Helium, Steam – experimental program)Anglais_Book

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HI: Horizontal Induction (tests as part of a study of ﬁssion product emission by fuel when
its temperature rises)
HOF: Human and Organizational Factors
HRO: High Reliability Organizations
HRP: HALDEN Reactor Project (research aimed at improving nuclear power plant safety)
HT: High Temperature
HYCOM: integral large scale experiments on HYdrogen COMbustion for severe accident
code validation (European project)
HYDRA: name of an experimental device on the GALAXIE platform for research into
controlling ﬁre risks in nuclear facilities
HYDRAZIR: research program to study fuel rod behavior in a loss-of-coolant accident
ICE: Interaction corium-eau (Corium-Water Interaction, experimental program)
IGR: Impulse Graphite Reactor (Russian research reactor – study of reactivity accidents)
IMPACT: name of an experimental program as part of research and development on the
behavior of engineered structures subject to an impact
INB: installation nucléaire de base (Basic Nuclear Installation)
INCEFA: INcreasing Safety in NPPs by Covering gaps in Environmental Fatigue Assessment, European project
INSAG: International Nuclear Safety Group
InSa: Interferometric Synthetic aperture radar (technique used in geodesy and remotesensing)
IRIS 2010: Improving Robustness assessment of structures Impacted by missileS (international benchmarking on the behavior of engineered structures subject to an impact)
IRMA: Irradiation matériaux (IRradiation of MAterials, a CEA experimental facility forstudying radiation/material interaction mechanisms)
IRWST: In-containment Refueling Water System Tank (borated water tank located inside
the EPR containment building)
ISAFES: Interactive Seismic Analysis of Fragilities of Equipment and Structures
ISIS: name of a software system for analysing the ventilation of a ﬁre and airborne
contamination
ISP: International Standard Problem
ISR: ingénieur de sûreté-radioprotection (Safety and Radiation Protection Engineer)
ISTP: International Source Term Program
ITER: International Thermonuclear Experimental Reactor
IVANA: name of a VUEZ experimental facility used for research into cooling water
recirculation in accident conditions
IVMR: In-Vessel Melt Retention
IVR: In-Vessel Retention
KALI H2: name of a CEA experimental facility for studying hydrogen risk

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KARISMA: KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment (international benchmarking carried out following an earthquake affecting the Kashiwasaki
Kariwa nuclear power plant in Japan)
KMS: Russian experimental facility – hydrogen risk
KONVOI: name of a pressurized water reactor of German design
KROTOS: name of a CEA experimental facility for studying the interactions between amolten mixture and a coolant (steam explosions)
LDV: Laser Doppler Velocimetry
LII: Laser Induced Incandescence
LOBI: name of an experimental facility at the Ispra Joint Research Centre, used forstudying the thermohydraulic behavior of a nuclear reactor in an accident situation
LOCA: Loss-of-Coolant Accident
LOFT-FP: Loss Of Fluid Tests–Fission Product (research project on core temperature riseand core melt in a reactor vessel due to a loss of coolant)
LS-DYNA: name of a rapid dynamic simulation code for studying structures
LSTF: Large Scale Test Facility (Japanese experimental facility – study of the thermohydraulic behavior of a nuclear reactor in an accident situation)
MAAP: Modular Accident Analysis Program (software [or software system] for thesimulation of physical phenomena that occur during a core melt accident in a pressurizedwater reactor)
MACE: Melt Attack and Coolability Experiments (test program on corium-concreteinteraction as part of research into reactor core melt accidents)
MAEVA: Maquette échange vapeur air (Steam Air Exchange Model, EDF experimental
model for studying on a large scale the behavior of a containment in a loss-of-coolant
accident)
MAFFé: name of an oven on the EDF experimental platform used for research into
controlling ﬁre risks in nuclear facilities
MANON: name of an experimental loop used for research into cooling water recirculationin accident conditions
MARC: Masse, amortisseur, raideur, critère (Weight, Damper, Stiffness, Criterion)
MASCA: name of a test program conducted as part of research into reactor core melt
accidents
MC3D: name of a 3D multi-phase thermal hydraulics code for simulating the interaction
between molten materials and a coolant
MCCI: Molten Core-Concrete Interaction
MELCOR: name of a software (or software system) for simulating the physical phenomena that occur during a core melt accident in a pressurized water reactor
MFPR: Module for Fission Product Release (code for simulating ﬁssion product behavior)
MILONGA: name of an experimental platform developed by EDF for research into
controlling ﬁre risks in nuclear facilities
MIRE: Mitigation des rejets à l’environnement en cas d’accident nucléaire (Mitigation of
Releases into the Environment in the Event of a Nuclear Accident, research program toAnglais_Book

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XIX

study and improve the mitigation of radioactive releases in the event of a reactor coremelt accident)
MISS3D: Modélisation de l’interaction sol-structure en trois dimensions (3D Modelling of
Ground-Structure Interaction, a simulation software)
MISTRA: name of a CEA experimental facility – studies of hydrogen risk
MITHYGENE: Mitigation hydrogène (Hydrogen Mitigation, improving knowledge ofhydrogen risk and how to manage it in a reactor core melt accident)
MIT3BAR: Évaluation et mitigation du risque de percement de la troisième barrière deconﬁnement des centrales nucléaires (Assessment and Mitigation of the Risk of Third
Containment Barrier Breakthrough at Nuclear Power Plants)
MOCKA: name of a KIT experimental facility – study of corium-concrete interaction in areactor core melt accident
MOBY DICK: name of a series of analytical tests conducted as part of the study of twophase thermal hydraulics
MOX: Mixed Oxide Fuel (UO2 + PuO2)
MOZART: Mesure de l’oxydation du zirconium par l’air en température (Measurement of
Zirconium Oxidation by Air at Temperature, a program of analytical tests to study theoxidation of fuel rod cladding in the presence of air)
MPE: Maximum Probable Earthquake
MRBT: Multi-Rod Burst Test (ORNL experimental facility – study of fuel rod behavior)
NORS: NOkia Research Simulator
NPPs: Nuclear Power Plants
NUREG: Nuclear Regulatory Report, U.S.NRC
NUSMOR: NUgenia Small Modular Reactor with passive safety features, European
research project
NRU: National Research Universal (Canadian research reactor)
NSRR: Nuclear Safety Research Reactor, Japan
NYX: name of an experimental device on the GALAXIE platform, used for research into
controlling ﬁre risks in nuclear facilities
OCC: Outage Control Centers
ODOBA: Observatoire de durabilité des ouvrages en béton armé (Monitoring Center for
the Durability of Reinforced Concrete Structures, a research project to study the ageingof structures and the pathologies affecting them)
OLHF: OECD Lower Head Failure (research program to study failure of the lower part of areactor vessel)
OMEGA: name of a series of analytical tests conducted as part of studies of two-phasethermal hydraulics
OSIRIS: name of a CEA research reactor
PACTEL: Parallel Channel Test Loop (Finnish experimental facility for studying thethermohydraulic behavior of a nuclear reactor in an accident situation)
PANDA: name of a PSI experimental facility used for research into hydrogen risk

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Current state of research on pressurized water reactor safety

PARIS: name of an experimental program conducted as part of the study of radioactive
iodine behavior in a reactor core melt accident
PASSAM: Passive and Active Systems on Severe Accident source term Mitigation (multipartner research project on passive and active systems to mitigate releases in a reactor
core melt accident)
PATRICIA: name of a research program on reactivity accidents
PBF: Power Burst Facility, INL, USA
PCCV: Prestressed Concrete Containment Vessel (SNL experimental model – assessmentof leaks of air and steam through cracks in conditions representative of a reactor
containment)
PCMI: Pellet-Cladding Mechanical Interaction
PEARL: name of an IRSN experimental facility for testing debris bed reﬂooding
PERFECT: name of a European research program – corrosion of stainless steel under stressand under irradiation
PERFORM60: Prediction of the Effects of Radiation FOR Pressure Vessel and in-core
Materials using multi-scale Modelling – 60 years foreseen plant lifetime (irradiation
European research and development project on metallic components subject toirradiation)
PERFROI: Étude de la perte de refroidissement (Loss-of-Coolant Study, an experimental
project aimed at ﬁlling the gaps in knowledge of reactor core cooling in a loss-of-coolant
accident)
PERICLES: name of a series of analytical tests conducted as part of a study of two-phasethermal hydraulics
PGA: Peak Ground Acceleration
PHEBUS: name of a CEA experimental reactor
Phebus-CSD: international research program to study severe fuel degradation, using tests
carried out in the PHEBUS reactor
Phebus-FP: international research program to study the behavior of ﬁssion products,
using tests carried out in the PHEBUS reactor
PHENIX: name of a CEA prototype nuclear power reactor (and experimental reactor), a
fast neutron reactor using liquid sodium as a coolant
PICSEL: Propagation de l’incendie de combustibles solides dans un environnementlaboratoires et usines (Propagation of Solid Fuel Fires in Laboratories and Factories,
research programs into ﬁres in electrical cabinets, looking at safety in fuel reprocessing
plants)
PIA: plan d’investissement d’avenir (Investment in the Future Program)
PIV: Particle Image Velocimetry
PKL: Primärkreislauf (reactor coolant system, German research projects and large-scale
test facility for studying the thermohydraulic behavior of a nuclear reactor in an accident
situation)
PLUTON: name of an experimental facility on the GALAXIE platform used for research
into controlling ﬁre risks in nuclear facilities

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Page 21XXIPRELUDE: Préliminaire sur le renoyage expérimental d’un lit de débris (Preliminary Studyon the Experimental Reﬂooding of a Debris Bed, an IRSN experimental facility for
conducting feasibility studies and qualiﬁcation studies of the instrumentation used by the
PEARL program on reﬂooding debris beds)
P2REMICS: name of a simulation code for studying hydrogen risk
PREMIX: name of a KIT experimental facility for studying steam explosion in a reactor
PRENOLIN: Amélioration de la prédiction des effets non linéaires induits par les mouvements sismiques forts (Better Prediction of Non-Linear Effects Induced by Strong Seismic
Motion, benchmarking process)
PRISME: Propagation d’un incendie pour des scénarios multi locaux élémentaires (Spreadof a Fire for Multi-Room Elementary Scenarios, international research program)
PROGRES: Program expérimental analytique sur le renoyage de lits de débris (Analytical
Experimental Program on Debris Bed Reﬂooding)
PROMETRA: Propriétés mécaniques en transitoire (Mechanical Properties in a Transient,research program to study the mechanical behavior of fuel cladding in a reactivity
accident)
PSB-VVER: Russian experimental facility used to study the thermohydraulic behavior of a
nuclear reactor in an accident situation
PSA: Probabilistic Safety Assessments
PTR: système de traitement et refroidissement de l’eau des piscines (FPCPS water tank)
PWR: Pressurized Water Reactor
QUENCH: name of a KIT experimental facility used for integral core reﬂooding tests
RADIOSS: name of a fast dynamic code for analysing structuresRASPLAV: name of a test program run as part of research into reactor core melt accidents
RB: Reactor Building
RCC-G: règles de conception et de construction des ouvrages de génie civil (Rules on the
Design and Construction of Civil Engineering Structures)
RCC-M: règles de conception et de construction des matériels métalliques (Rules on the
Design and Construction of Metallic Equipment)
RCCV: Reinforced Concrete Containment Vessel (SNL experimental model – assessmentof leaks of air and steam through cracks in conditions representative of a reactor
containment)
RCS: Reactor Coolant system
REBEKA: name of a KfK experimental facility (Germany) – study of fuel rod behavior in a
LOCA
RECI: Recombineur et iode (Recombiner and Iodine, tests to quantify experimentally theconversion rate of metal iodides into iodine on the basis of temperature)
REKO: name of a KfJ experimental facility – study of hydrogen risk
REPASS: Reliability Evaluation of Passive Safety Systems (European research project)RESOH: Recherche en sûreté, organisation et hommes (Research on Safety, Organizationand Humans, a chair devoted to safety management in hazardous industries, particularly
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Current state of research on pressurized water reactor safety

RG: Regulatory guide, USA
RHRS: Residual Heat Removal System
RIA: Reactivity Injection Accident
RJH: réacteur Jules Horowitz (Jules Horowitz Reactor, France)
ROSA: Rig Of Safety Assessment (research projects to study the thermohydraulicbehavior of a nuclear reactor in an accident situation)
ROSCO: name of a series of analytical tests to study two-phase thermal hydraulics
RSE-M: règles de surveillance en exploitation des matériels mécaniques (Rules for the
Monitoring of Mechanical Equipment in Operation)
RSNR: Recherche en matière de sûreté nucléaire et de radioprotection (Research Projects
in the Field of Nuclear Safety and Radiation Protection)
RT: Release of Transuranics
RTF: Radioiodine Test Facility (AECL experimental facility – study of radioactive iodine
behavior in a reactor containment during a core melt accident)
RUT: name of a Russian experimental facility – study of hydrogen risk
RWST: Refueling Water Storage Tank
SAFEST: Severe Accident Facilities for European Safety Targets
SARNET: Severe Accident Research NETwork of excellence (international network of
excellence on core melt accidents)
SATURNE: name of an experimental facility on the GALAXIE platform used for research
into controlling ﬁre risks in nuclear facilities
SCANAIR: Système de codes pour l’analyse d’accidents d’insertion de réactivité (System of
Simulation Software for Analysing Reactivity Injection Accidents)
SCARABEE: name of a CEA research reactor (used for studying accident situations in
FNRs)
SEFLEX: (Fuel Rod) Simulator Effects in Flooding Experiments (experimental program to
study fuel rod behavior in a LOCA)SERENA: Steam Explosion REsolution for Nuclear Applications (research program to
study and develop simulation tools for steam explosion)
SETH: SESAR THermalhydraulics (international research program on hydrogen risk)
SFD: Severe Fuel Damage (experimental program on core temperature rise and core melt
inside a reactor vessel)
SFP: Spent Fuel Pool
SFR: Sodium-cooled Fast neutron Reactors
SG: Steam Generator
SIGMA: Seismic Ground Motion Assessment (research and development program)
SILOE: name of a CEA research reactor
SIMIBE: name of tests conducted on leaks through a crack in a reactor containment
SINAPS@: Séisme et installations nucléaires, améliorer et pérenniser la sûreté (Earthquakes and Nuclear Facilities, Improving and Protecting Safety, a research project onseismic hazards and the vulnerability of nuclear components and structures)Anglais_Book

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SIROCCO: name of an oven on the GALAXIE platform, used for characterizing malfunctions of electrical equipment subjected to thermal stress
SIS: Safety Injection System
SMA: Seismic Margins Assessment
SMART: name of an instrumented CEA model used for assessing seismic motion transferred to equipment
SMD: SUPER MOBY DICK (name of a series of analytical tests conducted as part of thestudy of two-phase thermal hydraulics)
SME: Seismic Margin Earthquake
SMR: Small Modular Reactors
SNETP: Sustainable Nuclear Energy Technology Platform
SOA: State-Oriented Approach
SOAR: State-of-the-Art-Report, OECD/NEA
SOFIA: Simulateur d’observation du fonctionnement incidentel et accidentel (Observation
Simulator for Incidental and Accidental Operation, an engineering simulator to improve
the nuclear safety of pressurized water reactors)
SPARK: software for simulating the operation of passive autocatalytic recombiners
SPERT: Special Power Excursion Reactor Tests (American research reactor used forstudying reactivity accidents)
SPLASH: name of a series of tests on thermal fatigue and their test equipment
SPOT: name of a Russian experimental facility – study of hydrogen risk
SSI: Soil-Structure Interaction
SSG: Speciﬁc Safety Guide, IAEA
SSWICS: Small Scale Water Ingression and Crust Strength (ANL experimental platform
for research into reactor vessel failure and basemat erosion due to corium)
STARMANIA: name of an IRSN experimental facility for conducting research into
controlling ﬁre risk in nuclear facilities
STEM: Source Term Evaluation and Mitigation (research program on the behavior ofradioactive products likely to be released into the environment during a core melt
accident)
STORM: name of an Ispra Joint Research Centre experimental facility used for studying
the transfer of radioactive products into reactor systems in a core melt accident
STL: sonde tournante longue (Long Rotating Probe)
STT: sonde tournante transversale (Transverse Rotating Probe)
STYX: name of an experimental facility on the GALAXIE platform used for research into
controlling ﬁre risks in nuclear facilities
SUPERCANON: name of a series of analytical tests conducted as part of the study of twophase thermal hydraulics
SUPERPHENIX: name of an EDF nuclear power reactor, a fast neutron reactor using liquidsodium as a coolant

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Current state of research on pressurized water reactor safety

SUW: Scale-Urania-Water (Winfrith experimental facility – study of steam explosion in areactor)SYLVIA: Système de logiciels de simulation pour l’étude de la ventilation, de l’incendie etde l’aérocontamination (Software System for Analysing the Ventilation of a Fire andAirborne Contamination)
TAGCIS: Trempe en APRP de gaine de combustible à irradiation simulée (Quench during a
LOCA of fuel rod cladding that has undergone simulated irradiation, a research program)
TAGCIR: Trempe en APRP de gaine de combustible irradiée (Quench during a LOCA ofirradiated fuel rod cladding, a research program)
TAMARIS: Tables et moyens d’analyses des risques sismiques (Tables and Means ofAnalysing Seismic Risks, a CEA experimental platform)
TANDEM: Tsunami in the Atlantic and the English ChaNnel: Deﬁnition of the Effects
through numerical Modeling
THAI: Thermal-hydraulics, Hydrogen, Aerosols and Iodine (experimental installation of
Becker Technologies)
THETIS: name of an experimental program to study fuel rod behavior
THINS: Thermal-hydraulics of Innovative Nuclear Systems (European research project)
TIB: Total Instantaneous Blockage
TMI: Three Mile Island, USA
TMI-2: Reactor 2 at the Three Miles Island NPP, USA
TOFD: Time Of Flight Diffraction
TONUS: name of a simulation code for assessing hydrogen risk in core melt accident
conditions
TORPEDO: name of a KIT experimental facility – study of hydrogen risk
TOSQAN: name of an IRSN experimental facility used for simulating the thermohydraulicconditions in a nuclear reactor containment during a core melt accident
TRANSAT: name of a program of analytical tests conducted as part of research into thetransfer of radioactive products into reactor systems during a core melt accident
TREPAM: name of an experimental program to study corium-concrete interaction
TRIGA: Training, Research, Isotopes, General Atomics (pool-type research reactor)
TROI: Test for Real cOrium Interaction with water (KAERI experimental facility used to
study the interaction between fuel and coolant)
TSO: Technical Safety Organization
TUBA: name of a program of analytical tests to study the transfer of radioactive products
into reactor systems during a core melt accident
ULPU: An IVR-related full-scale boiling heat transfer facility at University of California,
Santa Barbara (a UCSB experimental facility for research on the possibility of retaining
the corium in the reactor vessel during a reactor core melt accident)
UMR: unité mixte de recherche (Joint Research Unit)
UNGG: uranium naturel-graphite-gaz (Natural Uranium-Graphite-Gas, gas-cooled
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XXV

VD: visite décennale (Ten-Yearly Outage Program)
VEGA: Veriﬁcation Experiments of radionuclides Gas/Aerosol release (tests to study theemission of ﬁssion products by fuel when its temperature rises)
VERCORS: CEA facility in Grenoble, used to study the release of ﬁssion products by
irradiated fuel subjected to a temperature increase
VERCORS: Vériﬁcation réaliste du conﬁnement des réacteurs (Realistic Veriﬁcation of theContainment of Reactors, an EDF model and a series of tests to assess leaks through acontainment in an accident situation)
VERDON: name of the CEA experimental facility that replaced VERCORS in Grenoble
VI: Vertical Induction (tests to study the emission of ﬁssion products by fuel when its
temperature rises)
VIKTORIA: name of an experimental loop for conducting research into cooling water
recirculation in accident conditions
VITRA: name of an experimental loop for conducting research into cooling water
recirculation in accident conditions
VULCANO: Versatile UO2 Laboratory for COrium ANalysis and Observation (CEA
experimental facility for research into reactor vessel failure and basemat erosion due
to corium in a reactor core melt accident)
VVER: Vodo-Vodianoï Energuetitcheski Reaktor (Russian water-cooled, water moderated
nuclear power reactor)
VWU: VULCANO Water-Uranium
WUMT: tests to study steam explosion in a reactor
XRD: X-Ray Diffraction

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Foreword
This summary on the current state of research in the ﬁeld of pressurized water reactor
safety is a collective effort by a team of authors from IRSN, the French national Institutefor Radiological Protection and Nuclear Safety. Jean Couturier and Michel Schwarz(retired from IRSN) are the main authors.
Jean Couturier was responsible for coordinating the project.
The following authors contributed to the work:– Chapter 5: Sébastien Roubaud, Caroline Lavarenne, Jean-Marie Mattei,– Chapter 7: Laurence Rigollet,– Chapter 8: Oona Scotti, Christophe Clément, Maria Lancieri, Céline Gelis,– Chapter 9: Didier Jacquemain, Ahmed Bentaib,– Chapter 10: Georges Nahas, François Tarallo, Gilbert Guilhem, Gérard Cattiaux,Benoît Durville, Christian Mun, Christine Delaval, Thierry Sollier, Jean-Marc
Stelmaszyk,– Chapter 11: François Jeffroy, Nicolas Dechy, Olivier Chanton, Daniel Tasset,
Isabelle Pichancourt.
The following experts were involved in proofreading in their specialist ﬁelds: François
Barré, Gianni Bruna, Jean-Michel Évrard, Richard Gonzalez, Olivier Loiseau, Daniel
Quéniart, Didier Vola.
Georges Goué and Odile Lefèvre prepared the work for publication.

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Contents
Preface ................................................................................................................................................

III

The authors ......................................................................................................................................

V

List of abbreviations.....................................................................................................................

VII

Foreword ............................................................................................................................................

XXVII

Chapter 1
Introduction
Chapter 2
Some of the Research Facilities Favored by IRSN
in the Field of Nuclear Power Reactor Safety
2.1

The CABRI reactor .............................................................................................................

9

2.2

The PHEBUS reactor .........................................................................................................

11

2.3

The GALAXIE experimental facility ............................................................................

11

2.4

Other facilities.....................................................................................................................

12

Chapter 3
Research on Loss-of-coolant Accidents
3.1

Two-phase thermal-hydraulics ....................................................................................

17

3.2

Fuel rod behavior ...............................................................................................................

20

Chapter 4
Research on Reactivity-initiated Accidents

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Current state of research on pressurized water reactor safety

Chapter 5
Research on Recirculation Cooling under
Accident Conditions
5.1

Operating experience feedback and research topics .........................................

36

5.2

Past research programs and lessons learned ........................................................

40

5.3

Ongoing research programs ..........................................................................................

42

5.4

Simulation .............................................................................................................................

44

Chapter 6
Research on Spent Fuel Pool Uncovery Accidents
Chapter 7
Research on Fires
7.1

Fire risks at nuclear installations ................................................................................

53

7.2

Organizations involved in research on ﬁre ............................................................

56

7.3

Research facilities, simulation tools ..........................................................................

57

7.4

The main research programs and their contributions ......................................

60

Chapter 8
Research on Hazards associated to Natural Events
8.1

Earthquakes ..........................................................................................................................

70

8.2

External ﬂooding ................................................................................................................

75

Chapter 9
Research on Core Melt Accidents
9.1

Heating of the core and core melt inside the vessel .......................................

79

9.2

Reactor vessel melt-through and basemat erosion by themolten corium .....................................................................................................................

89

Dynamic loading of the containment by a sudden increase ininternal pressure .................................................................................................................

99

9.3.1 Steam explosions ................................................................................................

99

9.3.2 Hydrogen-related risks .....................................................................................

102

9.3

9.4

9.3.3 Direct containment heating ...........................................................................

111

Radioactive releases ..........................................................................................................

113

9.4.1 Emission of radioactive products by fuel ................................................

113

9.4.2 Transporting of radioactive products into the reactor’s
circuits ......................................................................................................................

115

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Contents

XXXI

9.4.3 Radioactive product behavior in the containment .............................

117

9.4.4 Contribution of the Phebus-FP program on the various processes
involved in radioactive releases ...................................................................

120

9.4.5 Aspects to be explored ....................................................................................

121

Chapter 10
Research on the Behavior of Components
Important to the Safety of NPPs and More
Particularly the Aging of Such Components
10.1 R&D on metallic components .....................................................................................

127

10.1.1 Research on thermal fatigue .........................................................................

130

10.1.2 R&D on non-destructive testing .................................................................

134

10.1.3 Studies and research on the seismic behavior of overheadcranes .......................................................................................................................

144

10.2 R&D on civil engineering structures .........................................................................

145

10.2.1 Development of constitutive equations for civil engineering
structures................................................................................................................

147

10.2.2 R&D on the behavior of civil engineering structures underseismic loads .........................................................................................................

148

10.2.3 R&D on the behavior of civil engineering structures incases of collision .................................................................................................

153

10.2.4 R&D on the behavior of containments during coremelt accidents ......................................................................................................

154

10.2.5 R&D on containment aging ...........................................................................

161

10.3 Research on polymers .....................................................................................................

165

Chapter 11
Research in the Field of Human andOrganizational Factors and, More Broadly,in Human and Social Sciences
11.1 From the consideration of human factors in safety to studies inhuman and social sciences ............................................................................................

169

11.2 Studies and research on reactor control room design .....................................

172

11.2.1 The post-TMI period ..........................................................................................

172

11.2.2 The N4 series: the ﬁrst computerized control room .........................

174

11.2.3 The EPR project: the move toward increased automation .............

178

11.2.4 Contributions and outlook .............................................................................

181

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11.3 Studies and research on the organization and management of safety at
EDF's NPPs ............................................................................................................................

181

11.3.1 Organization of maintenance activities ....................................................

183

11.3.2 Contractor management .................................................................................

187

11.4 Research on human and social sciences: the AGORAS project ...................

190

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Chapter 1
Introduction

Preventing accidents liable to affect a nuclear facility and, in keeping with a defensein-depth approach, mitigating their impact, call for robust methods based on the state of
the art, operating experience feedback and regulations, and the demonstration by thelicensee that the provisions and measures it implements are effective. This demonstration and its assessment, carried out in France by IRSN as part of its assessment activities1,
draw on a body of scientiﬁc knowledge that has grown considerably after nearly forty
years of research and the related development work, particularly in the ﬁeld of simulation
codes.
This publication describes the current state of this research and development
conducted by IRSN, alone or in collaboration with other organizations, the lessons
learned and how they have contributed to safety at French nuclear power reactors
and, lastly, the work that is still underway or planned to consolidate and push backthe boundaries of knowledge. This summary does not claim to be exhaustive, and someR&D work is only quoted or mentioned brieﬂy at the end of this publication. Also, someof the research discussed here, such as the work on seismic, ﬂooding and ﬁre hazards, is of
beneﬁt to all nuclear facilities, not only to nuclear power reactors.
Nuclear reactors are large, complex machines. Many complex physical phenomenainteract as an accident unfolds. Studying these phenomena requires knowledge in manyﬁelds of physics, including neutronics, heat transfers, ﬂuid mechanics, structural mechanics, metallurgy, radiation chemistry, etc. That being the case, it is clear that the results ofexperimental research, which is generally carried out at a reduced scale and is simpliﬁed
to varying degrees, cannot be used directly for demonstration purposes. Physical models
1.

Deﬁned in Decrees No. 2002-254 of February 22, 2002 and No. 2016-283 of March 10, 2016.

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must therefore be developed, based on academic knowledge or targeted experiments,
also called separate effects experiments (see the Focus at the end of this introduction).
These models must then be integrated in computer codes used to simulate all thephenomena considered essential for a clear understanding of how the accident unfolds atreactor scale.
More complex experiments simulating real conditions as closely as possible, but
carried out at an intermediate scale, also known as integral experiments (see the Focus at
the end of this introduction) are conducted to assess the relevance of the calculations
performed using these simulation codes. In some cases, the experiments highlight areas
where further research is required to supplement and deepen knowledge. Accidents suchas those that occurred in reactor 2 of the Three Mile Island nuclear power plant in the
United States on March 28, 1979, and in the reactors of the Fukushima Daiichi nuclear
power plant in Japan on March 11, 2011, have also provided unique data. Despite gaps due
to lack of instrumentation, this data nonetheless helps to assess the performance of
simulation codes in terms of their ability to predict how real accidents unfold. The
accident that occurred in reactor 4 of the Chernobyl nuclear power plant on April 26,
1986 raised questions as to the suitability of models – and therefore the ability ofcomputer codes – for predicting radioactive release from western nuclear power plants inthe event of core melt, as well as the validity of the safety criteria adopted until then forreactivity-initiated accidents.
In spite of the considerable progress made over the past forty years in understanding
the phenomena involved, more in-depth knowledge is still required in some areas relating
to nuclear safety because of changes in reactor operating conditions, the use ofinnovative technology, and the search for more accurate scientiﬁc approaches, all inan environment of increasingly ﬁerce industrial competition. Assumptions thought to be
prudent have often been made at the design stage, sometimes making it necessary tocheck and determine how conservative they are and what margins they really allow.
Furthermore, the possible occurrence of core melt was taken into account right fromthe design stage of new-generation reactors, like the European Pressurized water Reactor(EPR), which was not the case for reactors commissioned during the 20th century, even if
they did subsequently beneﬁt from improvements aimed at taking this risk into
consideration. Consequently, extending the operating life of these reactors beyond forty
years cannot be contemplated without a signiﬁcant improvement in safety, in terms of
preventing core melt accidents and mitigating their impact. This heightens the need forresearch in this area. The complementary safety evaluation (ECS) carried out in the wakeof the accident at the Fukushima Daiichi nuclear power plant underscored the need forfurther research on external hazards that could potentially lead to core melt accidents,
the risk and consequences of hydrogen explosions, and radioactive release ﬁltering.
In France, the main organizations involved in nuclear safety research are IRSN, theFrench Alternative Energies and Atomic Energy Commission (CEA), EDF – French power
utility – and, to a lesser degree, AREVA (Framatome before 2006), the designer. Academic
research organizations, such as universities, engineering schools and CNRS2, the2.

French National Center for Scientiﬁc Research.

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3

French National Center for Scientiﬁc Research, are also involved in promoting a fuller
understanding of basic phenomena.
While it is for the licensee to provide the French Nuclear Safety Authority (ASN) withthe technical bases used in its safety demonstration, including the related research
results, it is essential for IRSN to also carry out its own research. It is this work that allows
it to develop and maintain its expertise in complex scientiﬁc themes that must be
perfectly mastered in order to perform relevant, independent assessments of thelicensee's safety demonstrations, and make a positive contribution to advancing safety.
IRSN conducts experimental research programs in its laboratories, or has them carriedout at partner laboratories (CEA's in particular), beneﬁts from partners' work and,
whenever possible, develops its own models and simulation codes. It validates thesimulation codes (or has them validated) that are vital to the research work and studies
carried out in support of its assessment activities.
The knowledge thus acquired and the developed and validated simulation codes alsoenable IRSN to provide the public authorities with scientiﬁc and technical support in theevent of emergency. It was thus able to provide the authorities, general public and media
with precise updates on events at the Fukushima Daiichi nuclear power plant following
the accident.
Some research allows IRSN to clarify or consolidate its technical examination of
topics that could help to signiﬁcantly enhance safety, in nondestructive testing of reactor
systems and components for example.
One speciﬁc feature of nuclear safety research is the period of time required before
results can be used for assessment purposes. In general, this is relatively long, about ten
years, for example, for research involving the use of a nuclear reactor, or requiring thedesign of a technologically innovative experimental setup. This is due to the complexityof the experimental facilities to be designed and implemented, particularly those thatmust be installed in nuclear research reactors, and to the time required for post-mortem(or post-test) examinations at specialized laboratories, especially when testing involves
the use of radioactive material. It is therefore important to plan ahead to ensure that therequired knowledge and simulation codes are available when needed. For licensees and
IRSN experts alike, such forward planning is crucial if they are to be ready for major
milestones such as reactor safety reviews during periodic outage programs, or examinations of safety conditions for extending the operating life of reactors beyond forty
years (EDF's DDF project on "operating life").
With regard to its cost, a great deal of this research – especially experimental
programs – is carried out under cooperation agreements between industry and IRSN and
its foreign counterparts, with each partner being free to use the results of the research for
its own purposes.
IRSN is a partner in more than a dozen projects selected by the French National
Research Agency (ANR), notably in the framework of the call for projects "Nuclear Safety
and Radiation Protection Research" launched in the wake of the Fukushima Daiichi
accident. But the cooperative research is not limited to France. It will be seen later thatthe main countries that designed and built nuclear power reactors – the United States,Anglais_Book4

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Canada, Japan, Germany, the United Kingdom, Switzerland and Russia – carried out
research programs on nuclear safety and continue to do so. IRSN receives funding frominternational partners under various bilateral and multilateral agreements for most of its
research programs and, in exchange, has access to the results of foreign programs towhich it makes a ﬁnancial contribution.
The Organization for Economic Co-operation and Development/Nuclear Energy
Agency3 (OECD/NEA) plays a very important role by assisting "its member countries
in maintaining and further developing, through international co-operation, the scientiﬁc,technological and legal bases required for a safe, environmentally friendly and economicaluse of nuclear energy for peaceful purposes". It works to achieve a consensus amongexperts on the current state of knowledge (e.g. State-Of-the-Art Reports [SOAR]), gaps
to be ﬁlled, and future research priorities. It organizes international benchmark exercises
to compare various simulation codes with experimental results (International Standard
Problems [ISP]). These exercises always provide a great deal of information. It also helps
to structure international research projects – referred to as Joint Projects, proposed by
one of its members – by bringing together other partner members to obtain the required
funding, and thus advance knowledge in areas where it considers this is necessary. With
regard to OECD/NEA activities, IRSN is particularly involved in the work of theCommittee on the Safety of Nuclear Installations (CSNI), whose mission is to assist
member countries in maintaining and further developing the scientiﬁc and technicalknowledge base required to assess the safety of nuclear reactors and fuel cycle facilities
and, more directly in its various working groups:– Working Group on Risk Assessment (WGRISK),– Working Group on Fuel Safety (WGFS),– Working Group on Analysis and Management of Accidents (WGAMA),– Working Group on Integrity and Ageing of Components and Structures (WGIAGE),– Working Group on Human and Organizational Factors (WGHOF),– Working Group on Fuel Cycle Safety (WGFCS),– Working Group on External Events (WGEV),– Working Group on Electrical Power Systems (WGELE).
The European Commission also makes a substantial contribution to the funding ofinternational research and development projects relating to nuclear safety. Calls for
projects on what are considered as priority themes are issued in the Euratom part of themulti-year research and development Framework Programs (FP), which began in 1984.
These are carried out within a cooperative framework, and generally involve industrial
partners, nuclear reactor operators, assessment organizations, and research laboratories.

3.

As of January 1, 2015, NEA membership included 31 countries in Europe, North America and the AsiaPaciﬁc region.

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5

It should also be noted that IRSN is involved in European bodies whose task is to steer
prestandardization research in various ﬁelds (metal structural mechanics, civil engineering, instrumentation and control software).
The European Sustainable Nuclear Energy Technology Platform (SNETP), which brings
together representatives from the nuclear industry, research, safety organizations, associations, and NGOs4, was set up in 2007 to formulate a collective vision of the contributionthat nuclear ﬁssion could make towards the transition to a low-carbon energy mix by 2050.
Within this context, the SNETP has published various documents, including a "Strategic
Research Agenda" (January 2009) and a "Strategic Research and Innovation Agenda"(February 2013), which addresses a number of safety issues relating to Generation II and III
water reactors, as well as a special document entitled "Identiﬁcation of Research Areas in
Response to the Fukushima Accident" (January 2013). These documents are used to help
the European Commission to deﬁne its research and development Framework Programs.
Note that in 2012, SNETP stakeholders (except for the NGOs) working on Generation II and
III reactors joined with other existing organizations or networks (including SARNET (Severe
Accident Research NETwork of excellence) for core melt accidents, NULIFE, and ETSON), toform NUGENIA (NUclear GENeration II & III Association), an association under Belgian law.
Two publications of this association are worthy of mention: "NUGENIA Roadmap –Challenges & Priorities" (October 2013) and "NUGENIA Global Vision" (April 2015).
IRSN is closely involved in both SNETP and NUGENIA.
There now follows an overview of the main safety-related research programs, inparticular those carried out by IPSN then IRSN, as well as an outline of the lessons
learned from results. The main foreign research programs in the same ﬁeld are alsopresented, although by no means exhaustively.
The ﬁrst chapter of this publication is devoted to research on the design-basis
accident taken into account for engineered safety equipment used in pressurized water
reactors (PWR), namely a main pipe break in the reactor coolant system (loss-of-coolant
accident [LOCA]). The ﬁrst research programs on nuclear safety, which started in the
1970s, focused on this area with the aim of improving knowledge and developing
computer tools for studying this accident.
For each of the research and development themes discussed in this publication –positioned in the "deterministic" safety demonstration in Figure 1.1 on the next page – thesafety objectives of the safety demonstration are indicated, together with the issues or
difﬁculties that the demonstration might have raised, or still raises today given thecurrent state of knowledge.
First, a brief description is provided of some of the research instruments favored by IRSN in
the ﬁeld of nuclear power reactor safety and which will be mentioned later in the publication.
As a general rule, it was decided to refer only to a few documents regarded as themost signiﬁcant, or as having the advantage of providing an overview of the knowledge
acquired in some of the areas of interest (e.g. some OECD State-Of-the-Art Reports), andwhich generally include references to many scientiﬁc documents.
4.

Non-Governmental Organization.

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Current state of research on pressurized water reactor safety► Research carried out as part of strategic guidelines and programs
The research themes and topics referred to in this publication concern strategic
guidelines and scientiﬁc programs aimed at ensuring that ﬁrst IPSN, then IRSN, has hadaccess to the best available knowledge for their assessments and other activities. Thesescientiﬁc guidelines and programs have, of course, been determined by issues raised
during safety analysis, by changes in reactor design, lesson learned from operating
experience feedback (from incidents in particular), the accidents that have occurred (TMI,
Chernobyl and, more recently Fukushima Daiichi), as well as research results. IRSN's
scientiﬁc strategy can now be consulted by the public. Pressurized water reactor safety is
directly addressed in reference [1], published in October 2015, which considers it in termsof seven "priority scientiﬁc issues", including one concerning cross-cutting human andorganizational aspects, corresponding to the themes and topics discussed in the present
publication.

#FOCUS...........................................................................................................................................................................................................................

Separate-effects or Analytical Experiments versus
Integral Experiments
Separate-effects experiments are very well-instrumented experiments used forthe detailed study of a particular physical phenomenon to precisely determine thephysical laws that govern it. The "experimental designs" allow systematic variationof the important physical parameters. For example, the oxidation kinetics of fuel rod
cladding is determined by placing a cladding sample in a test section with steam
ﬂowing through it inside a controlled temperature oven. The sample is hung fromscales and the required information is obtained by measuring the increase in its
weight at different temperatures over time.
Integral experiments, on the other hand, are more concerned with reproducing
the complexity of the phenomena under study, even if this often means that themeasurement of the physical variables is broader and less precise. For example, thedamage to a fuel assembly following loss of cooling was studied by irradiating abundle of 20 fuel rods in the PHEBUS reactor, in a heat-insulated test section withsteam ﬂowing through it. The physical phenomena involved are: nuclear ﬁssion
reactions that cause the fuel to overheat; cladding oxidation beyond a certain
temperature, which has the result of consuming steam, releasing hydrogen, andgenerating heat; and heat transfers in the bundle through conduction, convection andradiation. The phenomena can only be studied by measuring the reactor power, theﬂow of steam passing through the test section, local temperatures in the fuel and thetest section, as well as total hydrogen production.
......................................................................................................................................................................................................................................................................................

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Figure 1.1 Research and development themes discussed (or simply mentioned) in this publication. © Georges Goué/IRSN.

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Reference[1] La stratégie scientiﬁque de l’IRSN, October 2015: http://www.irsn.fr/FR/IRSN/presentation/Documents/IRSN_Strategie-scientiﬁque_2015.pdf

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Chapter 2
Some of the Research
Facilities Favored by IRSN
in the Field of Nuclear Power
Reactor Safety

2.1. The CABRI reactor
The CABRI safety test reactor, together with SCARABEE, its former twin reactor now
disused, is the basic nuclear installation (INB 24) at the Cadarache research center. CABRI
is operated by CEA for IRSN.
CABRI was commissioned in 1964. It is a pool-type reactor consisting of a 25 MW,
water-cooled driver core5 that uses slightly enriched uranium oxide (UO2) fuel since the1970s.
It is used to study the impact on fuel of reactivity-initiated accidents that may occur
in various types of reactors. The experiments consist in very rapidly increases of reactor
power above its normal operating level. The power peaks, which last only a few
milliseconds, may reach 20 GW and are achieved by the more or less rapid and sequential
emptying of control rods ﬁlled with a neutron-absorbing gas, the helium-3 isotope. The

5.

The driver core produces neutrons for conducting experiments on the test fuel, which is placed in aloop that contains its own coolant, and crosses the core along its central axis.

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65 cm width and 80 cm high reactor core is made up of 40 fuel rod assemblies designed
to withstand the impact of these power peaks.
From 1978 to 2001, CABRI was used for research on reactivity-initiated accidents insodium-cooled fast neutron reactors (SFR), using single fuel pins (or rods) placed in a
sodium test loop. From 1993 to 2000, the sodium loop was also used for carrying out
tests to simulate the ﬁrst phase of a reactivity-initiated accident in pressurized water
reactors. These were known in France as REP-Na (i.e. PWR-sodium) tests. CABRI is now
equipped with a pressurized water loop for the purposes of a new research program
comprising ten tests on pressurized water reactors, under conditions that are more
representative of a PWR6. This program, called the CABRI International Program (CIP),
was launched in 2000 (Figure 2.1).
The CABRI reactor includes a device called a hodoscope, which consists of several
tens of ﬁssion and recoil proton ionization chambers positioned at the end of a collimator
passing through the core. The hodoscope is designed to detect and precisely measure thesolid or molten fuel movements in the test rods during the experiment.
SCARABEE, which is no longer used, shared its cooling systems with CABRI.
Unlike CABRI, it was not used to produce rapid power transients. Equipped with asodium loop with a larger diameter than that of the CABRI loop, it was used mainly
for the study of hypothetical accidents involving fuel assembly blockage andFigure 2.1 Diagram of the CABRI safety test reactor with its pressurized water loop.

6.

The pressure and temperature in the water loop are respectively 155 bar and 350° C.

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11

meltdown in high-power fast neutron reactors (SUPERPHENIX, “RNR 1 500”
project), from 1983 to 1989. The tests were performed on small assemblies
comprising up to 37 fuel pins.

2.2. The PHEBUS reactor
PHEBUS was an experimental nuclear reactor (INB 92) also located at the Cadarache
center and operated by CEA. Commissioned in 1978, it was intended for research on
accidents liable to affect pressurized water reactors. It has not been used since 2010.
PHEBUS was a pool-type reactor consisting of a 40 MW drive core, equipped with acooling tower that enabled it to operate at high power for several days, unlike CABRI,
whose operating time cannot exceed tens of minutes. It was used to provide neutron ﬂux
to heat the test fuel that was inserted into the pressurized test loop at the center of thecore and produce radioactive substances, whose behavior was studied during a subsequent phase of the experiment.
It was mainly used as the experimental tool for a Fission Product program (PhebusFP), which was aimed at studying what became of ﬁssion products from a pressurizedwater reactor core in the event of core melt. This program was launched by IPSN in 1988
under a partnership with EDF and the European Commission, and with a number of othercountries (United States, Canada, Japan, South Korea and Switzerland). It lasted from1988 to 2010. PHEBUS had previously been used for the Phebus-LOCA and Phebus-CSD
programs.
CEA plans to ﬁle an application with the French Nuclear Safety Authority (ASN) in2017 for the ﬁnal shutdown of PHEBUS.

2.3. The GALAXIE experimental facility
GALAXIE is an IRSN-operated experimental facility that is also located at the
Cadarache center. It groups together experimental facilities with various capabilities,
designed for conducting experimental research on ﬁre risk control in nuclear facilities (seedetails in Chapter 7). A brief description of the two main GALAXIE test facilities is given
below.

► DIVA (Fire, Ventilation and Airborne Contamination Device)
The DIVA facility (Figure 2.2) is used for performing tests on ﬁres that may occur atfuel cycles facilities as well as pressurized water reactors. The tests are performed in
conﬁgurations involving several ventilated rooms. DIVA consists of three 120 m3 rooms, a
150 m3 corridor, another room with a volume of 170 m3 on the second ﬂoor, and aventilation network. It is very well-instrumented, with up to 800 measuring channels. Its
reinforced concrete structure and its equipment have been designed to withstand a gas
pressure range of 100 hPa to + 520 hPa. Leakage between rooms and ventilation
conﬁgurations can be adjusted to study ﬁre in situations involving several conﬁned andventilated rooms.

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Current state of research on pressurized water reactor safety

Figure 2.2 The DIVA facility © Jean-Marie Huron/Signatures/IRSN.

► SATURNE
The SATURNE calorimeter cone (Figure 2.3) is a facility used to determine characteristic variables of a ﬁre, such as its heat release rate and mass ﬂow. The capacity of thehood and ventilation network (maximum capacity of 30,000 m3/h) is suitable forstudying ﬁres with heat release rates of up to 3 MW.

2.4. Other facilities
IRSN uses ﬁve other experimental facilities in the ﬁeld of pressurized water reactor
safety.

► The CALIST experimental facility (Saclay)
The CALIST (Characterization and Application of Large and Industrial Spray Transfer)experimental facility was designed to characterize the size and velocity of drops from one
or more spray nozzles and for the study of air entrainment by the sprays.

► The TOSQAN facility (Saclay)
The TOSQAN facility is used to simulate the thermal-hydraulic conditions inside thecontainment building of a nuclear reactor in the event of a core melt accident. This facility
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13

Figure 2.3 The SATURNE facility. © Olivier Seignette/Mikaël Lafontan/IRSN.

distribution in a reactor containment: wall condensation, exchanges induced by thecontainment sump or the containment spray system.

► The CHIP facility (Cadarache)
The CHIP7 facility is used for physical-chemical studies. It consists of ovens in which
chemical reagents are placed, and temperature-controlled axial proﬁle tubes throughwhich these reagents and a carrier gas are injected. It is used in the CHIP research
program, which focuses on iodine chemistry far from thermodynamic equilibrium (impact
of chemical kinetics) in the reactor coolant system of a water reactor during a core melt
accident.

► The EPICUR facility (Cadarache)
The EPICUR8 facility is a panoramic irradiator composed of cobalt sources (60Co). It
emits γ radiation characterized by an average dose rate of 10 kGy/h. It consists of anexperimental chamber used to study the effect of radiation emitted by the radioactive
substances found in the containment of a pressurized water reactor during a core melt
accident. It is used for the program of the same name, which studies iodine chemistry in areactor containment in an accident situation.

7.
8.

Iodine Chemistry in the Reactor Coolant System.
Physical and Chemical Studies of Contained Iodine under Radiation.

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Current state of research on pressurized water reactor safety► The IRMA Irradiator (Saclay)
IRMA9 is a panoramic irradiator used to irradiate materials. It consists of cobalt
sources (60Co) and emits γ radiation characterized by an average dose rate of 20 kGy/h.
Various scientiﬁc and technical experiments are carried out at the facility. By nature, they
concern the study of radiation/material interaction mechanisms and, more especially, theresponse of materials and components to gamma radiation and any related degradation.
IRMA is also used in the study and design of biological shielding and protection for various
types of nuclear facilities. Its research potential (high γ ﬂux) is also compatible withaccident studies.

9.

IRradiation of MAterials.

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Chapter 3
Research on Loss-of-coolant
Accidents

Loss-of-coolant accidents (or LOCAs) are among the design-basis operating conditions to be considered for pressurized water reactors, in line with the deterministic safety
approach. They are hypothetical postulated accidents (Plant Condition Category 410), in
which the initiating event is a break in the reactor coolant pressure boundary. A break ofthis type leads to a pressure drop, of varying suddenness, and a loss of water. This
inexorably causes the fuel rods to heat up, owing to the residual heat given off by thefuel, and in spite of the automatic control rod drop stopping any further nuclear ﬁssion
reaction. This heating phenomenon must remain limited to ensure that fuel damage does
not adversely affect reactor core cooling and does not lead to meltdown. Loss-of-coolant
accidents are design-basis accidents considered, in particular, for the safety injection
system (SIS), which is designed to inject water at varying ﬂow rates, some mechanicalcomponents in the reactor coolant system, and the reactor containment building.
In the 1970s, safety criteria were deﬁned for fuel rod cladding (the ﬁrst barrier toensure containment), based on the knowledge available at that time. They appear inUnited States regulations, in particular in 10 CFR (Code of Federal Regulations) 50.46 andits Appendix K, issued in 1974, and were adopted in France for the construction of the ﬁrst
nuclear power reactors under license to Westinghouse. These regulations were the resultof long years of discussions between the Atomic Energy Commission (AEC), theforerunner to today's United States Nuclear Regulatory Commission (U.S.NRC), andthe licensees of American nuclear power plants. Since 1974, however, reactor operating
10.

See "Elements of Nuclear Safety", J. Libmann, EDP Sciences, 1996, Chapter 3.

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Current state of research on pressurized water reactor safety

conditions and fuels have changed (increased burnup, new fuel rod cladding materials,etc.), and this has led to the emergence of various research and development programs,
described in the following pages.
Loss-of-coolant accidents involve complex phenomena in the reactor in three areas:– thermal-hydraulics in the reactor coolant system,– mechanical behavior of structures inside the reactor vessel,– fuel rod thermomechanics.
The mechanical behavior of internal structures is an especially complex aspect ofLOCAs, owing to the dynamic and dissymmetrical nature of water decompression in
the reactor coolant system. The pressure drop, which ranges from 50 to 80 bar,
propagates as a decompression wave through the reactor coolant system at the speedof sound in water (about 1000 m/s), and reaches the vessel via a single nozzle. This
imposes signiﬁcant mechanical loading on the vessel internals and fuel assemblies thatmust keep their geometry to ensure that rod drop can be made to shut down thereactor, and that the core cooling function remains operational. Ever since pressurizedwater reactors have been designed in France, the double-ended guillotine break11 on areactor coolant pipe (known as a 2A break) has been systematically postulated to
study some of the consequences of a LOCA (core cooling capability, containment
integrity, radiological consequences), but not others (mechanical strength of vesselinternals and fuel assemblies, for example), for which limited guillotine breaks are
considered. In this respect, speciﬁc devices to limit pipe deﬂection in the event ofbreaks are ﬁtted on all pressurized water reactors currently in operation. In addition,
for the break postulated at the reactor coolant pump outlet, where there is no device
for restrain a pipe deﬂection, the break cross-section is limited by the rigidity of thecold leg.
Research on loss-of-coolant accidents has focused on two main areas:– the study of two-phase thermal-hydraulic phenomena encountered during
reactor coolant system draining, core reﬂooding, and fuel rod rewetting;– the study of cladding and fuel behavior under accident conditions of this type.
Knowledge of these phenomena has been considerably improved over the past
30 years. This has led to the development of sophisticated simulation software codesused to study these accidents at reactor scale, and to determine whether safety criteria
are met with sufﬁcient margins as regards remaining uncertainties and, where possible,
even improve these margins.

11.

This is a complete (360°) pipe break where the two separated sections are pulled apart, thus
maximizing the leak rate.

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3.1. Two-phase thermal-hydraulics
French research in this area, carried out for the most part through cooperation
between EDF, CEA, Framatome and IPSN, resulted in the development of the CATHARE12
simulation code [1]. This code provides detailed behavioral models of the water ﬂowing
through the reactor coolant and secondary systems of a pressurized water reactor, fromnormal operating conditions up to the limits of standard design-basis conditions, i.e. fueldamage. In order to cover the widest possible range of thermal-hydraulic conditions, theliquid and vapor phases are processed separately, using a set of six equations (conservation of mass, momentum and quantity of energy). Initially, modeling was mostly onedimensional, although in some parts of the reactor, such as the reactor core, ﬂows may
not be unidirectional, owing to non-uniform power distribution or changes in geometry
during the accident. In the 2000s, the code was upgraded to include the possibility ofmodeling multidimensional ﬂows and describing the behavior of droplets during core
reﬂooding separately from that of the continuous liquid and vapor phases [2]. As it travels
at high speed along the cladding, the vapor produces droplets due to a shear phenomenon; these circulate within the vapor ﬁlm and contribute to heat transfers.
Designing this type of software to obtain a detailed description of thermal-hydraulicphenomena calls for precise knowledge of the physical laws governing mass, momentum
and energy transfers at the interfaces between each phase and between these phases andthe walls. For this reason, many analytical tests were performed as part of thedevelopment of this code. Most of them took place between the 1980s and 1990s in
special, highly instrumented facilities built by CEA on its Grenoble site. The following is a
non-exhaustive list of experimental facilities:– CANON and SUPERCANON, for the study of depressurization and vaporization ofwater, ﬁrst in a circular conﬁguration, then in a geometrical conﬁguration
representative of a fuel assembly;– MOBY DICK and SUPER MOBY DICK (SMD), for the study of two-phase ﬂows
passing through openings representative of the breaks studied in the reactor
coolant system;– OMEGA and APHRODITE (EDF/Chatou), for the study of ﬁlm boiling around thefuel rods;– DEBORA, for the study of ﬂows under boiling conditions;– SMD, for the study of friction at the liquid-vapor interface;– PERICLES-2D, for the study of core uncovery and reﬂooding; in particular, it
implements a mockup of three assemblies of different power levels;– ROSCO, for the study of the very ﬁrst moments of core reﬂooding, which arecharacterized by an unstable ﬂow rate, as observed under experimental conditions
during large-scale experiments on reactor models (see further on).

12.

Advanced Thermohydraulics Code for Water Reactor Accidents.

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In the 1980s, CEA, with support from EDF, Framatome and IPSN, designed the
BETHSY facility on its Grenoble site, with the aim of checking whether the CATHARE code
was able to reliably predict the behavior of a reactor in an accident situation. BETHSY wasa mock-up of the reactor coolant system of a 900 MWe reactor. It was a full-scale modelin terms of the height of the different components, while volumes were represented on a
1/100 scale (Figure 3.1). It consisted of three loops, each equipped with a pump and asteam generator, together with the secondary system components considered essential
for thermal-hydraulic studies. The facility was designed for pressures of 17.2 MPa in thereactor coolant system and 8 MPa in the secondary system. The reactor core wasrepresented on a 1/100 scale by 428 electrically heated rods with stainless steel cladding.
Their power output was 3 MW, i.e. roughly 10% of the nominal power output of a reactor
on the scale considered, thus allowing simulation of the residual heat in the core just afterrod drop. All the engineered safety systems were reproduced, including the high- andlow-pressure injection systems, the accumulators and the secondary system safetyvalves. Breaks could be simulated at various points in the reactor coolant system: in thecold leg, in the hot leg, at the top of the pressurizer, and in the steam generator. Morethan 1000 measuring channels monitored changes in key parameters during the tests(temperature, pressure, rate and direction of ﬂow, void ratio, etc.).

Figure 3.1 The BETHSY loop (900 MWe PWR, three loops). © Georges Goué/IRSN-Source CEA (left),
CEA (right).

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In all, more than 80 tests were performed between 1987 and 1998. They did not focus
only on large-break LOCAs (complete pipe break). Other accident scenarios were studied,including those involving small or intermediate breaks, or nitrogen injection in the reactor
coolant system after total drainage of the accumulators, or loss of cooling during an
outage when the reactor coolant system is partially drained. The facility was also used toperfect control procedures (as part of the new state-oriented approach) to be implemented by licensees during the different phases of an accident to bring the reactor to asafe state.
In 1995, the international community selected test 6.9c as the basis for the OECD/
NEA international benchmark exercise on simulation software, called "International
Standard Problem (ISP) 38". This involved the study of a loss-of-coolant scenario during
an outage, when pressurizer and the steam generator outlet plenum manways are open
for maintenance work. ISP 38 brought together research organizations and licensees
from 18 countries and compared the results of ﬁve different software codes withexperimental results. It revealed that the main physical phenomena were effectively
reproduced by the codes considered, although variations were observed due to faults inmodeling multidimensional effects, particularly in the low-pressure range.
Similar facilities were built overseas in the 1970s and the early 1980s to study thethermal-hydraulic behavior of various nuclear reactor types in accident situations. The
most important of these are listed below:– Loss-of-Fluid Test, or LOFT, United States. A scale model of a 1000 MWe reactor(volume and power 1/50, height 1/2), comprising two loops, the only facility of its
type with a nuclear core for heating water;– Primärkreislauf, or PKL, Germany. A scale model of a 1300 MWe KWU reactor(volume and power 1/145, height full scale), comprising four loops and 314
electrically heated rods;– Large Scale Test Facility, or LSTF, Japan. A scale model of a 1100 MWe reactor(volume and power 1/48, height full scale), comprising two loops;– LOBI, European Community Joint Research Center, Ispra, Italy. A scale model of a
1300 MWe KWU-Siemens reactor (volume and power 1/700, height full scale). It
comprised two loops and 64 electrically heated rods;– PSB-VVER13 (Russia). A scale model of a 1000 MWe VVER-type reactor (volume
and power 1/300, height full scale) comprising four loops;– Parallel Channel Test Loop, or PACTEL, Finland. A scale model of a 400 MWe
VVER-type reactor (volume and power 1/305, height full scale), comprising threeloops and 144 electrically heated rods.
International agreements, generally negotiated in connection with research work
supported by OECD/NEA, gave CATHARE development teams access to many experimental results obtained during various research programs carried out at these facilities.
13.

Vodo-Vodianoï Energuetitcheski Reaktor.

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Current state of research on pressurized water reactor safety

In all, nearly 300 tests were analyzed and used to qualify the simulation code on a broad
spectrum of cooling accidents.
Most of these facilities were shut down and decommissioned. Some uncertainties
remained, however. In some conﬁgurations, for example, water slugs containing no boric
acid can form as a result of steam condensing in the steam generator tubes during a
LOCA, for example. They can then be transported to the core where they are liable toinitiate a criticality accident. In addition, performance checks had to be carried out on thepassive cooling systems, based on natural convection, that were contemplated for use insome Generation III reactors. The studies aimed at eliminating these uncertainties called
for the use of Computational Fluid Dynamics (CFD) codes, which are multidimensional
codes that provide very detailed ﬂow descriptions. It was, of course, important to checkthe predictive capabilities of these advanced tools through large-scale tests using
systems that were as realistic as possible. For this purpose, the OECD/NEA backed
new research projects in the PKL and LSTF facilities, in most cases with French
involvement. These included:– the PKL (2004–2007), PKL-2 (2007–2011) and PKL-3 (2012–2015) projects aimed
at studying phenomena such as boric acid dilution in various situations, andnatural convection in the event of loss of cooling during an outage, or to
reproduce beyond-design-basis situations involving delayed safety injections
for assessing safety margins;– the ROSA14 (2005–2009) and ROSA-2 (2009–2012) projects for studying thermal
stratiﬁcation and natural convection phenomena, and testing new cooling procedures in accident situations.
Tools simulating the behavior of the nuclear steam supply system were needed inorder to train teams likely to be required in the event of nuclear emergency. The SIPA
simulator was developed by IPSN in the 1990s using CATHARE modules. It has since been
replaced by the SOFIA15 simulator, which was jointly developed by AREVA and IRSN. It is
used by IRSN in particular to produce accident scenarios used during national emergency
exercises.

3.2. Fuel rod behavior
Research on fuel rod behavior during a loss-of-coolant accident has mainly focusedon the following phenomena [3,4,5]:– steam-induced oxidation of Zircaloy (zirconium alloy) cladding that modiﬁes themechanical properties of the cladding and generates hydrogen and heat;– cladding swelling and rupture;

14.
15.

Rig Of Safety Assessment.
Observation Simulator for Incidental and Accidental Operation.

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21– the mechanical behavior of oxidized cladding with regard to the thermal shock
induced by reﬂooding and to other phenomena during core cooling over the
longer term;– fuel pellet behavior inside the ballooned cladding, as the ceramic breaks up due tothe stress produced by reactor operation.
The earliest research work on zirconium oxidation due to steam at high temperatures
was carried out in the 1950s by Bostrom and Lemmon in the United States. The physicalprocess is quite complex. It involves the adsorption and dissociation of water molecules
at the cladding surface, the formation of O2– ions in the zirconia (zirconium oxide) layerthat forms around the edge of the cladding, with the ions diffusing to the oxide-metalinterface, where they contribute to the formation of zirconia or continue to diffuse
through the metal. At constant temperature, the zirconia layer builds up according to alaw that is parabolic with time (m2 = Kt, where m is the mass of oxidized zirconium by unit
area and t the time), which indicates that ion diffusion in the zirconia is the predominantphenomenon. The activity of the process increases with temperature, and the reaction
rate K is temperature-dependent according to an Arrhenius equation (K = e-c/RT, where cis a constant, R the ideal gas constant, and T the temperature in Kelvin).
Baker and Just at Argonne National Laboratory (ANL, United States) reviewed theexperimental results and completed them by performing other experiments at higher
temperatures, using hot ﬁlaments. Their correlation giving reaction rate versus temperature has been considered as a reference ever since.
A great deal of other research carried out in the 1970s backed up their work using
different methods and zirconium alloys: Cathcart and Powell at Oak Ridge National
Laboratory (ORNL, United States), Brown and Healey of the Central Electricity Generating Board (CEGB, United Kingdom), Urbanic and Heidrick of Atomic Energy of Canada
Limited (AECL, Canada), Lestikow and Schanz of Kernforchungszentrum Karlsruhe (KfK16,
Germany), and Prater and Courtright at Paciﬁc Northwest National Laboratory (PNNL,
United States). Other more precise correlations of the oxidation reaction rate wereobtained making a distinction between the various crystal systems of zirconia (monoclinic, tetragonal and cubic) according to temperature.
Depressurizing the reactor coolant system and emptying the reactor core causes the
cladding to heat up and leads to an increase in the difference between the pressure insideand outside the fuel rods. At around 700 °C, the Zircaloy sees a sharp decline in its
mechanical strength, but remains very ductile, which causes strain of up to nearly 50%
before rupture. The surface area of cladding exposed to steam oxidation can therefore
grow considerably, whereas the cladding becomes thinner. After rupture, the innersurface of the cladding is exposed to steam and can also be subject to oxidation.
The mechanical phenomenon was studied in the 1980s in CEA's EDGAR facility at the
Saclay center. Some 500 tests were performed on tubes made of various zirconium alloys
and directly heated by the Joule effect. This heating method ensured that temperature
was uniformly distributed throughout the tube. These tests determined the creep and
16.

Became after FzK (Forschungzentrum Karlsruhe) and then KIT (Karlsruher Institut für Technologie).

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rupture elongation laws according to tube temperature and the temperature andpressure ramps. In particular, the results highlighted the effect of heating kinetics on
the metallurgical phase change (from phase α to phase β) in the zirconium, which occurs
between 800 °C and 1000 °C, and signiﬁcantly modiﬁes rupture elongations. These laws
are used in the CATHARE code module for calculating the mechanical behavior of fuelrods during a loss-of-coolant accident.
Other research was also carried out on this phenomenon in the 1970s at the REBEKA
facility of KfK in Germany and ORNL's MRBT17 facility in the United States. The heating
method was the main difference between these tests and those carried out at the EDGAR
facility. In these tests, the zirconium alloy tubes were equipped with internal electricalheaters placed inside unheated cases. The heating element was not centered inside thetube, as a result of which tube temperature was not completely uniform. This reﬂects
reality more closely, at least for low-burnup fuel rods, where a gap remains between thefuel pellets and the inner surface of the cladding. The tests showed that a slight azimuthal
temperature gradient (around 10 degrees) could have a signiﬁcant impact on claddingstrain, with creep developing mainly at the hottest point. This leads to local ballooning
and lower total rupture elongation.
Further tests were carried out at these facilities with a geometry involving the use of
several rods arranged in a square-pitch conﬁguration (5 × 5 and 7 × 7 assemblies in
REBEKA, 4 × 4 and 8 × 8 in MRBT), with some unheated tubes. This was because someassemblies in a reactor are equipped with guide tubes through which the absorber rods inthe control rod assembly can slide into the core. Similar tests were performed in Japan
during the same period. They revealed that once the rods came into contact with one
another, cladding swelling spread in an axial direction. They also produced the highest
fuel assembly blockage rates (up to 90% during an MRBT test).
The shape of the blocked areas in the assemblies depends on the degree of
coplanarity of cladding circumferential strain and the axial extent of this strain. Thereis reason to fear that this shape may not be compatible with correct cooling of theseareas after core reﬂooding. This phenomenon was studied at various facilities underexperimental conditions at the end of 1970s and in the early 1980s. Examples include the
FLECHT-SEASET18 program conducted by Westinghouse Electric Corporation Nuclear
Energy Systems (United States), the FEBA19 and SEFLEX20 programs carried out by
Forschungzentrum Karlsruhe (Germany – FzK) and the THETIS and ACHILLES programs of
the United Kingdom Atomic Energy Authority in Winfrith (United Kingdom). The
experimental facilities consisted of assemblies of several electrically heated rods (upto 163 rods of normal length in the FLECHT-SEASET program). A group of these rods
included sections with a larger diameter in some places to reproduce the clad swelling
anticipated in the event of a LOCA. The main parameters studied were the locally blocked

17.
18.
19.
20.

Multi-Rod Burst Test.
Full-Length Emergency Core Cooling Heat Transfer–Separate Effects tests And System-Effects
Tests.
Flooding Experiments with Blocked Arrays.
(Fuel Rod) Simulator Effects in Flooding Experiments.

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fraction of the straight section, the reﬂooding rate and the type of reﬂooding (forced, or
passive gravity-driven).
Multidimensional thermal-hydraulic computer tools, designed to take into accountlocal changes in geometry, were required for the analysis of these different tests. It was
found that in areas subject to signiﬁcant blockage (90% of the straight section blocked),
the cladding could be cooled during reﬂooding, provided that the height of the blockage
was limited (less than 10 cm). Note that all these experiments were carried out with astationary heating element inside the rods. The conclusions of these studies could becontested because, in reality, nuclear fuel breaks up after several months of operation
and, as will be seen in some of the studies mentioned further on in this document, fuelfragments may ﬁll some of the gap inside the ballooned cladding. In France, IRSN has
been developing the DRACCAR21 multidimensional computer code since 2007 forstudying local blockage rates and all the above phenomena. The code has alreadybeen used in the examination of the safety analysis reports submitted by EDF in
connection with the planned update of the LOCA study baseline22.
Many experimental studies were carried out in the early 1970s on the mechanicalbehavior of oxidized cladding with regard to the thermal shock produced during fuel rod
reﬂooding. Argonne National Laboratory and Oak Ridge National Laboratory in the
United States were prominent in this area. These studies produced the experimental dataon which the maximum temperature of 1204 °C and the maximum equivalent oxidized
cladding fraction of 17% in U.S. regulations (10 CFR 50.46 and its Appendix K) werebased. The aim of these criteria was to provide a minimum requirement regarding
cladding ductility at temperatures above 135 °C, which is the water saturation temperature during core reﬂooding.
As the experimental studies concerned new cladding, it was naturally necessary toinvestigate the possible impact of in-service corrosion on the above values. While it isinside the reactor core, the cladding oxidizes, and the hydrogen that results from thedissociation of the water molecules migrates through the thickness of the cladding.
Beyond certain concentrations, the hydrogen in the solution reacts with zirconium toform hydride precipitates that embrittle the cladding material.
These phenomena were the subject of experimental studies in France as part of theresearch programs conducted at the CEA Grenoble center from 1991 to 2000, in
partnership with EDF and IPSN. The TAGCIS23 program studied the effect of a zirconia
layer around the edges of the cladding. Some 400 new cladding samples were placed
inside a pressurized water loop until a representative outer layer of zirconia had formed.
The samples were then heated by steam up to high temperatures, before being plunged
into a water tank. Various parameters were studied, such as the temperature increaserate, the maximum temperature reached, the initial zirconia thickness, and simultaneousoxidation of the outer and inner cladding surfaces. Some tests were reproduced using
21.
22.
23.

Deformation and Reﬂooding of a Fuel Rod Assembly during a Loss-Of-Coolant Accident.
Study method including, in particular, the criteria to be met.
Quench during a LOCA of fuel rod cladding that has undergone simulated irradiation.

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samples from cladding that had been used in a reactor as part of the TAGCIR24 program,with some cladding coming from rods with a maximum burnup of 60 GWd/tU.
In order to study the effect of hydride formation separately, new cladding samples
undergo a treatment in order to load them with hydrogen. They were then oxidized bysteam at high temperature (CODAZIR program) and some of the samples were quenched(HYDRAZYR program).
Analysis of all these results showed that the effects of in-service corrosion on thecladding were slight, in terms of both oxidation kinetics and mechanical strength during
rod reﬂooding.
However, other research carried out by the Japan Atomic Energy Research Institute(JAERI25) in Japan since then found that the secondary hydriding phenomena occurring athigh temperature on the inner surface of cladding after rupture considerably reduced
cladding ductility during reﬂooding. Furthermore, this research highlighted the importance of reproducing the axial stress applied to the rods as a result of their bowing andblocking in the grids when the core is reﬂooded. IRSN and JAEA26 continue to study these
phenomena, with research focusing particularly on the types of stress to be considered.
The results of these studies and research will be used to adjust the values of these criteria,
especially for cladding that has undergone signiﬁcant in-service corrosion.
Meanwhile, industrial research has focused on developing new zirconium alloys thatoffer improved resistance to in-service corrosion, consistent with longer reactor operating lives. These are zirconium and niobium alloys with additional elements destined toreplace the "traditional" Zy-4 alloy. The new alloys are called ZirloTM, E110 and M5TM,
respectively developed by Westinghouse in the United States, by Russia, and by AREVA in
France. Some of the safety tests mentioned above were, of course, repeated using these
new materials for an accurate assessment of the margins obtained with regard to
postulated events.
Given the complexity of the phenomena studied, it was important to conduct
experiments that reproduced the conditions of a large-break LOCA on a large scale
and as realistically as possible. The PHEBUS experimental reactor mentioned earlier was
built in France in 1970. The Phebus-LOCA program was designed to study the loss-ofcoolant accident caused by an instantaneous double guillotine break in the largest pipe in
the reactor coolant system, as well as the effectiveness of emergency cooling. Tests were
performed from 1979 to 1984, three with a single rod, and 22 with 25 rods that had neverbeen used in a reactor, arranged in a square-pitch conﬁguration, and with a ﬁssile heightof only 80 cm. The tests were aimed at observing fuel behavior in the limit cases ofengineered safety system operation. The results showed that damage was limited, evenunder pessimistic conditions, and that cladding ballooning did not prevent core cooling in
the case of fresh fuel (Figure 3.2). This can be explained by the fact that as cladding

24.
25.
26.

Quench during a LOCA of irradiated fuel rod cladding.
JAEA (Japan Atomic Energy Agency) is the result of the merging of JAERI and JNC (Japan Nuclear
Cycle Development Institute) in 2005.
Japan Atomic Energy Agency.

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Figure 3.2 Phebus-LOCA – sectional view (post-mortem) of a test fuel cluster after a characteristic
LOCA temperature transient. © IRSN.

temperatures were not uniform on the horizontal plane, balloon size was less than thatobserved during the experiments carried out in the EDGAR facility mentioned earlier.
Phebus-LOCA test 218 was used as the reference for a computer code benchmark test
organized by the OECD/NEA (ISP 19).
Other integral experiments were carried out in foreign reactors, some of which
involved portions of rods removed from a power reactor:– PBF-LOC tests performed in the Idaho National Laboratory (INL) Power Burst
Facility (United States) on a single rod irradiated up to 16 GWd/tU;– tests performed in the FR2 reactor at the KfK center (Germany), on a single rod
irradiated up to 35 GWd/tU;– tests carried out in the ESSOR27 reactor at the Joint Research Center (Ispra, Italy),using a single, unirradiated rod;
– FLASH tests performed in the SILOE reactor in Grenoble, with a single rod, which
was irradiated up to 50 GWd/tU in one test;– MT tests carried out in the NRU (National Research Universal) reactor at AECL's
Chalk River center (Canada) using assemblies of 32 unirradiated rods that had
preserved their original length.
On analysis, the results of these experiments revealed that the fuel in the rods
removed from power reactors breaks up and tends to relocate and ﬁll the gap left bycladding ballooning. These observations were recently conﬁrmed by tests carried out in
27.

ORGEL Reactor Test.

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the reactor at the Halden center in Norway. The tests in question were carried out as partof the HRP28 LOCA program, under the aegis of OECD/NEA from 2003 to 2012. In all,13 tests were performed with a single rod, under conditions representative of a largebreak LOCA. The rod samples studied came from Western and Russian pressurized water
reactors, and from boiling water reactors. The burnups studied ranged from 50 to90 GWd/tU. Results analyses conﬁrmed that at the time of cladding rupture, fuelfragments were relocating due to the effect of rod depressurization. Much of the mosthighly irradiated fuel (90 GWd/tU) had even spread outside the rod through the break.
Research on this phenomenon is in progress at the Studsvik laboratory in Sweden, whereseparate-effects tests are carried out on irradiated fuel to study post-rupture fueldispersion. At Halden, research work continues, focusing in particular on rods from
EDF nuclear power plants that have undergone ﬁve or six irradiation cycles.
So far, no integral experiment has been carried out inside a reactor using an irradiatedrod assembly. The conclusions of the blocked area cooling tests mentioned earlier could
be called into question if the balloons are ﬁlled by some of the fuel contained in the rods.
In 2013, IRSN, with EDF support and the participation of two CNRS researchlaboratories (LEMTA29 in Nancy and INSA-LamCoS30 in Lyon), launched a research
program called PERFROI31. Scheduled to last six years, the program is jointly funded by
the French National Research Agency (ANR), as part of the "Investment in the Future"
program and, more speciﬁcally, the call for research projects in the ﬁeld of nuclear safety
and radiation protection, issued in the wake of the Fukushima accident in 2012.
Research in this area focuses on the study of dreaded blockages and the possibility ofcooling them under reactor core reﬂooding conditions. It includes experimentation andmodeling work to validate the DRACCAR simulation code by 2020. The program is built
around two main topics: cladding mechanical properties and two-phase ﬂow.
The ﬁrst part of the program comprises creep and rupture tests performed in atemperature range of 600 °C to 1100 °C, using specimens of various zirconium alloys,some of which are pre-oxidized and pre-hydrided to simulate the different stages of inservice corrosion. In addition to these tests, swelling and rupture experiments will be
carried out on segments of cladding arranged in a more realistic, but ﬁnely instrumented,
conﬁguration. These will include several rods used to study the effect of contact between
neighboring rods on rupture, together with the extent of ballooning, particularly in theaxial direction.
The second part comprises thermal-hydraulic tests that implement instrumented
assemblies of 49 rods, simulating as realistically as possible anticipated blockage and fuelrelocation phenomena. The technological challenge consists in developing electricalheating elements capable of realistically simulating nuclear power distribution in
deformed rods partly ﬁlled with fuel fragments. The main parameters studied are:
28.
29.
30.
31.

Halden Reactor Project.
Laboratory of Energy and Theoretical and Applied Mechanics.
French National Institute of Applied Sciences - Laboratory for the Mechanics of Contacts andStructures.
Loss-of-Coolant Study.

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balloon geometry, balloon overpower temperature and ﬂow rate of injected water, andpressure. Some geometries studied will be similar to those already covered in earlier
research programs (in particular the THETIS program mentioned earlier) for the purposeof comparison.

References[1] F. Barré and M. Bernard, The CATHARE code strategy and assessment, Nuclear
Engineering and Design, 124, 257–284, 1990.
[2] P. Emonot, A. Souyri, J.L. Gandrille, F. Barré, CATHARE-3: a new system code forthermal-hydraulics in the context of the NEPTUNE project. Nuclear Engineering and
Design, 241, 4476–4481, 2011.
[3] C. Grandjean, A state-of-the-art review of past programs devoted to fuel behavior under
LOCA conditions. Part One. Clad swelling and rupture. Assembly Flow Blockage, 2005.
http://www.irsn.fr/EN/Research/publications-documentation/Publications/DPAM/SEMCA/Documents/IRSN_review-LOCA-Part1.pdf.
[4] C. Grandjean, A state-of-the-art review of past programs devoted to fuel behavior
under LOCA conditions. Part Two. Impact of clad swelling upon assembly cooling,
2006. http://www.irsn.fr/EN/Research/publications-documentation/Publications/
DPAM/SEMCA/Documents/IRSN_review-LOCA-Part2.pdf.
[5] Nuclear Fuel Behavior in Loss-of-Coolant Accident (LOCA) conditions. State-Of-theArt Report, OECD 2009, NEA No 6846, 2009.

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Chapter 4
Research on Reactivity-initiatedAccidents

The most severe design-basis reactivity-initiated accident (RIA) considered for apressurized water reactor in terms of uncontrolled nuclear reaction is the control rod
ejection accident. This accident is in plant condition category 4. In terms of safety, theaim is to ensure core coolability and avoid a core melt accident and, at the same time,check that this accident would not impair the integrity of the reactor vessel and coolant
system were it to occur.
In the event of control rod drive mechanism failure, the control rod ejection accident
is due to the difference in pressure between the reactor coolant system (155 bar) and thecontainment (at atmospheric pressure). This violent ejection causes a local runaway
effect in the nuclear reaction for several tens of milliseconds (power pulse), leading to a
rapid increase in fuel temperature. Neutron feedbacks limit the power transient before
the reactor trip (intact control rod drop), which occurs in a second phase.
A speciﬁc safety criterion relative to rod ejection was deﬁned in the 1970s based on
American tests, limiting the deposited energy (enthalpy) in the fuel (generally expressedin cal/g) during the reactivity transient. In the early 1990s, however, following theChernobyl accident and, more especially, in view of the gradual increase in fuel assembly
burnups, the international scientiﬁc community began to question the validity of thiscriterion, which was deﬁned for moderate burnups.
It was within this context that Japan and France developed research programs,
including IPSN's tests in the CABRI reactor. These were principally aimed at improving
understanding of the physical phenomena that could lead to cladding leaks, and theAnglais_Book30

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ejection of fuel fragments into the reactor coolant system, which could be detrimental tocore cooling.
The OECD report cited in the reference section [1] is a State-Of-the-Art Report on theknowledge of RIAs as of 2010.
Two types of cladding failure can occur during an RIA:
– failure due to pellet-cladding mechanical interaction (PCMI), which can occur
during the very ﬁrst instants of the power excursion, with the heated fuelsuddenly expanding more rapidly than the cladding, which is still cold;– this can be followed (after a few hundred milliseconds) by post-DNB32 failure,
when the cladding bursts due to a degradation in the cladding-coolant heatexchange coefﬁcient, followed by a sharp increase in cladding temperature, as
well as rising pressure inside the rods as the ﬁssion gases initially trapped inside
the fuel are released.
In view of the risks induced by burst cladding and the dispersal of fragmented fuel [2]– which could lead to water vaporization and rod cooling blockage – the U.S.NRC adopted
the "conservative maximum limit" of 280 cal/g of UO2 for the energy deposited33(enthalpy) during an RIA-type power transient in its Regulatory Guide 1.77. The aim wasto ensure that "core damage will be minimal and that both short-term and long-termcore cooling capability will not be impaired". The adopted value was deduced fromexperiments performed using unirradiated and slightly irradiated fuel (up to 32 GWd/tU)
as part of the Special Power Excursion Reactor Tests (SPERT [1969–1970] in stagnant
water and at ambient temperature). In the early 1980s, however, in light of continued
experiments conducted under more realistic conditions at the INL Power Burst Facility
reactor (PBF [1978–1980], circulating water, representative temperature and pressure
conditions, but with burnup only reaching 6 GWd/tU), it was found that a value of
280 cal/g was not sufﬁciently conservative. Criteria were therefore revised downwards: inEurope, values of 220 cal/g and 200 cal/g were adopted for fresh and irradiated fuelrespectively.
Increasing burnup to 52 GWd/tU or beyond, which is the value considered34 byvirtually all nuclear power plant licensees in the world, including EDF, can impair themechanical properties of Zircaloy cladding (see Section 3.2), while the fuel contained
inside the rod has undergone signiﬁcant changes induced by reactor operation, such as
fragmentation, and an increased quantity of trapped ﬁssion gases.
IPSN carried out a program of 14 tests in the CABRI reactor (Cabri REP-Na, 1993–
2002) to validate this value. The tests used fuel from nuclear power plants at burnups
ranging from 33 to 76 GWd/tU. Four of the tests used MOX35 fuel, a mixture of uranium
and plutonium oxides.

32.
33.
34.
35.

Departure from Nucleate Boiling.
Average radial value.
Adopted and authorized in some countries.
Mixed Oxide Fuel (UO2 + PuO2).

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As mentioned in Chapter 2, the CABRI reactor is able to generate power peaks
representative of those that could be encountered during a reactivity-initiated accident ina PWR. It achieves this through the depressurization of rods that have been pre-ﬁlledwith a neutron-absorbing gas (3He). The test fuel rod is placed in a test device andinserted in a cell at the center of the reactor. Almost all the meter-long test rods used for
the Cabri REP-Na program were made from fuel rods removed from EDF-operated
reactors. An instrument called a hodoscope for observing the fuel, and more conventional
instruments (for taking ﬂow rate, temperature, pressure, and acoustic measurements)
were used to determine the precise moment when cladding burst occurred, estimate thequantity of fuel dispersed and measure the resulting pressure wave. Until 1992, the CABRI
facility was used to carry out safety tests on fuel used in fast neutron reactors, and wasequipped with a sodium loop. The tests were therefore performed with sodium ﬂowing
around the test rods. This was considered acceptable for the study of the predominantly
mechanical phenomena occurring during the ﬁrst few tens of milliseconds of the power
excursion, when there is little effect on cladding temperature.
IPSN then launched a new program of experiments calling for a complete overhaul ofthe facility to study the phenomena occurring after the ﬁrst few hundred milliseconds(cladding dryout and rupture), as well as the consequences on reactor structures in terms
of pressure wave caused by the possible dispersal of fuel in the coolant. This was the
OECD/NEA project called CABRI International Program (CIP, 2000–2015), carried out in
partnership with EDF and many foreign safety and industrial organizations. Twelve tests
are scheduled under the program, two of which were performed in 2002 in the previous
facility using very high-burnup fuel (75 GWd/tU). The facility was then extensively
overhauled. The overall seismic resistance of the reactor was increased, a new cell was
installed and connected to a pressurized water system designed to reproduce representative thermal-hydraulic conditions, and many components were replaced.
Two other CIP tests (CIP3-1 and CIP3-2) are scheduled to focus on the study of postrupture phenomena. Further tests are planned with low deposited energy on rods whose
cladding has been embrittled before the tests as part of a post-CIP program. Tests such asthese would help to conﬁrm deposited energy criteria by examining whether fueldispersal in the coolant would not be detrimental to core coolability36, even if thecladding is embrittled or already ﬂawed.
CEA performed a series of analytical tests for IPSN and EDF on the mechanicalbehavior of cladding samples taken from fuel rods removed from nuclear power reactors(as part of the PROMETRA37 program at the Saclay center) and on critical heat ﬂux
phenomena during rapid wall heating (determination of critical heat ﬂux under transientconditions as part of the PATRICIA program at the Grenoble center).

36.

37.

IPSN had already considered this question in a similar way in the early 1980s during technicalassessments in connection with the SUPERPHENIX reactor, for the speciﬁc case of inadvertentcontrol rod withdrawal, especially as this event was classed in plant condition category 2 at thattime.
Mechanical Properties in a Transient.

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Other foreign programs with high-burnup fuels worthy of mention here include:– the tests performed in Russian reactors: IGR (Impulse Graphite Reactor), 47 to
49 GWd/tU, from 1990 to1992; BIGR, 47 to 60 GWd/tU, from 1997 to 2000;– the sixty or so tests performed in the JAEA Nuclear Safety Research Reactor38(NSRR) in Japan, using PWR, BWR (Boiling Water Reactor) and MOX fuels over aburnup range of 20 to 77 GWd/tU, from 1975 to 2011.
In all, nearly 140 tests were performed in reactor on high-burnup fuels. Analysis of theresults shows that the fuel can be dispersed in the coolant at deposited energy levels well
below the 200 cal/g of UO2 criterion with a value of around 120 cal/g once the burnup
exceeds 40 GWd/tU. This dispersal occurs after sudden cladding rupture due to:– stresses applied by the fuel, the volume of which tends to increase with heating(thermal expansion and internal pressurization by ﬁssion gases), bearing in mind
that the initial internal gap between the fuel and cladding is ﬁlled as soon as
burnup exceeds a few tens of GWd/tU;– impaired mechanical properties of cladding caused by the formation of zirconium
hydrides that embrittle it; the hydrides are formed when some of the hydrogen
released by the decomposition of water at the cladding surface in the reactor isdiffused inside the cladding.
However, the tests also showed that results were highly sensitive to the type of
cladding material. The worst results were obtained with Zircaloy-4. The rod clad with thisalloy used during the Cabri REP-Na 1 test (64 GWd/tU) exhibited traces of spalling39 of
the outer zirconia layer that formed while the rod was inside the reactor. It ruptured witha rise in enthalpy of only 30 cal/g of UO2, and about 2% of the fuel was dispersed in thesodium. More recently developed alloys (low-tin Zircaloy, ZirloTM, and M5TM) are lesssensitive to hydriding and are more resistant, even at high burnup.
Only 13 tests have been performed in the world using MOX fuel. The results of Cabri
REP-Na tests indicate that, energy levels being equal, the cladding would be exposed togreater stress and, in the event of rupture, greater quantities of fuel would be dispersed.
The mechanisms that might explain this difference in behavior have not yet been clearly
identiﬁed and further experiments are planned to study them as part of the CABRIInternational Program.
In 2010, JAEA also launched an international research program called “Advanced Light
Water Reactor Performance and Safety-II” (ALPS-II) in the NSRR reactor to improve
38.

39.

The TRIGA (Training, Research, Isotopes, General Atomics) pool-type reactor, designed and built byGeneral Atomics. It has been in use since 1975. The power excursion is produced by the rapid
ejection of neutron-absorbing rods. The fuel, a uranium-zirconium hydride alloy, enriched to about
20% with uranium-235, is designed to rapidly halt the excursion. The power peaks are not very
wide (from 4.4 to 7 ms at mid-power), unlike in the CABRI reactor, where the transient rod
depressurization valves can be adjusted to vary the peak width from 10 to a few hundred
milliseconds.
Spalling is the localized loss of some of the zirconium oxide (or zirconia) layer that forms at thesurface of the rod while it is in the reactor.

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33

knowledge of high-burnup and MOX fuels. This program is a sequel to the ﬁrst ALPS program
conducted from 2002 to 2010, with 14 tests on high-burnup fuels (67–77 GWd/tU) and MOX
fuels (45–59 GWd/tU).
The tests performed in the NSRR use an instrumented experimental capsule that
accepts reconstituted rods ﬁlled with fuel up to a height of 120 mm. A single rod is tested
during a test, with the cladding surrounded by stagnant water, which was initially atambient temperature and pressure in all tests before those of the ALPS program. The
instrumentation is used to measure cladding and coolant temperature, detect the
moment of cladding rupture, and measure the mechanical energy developed during
fuel dispersal and water vaporization.
For the ALPS program, JAEA designed a high-temperature (HT) capsule capable ofoperating at 280 °C and 7 MPa, values that are closer to real conditions. Six tests were
performed in this type of capsule. Results clearly revealed the effect of initial cladding
temperature. The cladding, which is made more brittle by the presence of zirconia
hydrides in particular, bursts at low temperature at signiﬁcantly lower energy levels.
The ALPS-II program should comprise 12 to 14 tests performed with samples of rodsthat have been used in European reactors (including, for France, a UO2 rod with a burnup
of 76 GWd/tU from reactor 5 of the Gravelines NPP, and a MOX rod with a burnup of
61 GWd/tU from reactor B3 of the Chinon NPP, both with M5TM alloy cladding). Half thetests will be performed in an HT capsule. The program also comprises four to six
analytical tests, called the Fission Gas Dynamics (FGD) tests, carried out by JAEA in
partnership with IRSN, which contributes to the design of the experimental device. These
tests are aimed at measuring the ﬁssion gas quantities released during a power excursion
by fuels from French reactors, and an experimental fuel irradiated to 130 GWd/tU.
This research program came to a halt when Japan's reactors were shut down in thewake of the Fukushima accident.
As in many other ﬁelds, simulation codes are needed to acquire a full understandingof phenomena and transpose experimental results for the study of reactivity-initiatedaccidents at reactor scale. The SCANAIR40 simulation code developed by IRSN isdesigned to calculate temperature and stress ﬁelds in fuel and cladding. It also calculatespressures due to ﬁssion gases inside the fuel. Meanwhile, more academic research is
conducted with CNRS (MIST41 laboratory, a laboratory "without walls", jointly funded by
IRSN, CNRS and the University of Montpellier) to determine crack propagation laws
inside hydrided cladding, and deﬁne rupture criteria.
Most countries operating pressurized water reactors consider that fuel-related
criteria should be updated with regard to RIAs, and jointly fund experimental programs
aimed at learning more about fuel behavior (CIP [IRSN] and ALPS [JAEA]). In France, talks
between IRSN, the French Nuclear Safety Authority (ASN) and EDF have been ongoingsince the early 2010s to discuss EDF's proposal for a "decoupling domain" aimed at
precluding the occurrence of cladding rupture due to pellet-cladding mechanical40.
41.

System of Simulation Software for Analysing Reactivity Injection Accidents.
Micromechanics and Structural Integrity Laboratory.

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Current state of research on pressurized water reactor safety

interaction in the event of a reactivity-initiated accident, with or without cladding
spalling. The following parameters are considered:– average rod burnup,– maximum thickness (azimuthal average) of the external oxide layer,– enthalpy variation,– mid-height pulse width,– maximum cladding temperature.

References[1] Nuclear Fuel Behavior under Reactivity-initiated Accident (RIA) Conditions. StateOf-the-Art Report, OECD 2010, NEA No. 6847, 2010.
[2] Nuclear Fuel Safety Criteria Technical Review, Second Edition – OECD 2012, NEA No.
7072, 2012.

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Chapter 5
Research on Recirculation Coolingunder Accident Conditions

If a leak occurs in a PWR reactor coolant system and is not compensated by thechemical and volume control system (CVCS), reactor core cooling is ensured by injecting
borated water. This is done by the safety injection system (SIS). In order to remove
residual heat and preserve the integrity of the reactor containment building, thecontainment spray system (CSS) may also be required. Sodium hydroxide is added tothe spray water to favor retention inside the reactor building of radioactive substancessuch as iodine-131.
The borated water required for these operations is initially drawn from the tank of thetreatment and cooling system of the water of the spent fuel storage pools (RWST42),
which has an approximate capacity of 1600 m3 in a 900 MWe pressurized water reactor.
When the low level threshold is reached in this tank, the SIS and CSS automaticallyswitch over to recirculation mode. In this mode, they take in water that has beencollected in the sumps located at the bottom of the reactor building (Figure 5.1). This
recirculation mode may be required over very long periods of time to cool the fuelassemblies. It must be highly reliable as it is fundamental in avoiding fuel assembly
damage and preventing a core melt accident.
The sumps at the bottom of the reactor building are equipped with a strainer systemdesigned to ensure that the quality of water downstream of the strainers is compatible withthe operation of SIS and CSS components, and with fuel assembly cooling. Debris can begenerated as a result of the break (destruction of equipment by shock wave or jet) or
42.

Refueling Water Storage Tank.

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Figure 5.1 Simpliﬁed diagram of SIS and CSS operation in direct injection mode (left) and recirculation
mode (right) – the component cooling water system (CCWS) is also responsible for cooling some pumps
in the systems shown. © IRSN.

ambient conditions in the reactor building (temperature, irradiation, etc.), or it may alreadybe present in the reactor building (dust, etc.). The ﬂow of water (from the break in thereactor coolant system or CSS washdown) can then transport this debris to the sumps.
Depending on its characteristics (size and weight), some of this debris may be carried
as far as the strainers and form a debris bed, which can lead to strainer clogging as a resultof physical or chemical effects. The main risk involved is the impeding or obstruction ofwater recirculation.
Strainers must be sized (in terms of surface area and mesh size) to:– prevent SIS and CSS pump cavitation failure. This risk is reduced by the use of
sufﬁciently large screen surface areas;– ensure the operation of components located downstream of the strainers, as well
as fuel assembly cooling. This can be guaranteed by a suitable strainer mesh size.

5.1. Operating experience feedback and research topics
Risks relating to the loss of recirculation cooling capability in light water reactors wereidentiﬁed in the 1970s, and were the subject of the U.S.NRC Regulatory Guide RG 1.82
issued in 197443.
43.

In 1974, the U.S.NRC published its "Regulatory Guide 1.82", which describes the methods andpractices acceptable to U.S.NRC staff for addressing all the issues generally related to emergency
recirculation cooling after a loss-of-coolant accident and applicable to PWR- and BWR-type lightwater reactors. The guide was gradually updated to reﬂect operating experience feedback and new
knowledge resulting from research and development. This initial version of the Regulatory Guide
was applied to French nuclear power plants designed by Westinghouse. The version currently inforce is Revision 4 released in March 2012 [1].

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37

During the early 1990s, however, several incidents occurred in boiling water reactors(at Sweden's Barsebäck plant, and at the Perry and Limerick plants in the United States),
which raised new questions on the risk of strainer clogging. The incident that affectedreactor 2 at the Barsebäck BWR plant in Sweden [2] on July 28, 1992 was caused by the
inadvertent opening of a valve. The resulting steam jet produced 200 kg of ﬁbrous debris,composed mostly of insulating material. Half of it was carried to the suppression pool orwetwell44, leading to a signiﬁcant increase in head loss across the emergency core coolingsystem strainers after 70 minutes. Given the modest size of the equivalent break(corresponding to the valve diameter), the quantity of debris appeared far greaterthan could have been expected based on RG 1.82 in its ﬁrst revision in November 1985.
In view of the incidents mentioned above, international investigations focused
especially on boiling water reactors. Research results concerning this type of reactor
revealed that the quantity of debris generated by a pipe break in the reactor coolant
system could exceed by far the quantities considered until then, based on earlier
research. They also showed that the debris could be ﬁner (and therefore more easilytransportable) and that some combinations of debris (ﬁbers with particles, for example)
could augment the risk of strainer clogging. These results can be found in many
documents, including the report NUREG45 6224 issued in 1995 [3] on the study ofstrainer blockage in BWR engineered safety systems.
This led to new studies being initiated in the late 1990s by industry, laboratories andresearch organizations, with active involvement on the part of IPSN.
Research work also revealed further issues relating to the chemical effects associatedwith the pH and temperature of the solution, and their impact on the risk of strainer
blockage, as well as the effect of debris passing through the strainers on the functional
performance of SIS and CSS engineered safety system components, and on fuel assembly
cooling.
Research carried out since 1974, and incident analysis have raised many questions, forthe most part in studies on the following topics.

► Characterization of debris likely to reach the strainers
It is essential to characterize the debris (type, quantity and size) likely to betransported to the strainers in order to assess: the risk of strainer clogging; the operationof components located downstream of the strainers; and fuel assembly cooling.
By far, the greatest amount of debris is directly produced by the shock wave
generated by the break. In particular, it comes from the destruction of insulating material(glass ﬁber, Microtherm, etc.) on equipment located near the break (such as piping andsteam generators), as well as paint and concrete. The quantity of debris depends
especially on the pressure ﬁeld induced by the break and the strength of the material

44.
45.

Pool located at the bottom of the reactor building from which the engineered safety systems draw
water.
Nuclear Regulatory Report.

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38

Current state of research on pressurized water reactor safetyof which the debris is composed. Size and shape varies (ﬁbers a few millimeters in length,
particles a few micrometers in diameter).
Other debris is generated by ambient conditions (temperature, humidity, irradiation)in the reactor building. It may be composed of chips (a few mm2) or particles (a few μm in
diameter) of worn or damaged paint.
Lastly, some debris, such as dust or grease, is found in the reactor building before theaccident. Referred to as latent debris, it can be entrained in streams of water and must be
taken into consideration in strainer design.

► Debris transport to strainers
Some debris can be transported (Figure 5.2) to the bottom of the reactor building in
the water ﬂowing through the break and in containment spray washdown. Other debris
can be trapped in retention zones – for example on ﬂoors and gratings. This is calledvertical debris transport.
During the switchover to recirculation mode, the debris transferred to the bottom of
the reactor building can, depending on its size or weight, be transported to the SIS andCSS sump strainers. This is referred to as horizontal transport.

Figure 5.2 Illustration of debris transport to sump strainers (900 MWe PWR containment building).
© IRSN-source EDF.

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39► Assessment of strainer head loss and clogging risk
The debris transported to the strainers may either agglomerate in the strainer meshand form a debris bed over the entire screen surface area (Figure 5.3), thereby increasing
the head loss across the strainer terminals, or, depending on its size or the permeability ofthe debris bed formed, be transported downstream of the strainer, and induce "downstream" effects. Strainer head loss is determined using correlations, such as the NUREG
6224 correlation, that are used to check that there is no risk of pump cavitation failure.
In this respect, research has demonstrated that for some strainers, just a small
quantity of mixed debris (ﬁbers and particles) can lead to signiﬁcant head loss acrossstrainer terminals in recirculation mode (Figure 5.4), and thereby increase the risk ofpump cavitation. This is known as the thin ﬁlm effect.

► Impact of chemical effects on strainer head loss
Research work, in particular that carried out by IRSN, has shown that physicalchemical phenomena can lead to the formation of precipitates (crystals, gels) in thedebris bed. The precipitates can build up and cause an increase in strainer head loss. Thesephenomena depend not only on water pH and temperature, but also on the characteristics of the debris in the containment, especially in the water at the bottom.

Figure 5.3 Examples of ﬁbrous and particulate debris covering screen surfaces (qualiﬁcation tests on SIS
strainers for the EPR). © IRSN.

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Figure 5.4 Diagram of strainer head loss peak or thin ﬁlm effect [6].

► Downstream effects
Depending on strainer system characteristics (mesh, screen surface area), the debris
passing through the strainers, and the chemical substances in the water, can lead tophysical and chemical phenomena, such as soiling, clogging and erosion. This can impair
the performance of SIS and CSS components, such as pumps, diaphragms, heatexchangers, valves and safety valves, and compromise fuel assembly cooling in thecore (partial blockage of fuel assembly grids).

5.2. Past research programs and lessons learned
Following the BWR incidents that occurred in the 1990s, a considerable number ofstudies and research and development programs were conducted around the world. Thiswas particularly the case in France with work carried out by EDF and IPSN (then IRSN), to
obtain a fuller understanding of the phenomena involved and assess strainer systemdesign.
As part of a knowledge-sharing initiative, OECD/NEA published a knowledge-basereport [4] in 1995, following the incident at the Barsebäck plant. An updated version of
the report was issued in 2013 [5]. More than ten laboratories and research organizations
worked – or continue to work – on these topics for safety authorities and TSOs46 (U.S.
NRC, IRSN, etc.), designers and industry (e.g. General Electric, AREVA, EDF). Progress insome of this work is reported in a number of documents [6,7,8]. IPSN collaborated with46.

Technical Safety Organization.

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41

VUEZ and the University of Trenčín (Slovakia), and with EREC (Russia) on a series of tests
in experimentation loops, described in greater detail below.
EDF took part in the Nuclear Energy Institute (NEI) study and research program onrecirculation issues that led to the publication of a guide [6, 7] in 2004 describing
methods for evaluating strainer design for PWRs.
In an initial research program, IPSN deﬁned and funded a number of experiments,
most of which were carried out by VUEZ and EREC from 1999 to 2003. IPSN supervised
the experiments. The following topics were studied:– efﬂorescence of debris induced by water ﬂow: 77 tests were performed in the
ELISA facility set up by IPSN at VUEZ to study the impact of insulation type andwater ﬂow rate, temperature and quality, as well as how solid particles affect headloss due to the presence of insulation material on a strainer screen;– vertical debris transport and crushing caused by obstacles: 30 tests were carriedout in the IVANA facility, also installed at VUEZ, to assess the size of debris
generated, based on the initial size of debris, spray water ﬂow rate, and the type of
insulating material considered;– horizontal debris transport rates and debris settling in the containment: 52 tests
were performed in the VITRA facility, set up by IPSN at EREC, to study the waterﬂow rates below which debris, depending on its characteristics, settles on thehorizontal ﬂoors of the containment building, and thus does not contribute tosump strainer clogging;– strainer obstruction mechanisms: following a campaign of 15 preliminary tests,
11 full-scale tests were performed at the MANON facility, set up by IPSN at VUEZ,
to assess the quantities of debris likely to seriously compromise the performanceof water recirculation pumps.
In 2003, IRSN shared to the French Nuclear Installations Directorate the lessons
learned from international research and, in particular, the tests mentioned above. The
ﬁndings of this work raised the issue as to whether the possibility of sump strainer
clogging could lead to loss of recirculation cooling in PWRs under accident conditions. In
2004, EDF decided to make a number of engineering changes to its nuclear power plants
in operation, based on the NEI guide mentioned earlier [6,7] and the results of the ensuingresearch. These changes were implemented on reactors from 2005 to 2009. The newstrainer design is now based on methods and practices developed by EDF that are speciﬁc
to French reactors but also comply with the design provisions of the latest version of
U.S.NRC Regulatory Guide RG 1.82.
Until the early 2000s, the mesh size of the strainers installed in French nuclear power
varied with the reactor series. The strainers were vertical panels installed in the sumps in acircumferential arrangement. Based on research and development ﬁndings, two changes(decided in 2004) were implemented:– removing Microtherm insulation, because this material can generate very small
particles, making it unacceptable in terms of the risk of sump strainer clogging;Anglais_Book42

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Current state of research on pressurized water reactor safety– ﬁtting strainer systems with new strainers with a signiﬁcantly larger surface area
(up to 48 times that of the previous strainers).
Note that for the EPR, the strainer system adopted consists of two47 lines of strainers
installed across the path of the debris:– basket strainers at the edge of the Internal Reactor Water Storage Tank (IRWST),
which provides a reserve of borated water in the containment, at the ﬂoor
openings in large components;– strainers in the middle section of the IRWST (In-containment Refueling Water
System Tank), where the spray lines leading to the pumps are installed.
Following the studies it carried out from 1999 to 2003, IRSN decided to launch a joint
program with VUEZ and the University of Trenčín to study the chemical effects of theﬁbrous bed deposited on the strainers.
The aim was to assess how the formation of precipitates impacted strainer head loss,
with engineered safety systems operating in recirculation mode, following a reactor
coolant system break. The program was in two parts. The purpose of the ﬁrst part was todetermine the concentration of precipitates that might form. The second sought todetermine the impact of precipitate formation on ﬁbrous bed head loss, based on tests
carried out in the ELISA test loop.
The program conﬁrmed that precipitates could form in a ﬁbrous bed and that thissigniﬁcantly affected strainer head loss. Tests also showed that temperature had aconsiderable impact on the formation, type and development of precipitates. A strainer
head loss computational model was developed based on the above results. The model isalso designed to provide ﬁbrous bed porosity data.
IRSN initiated another research program to conﬁrm the above model and consolidate
knowledge concerning long-term corrosion in different types of insulating material andthe precipitates that could form as a result. The program was performed using six test
loops called ELISA Babies, derived from the design of the ELISA loop.

5.3. Ongoing research programs
As of 2015, many research programs were still in progress both in France (EDF and
IRSN) and abroad. These are mostly concerned with studying the effects of debris on
engineered safety system components located downstream of the strainers, and on fuelassembly cooling. Some research programs are also aimed at modeling physical andchemical phenomena more precisely to substantiate some strainer designs.
In 2015, for example, IRSN decided to launch new research programs involving theuse of VIKTORIA, a new test loop installed at VUEZ. The aim is to evaluate the design ofstrainer systems used in currently operating reactors and those intended for use in the
EPR, in terms of chemical effects and the effects of debris passing through the strainers.

47.

In addition to trash racks installed on the ﬂoor of large components.

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43

Figure 5.5 View of the VIKTORIA test loop © Brano Valach/IRSN.

VIKTORIA is an integral loop (Figure 5.5) that can simulate all the key physical andchemical phenomena for analyzing questions relating to strainer systems. It can be usedfor various types of pressurized water reactors for the following purposes:– studying effects upstream of strainers including:strainer head loss,

impact of chemical effects on this head loss,gas formation,the effect of back-ﬂushing, i.e. injecting water inside the strainers to unclog them;– studying physical and chemical effects downstream of the strainers, which entails
characterizing the debris (quantity, type, size) passing through the strainers, andits impact on fuel assemblies and other components, such as heat exchangers anddiaphragms.
Otherwise, research and studies carried out so far have not considered core melt
situations. In the EPR, however, a core melt situation would require the use of theContainment Heat Removal System (CHRS), which is equipped with strainers. The types
and quantities of debris generated in such situations could differ from those considered indesign-basis transients owing to the extreme conditions (temperature, irradiation)encountered in the event of core melt. This could lead to changes in the local chemistryof the radionuclides trapped in the debris (as a result of water radiolysis) and cause

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precipitates. IRSN has begun an exploratory study on this topic that could lead on tomore speciﬁc research over the coming years.

5.4. Simulation
The subject of debris, and its impact on recirculation cooling following a break in thereactor coolant system, involves a wide range of generally three-dimensional phenomena
and disciplines, such as hydraulics and chemistry. There is a gradual increase [8] in the useof CFD simulation tools for studying issues such as debris transport, the deposition ofdebris on strainers, variation in head loss through a debris bed, and the risk of air andincondensable gas entrainment.
IRSN has used this type of tool for studying the representativeness of EDF tests
carried out to substantiate the design of new strainers. These include tests on chemical
effects. Based on the initial results of these tests, EDF considered that chemical effects
had no impact on strainer head loss, owing to signiﬁcant settling observed at the bottomof the test loop, and the fact that a stable debris bed could not be formed on the strainer.
IRSN simulations, for example, revealed that debris was drawn down upstream of theloop strainer (Figures 5.6 and 5.7). The representativeness of these tests for the case of areactor therefore remained to be demonstrated.

Figure 5.6 View of jets in the EDF/CEMETE model (velocities greater than 6.5 cm/s) where x is the
direction of ﬂow. © IRSN.

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Figure 5.7 View of streamlines in the direction of ﬂow at the middle of the EDF/CEMETE model (in thevertical plan) – the injection nozzles are on the left. © IRSN.

References[1] Water Sources for Long-term Recirculation Cooling Following a Loss-of-Coolant
Accident, Regulatory Guide, RG 1.82, Revision 4, March 2012.
[2] Proceedings of the OECD/NEA Workshop on the Barseback Strainer Incident (1994:
Stockholm, Sweden), Report NEA/CSNI/R(1994)14.
[3] Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA
Generated Debris, October 1995, NUREG/CR-6224, SEA No. 93-554-06-A:1.
[4] Pressurized Water Reactor Sump Performance Evaluation Methodology, Guide NEI
04-07, Revision 0, Vol. 1, December 2004.
[5] Knowledge Base for Emergency Core Cooling System Recirculation Reliability, Report
NEA/CSNI/R(95)11.
[6] Update Knowledge base for Long-term Core Cooling Reliability, Report NEA/CSNI/R(2013)12.
[7] Safety Evaluation by the Ofﬁce of Nuclear Reactor Regulation Related to NRC
Generic Letter 2004-02, Guide NEI 04-07, Revision 0, Vol. 2, December 2004.
[8] Evaluation of Long-term Cooling Considering Particulate, Fibrous and Chemical
Debris in the Recirculating Fluid, Westinghouse Report, WCAP-16793-NP, Revision 2,
October 2011.

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Chapter 6
Research on Spent Fuel Pool
Uncovery Accidents

Nuclear power reactors have a spent fuel pool where spent fuel assemblies are stored
until they are ready to be transported to the reprocessing plant, in other words, when
they have lost enough of their residual heat. During reactor outages, the spent fuel poolcan also be used to temporarily accommodate fresh or already irradiated fuel assemblies
for loading into the reactor core before it is restarted. The spent fuel pool is about 12 m
deep. The fuel assemblies are stored in racks placed on the bottom of the pool, underabout seven meters of water, thus providing workers with sufﬁcient protection against
ionizing radiation under normal operating conditions. The pool also serves as a major
source of water for cooling fuel assemblies in the event of a cooling system failure.
A spent fuel pool can accommodate a large number of spent fuel assemblies (from
300 to 600 depending on the type of reactor) containing radioactive substances. It is
located in a fuel building (BK) with a dynamic conﬁnement system ensured by theventilation system.
The reactor cavity in the reactor building (RB) and the spent fuel pool in the fuelbuilding (BK) are shown in Figure 6.1.
It was considered in the safety demonstration carried out at the time of design that, inthe event of pool cooling system failure, and given the low residual heat of the spent fuelassemblies, there would be enough time to implement safety measures before the fuelwas uncovered. That is why, in the event of loss of fuel rod integrity, the spent fuel poolbuilding does not have the same radioactive conﬁnement capability as the reactor
containment building.

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Figure 6.1 Cutaway view of the reactor cavity in the reactor building (RB) and the spent fuel pool in thefuel building (BK) in a 900 MWe CPY48-series plant.

Various design features, such as vacuum relief valves on the fuel pool cooling andpuriﬁcation systems (FPCPS), were implemented to preclude the risk of the pools
emptying. These features were signiﬁcantly reinforced following several incidents(loss of tightness of doors and gates, drainage induced by lineup errors, unsatisfactory
performance in vacuum relief valves, foreign matter blocking these valves) or inconnection with safety reviews.
The Fukushima Daiichi accident conﬁrmed the importance of examining all thepossible strategies for guaranteeing fuel assembly cooling, especially now that fuel andfuel pool operating conditions have changed considerably since design (fuels more highly
irradiated, existence of MOX fuels).
If it becomes impossible to ensure sufﬁcient fuel assembly cooling, the dreaded cliffedge effect is a runaway exothermic oxidation reaction in fuel cladding due to air andsteam, leading to cladding deterioration and signiﬁcant radioactive release. In this event,very little radioactive iodine-131 should be released, given the time the fuel is stored after
removal from the reactor core. On the other hand, a very large quantity of ruthenium
should be released. This is a particularly radiotoxic element, even if its radioactive half-life
is considerably shorter than that of cesium (see Section 9.4).

48.

Second serie of French power reactors of 900 MWe (EDF).

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Boiling water in the storage racks could also cause the fuel assemblies in the rack toreturn to criticality. This could lead to rod failure, human radiation exposure andradioactive release.
Assessing the risks associated with loss-of-coolant or loss-of-cooling accidents
entails modeling convection phenomena that are mixed (forced and natural); threedimensional (owing to the shape of the storage racks and the non-uniform distribution ofresidual heat in the spent fuel assemblies in the pool); and involving multiple ﬂuids(water, steam and air during the uncovery phase). Oxidation phenomena of zirconium
alloys in contact with mixtures of steam, oxygen and nitrogen must also be considered[1]. These oxidation reactions generate heat and, possibly, hydrogen, with a related risk of
explosion. Research work also focuses on the effectiveness of engineered safety systems
according to the scenarios considered (for example, risk of pump cavitation if steam or air
is entrained in the systems, effectiveness of spray systems for cooling uncovered fuelassemblies).
Although tests carried out in the 1950s already showed that the kinetics of zirconium
oxidation by air at high temperatures differed considerably from that of oxidation bysteam, air-induced oxidation of zirconium alloys only recently became the focus ofdetailed studies.
The zirconium oxidation reaction due to oxygen releases about twice as much heat
(1101 kJ/mole compared with 528 kJ/mole when steam is the oxidizing agent). Zirconium
also reacts with nitrogen. The resulting nitride can then react with oxygen and also
release considerable heat (736 kJ/mole). If the conditions are suitable for initiating thesechemical reactions (air in the storage racks and temperature around 600 °C), the heatgenerated could lead to runaway temperatures and damage to the fuel rods.
In France, IRSN carried out air-induced oxidation tests on Zircaloy-4 and M5TM
cladding samples, a few centimeters in length, as part of the International Source Term
Program (ISTP, 2005–2013, MOZART49 test series). The studies were carried out in atemperature range between 600 °C and 1100 °C. A thermogravity system was used tocontinuously monitor the increase in sample weight, and thereby deﬁne oxidation
kinetics laws. These were then incorporated in the ASTEC50 (see chapter 9 on severe
accidents) simulation code used to calculate the different stages of the accident.
Other analytical tests on different alloys were performed at more or less the sametime in the United States and Germany. These differed in terms of the type of alloys
studied, sample geometry (cladding sections with open or closed ends), the initial state of
the materials (fresh or pre-oxidized and pre-hydrided to simulate the effects of in-service
oxidation), and the type of oxidizing ﬂuid studied (pure nitrogen, mixture of nitrogen andsteam, or mixture of air and steam).
These tests, and those performed by the Karlsruhe Institute of Technology (KIT) in
Germany, conﬁrmed that cladding oxidation in air could lead to more serious consequences than oxidation by steam. This is because the kinetics of the oxidation reaction
49.
50.

Measurement of Zirconium Oxidation by Air at Temperature.
Accident Source Term Evaluation Code.

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accelerates as from a certain oxidation rate. First, a dense layer of zirconium oxide isformed that limits oxygen diffusion to the metal. Its thickness increases with the squareroot of time (see Section 3.2). The kinetics then accelerates further when this dense oxide
layer cracks. The air then reaches new areas of unoxidized metal, and the oxidation
reaction proceeds in a linear manner with time. This phenomenon – known as breakaway– is not very well understood at the present time. One explanation might be the change inthe zirconia crystal system (from monoclinic to tetragonal) at around 1000 °C (seeSection 2.2). However, another hypothesis was put forward to explain the speciﬁc role ofnitrogen and the formation of dense nitrided compounds (Zr-O-N or ZrN), the oxidationof which could cause the zirconium oxide layer to crack at temperatures below 1000 °C.
Integral tests were also carried out by Sandia National Laboratories in the UnitedStates. These were part of a program called "Spent Fuel Pool Heatup and Propagation
Project" (2004–2006), set up to study BWR fuel assembly damage51. This was followedby the OECD/NEA "Sandia Fuel Project" (2009–2013), devoted to the study of PWR fuelassembly damage, bringing together 13 countries, including France. The purpose was toobtain thermal-hydraulic data on natural convection cooling of fuel rods by ambient air
in the event of loss of cooling, in order to validate accident computer codes (such as
ASTEC and a dedicated version of DRACCAR).
Experiments were carried out on a full-length, ﬁnely instrumented assembly comprising 17 x 17 Zircaloy-clad, magnesium oxide52 heater rods. First, nondestructive tests
were carried out for natural convection in air. These were followed by a destructive test
that was carried out until the onset of runaway oxidation (also called zirconium ﬁre) and
its axial propagation within the assembly. In a second test campaign, the heatedassembly was surrounded by four unheated assemblies to study the radial propagationof runaway oxidation to surrounding, lower power assemblies. Some of the rods in these
unheated assemblies were also pressurized to simulate ballooning effects (possibility ofoxidation accelerating after spalling of the oxide layers formed). It was seen that runaway
oxidation did propagate to the surrounding assemblies. Following this research, core melt
accident simulation codes, such as ASTEC, beneﬁted from improved models for describing this phenomenon more accurately.

► Planned programs
In 2013, IRSN launched a research program, called DENOPI53, to learn more about
these phenomena. The following CNRS research laboratories also participate in the
program: PROMES – Processes, Materials and Solar Energy, laboratory in Perpignan –,LEPMI – Laboratory of Electrochemistry and Physico-chemistry of Materials and Interfaces in Grenoble –, ARMINES-SPIN – Carnot M.I.N.E.S Institute – Center for Industrial
and Natural Processes that is part of the École des Mines in Saint-Etienne; and LVEEM –Vellave Laboratory on the Development and Study of Materials in Le Puy-en-Velay.

51.
52.
53.

Unlike PWR fuel rod assemblies, BWR assemblies are arranged inside Zircaloy channels.
Magnesium oxide is not radiotoxic and its thermal properties (conductivity, speciﬁc heat, etc.) aresimilar to those of uranium oxide.
Spent Fuel Pool Water Uncovering.

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51

Scheduled to last six years, the program is jointly funded by the French National Research
Agency (ANR), as part of the "Investment in the Future" program and, more speciﬁcally,
the call for research projects in the ﬁeld of nuclear safety and radiation protection(RSNR), issued in the wake of the Fukushima Daiichi accident in 2012.
Work scheduled as part of DENOPI includes experiments, modeling and simulation code
validation, with the aim of learning more about the different phases of accidents of loss ofheat removal or uncovery of fuel assemblies stored in the spent fuel pool. The program adopts
an analytical approach seeking to expand knowledge in the following three areas:– natural convection cooling at pool scale,– thermal-hydraulic behavior at fuel assembly scale in the event of fuel uncovery,
and effectiveness of spraying,– acceleration mechanisms observed in fuel cladding oxidation in contact with an airand steam mixture.
Natural convection in the spent fuel pool will be studied (2015–2017) using a 1/5
model of a spent fuel pool, storage racks and the fuel assemblies they may contain.
Results will also be interpreted using computational ﬂuid dynamics tools designed tosolve equations governing ﬂuid movement.
In 2018–2019, the study of thermal-hydraulic behavior in a fuel assembly will use afull-length model of a fuel assembly and its storage rack, under conditions that represent
the various phases of the accident, namely loss of cooling, pool water boiling, fueluncovery, and resumption of cooling. Fluid velocities in the gaps between rods andoxygen concentrations will be measured to obtain very precise experimental data thatwill be fed into a database to validate the models used in accident calculations.
Cladding oxidation and the related runaway phenomena will be studied, together withthe presumed role of nitrides, using high-performance laboratory technology including:– in situ X-ray diffraction (XRD) in a steam-rich gas ﬂow will be used for real-timemonitoring, at reaction temperature, of changes in compounds formed at hightemperatures (500 °C to 1200 °C);
– Raman micro-imaging will be used to study the microstructure and compositionof the zirconium oxide layer formed.
Integral tests are also planned using simulated fuel assemblies in KIT's QUENCH
facility as part of the Severe Accident Facilities for European Safety Targets (SAFEST)
program, which is partly funded by the European Union.
Lastly, IRSN coordinates the AIR-SFP project (2015–2017), which is jointly funded by
the European Union, following a call for research projects as part of NUGENIA+54. The

54.

NUGENIA+ manages research for the development and safety of Generation II and III nuclear
power reactors on behalf of NUGENIA (see Chapter 1). This association has more than a hundred
members, including industrial companies, nuclear licensees, technical safety organizations, andresearch laboratories, for the most part in Europe.

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purpose of this project is twofold: to assess the ability of different codes used insimulating reactor core melt accidents, such as ASTEC, to predict the different phases ofan accident caused by fuel uncovery in a spent fuel pool; and to prepare a research
program to ﬁll the most signiﬁcant gaps in knowledge. Fourteen countries are involved inthis project, the results of which will lead to a better understanding of these accidents,and help to prevent them and mitigate their consequences.

References[1] Status Report on Spent Fuel Pools under Loss-of-Coolant Accident Conditions.
Nuclear Safety NEA/CSNI/R(2015)2, May 2015.

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Chapter 7
Research on Fires

7.1. Fire risks at nuclear installations
Fire is a major risk to be taken into account for nuclear installation safety. On March
22, 1975 a cable ﬁre broke out in Unit 1 of the Browns Ferry power plant in the US, as aresult of a leaktightness test of a polyurethane foam cable sleeve through a wall,
performed by candle. Although the operators involved immediately proceeded to put out
the ﬁre, unbeknown to them it spread along the cables on the other side of the wall. It
caused a loss of control of some equipment important for the safety not only of Unit 1
but also of the neighboring Unit 2. The operators' reactions to this loss of control
contained the incident before it could degenerate into a more severe accident. In October
1989, a turbine blade failure at the Vandellos power plant in Spain (UNGG-type
reactor55) had multiple consequences: a hydrogen leak and explosion, turbine lubricating
oil ﬁre, loss of power and of compressed air for regulating several items of equipment
involved in residual reactor heat removal, major basement ﬂooding (including of thereactor building basement), etc. The ﬁre, which lasted for more than four hours, wasbrought under control with the help of ﬁreﬁghting units from within a 100 km radiusaround the power plant. If the cooling system had stopped working, the 3000 metric tonsof graphite would have caught ﬁre. The reactor has not been restarted, mainly because of
the high cost of the modiﬁcations necessary to improve safety.

55.

Reactor running on graphite-moderated, gas-cooled natural uranium fuel (UNGG: Natural
Uranium-Graphite-Gas).

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Fires do happen from time to time in French nuclear reactors, with consequences ofvarying degrees of severity. Some examples are:– an incident in July 1999 at Bugey nuclear power plant in which a single cause (anelectrical fault in a board at the pumping station) was responsible for twoelectrical faults and two ﬁres in Unit 3, which started almost simultaneously
in different, geographically separate ﬁre compartments;– an electrical cable ﬁre in 2004, caused by overheating in an opening between theturbine hall and the electrical building of Unit 2 at the Cattenom power plant;– in 2012 an oil ﬁre broke out in the reactor coolant pump of the reactor building for
Unit 2 at the Penly nuclear power plant.
These examples demonstrate the vital importance of considering ﬁre risks.
Fire is taken into account in the design of pressurized water reactors as both aninternal and an external hazard. As a general rule, hazards of this kind should not lead toreactor accidents or endanger the operation of safety systems designed to manage
accidents. As part of the defense in depth principle and the deterministic safety approach,controlling ﬁre risk relies on preventive measures – particularly limiting the heating loadin buildings –, on the detection of any ﬁres that do start, and on measures to limit theconsequences of ﬁres, particularly by means of the compartmentation56 of buildings andthe installation of extinguishing systems. Moreover, the overall risk of core melt in which
an internal ﬁre is the originating event is assessed as part of special level 1 probabilistic
safety assessments ("PSA Fire" developed by EDF and IRSN), which take account of
scenarios where compartmentation elements fail. This type of assessment is all the more
important given that reactor operation can reveal compartmentation anomalies or nonconformities (openings not properly sealed, etc.).
A good knowledge of all the phenomena that can come into play when ﬁre breaks out
in a room is therefore necessary: the heat level in the room, the increase in gas pressure in
the room, the production of (burned or unburned) hot gases and smoke, transfer to otherrooms (particularly if the compartmentation fails), etc.
Analysis of data from the OECD Fire database [1] shows that the majority of ﬁres(around 50%) are caused by electrical sources. Electrical and electronics cabinets are oneof the biggest contributors, causing more than 10% of ﬁres, regardless of type (low,56.

Nuclear installations are designed so that ﬁre can be contained within a deﬁned perimeter and that
inside this perimeter the consequences of the ﬁre can be controlled. The perimeter is delimited byphysical walls, or may even be separated by sufﬁcient space. For the safety demonstration, it isassumed that all equipment within a perimeter on ﬁre has been lost (i.e. is unavailable or has
failed). The layout of the buildings, the delimitation of these perimeters and the installation of theequipment are all designed so that, if the equipment in the train of a redundant system is affectedby ﬁre, the equipment in the other trains cannot be affected by the ﬁre.

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Figure 7.1 Example of a cableway conﬁguration in an installation. @ Georges Goué/IRSN.

medium or high voltage). Electrical cabinets and cableways (Figure 7.1) are therefore of
speciﬁc interest as regards research on ﬁre risks, for several reasons:– they can cause ﬁres to start;– they can help ﬁres to spread;– conversely, they can also be "targets" if a ﬁre breaks out, so their vulnerability
should be reduced as far as possible if they are important for reactor safety.
Most of the research on ﬁre aims to assess the characteristics of the ﬁres themselves,particularly estimating heat output, and the consequences for an installation. The
consequences of particular interest are the effects of gas pressure and temperaturein rooms, transfers of gases (produced by combustion or unburned) and of soot between
rooms, through doors or openings, or through ventilation systems, and the effects ofthese transfers on containment systems and electrical equipment.
Studying the development of ﬁre inside a nuclear installation and its consequences
for equipment important for safety is particularly complex because of the conﬁnement of
the rooms and the presence of mechanical ventilation. The oxygen concentration in aroom on ﬁre decreases then stabilizes, generally at a value where the ﬂow of oxygen
being consumed by the ﬁre is balanced by the ﬂow being drawn in by the ventilation. The
degree of conﬁnement of installations, the characteristics of ventilation systems (airrenewal rate in the rooms, air ﬂow resistance) and how they behave in a ﬁre (shutdown,
closure of ﬁre dampers) are decisive in determining the heat output of the ﬁre, how longit burns for and how it spreads to other ﬁre sources, or even to other rooms.
Since the early 1980s, OECD/NEA/CSNI documents [2, 3, 4] have reported international concerns about the safety of power reactors and evolving knowledge of ﬁre risks

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Current state of research on pressurized water reactor safety(including as part of the development of probabilistic safety assessments). Since 2006,
IRSN has run programs, particularly the international PRISME57 projects guided by the
OECD/NEA, to improve knowledge of ﬁres in conﬁned, ventilated spaces representativeof nuclear installations.

7.2. Organizations involved in research on fireResearch on ﬁre is not unique to the nuclear sector. In France, there are manydifferent organizations conducting research and development related to ﬁre. In particular, the following four university laboratories are helping to run CNRS' ﬁre researchgroup:– IUSTI58 (UMR59 6595) in Marseille (compartment ﬁres and forest ﬁres),– P-PRIME (UMR 9028) in Poitiers (combustion of solids, ﬁre and smoke in open andconﬁned spaces),– CORIA60 (UMR 6617) in Rouen (metrology of soot),– LEMTA (UMR 7563) in Nancy (thermal radiation, fuel emissivity measurements).
IRSN, EDF and the French General Directorate for Armament (DGA) are also carryingout research in this ﬁeld (there are similarities between ﬁres on ships and submarines andnuclear ﬁres), as are French technical centers such as CNPP61, CTICM62, INERIS63,
CERIB64 and CSTB65, which are conducting technological tests, sometimes on a large
scale, particularly for EDF.
The research begun more than 20 years ago by IPSN, then IRSN, concerns ﬁres ofinternal origin in conﬁned, ventilated rooms similar to those found at fuel cycle
laboratories and plants (glove boxes, etc.) and in nuclear reactors. However, we will
limit the following discussion to research looking speciﬁcally at nuclear reactors.
The experimental resources available to IRSN enable it to run a complete study of aﬁre in two stages: in the ﬁrst stage, its main characteristics (heat output of the ﬁre,combustion heat, pyrolysis rate, combustion products, etc.) are determined in an open
atmosphere in a device known as a cone calorimeter. Then, in the second stage, tests are
run in ventilated rooms representative of those found in nuclear installations. These tests
are used to assess the effect of conﬁnement and mechanical ventilation on thedevelopment of a ﬁre.

57.
58.
59.
60.
61.
62.
63.
64.
65.

Spread of a Fire for Multi-Room Elementary Scenarios.
French University Institute for Industrial Thermal Systems.
Joint Research Unit.
French Aerothermochemistry Research Complex.
French National Center for Prevention and Protection.
French Industrial Technology Center for Construction in Metal.
French National Institute for the Industrial Environment and Risks.
French Concrete Industry Study and Research Center.
French Construction Science and Technology Center.

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IRSN is conducting or has conducted research with various partners: industrial
partners (AREVA, EDF, ENGIE-Tractebel Engineering, Vattenfall, etc.), universities(Aix-Marseille, Rouen, Edinburgh, Ghent, Maryland, Lund, Aalto), research bodies(CNRS, INERIS, DGA, LNE66, etc.) and international organizations (BelV67, GRS68,
HSE69, NRA70, VTT71, CNSC72, CSN73, etc.).
In addition, ETIC74, a "virtual" laboratory for studying ﬁres in conﬁned spaces, wasestablished in 2010 as a joint venture between IRSN and IUSTI (University Institute forIndustrial Thermal Systems, a joint research unit CNRS/Universities of Provence and of
the Mediterranean).

7.3. Research facilities, simulation tools
At Cadarache IRSN has the GALAXIE experimental platform, which offers experimental facilities of various capacities. The platform was originally built to conduct
research in the 1970s to 1990s on sodium ﬁres, sodium being a coolant used for the
PHENIX and SUPERPHENIX fast neutron reactors. The GALAXIE test facilities were
modiﬁed in the late 1990s and new experimental equipment was added from 2000 toadapt it for research into conventional ﬁres in fuel cycle installations and France's nuclear
power reactor ﬂeet.
The GALAXIE platform consists of:– DANAIDES75, an installation for performing analytical tests on the separate andcombined effects of heat and soot on the operation of different electrical devices(relays, circuit-breakers, etc.);– a 0.3 MW cone calorimeter for measuring the combustion heat of different
materials in open air, and a radiant panel enabling equipment to be subjected to aﬁxed ﬂow of heat, in order to study its degradation; together the devices
constitute the CARINEA facility;– a larger-scale 3 MW cone calorimeter in the SATURNE tower (2000 m3), forstudying the combustion in open air of equipment from nuclear power plants,
such as electrical cabinets and electrical cableways;

66.
67.

French National Laboratory for Metrology and Testing.
Belgian Federal Agency for Nuclear and Radiological Inspections of nuclear installations (hospitals,
universities, radiological facilities, etc.).
68. Gesellschaft für Anlagen – und Reaktorsicherheit (reactor safety organization, Germany).
69. Health and Safety Executive (UK).
70. Nuclear Regulation Authority (Japan).
71. Technical Research Center (Finland).
72. Canadian Nuclear Safety Commission.
73. Consejo de Seguridad Nuclear (Spain).
74. Laboratory for the Study of Fire in a Contained Environment.
75. Analytical Equipment for Studying Electrical Malfunctions caused by Soot during a Fire.

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Current state of research on pressurized water reactor safety– a controlled-atmosphere calorimeter, CADUCEE, for studying the effect of oxygen
depletion on the combustion of different types of fuel, the heat radiated and sootproduction;– the 400 m3 PLUTON box connected to a ventilation system, which can be used formaking large ﬁres (up to 5 MW) with different ventilation conﬁgurations. The
HYDRA device (2.4 m high, 3.6 m long and 2.4 m wide) is currently installed in thisbox and can be used to study soot movements using laser velocimetry throughthe opening of a door for different ventilation conﬁgurations. Small-scale devices
such as NYX and STYX can also be installed in the box; they are designed forstudying ﬂows of smoke through openings or doors;– the DIVA device consists of three rooms of 120 m3, a 150 m3 corridor and a ﬁrstﬂoor room of 170 m3, connected to a ventilation system with variable conﬁgurations; the device can withstand negative pressures and overpressures within therange 100 hPa to + 520 hPa.
Figures 7.2-a and 7.2-b show the how the test facilities ﬁt into the different topics
being studied; Figure 7.2-c shows the DIVA facility.
All these devices are ﬁtted with signiﬁcant amounts of instrumentation (up to 800
measuring channels in the case of DIVA) for measuring the main characteristics of ﬁres
(temperature, pressure, gas concentrations [combustion and pyrolysis products], sootconcentrations, total and radiative ﬂows to the walls) and to take samples to be analyzed
after the test (soot composition and granulometry). Video recordings are also madeduring the tests.

Figure 7.2-a Positioning of the different types of ﬁre characterization test in relation to the test facilities.
@ Laurence Rigollet/IRSN.

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Figure 7.2-b Positioning of the different types of "target" behavior test in relation to the test facilities.
@ Laurence Rigollet/IRSN.

As in other nuclear safety ﬁelds, assessing the consequences of a ﬁre in a nuclear
reactor requires the use of simulation codes incorporating models developed andvalidated on the basis of tests. Their capacity to simulate real ﬁres in the conﬁgurations
found in nuclear installations is veriﬁed by means of comparisons with large-scale tests
performed in facilities reproducing these conﬁgurations (conﬁnement, ventilation) as
closely as possible. Obviously these tools are essential for analyzing the risks associatedwith ﬁre and for studying ﬁre scenarios to support IRSN's development of its own "PSA
Fire" probabilistic safety assessments for France's nuclear power plant reactors.

Figure 7.2-c The DIVA installation. © IRSN.

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Figure 7.3 Three-dimensional temperature mapping resulting from the simulation of a test at the DIVA
facility. © IRSN.

IRSN is developing and validating two types of simulation tool:– the SYLVIA76 code, which models the burning room as two homogeneous zones
with a boundary between them (a ﬂat horizontal surface) that moves over time.
The rooms are connected to one another by doors (possibly ﬁre doors), which areopen or closed (leaks are modeled), and by the ventilation system. The ventilation
system can be modelled in its entirety, with ﬁre dampers, HEPA77 ﬁlters, control
dampers, fans, etc. Correlations of the mass and heat exchange between thezones, ﬂames and walls complete the mass and energy balance equations for eachzone. Because of its short calculation times, this code is used by IRSN for studies insupport of its safety expertise and for its probabilistic safety assessments of ﬁre;– the ISIS (CFD-type) code, which models three-dimensional, non-stationary,marginally compressible, turbulent, reactive or chemically inert ﬂow ﬁelds; it
can be used to calculate combustion, thermal transfers and soot transport in large
rooms, either ventilated naturally or conﬁned and mechanically ventilated. Only
the intake and extraction branches of the ventilation system are modeled (seeFigure 7.3 for an illustration of the results achieved with ISIS).
IRSN has coupled the SYLVIA and ISIS codes together, giving it the beneﬁt in a single
tool of the precision of ISIS, which performs 3D simulations of ﬁres in rooms, and SYLVIA's
ability to describe a whole installation with a complete ventilation system connecting all
the rooms. No other coupling of this kind exists anywhere else in the world.

7.4. The main research programs and theircontributions
Major progress has been made since 1990 in terms of knowledge of conﬁned,
ventilated ﬁres and how to model them, in particular due to research programs run
76.
77.

Software System for Analysing the Ventilation of a Fire and Airborne Contamination.
High Efﬁciency Particulate Air.

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in collaboration with AREVA on fuel reprocessing plant safety (FLIP78 program on solvent
ﬁres and PICSEL79 program on electrical cabinet ﬁres). Further progress has also been
made through the international PRISME and PRISME 2 programs (2006–2011 and 2011–2016), run by IRSN under the aegis of the OECD/NEA, focusing on nuclear reactor safety.
These programs have been used to validate the ISIS and SYLVIA codes, which are able
to a sufﬁcient degree of precision of simulating scenarios involving ﬁre in a conﬁned
environment with managed ventilation (shutdown of the air supply followed by air
extraction after a certain period of time).
During the large-scale tests, signiﬁcant variations in gas pressure in the rooms(overpressure when the ﬁre starts, negative pressure when it is extinguished and, in someconﬁgurations, wide oscillations) have been observed. These gas pressure variations cansubject compartmentation systems (ﬁre doors and dampers) to stresses beyond thosethey were designed for, enabling ﬁre to spread within an installation. This gas pressure
variation is linked to the ﬁre's conﬁnement and the ventilation system's resistivity. The
oscillations and instability of combustion are due to the under-oxygenation of the ﬁre,leading to pyrolysis gas combustion. These phenomena could also occur in a nuclear
power plant. The tests have also identiﬁed the effect of soot on the operation of electricalor electronic equipment.
The PICSEL program run in collaboration with AREVA between 2004 and 2011 studied
ﬁres in electrical cabinets and their consequences in experiments in the SATURNE facility(ﬁres in open atmospheres) and the DIVA facility (ﬁres in conﬁned, ventilated rooms),
enabling them to be modeled. These results, obtained during a program looking more
speciﬁcally at the conﬁgurations of fuel cycle installations, can be transposed to nuclear
reactors.
One of the ﬁndings of this program on electrical cabinet ﬁres (Figure 7.4) concerns
the heat output of these complex ﬁres (involving multiple components and manydifferent types of material). In particular, the tests showed that the heat output ofan electrical cabinet ﬁre with the cabinet doors open was 10 times greater than that of anelectrical cabinet ﬁre in which the doors were closed. This difference is due to the factthat soot clogs the door vents, preventing they entry of oxygen and therefore combustion, in the electrical cabinet with its doors closed. These are the ﬁrst electrical cabinet
ﬁre tests run in a conﬁned, ventilated atmosphere; a few tests had been done previously,at the Sandia National Laboratories (SNL) in 1987 and by VTT in 1994, but they had only
evaluated the heat output of this type of ﬁre in an open atmosphere.
The PRISME program consisted of 24 tests run in the DIVA facility, plus a further
13 more analytical tests in SATURNE. It produced results on the propagation of smokeand hot gases in the adjacent rooms to a burning one, on how long it would take for thecables in the burning room to malfunction, on how the ﬁre dampers in the ventilation
system would work and on how the ventilation system should be managed to prevent
pressure effects that might damage the compartmentation system. The PRISME program
78.
79.

Interaction of Liquid Fires with a Wall.
Propagation of Solid Fuel Fires in Laboratories and Factories.

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Figure 7.4 Fire in a type of electrical cabinet used in fuel cycle installations, tested at the SATURNE
facility as part of the PICSEL program. © Florent-Frédéric Vigroux/IRSN.

provided a better understanding of the effect of ventilation (and therefore underoxygenation of the ﬁre) on the heat output from a ﬁre breaking out in a conﬁned,
ventilated room, and particularly of how long the ﬁre would last. Depending on the air
renewal rate, the ﬁre could rapidly go out because of a decrease in the oxygen
concentration in the burning room. But the tests run under thePRISME program showedthat an equilibrium could be reached between the air coming from the ventilation air
supply and the heat output of the ﬁre: in this case, any combustible material would be
consumed more slowly than in the open air (lower heat output) and more completely. For
example, the same ﬁre could burn 2.5 times longer in a ventilated room with an hourly air
renewal rate of 4.7 than in open air. Correlations and analytical models of pyrolysis in a
conﬁned, ventilated environment (under-oxygenated ﬁres) have been developed andvalidated. These models correct the heat output obtained for a ﬁre in open air (measured
under the SATURNE hood) by a factor taking into account the under-oxygenation rate ofa ﬁre in a conﬁned space.
The PRISME program was also used to quantify the effect on smoke propagation of"mixed" convection, which combines the forced convection created by ventilation with
the natural convection induced by the steep vertical temperature gradient of the gaseousatmosphere in a room ("fresh" air on the ground and hot smoke at the ceiling). The
mechanical ventilation of the burning room can signiﬁcantly alter the gas ﬂows thatAnglais_Book

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would naturally develop at an open door between two rooms. Depending on the way theventilation system is adjusted, the mechanical ventilation helps to cause an imbalance inthe ﬂows entering and leaving the burning room, and to change the position of theneutral plane (the height where ﬂow rates are zero due to ﬂow reversal) at the door.
All the data collected during this program has been used to assess the ability ofcomputer codes to simulate different ﬁre scenarios. The new models, particularly for
pyrolysis, have been integrated into the various partners' codes and validated by meansof experimental data obtained during this program.
The PRISME 2 project (2011–2016) was launched for the purpose of studying ﬁrerelated themes to complement those studied in the ﬁrst project (PRISME), such as smoke
propagation between rooms on top of one another, the ﬁre propagation between
cableways and the performance of sprinkler systems. The PRISME 2 program consistsof 22 tests at the DIVA facility and approximately 20 tests in the SATURNE calorimeter.
The tests, involving a simple ﬁre (a pool of liquid) in a ventilated room which is either
closed or is connected to an upper room by an opening, revealed a phenomenon of highamplitude and very low frequency oscillations, of the thermodynamic variables in thefacility, which suggest combustion instability that is highly correlated with the room's
ventilation and with oxygen transfer from the air supply vent to the combustion zone.
Understanding these oscillations required further experimental research to identify the
conﬁguration and scenario parameters causing them. Results have been obtained for thetransfer of smoke in these conﬁgurations in relatively stable combustion regimes (i.e.
without wide oscillations), providing new data for validating the correlations for smoke
transfer through a horizontal opening between closed, ventilated rooms and forvalidating detailed CFD-type codes.
Several types of cable were tested during the tests in DIVA (following theircharacterization in an open atmosphere under the SATURNE hood): cables with ahalogenated ﬂame retardant and cables with a non-halogenated ﬂame retardant(components reducing the spread of a ﬁre affecting an electrical cable). The integral
tests in DIVA revealed that the propagation of ﬁre along ﬁve horizontal cableways orfrom an electrical cabinet to the cableways above it depended heavily on the type ofcables and the air renewal rate in the room on ﬁre. The intensity and duration of the ﬁre in
a conﬁned, ventilated environment could not be correlated with the degree of ﬁreresistance of the cables. So in certain conditions, supposedly ﬁre retardant cables burned
up completely in a ﬁre lasting a long time whereas supposedly non-ﬁre retardant cablescaused the ﬁre to go out early on due to a lack of oxygen (in this case, the mass of burned
cable was low as a result of the simultaneous consumption of oxygen by the ﬁre in thecabinet and the ﬁre in the electrical cables). This behavior, which may seem paradoxical,
is explained simply by the fact that if the runaway of a ﬁre in a conﬁned environment is
too sudden, this can cause a drop in the oxygen concentration, which is not compensatedfor by the air supplied by the ventilation. This produces conditions in which combustion at
the ﬂame front stops. These tests also revealed that the sudden reignition can occur(slow deﬂagration but with a signiﬁcant peak in gas pressure) of unburned gases thathave accumulated in the room. The results are new and original, and underline theimportance and value of continuing to study the combustion of cableways in conﬁned,

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ventilated facilities for the development of combustion models (simpliﬁed, such ascorrelations, and more detailed, such as porous environments for CFD codes), taking
account of the speciﬁc characteristics of the cableways in question (cable type, size,
spacing between cableways, etc.).
The PRISME 2 program also provided data on the use of extinguishing systems. During
these sprinkler tests, contact between the droplets and the ﬂame zone above the tray ofburning oil and walls was avoided so that the effect of the sprinkler on controlling the ﬁre
could be studied without directly extinguishing it and avoiding the edge effects
associated with the walls. This data can be used to validate simulation codes and toassess the codes' ability to simulate the phenomena identiﬁed during the tests, e.g.
cooling of the gases in the room due to contact with the water droplets and theirvaporization, vigorous mixing of the gases in the room producing a uniform gas
distribution and therefore temperature and oxygen concentration and a signiﬁcant
increase in combustion of the burning liquid (measured by loss of mass by the poolof liquid over time).
Industrial protection aimed at preventing the spread of ﬁre along a cableway was
installed and tested under the SATURNE hood on a set of three cableways, one on top of
the other (Figure 7.5). The ﬁre did not cross the barrier created by this protection, but thedemonstration at this stage is not fully established because there was some pyrolysis of
the cables downstream of the protection and the ﬂames could have crossed the barriers if
there had been more cableways present or if the ﬁre had been in a conﬁned environment
so that the cables downstream of the protection had been preheated.
The speed of propagation along inclined cableways was measured and compared withthe speed along horizontal cableways.

Figure 7.5 An experiment involving a ﬁre in a set of cableways, one on top of the other. © FlorentFrédéric Vigroux/IRSN.

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A test of ﬁre propagation from an electrical cabinet with open doors to adjacentcabinets and to cableways passing above the electrical cabinets was performed in the
DIVA facility. The test revealed that the ﬁre spread to one of the cabinets adjacent to theburning cabinet but that the spread of the ﬁre was less intense than during a similar test
in a previous campaign, without cabinets adjacent to the electrical burning cabinet.
The results of the tests are being analyzed by the international bodies involved in the
PRISME and PRISME 2 programs (Germany, Belgium, Canada, Spain, Finland, France(IRSN, EDF, DGA, Marseille University), Japan, Sweden, the UK, South Korea, the USA,and the Netherlands, though the last three only for PRISME). Code benchmark exercises
have been organized by IRSN in the context of a working group related to the PRISME andPRISME 2 programs: the various partners have compared the results of ﬁre simulations
with the experimental data. A sensitivity study carried out with six different simulation
codes, involving six input parameters (heat release rate of the ﬁre, radiative fraction of
the ﬂame, thermal properties of the walls, etc.) revealed that the ﬁre's heat release rate isalways the dominant parameter, showing that efforts to improve its modeling need tocontinue.
In another example of assessment of the relevance of simulation codes, the CNRS ﬁreresearch group organized a benchmark exercise of multi-dimensional computer codesthat use ﬁelds (ISIS code mentioned above, SATURNE code developed by EDF, and FDS80
code developed by NIST81 in the United States), based on a ﬁre test in a hotel room. The
exercise showed that results produced by the different codes were very widely dispersed
as regards the values measured during the test. Experts attributed this dispersal inparticular to the difﬁculty of choosing an appropriate combustion model for calculating
the ﬁre's instantaneous heat release rate.
These results demonstrate the need to continue the research in order to achieve
better modeling of the characteristics of ﬁres by developing pyrolysis and combustion
models in under-oxygenated and complex ﬁre conditions. It is also necessary to increaseknowledge particularly of the instability of combustion as a result of the production andignition of unburned gases, as observed during the study of cable ﬁres in an underoxygenated environment82 in the PRISME 2 program, and of about the sudden pressure
variations caused by this, so that the predictive capabilities of the simulation tools can be
improved. The layout of cableways (horizontal, vertical, mixed, near a wall, etc.) and thecables they carry (loose, tight, etc.), as well as the type of cables, constitute a set ofparameters with effects on ﬂammability and the propagation of ﬁre that are complex tomodel and require new experimental data. The effectiveness of measures to protect
cables from catching ﬁre (wrapping in ﬁre-resistant tape, enclosure, etc.) is also worth
studying experimentally. The ﬁrst tests on ﬁre-resistant taping were run during the
PRISME 2 program. They acquired data on the ﬁre resistance of this taping under realistic
conditions.
80.
81.
82.

Fire Dynamics Simulator.
National Institute of Standards and Technology.
Situation arising as a result of a conﬁned, ventilated atmosphere; the phenomenon can also occur
as a result of the accidental heating of cables, e.g. through the Joule effect, causing the release ofpyrolysis gases.

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In addition to these research programs using integral tests, academic research is also
being carried out on the modeling of pyrolysis, a phenomenon strongly linked toconditions around the ﬁre, soot production and radiation from the ﬂames. The models,developed particularly at the ETIC joint research laboratory, have been validated in
medium-scale tests, especially in the CADUCEE controlled-atmosphere calorimeter. For
this validation, appropriate instrumentation was developed to cover different aspects,
particularly ﬂowmeters and non-intrusive methods using a laser (LII83, PIV84, etc.). Laser
velocimetry techniques and PIV (visualization of the movement of particles in a movingﬂuid) can be used to ﬁnd out the velocity ﬁelds of ﬂows and to measure the local
intensity of turbulence. These measurements are essential to validate the airﬂow
calculations and therefore the calculations for smoke and heat transport within therooms. The PIV technique has been used to measure the velocity ﬁelds of ﬂows through adoor in the HYDRA device and at an opening in the STYX device.
Research being conducted on ﬁres is looking at the effects of a ﬁre on equipment
important for safety, in particular compartmentation components such as doors and ﬁredampers, and electronic and electrical equipment.
As far as the compartmentation components are concerned, airﬂow tests are
performed in IRSN's STARMANIA facility at the Saclay research center. In particular
this allows measurement of the effects of pressure on ﬁre protection components (doors,
dampers).
Electrical equipment malfunctions due to heat and soot were studied during theanalytical test programs CATHODE (2007–2009) in an oven named SCIROCCO, then
DELTA (2014–2015) using the DANAIDES device (Figure 7.6); the criteria deduced fromthese tests can be used to predict the operational limits of the tested equipment85 whena ﬁre breaks out in the room where it is located. The tests have identiﬁed three generalzones (in terms of temperature and soot concentration): one zone where the equipment
remains operational, one zone where malfunctions occur that are reversible, and one
zone where the malfunctions are irreversible. During tests on circuit boards, the damage
from soot appears to be signiﬁcant from a soot concentration of 1.5 g/m3, reducing thezone in which the equipment functions properly.
EDF is also performing analytical tests on the MILONGA experimental platform at the
Chatou center. In particular these tests consist of taking measurements using a smallcone calorimeter coupled with a Fourier transform infrared spectrometer and anelectrical low pressure impactor, of the gases and soot released by the combustionof different materials, particularly those used in cables. Malfunctions of electrical andelectronic equipment exposed to heat and smoke are also being studied in the MAFFé

83.

84.
85.

Laser Induced Incandescence. LII is used to measure soot concentration and to validate sootproduction models in an under-oxygenated environment. This measurement technique is not
currently used in ﬁre tests because it is the subject of academic research.
Particle Image Velocimetry.
Tested equipment: D125 circuit-breakers, MICOM P921 and VIGIRACK A326E electronic relays,
LOREME.

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Figure 7.6 The DANAIDES facility used for the DELTA program. @ Marc Piller/IRSN.

furnace. The results of these tests are used by EDF in its safety demonstrations, subject toassessment by IRSN.

References[1] W. Werner, A. Angner, M. Röwekamp, J. Gauvain, The OECD ﬁre database –conclusions from phase 2 and outlook. In: 20th International Conference on SMIRT,
11th International Post Conference Seminar on Fire Safety in Nuclear Power Plants
and Installations, Helsinki (Finland), 2009.
[2] Fire Risk Analysis, Fire Simulation, Fire Spreading and Impact of Smoke and Heat on
Instrumentation Electronics – State-Of-the-Art Report (SOAR), Report NEA/CSNI/R(1999)27.
[3] Proceedings of the specialist meeting on ﬁre protection systems in nuclear power
plants – GRS, Cologne, Germany, Report NEA/CSNI/R(1994)9, December 1993.
[4] Proceedings of the CSNI specialist meeting on interaction of ﬁre and explosion withventilation systems in nuclear facilities – Los Alamos National Laboratories, New
Mexico, USA, Report NEA/CSNI-83, April 1983.

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Chapter 8
Research on Hazards associatedto Natural Events

Taking natural events into account in the design of nuclear installations and during
their operation is essential for their safety. The general objective is that the safetyfunctions associated with their structures, systems and components are not harmed bysuch events. Scientiﬁc knowledge in the earth sciences ﬁeld (such as geology, hydrogeology, seismology, etc.) forms the basis for assessing the risks associated with natural
events, i.e. the characteristics to be taken into account in the design or testing of
installations (seismic movements of the ground, water level, etc.). Improving installation
safety particularly involves working to improve the reliability of our ability to grasp themechanisms at work in natural phenomena and their effects. This is the general objectiveof the research initiatives run in the last forty years or so by IPSN (then IRSN), particularly
as regards seismic risks.
Natural events, the subject of research and development by IPSN (then IRSN) either
on its own or in the context of collaborations, are mainly earthquakes, ﬂoods, andclimate-related hazards. This work aims to develop, improve and validate the tools(including databases) and methods used so that the hazards and their consequences canbe determined more accurately. Field studies, one-off experiments and ongoing instrumental measurements (seismographic network and GPS86) complement the theoreticalresearch and modeling.
As the ﬂood at the Le Blayais nuclear power plant site in December 1999 and theaccident at the Fukushima Daiichi nuclear power plant in March 2011 showed, nuclear
86.

Global Positioning System.

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safety can only be guaranteed at the cost of adequate protection of facilities against all
types of natural events, which means that the natural events that could strike them need
to be evaluated correctly, particularly those associated with seismic events and theoccurrence of ﬂoods. In Japan, the Tohoku earthquake and the resulting tsunami, and also
the earthquake in July 2007 at Chūetsu-oki not far from the Kashiwasaki-Kariwa nuclear
power plant87, and to a lesser extent the earthquake in Virginia (USA) in August 2011,
around 18 km from the North Anna nuclear power plant88, highlighted the importance of
the knowledge and methods on which the design of nuclear installations is based and but
also their limits. There is a broad international consensus on the need to extend theknowledge and assessment of natural events that can seriously affect nuclear sites – asevidenced by the conclusions of an international conference held by the International
Atomic Energy Agency (IAEA) in March 2012. This is particularly true in France, where
improving the way seismic risks and ﬂood risks are taken into account is one of thepriorities set by IRSN, ASN and the public authorities as a result of feedback from the
Fukushima accident, reﬂected in the call for RSNR projects issued by the French National
Research Agency (ANR) with funding from the Investment in the Future Program (PIA). In
this new context, there are two issues for research and development when it comes todeﬁning extreme hazards. One is obtaining knowledge of the phenomena (including
historical indices), and the other is ﬁnding methods for taking account of the associateduncertainties and reaching a deﬁnition of the maximum risks to be taken into accountwhen designing nuclear installations.

8.1. Earthquakes
Seismic hazard assessment practice in France for installations at special risk (nuclear
and chemical installations, dams, etc.) has historically been based on a deterministic
approach. This approach estimates the acceleration that would be produced by thestrongest historical earthquake as close as possible to the site, with an added safetymargin89. This margin is deﬁned by increasing the magnitude90 of the earthquake inquestion by a half-degree.
In general, seismic hazard is assessed by identifying possible aggressive earthquake"sources" and characterizing their potentiality. In low-seismicity areas where little is
known about faults, the deterministic approach considers regions as possible sources ofearthquakes. These regions are deﬁned according to their geological or seismic homogeneity. In regions with slightly higher seismicity (e.g. Provence, Alsace, Pyrénées), moreis usually known about faults and, most importantly, signiﬁcant earthquakes can be

87.
88.
89.
90.

This nuclear power plant was not damaged.
Shallow (6 km deep) earthquake of moderate magnitude (5.8), which was unexpected given thehistorical seismicity of the part of Virginia concerned.
The strongest historical earthquake is known as the MPE (Maximum Probable Earthquake); with themargin added it is known as the SME (Seismic Margin Earthquake).
The magnitude of an earthquake is a measurement of the amount of energy released at the
hypocenter of the earthquake. The magnitudes usually used these days are expressed as moment
magnitudes (denoted Mw).

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associated with these source faults (by their cartographic position). Nevertheless, seismic
hazard assessment is part of a larger process requiring answers to the following questions:– how big are the faults capable of producing earthquakes? The magnitude of anearthquake and, therefore, the danger it represents, is directly related to thisparameter. The size and extent of an active fault can be determined by surface
mapping, locating microseismic events, and depth imaging. This parameter canalso be assessed by characterizing major "prehistoric" earthquakes;– what is the deformation rate of the region affected by the faults considered? Thisis an important point for determining the periodicity of major earthquakes. In
addition to dating geological markers, spatial geodesics measurements (GPS,InSar91) also provide useful data for assessing the deformation rate.
The seismic hazard associated with a system of active faults can be assessed byconducting research and development to ﬁnd answers to the above questions.
IPSN, and later IRSN, pioneered many aspects of seismic risk research, such as
studying historical seismicity with the SisFrance database, analyzing deep geophysicaldata, paleoseismology, and seismic mapping and monitoring of active faults, in particularwith studies of the Moyenne Durance fault system, characterized by regular seismic
activity (reference document [1], 2007). The results obtained by these studies have beenused to characterize the microseismicity of the fault system and to produce a more
precise assessment of seismic risk (particularly in terms of magnitude). This will be
particularly useful for assessing or reassessing the seismic risk of the nuclear installationsat CEA's center in Cadarache.
Over the years, the research programs run have enabled IPSN (then IRSN) to develop
a network of scientiﬁc partnerships both in France and abroad. In the last few years,
IRSN's accumulated expertise has enabled it to beneﬁt ANR funding and has led torequests for assistance with site studies from a number of countries wishing to develop a
nuclear power generation ﬂeet (or strengthen the ﬂeet they already have). Its activities
have also gradually contributed to the deﬁnition of methods applicable to basic nuclear
installations, notably those laid down in the French Basic Safety Rule 2001-01. The
knowledge and data produced also feed into the expert appraisals performed by IRSN in
the frame of safety reviews led by the operators. They have also been widely used in thecomplementary safety evaluation (ECS) carried out in the wake of the Fukushima Daiichi
accident, in the presentation of the French report during the European peer review
process, and in reports on the "hard core" being introduced particularly at French nuclear
power plants (equipment that must withstand more severe hazards than the ones forwhich the installations were designed). As a result of the ENSREG92 recommendations, aprobabilistic approach to assessing risks with a low probability of occurrence has been
introduced in addition to the deterministic method.
IRSN has also used the knowledge obtained on seismic risk in its role as expert
supporting the French Ministry of the Environment during preparation of the seismic risk
91.
92.

Interferometric Synthetic aperture radar.
European Nuclear Safety Regulators Group.

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mapping of metropolitan France, which came into force on May 1, 2011. This new"zoning", which covers conventional buildings and "special risk" installations (apart from
nuclear installations and dams), is based not only on historic seismicity, which wasessentially the case with the previous zoning in 1991, but also on a probabilistic
assessment of seismic risk (estimation of the seismic movement likely to happen or
be exceeded according to a ﬁxed probability, typically 10%, over a time period of
50 years). However, to comply with the recommendations of French basic safety rules
2001-01, for nuclear installations IRSN regularly updates a "seismotectonic zoning",
mainly to incorporate knowledge acquired through research in general. Also in relation to
the recommendations in French basic seafety rules 2001-01, IRSN was involved in thecreation of a national database93 of indices of strong earthquakes that occurred in France
a very long time ago (several thousands to several tens of thousands of years), known as
paleoearthquakes. In the context of a memorandum of understanding with French
General Directorate for Risk Prevention (DGPR94), IRSN also recently developed a
database of active faults in a 50 km perimeter around any nuclear installations.
IRSN's research on seismic risk is also looking at:– identifying and understanding how faults function (in France and abroad). The
work aims to identify faults likely to cause earthquakes (active faults) and tocharacterize them (location, geometry, deformation rate). It is based on theanalysis of records and on the study of historical seismicity and paleoseismic
indices. Characterization of the activity of a fault (i.e. its ability to cause anearthquake) and the discovery of new active faults in France are necessary to form
the basis, conﬁrm or improve assessments of seismic risk. This research combines
various different approaches (study of morphological anomalies using digitalterrain models, Spot images, spatial imaging methods, geological studies ofterrain, and trenching studies to ﬁnd traces of paleoearthquakes). Since thestudy of the Moyenne Durance fault system mentioned above, IRSN's research
has continued in France on the Vuache fault (Jura), which in particular was
responsible for the earthquake that caused damage in Annecy in 1996, and onother faults in the Alpine Foreland, the Rhine Graben and the Channel and abroad(Ecuador, India, Spain);– predicting potential seismic movements: this work is done using empiricalapproaches (based on observations) and numerical simulations;– estimating the seismic response of the ground speciﬁc to a site (which comesunder the term "site effects" [Figure 8.1]). The research has revealed the particularimportance of "site effects", which can signiﬁcantly amplify seismic movements inthe frequency range of interest for concrete structures95. The research done in thisﬁeld in particular requires the acquisition of geotechnical and geological dataat the sites and the use of complex modeling that takes account in particular of
93.
94.
95.

See the website www.neopal.net.
French civil service department for risk prevention.
A few Hertz. "Site effects" can reduce ground movements at the highest frequencies and amplify
them at lower frequencies.

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Figure 8.1 Example of site effects in the lakes area of Mexico. © Jean-François Semblat.

non-linear behavior. "Site effects" refers to effects due to the type of ground on
which an installation is built (rock, sediment) and effects due to the particularconﬁguration of the site (also known as "site-speciﬁc effects", which refer for
example to sites in a sedimentary basin set into a rocky environment). The
Grenoble Basin has historically been France's test site for observing "site effects"
in deep valleys, following the drilling of a deep borehole in 1999 at IPSN's
instigation. IRSN has also worked, in collaboration with foreign institutes, in otherareas of active seismicity (Gulf of Corinth in Greece, Santiago Basin in Chile). The
current sites are in France, in the sedimentary basin of Cadarache, and in Greece,
near the town of Argostoli.
IRSN is actively represented within the main bodies involved in seismologicalresearch, such as the French Earthquake Engineering Association (AFPS) and the French
Seismographic & Geodesic Network (RESIF).
IRSN is also a partner in the SINAPS@96 project on seismicity and nuclear installations, with CEA (project coordinator), IFSTTAR97, École Centrale Paris, EDF, ENS98
Cachan, etc. SINAPS@ is a research project in which seismic risk is assessed as a whole,from the fault to concrete structures and to equipment. It aims to explore the inherent
uncertainties in assessing seismic risk and the vulnerability of structures and equipment.
The main ultimate objective is to identify, and indeed quantify, the seismic margins
resulting from the hypotheses used either when making design choices or when devising
the design strategy (conservative hypotheses, choice of materials, etc.). The main themesof the SINAPS@ project to which IRSN is contributing are:
1. Quantiﬁcation of the uncertainties associated with knowledge of the input dataand parameter hierarchy controlling the variability of seismic risk assessments inthe deterministic and probabilistic approaches.
96.
97.
98.

Earthquakes and Nuclear Facilities, Improving and Protecting Safety.
French Institute of Science and Technology for Transport, Development and Networks.
French higher education institution.

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2. The international benchmark exercise (code and data comparisons) known as
PRENOLIN99. The PRENOLIN project (2013-2015), reincorporated into SINAPS@
as a result of a request by ANR, aims to produce an approach that takes account ofground non-linearity phenomena in the "site effects" dimension of seismic risk, anapproach validated by comparison with ﬁeld observations. It should eventually bepossible to apply this approach to contexts of low and moderate seismicity, whichmeans using simulations, while making use of empirical data (recordings of small
earthquakes) for the areas concerned.
3. Numerical simulation of seismic movement close to faults. The aim of this part ofthe project is to complement the empirical prediction of seismic movement, forwhich there is a paucity of data close to faults. Numerical simulation is used to
explore the physical origins of the variability of seismic movement, which could
reduce the uncertainties due to the current lack of knowledge. Quantifying andreducing the uncertainty associated with near-ﬁeld seismic movement predictions
is a major challenge in estimating site-speciﬁc seismic risk.
4. Selection of pertinent seismic signals for assessing the earthquake resistance ofbuildings. This theme lies at the interface between seismology and civil engineering. The multiplicity of selection procedures appropriate to the risk on the one
hand, and the evolution of the methods for non-linear modeling of the behavior ofbuildings on the other, mean that the studies straddle both disciplines. The aim is
to provide information for the existing guides to be updated.
There is another research program too: the SIGMA program (Seismic Ground Motion
Assessment, 2011–2015) brought together by EDF to meet its own needs and those ofother licensees and designers (AREVA, CEA, ENEL100, etc.). The SIGMA program covers
the areas mentioned above: characterization of seismic sources, prediction of groundmotion, "site effects" and probabilistic methods. The ﬁndings and recommendations fordesigners and licensees are expected in around 2016–2017.
Lastly, one particular topic of investigation has been chosen for the seismic reviews ofFrance's nuclear power plant reactors: the identiﬁcation of active faults that could cause
ground surface or near-surface movements, commonly known as capable faults101. EDF
was asked102 to present a survey of current knowledge concerning the identiﬁcation ofcapable faults within a 25 km radius of sites103 and, where necessary, a methodology fortaking account of them when reviewing the seismic risk of sites. The treatment of this

99.
100.
101.
102.

103.

Better Prediction of Non-Linear Effects Induced by Strong Seismic Motion.
Ente Nazionale per l'Energia Elettrica.
On capable faults, see the IAEA guide SSG-1 from 2010.
Request made by ASN as part of the safety review associated with the third ten-yearly outage
program (VD3) of the 1300 MWe reactors, extended to VD4 900 MWe and VD2 “N4” (letter
CODEP-DCN 2014-051797 of December 18, 2014).
Investigation radius proposed in IAEA guide SSG-9 (near regional investigations). This guide gives
four spatial scales for progressively more detailed investigations: regional, near regional, site
vicinity and site area. The guide recommends ﬁnding evidence of capable faults going back to very
early periods, at least 10,000 years ago.

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subject prompted IRSN to launch some new research and development work withininternational groups (IAEA and INQUA104).

8.2. External flooding
As far as other natural events are concerned, the most work has been done on
external ﬂooding, with studies of the suitability of statistical methods for explaining
outlier events, extreme rainfall, the handling of heterogeneity in the statistical treatment
of data (particularly for river ﬂoods), the historical analysis of exceptional events(tsunamis on the Atlantic coast, etc.) and assessing the risk of percolation throughdikes. The knowledge acquired recently via research in this area is set out in the work inreference [2], the outcome of research by a group run by IRSN which brought together
representatives from many organizations (ANDRA105, AREVA, CEA, EDF, the French
Technical Maritime and River Study Center (CETMEF), the French National Company of
the Rhone river (CNR), the French Naval Hydrographic and Oceanographic Service(SHOM), France's national meteorological service (Météo-France), and the French
Technical and Inspection Ofﬁce for Large Dams (BETCGB). Specialists from the French
Geological Research Mining Bureau (BRGM) and the French National Center for
Agricultural Machinery, Rural Engineering, Water and Forestry (CEMAGREF) also participated in the group's activities, along with academics.
For each phenomenon (high sea levels including due to tsunamis, river ﬂoods,
rainwater and runoff, rising groundwater, dam breaches), the following were examined:– the basic data,– the physical parameters to be quantiﬁed (intensity, volume, water level, etc.),– the existing characterization methods (deterministic or statistical), identifying thelimitations of these methods,– the identiﬁcation and incorporation of uncertainties,– the inﬂuence of climate change,– the dependency between the different phenomena/events.
This report on the state of the art has served notably as the basis for a guide by ASN
for nuclear licensees, setting out recommendations for assessing and quantifying external
ﬂooding risks, and for deﬁning suitable protection measures106. The preparation of thisguide reﬂects the lessons learned from the partial ﬂooding of the Le Blayais nuclear
power plant during the December 1999 storm, prompting nuclear licensees to review thesafety of installations in view of ﬂood risks under stricter conditions than before and toincrease protection for their installations. Aside from these improvements, the report on
the state of the art [2] has allowed a detailed reﬂection process to take place based on
104.
105.
106.

International Union for Quaternary Research.
French National Radioactive Waste Management Agency.
ASN Guide No 13, Protection of Basic Nuclear Installations against external ﬂooding - Version ofJanuary 8, 2013.

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increased knowledge, ensuring external ﬂooding risks are taken into account morecomprehensively and more robustly. ASN's guide constitutes a reference text in France
not only for new nuclear installation projects but also for the ten-yearly safety reviews
for those in operation.
IRSN's ongoing work, especially in the post-Fukushima context, concerns methodological aspects such as:– statistical methods for determining extreme hydrometeorological events, paleosedimentary approaches;– modeling of ﬂows and runoff phenomena, with application to actual cases.
In addition, concerning the tsunami risk, IRSN is a partner in the TANDEM107 project(2014–2018) funded by ANR as part of RSNR research, with partners CEA, EDF, BRGM,
Ifremer108, Inria109 and Pau University. This research project aims to assess the effects ofa tsunami on the French coast, focusing particularly on the Atlantic and Channel coasts
where there are nuclear installations in operation. This project will deploy new numericalanalysis methods, which will be adapted and tested on the databases concerning the
March 11, 2011 tsunami off Tohoku in Japan. Once these methods have been validated,
they will be applied to the French coast in order to determine the impact of a tsunami ona similar scale. In the long-term, the conclusions should make it possible to produce new
guidance for assessing risks to nuclear installations.

References[1] Étude du potentiel sismique et de l’enracinement de la faille de la Moyenne Durance,
2007 scientiﬁc and technical report, IRSN, 2007.
[2] L'aléa inondation – État de l’art préalable à l’élaboration du guide inondation pour les
installations nucléaires, IRSN, Notices and reports/assessment reports/nuclear safety
series, 2013.

107.
108.
109.

Tsunami in the Atlantic and the English ChaNnel: Deﬁnition of the Effects through numerical
Modeling.
French Research Institute for Exploitation of the Sea.
French National Research Institute for the Computational Sciences.

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Chapter 9
Research on Core Melt Accidents

When the nuclear power plants with pressurized water reactors (PWRs) and boilingwater reactors (BWRs) were built, no speciﬁc design provisions were retained as regards
core melt accidents. It was not until the 1990s, with reﬂection and research on "newgeneration" reactors (Generation III, e.g. EPR, AP1000, etc.), and following the Three Mile
Island (TMI) and Chernobyl accidents, that core melt was taken into account in the designof third conﬁnement barrier components110. The objective is that the human andenvironmental consequences of a core melt accident should only lead to countermeasures "limited in space and time" (no permanent relocation of communities, nopermanent bans on the consumption of foodstuffs, etc.).
Nevertheless, from the ﬁrst deployment of PWRs and BWRs, research on reactor coremelt and its radiological consequences was begun, particularly in the United States. The
aim was to assess the risk to the public from a major accident causing radioactive releases
into the environment, at a nuclear power reactor of several hundred MWe, approximately
50 kilometers from a city (Theoretical Possibilities and Consequences of Major Accidents
in Large Nuclear Power Plants, WASH-740 in 1957 and WASH-1400 in 1975 [1]). However,it was mainly the accident at the Three Mile Island power plant in 1979, and especially thediscovery in 1985 that a large portion of the core had melted, that triggered a huge raft ofresearch and development programs globally in this ﬁeld.
After more than 30 years' research, vast progress has been made with understanding
and modeling the complex phenomena that could occur during a core melt accident.
Numerous programs run by the OECD/NEA (CSNI), the Phebus-FP (Fission Product)110.

Or components that could pose a risk to the containment, such as the main reactor coolant
system.

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program, the international ISTP program, and so on (we will come back to these) mademajor contributions to this progress. Models using the results of these programs havebeen integrated into codes for simulating these accidents and their impact on the reactor,
e.g. the ASTEC [2] code, developed in France111 by IRSN, in partnership until 2016 with its
German counterpart GRS, and used in its safety assessments and to build the models
used in level 2 probabilistic assessments112. Since 2004, under the European Commission's impetus, a research excellence network on severe accidents has been set up, known
as SARNET [3, 4]. This network, directed by IRSN, brings together the work of around
250 researchers from 44 research organizations and laboratories in 23 countries (EU andnon-EU). The network, which is now part of NUGENIA, provides a means of coordinating
the researchers' work and pooling their efforts to establish the state of the art ofknowledge, determining research priorities, setting up research projects and cooperatingon the development and validation of accident scenario computer codes, particularly the
ASTEC code (Figure 9.1), which has been made available to the network's members.
Since the early 2010s, as part of the safety reviews for the purposes of extending theservice life of power reactors to 60 years and the complementary safety evaluation (ECS)
conducted in the wake of the Fukushima Daiichi accident in Japan in March 2011, researchin the ﬁeld of severe accidents has focused more on studying physical measures toprotect the containment and limit radioactive releases. These aspects are covered atnational level by the ANR RSNR projects launched following the Fukushima Daiichi
accident, at European level as part of the research and development framework program(FP7) and Horizon 2020 (H2020) and at international level in the context of a number ofOECD projects. In particular, physical measures designed to contain the progression ofcore melt in the reactor vessel, keep the molten materials in the containment if it fails
and reduce radioactive releases through ﬁltration113 are being examined.
We will not describe here the research and development being conducted to model
the often complex thermohydraulic phenomena that lead to core uncovery114, whichwere discussed in part in the previous section. We strongly recommend that readers
wishing to understand in more detail the phenomenology of core melt accidents and therelated research should refer to IRSN's publication Nuclear power reactor core melt
accidents – State of knowledge115 published in 2013, or to the reports by the OECD or the
U.S.NRC on core melt accidents [5, 6].
111.
112.

113.
114.
115.

Other codes have been developed abroad, including MAAP (Modular Accident Analysis Program)
and MELCOR.
Level 2 PSAs are used to assess the nature and scale of radioactive releases outside thecontainment as a result of a core melt accident, with the corresponding frequencies, andcontribute to assessing the overall safety of the facility. They make it possible to verify thatthe estimated frequencies of accidents that could lead to large releases to the environment arevery low. See the document “Nuclear power reactor core melt accidents – State of knowledge” –Science and Technology Series – IRSN/EDP Sciences – 2013.
Where ﬁltered venting of the containment is necessary to prevent its failure due to excessively
high internal pressure.
The research and development on core melt described in this Chapter relates to an initiating event
such as a loss-of-coolant accident, aggravated so that it leads to core melt.
Science and Technology Series – IRSN/EDP Sciences – 2013.

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Figure 9.1 ASTEC code modules. © Didier Jacquemain/IRSN.

The different phases following core uncovery, if the accident is not brought undercontrol, are:– core heating and core melt inside the vessel,– reactor vessel melt-through and erosion of the concrete basemat of the containment by the molten corium,– loading of the containment by a sudden increase in internal pressure,– loading of the containment by a slow increase in internal pressure,– release of radioactive products outside the reactor (commonly referred to using
the term "source term").
In the early 1980s, research focused mainly on studying the ﬁrst phase.

9.1. Heating of the core and core melt inside the vessel
Many analytical tests were performed in the 1980s by different research laboratories,
particularly in the United States and Germany, to study separately the various physicalphenomena and ﬁnd out the laws governing them.
Initially, the oxidation of the zirconium alloy cladding by steam was the subject oflaboratory studies (see Section 3.2). This is an exothermic phenomenon that increases
once the temperature has passed 1000 °C. The hydrogen produced spreads inside thevessel then in the reactor building if a break occurs in the reactor coolant system. The
kinetics of cladding oxidation can be described by a parabolic law, expressing the fact thatat a given temperature, the thickening of the oxide layer acts as a barrier to the transfer

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Current state of research on pressurized water reactor safetyof oxygen to the remaining metal, slowing down the oxidation. It greatly increases above
1580 °C as a result of a change in the zirconia's metallurgical phase, making it lessresistant to oxygen diffusion within the metal itself. Correlations allow these phenomena
to be taken into account with a good level of conﬁdence.
The core of a water reactor is made up of many materials: several dozen metric tonsof uranium oxide, plus around ten metric tons of zirconium, but also stainless steel,
Inconel and materials designed to control the ﬁssion reaction. These are a silver, indium
and cadmium alloy (AIC) in 900 MWe reactors, boron carbide (B4C) and AIC in 1300 MWeand 1450 MWe reactors and in the EPR reactor – boron carbide only in BWRs. All thesematerials can interact with one another and form liquid phases at much lower
temperatures than their respective melting points. Uranium oxide, which melts at
2840 °C, can be dissolved at much lower temperatures than its own melting point inalloys where it is combined with metals with lower melting points (such as the zirconium
of fuel claddings).
Experiments on small samples of materials, particularly in Germany, the UnitedStates and Japan, have enabled these interactions to be observed. Phase and correlation
diagrams have been drawn up as a result, describing the reaction kinetics. Between 1200 °C
and 1400 °C, the materials absorbing the neutrons (AIC, B4C) react chemically with
their stainless steel cladding, forming a liquid mixture that in turn dissolves the zirconium.
Above 1760 °C, the zirconium metal melts and dissolves the zirconia and the uraniumoxide. Finally, the zirconia melts at 2715°C and also dissolves the uranium oxide. The
mixtures obtained can also react with steam if they are not completely oxidized, producing
hydrogen.
These phenomena of oxidation and chemical interaction between all the materials in areactor core have been modeled, usually empirically, and the models have been integrated
into simulation codes. In brief, the reactor core is split into meshes – representing at fuel rod
scale the average behavior of a set of fuel assemblies over a given height – and for eachmesh the code calculates the velocity and composition of the ﬂuid ﬂowing in the core, thetemperature reached, the oxidized cladding thickness and the fraction of molten material
and its composition. Above a certain fraction of molten material, trapped between the fuelpellets and cladding in the case of fuel rods or accumulated in a control rod, or if theremaining envelope of solid cladding is estimated not to be thick enough to contain themolten material, the liquid materials ﬂow by gravity and solidify in the lower, cooler regionsof the core. They then form metal plugs (mainly of AIC or a mixture of boron and steel)topped by ceramic plugs (Zr, U, O mixture).
Given the complexity of the physical and physico-chemical phenomena studied, thediversity of the materials present and the geometry of a reactor core, it was quickly
realized that it was necessary to perform integral tests on a larger scale with components
representative of a reactor core. The aim was to check that the interactions between the
different phenomena would not produce effects that had not been predicted by themodels stemming from the analytical tests.
Most of these tests were performed in experimental facilities, using a nuclear reactor
to heat the fuel being studied. This experimental method, which is fairly difﬁcult and veryAnglais_Book

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expensive to carry out, has the advantage of keeping the fuel (and only the fuel) heated
when it ﬂows to the bottom of the experimental equipment once melted, as in a real
accident.
In France, IPSN, and then IRSN, conducted two large-scale research programs in the
PHEBUS experimental reactor described in Section 2.1: the Phebus-CSD116 programs (onseverely degraded fuel) from 1986 to 1989 and Phebus-FP (on ﬁssion products) from1988 to 2012 (the last test was performed in 2004).
The Phebus-CSD program was used to study phenomena linked to zirconium
oxidation by steam, the associated hydrogen production, physico-chemical interactions
between the uranium oxide and zirconium, in a solid or molten state, and the effect oncore degradation of AIC control rods. The program consisted of six tests: B9, B9R, B9+,
C3, C3+ and AIC. The experimental devices each consisted of an assembly of 21 fuel rodsthat had not been irradiated, of the same type as those used in PWRs but shorter (a ﬁssile
length of 80 cm compared with 3.66 m in a real fuel rod), distributed in a square-pitch
arrangement and held in place by two Inconel grids. The assembly was contained inside a
steel tube, which in turn was inserted into the cell placed at the center of the reactor. A
thick thermal screen made from porous zirconium was used to isolate the assembly fromthe steel tube, cooled by the circulation of pressurized cold water.
A mixture of steam, hydrogen and helium, the composition of which varied from one
test to another depending on the test objectives, was injected at the foot of theassembly. The assembly was heated by gradually increasing the power of the PHEBUS
reactor. A borated steel neutron screen placed between the PHEBUS reactor core and thetest device was used to obtain an axial distribution of uniform power as in a power reactor
core.
Each device was ﬁtted with a very large amount of instrumentation for measuring all
the physical parameters essential for interpreting the tests, particularly high temperaturethermocouples (tungsten-rhenium, W/Re), which functioned at up to 2100 °C, and amass spectrometer for measuring the amount of hydrogen being produced at every
moment. An array of post-test examinations, both non-destructive (X-rays, gammascanning117) and destructive (photo macrography, analyses with a scanning electron
microscope), complemented the measurements taken during the experiments, so thatthe damage to the fuel rods could be quantiﬁed and the composition of the resolidiﬁed
mixtures could be determined.
The six tests carried out produced many experimental results, which were used fordeveloping and validating the core melt models integrated into the ASTEC simulation
code. In particular, they made it possible to observe how the oxidation of the zirconium
alloy cladding spreads axially with the temperature runaway above 1550 °C. The
progression of the oxidation front was characterized by the occurrence of steam

116.
117.

Severely Degraded Fuel.
Device for measuring the intensity of the gamma radiation emitted by a source depending on its
energy levels. Following calibration, analysis of the spectrum obtained can identify the radionuclide(s) responsible for the radiation and their masses.

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starvation, indicating that all the steam available to the cladding had been absorbed by
the cladding to oxidize the metal (the gas ﬂowing beyond the front did not contain any
more steam, only hydrogen). Very signiﬁcant results were obtained for interactions
between the core materials and the processes causing core melt (liquefaction of thezirconium on contact with the Inconel of the grids, dissolution of the outer layer ofzirconia and the fuel inside the cladding by the molten zirconium, liquefaction of thecontrol rods at around 1400 °C and damage to the adjacent fuel rods).
The Phebus-FP program [7] helped to reduce the uncertainty concerning theestimation of radioactive releases of a core melt accident at a light water reactor,
and to increase IRSN's assessment and emergency response expertise in this ﬁeld. The
program consisted of ﬁve tests during which the main physical phenomena governing
core melt, the transfer of ﬁssion products from the fuel to the reactor building, and their
behavior inside it were studied. It was conducted in collaboration with a number ofFrench and foreign partners (EDF, the European Commission and its Member States, theUnited States, Canada, Japan, South Korea and Switzerland). The deﬁnition of the tests,
engineering and experiments, and the analysis and interpretation of the experimental
results involved around 80 people over a period of approximately 15 years.
The program required major modiﬁcations to the PHEBUS facility. The reactor
building was strengthened against earthquakes and enlarged to accommodate a sealed
steel box containing a model of the reactor coolant system and containment of a PWR
and all the necessary equipment and instrumentation (Figure 9.2). The experimental
circuit reproduced, on a 1/5000 scale, the three vital components for this type of study ofa 900 MWe PWR: the core, the reactor coolant system and the containment.

Figure 9.2 Schematic view of the experimental circuit used for the Phebus-FP program. © IRSN.

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To represent the core, a similar device to the one designed for the Phebus-CSD
program was used. This was an assembly consisting of 20 fuel rods with a ﬁssile height of
1 m, and a rod in the center representative of the control rods used in power reactors, allarranged in the cell at the center of the PHEBUS reactor. The fuel used in the ﬁrst test had
not been irradiated before being used in the PHEBUS reactor, whereas the fuel used insubsequent tests came from the Belgian BR3118 reactor and an EDF reactor; their burnup
rates were between 23 and 38 GWd/tU. The instrumentation had been signiﬁcantly
improved, in particular with the use of ultrasonic thermometers that could measuretemperatures of up to 2900 °C.
The reactor coolant system model consisted of a section cooled to 150 °C simulating asteam generator tube with a break at the outlet from this tube. The ﬂuid leaking from thebreak was collected in a 10 m3 tank ﬁtted with a sump at its base, simulating a reactor
containment. The external surface of the tank was thermally isolated and there were cooled
cylinders at its center as a scale representation of the containment wall of a PWR on which thesteam could condense. Some of the cylinders and a plate immersed in the sump were coatedwith identical paint to the paint used in the PWRs in France's nuclear power plant ﬂeet.
The facility's strength was its instrumentation. As well as measuring ﬂow rate,pressure and temperature at several points in the circuit, it could also measure hydrogenconcentrations in the coolant system and the containment. It also included gamma-ray
spectrometers targeted at the tube intake, several points along the wall and the outlet
from the tube simulating the steam generator, the atmosphere and the walls of the tanksimulating the containment, as well as the sump, to continuously identify and measure
the different γ-emitting ﬁssion products in suspension in these areas or deposited on thewalls. A Maypack composed of selective ﬁlters, monitored continuously during the tests
by a gamma-ray spectrometer, was used to measure the concentrations of the different
forms of iodine in the containment atmosphere (aerosols or gases, and among the
gaseous species, molecular iodine or organic iodine). There was also an optical device forcontinuously measuring the aerosol concentrations in suspension in the containment.
Branch pipes at the intake and outlet of the tube simulating the steam generator, inthe sump and on the containment wall, meant that samples could be taken at various
points during the tests through ﬁlters, impactors (instruments for measuring the size andconcentration of the aerosols), thermal-gradient tubes (instruments used for working out
the chemical forms of different species in vapor form by recording the temperatures atwhich they condense on the walls) and in capsules. In the tanks simulating thecontainment, coupons collecting the aerosols deposited by settling were also exposedto the containment atmosphere for predetermined time intervals. After the tests,
radiochemical analyses were performed on all the samples taken, allowing identiﬁcationof the ﬁssion products and the different actinides carried or deposited, and enabling theirmasses to be calculated precisely.

118.

Belgian Reactor 3.

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Speciﬁc non-destructive examination techniques using transmission and emission
tomography119 were also developed for the program and provided valuable information
about the state of degradation of the fuel rods after the tests, and about the quantities ofsome radioactive products emitted by those rods.
The international Phebus-FP program, which ran from 1988 to 2012, consisted of ﬁve
tests. The main parameters studied in these tests are summarized in the table below.
Fuel type

FPT1210 Very lightly
irradiated
FPT1~20 GWd/tU
FPT2
Ditto
FPT3
Ditto
FPT4
Debris bed of
UO2 at38 GWd/tU
and of ZrO2

Steam flow
Material
simulating rate (g/s)
the control
rods

pH of the sump Sump/atmospherewatertemperature
difference120 (°C)

AIC

0.5–3

5

−18/−37

AIC
AIC
B4C
None

0.5–2.2
0.5 (+ boric acid)0.5
0.2–0.5 H2O
0–0.25 H2

5
9
5
N/A

−18/−38−18/+11−18/−3
N/A

Apart from test FPT4, each test consisted of three successive phases:– the ﬁrst phase of re-irradiation of the fuel, during which the PHEBUS reactor was
operating at full power for approximately a week. The fuel being studied wascooled by pressurized water circulation. The aim was to recreate in the fuel the
short-lived ﬁssion products that had disappeared since the fuel was unloaded
from its original reactor, such as iodine-131 (half-life of 8 days122), which wasessential for studying the releases;– a "degradation" phase lasting a few hours, during which, by gradually increasing
the PHEBUS reactor power, the temperature of the uncooled test fuel increased
until the materials liqueﬁed (at between 2300 °C and 2500 °C), causing the ﬁssion
products to be released and transported into the circuit and then into the tanks
simulating the containment. At the end of this phase, the PHEBUS reactor was
shut down;

119.

120.

121.
122.

Technique for reconstructing the volume of an object from a series of radiographic measurements(transmission) or gamma-ray spectrometry measurements (emission) performed on each unit
from the outside.
On the left of "/", temperature difference during the degradation phase; on the right of "/",
temperature difference during the long-term study of iodine behavior. A positive difference meansthat water is evaporating from the sump surface, which is likely to encourage the transfer of thevolatile iodine produced in the sump water to the containment atmosphere.
Fission Product Test.
The concentration decreases by half during the half-life.

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85– a “containment” phase, lasting for several days during which the variables ofinterest for understanding the deposition of radioactive products and the iodine
chemistry in the tank simulating the containment were measured.
The ﬁrst group of three tests (FPT0, FPT1 and FPT2) was used to study the effect of
the burnup rate and the oxygen potential (atmosphere rich in steam or conversely inhydrogen, surrounding the fuel rods) on oxidation and damage to the fuel rods, and on
the release and transporting of ﬁssion products in the circuit. These three tests were
performed with an AIC fuel rod. Boric acid was also injected with the steam during test
FPT2 (during normal reactor operation the boric acid serves to control its reactivity; in an
accident situation, the boric acid can react with some ﬁssion products such as cesium, andsigniﬁcantly alter the chemical balance with the different iodine species, encouraging theformation of the most volatile).
Test FPT3 was used to study the inﬂuence that the presence of boron carbide control rods
has on core degradation and ﬁssion product behavior. This type of material is used in EDF's
most recent PWRs (1300 MWe, 1450 MWe and EPR) in addition to AIC, as well as in BWRs.
Test FPT4 focused on the advanced phase of an accident, studying the release of lowvolatile ﬁssion products and transuranium elements from a debris bed of fuel andoxidized cladding brought to melting point. It consisted of a single temperature-increase
phase until a molten material pool was obtained. The ﬁssion products and actinides
emitted were sampled through a bank of ﬁlters situated above the debris bed andactivated in succession.
Tests FPT0, FPT1, FPT2 and FPT3 were used to reproduce fuel rod damage states that
had never yet been achieved experimentally, with signiﬁcant melting of the fuel (by up to
50% of its mass) and the formation of corium pools, as illustrated in Figure 9.3. The
numerous results obtained contributed signiﬁcantly to improving knowledge of coredamage mechanisms and substantially extended the scope of simulation code validation.
For example, the cladding dislocation criterion was revised to take account empirically ofthe mechanisms causing the zirconia outer layer of the cladding to fail, releasing themolten zirconium inside and causing oxidation and hydrogen production to stop locally.
The tests also signiﬁcantly improved knowledge of the effects of physico-chemicalinteraction between the fuel, the zirconium and the materials forming constituting thecontrol rods, leading to liquefaction of the fuel at around 2200 °C, a much lower
temperature than predicted by models. The core degradation models in simulation codessuch as ASTEC have beneﬁted extensively from these new results and their prediction
capabilities have been greatly improved by them.
The originality and quality of the tests performed at the PHEBUS facility was such that
the Phebus-CSD B9+ and Phebus-FP FPT1 tests were selected by the OECD/NEA for the
performance of international exercises to compare core melt codes (ISP 28 and ISP 46).
Abroad, large-scale programs have also been run, in chronological order:– the SFD (Severe Fuel Damage) program, consisting of four tests conducted
between 1982 and 1985 in the PBF (Power Burst Facility) reactor of the Idaho
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Figure 9.3 State of degradation of a fuel assembly after test FPT1; post-test non-destructive examinations of the tested fuel cluster: X-ray (left) and X-ray transmission tomography (right). (BFC: Bottom of
Fissile Column). © IRSN.

– the DF (Damaged Fuel) program, consisting of four tests conducted in 1984 in the
ACRR (Annular Core Research Reactor) reactor of the Sandia National Laboratory
in the United States;– the OECD/NEA LOFT-FP (Loss of Fluid Tests–Fission Product) project, consistingof two tests conducted from 1984 to 1985 in the LOFT reactor of the Idaho
National Laboratory in the United States;– the FLHT (Full Length High Temperature) program, consisting of four tests
conducted between 1985 and 1987 in the NRU (National Research Universal)
reactor of the Chalk River Laboratories in Canada;– the CORA program, consisting of around 20 tests conducted between 1987 and1997 at a non-nuclear facility of the German research center in Karlsruhe.
For experiments performed in nuclear reactors, the experimental devices consisted ofassemblies of between ten (DF) and several tens (PBF) of fuel rods (or indeed 100 in theAnglais_Book

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87case of the LOFT-FP project) of similar manufacture to the rods used in power reactors.
They were generally shorter, like those used in the Phebus-CSD and FP programs, because
the cores of the research reactors hosting the experimental devices were not as high, withthe exception of the FLHT program which, as the name suggests, used full length rods
with a ﬁssile height of 3.66 m in the large NRU reactor. In some tests, rods containing AIC
or boron carbide were included in the assembly, as well as fuel rods that had previously
been irradiated for a signiﬁcant period to obtain representative burnup rates.
In the CORA program tests, the fuel rods were simulated by zirconium alloy cladding
ﬁlled with alumina and heated internally using a tungsten core with an electrical current
running through it. The lower cost of these tests meant that many parameters could bestudied: temperature increase rate, ﬂow rate of the injected steam, ambient pressure,cladding materials, effect of the presence of AIC or boron carbide, geometry of thedifferent reactors (PWR, BWR, VVER), effect of quenching at the end of the test, etc.
However, this heating method meant that it was not possible to study the corium poolformation phase.
All the tests contributed to a better knowledge of cladding oxidation, hydrogenproduction and core melt phenomena. They delivered a wealth of experimental resultscovering a broad spectrum of situations, which are extremely useful for developing
models and validating simulation codes such as the integrated ASTEC, MAAP and
MELCOR codes.
In 2016, as shown by a benchmark exercise run by the OECD/NEA for an accident
scenario similar to TMI-2 [8], the integrated codes for core melt accidents are able tosimulate with a good degree of precision the progression of PWR core degradation up to
the reﬂooding phase. On the other hand, a benchmark exercise run by the OECD/NEA forthe damaged BWR reactors at the Fukushima Daiichi plant [9] identiﬁed major discrepancies in the calculation of the progression of core degradation in the three damaged
reactors and of hydrogen production, revealing current uncertainties as regards representing degraded conﬁgurations for BWRs. Partly these difﬁculties are due to the factthat core melt accident experiments have focused mainly on PWR fuel assembly
conﬁgurations. There has been very little study of BWR fuel assembly conﬁgurations;major uncertainties exist with regard to molten mixture relocation phenomena in thepresence of Zircaloy channels for water in the assemblies and the cruciform B4C-steel
control rod plates speciﬁc to BWRs. More detailed analyses are under way in order tounderstand the differences between these very different conﬁgurations.
Interestingly, none of the tests conducted abroad reached the temperature or fueldegradation levels observed in the Phebus-FP tests, making them particularly valuable.
On the other hand, some of the tests performed abroad (LOFT, CORA and QUENCH
programs – QUENCH succeeded CORA at the same facility from 1997) have been able tostudy the effects on cores molten to varying degrees, of reﬂooding with large ﬂows ofwater or steam.
The most important ﬁnding from these tests is that reﬂooding a core that has alreadybeen extensively damaged may, at least initially, considerably accelerate cladding
oxidation, hydrogen production and core melt. When the accident in Unit 2 at theAnglais_Book

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Three Mile Island power plant (TMI-2) happened, there is no doubt that water entering
the molten core caused any cladding that had not yet melted to rupture and the debris
bed observed in the upper part of the core to form. Experimental results tend to showthat core reﬂooding can halt the progression of an accident only if it happens very early
on, i.e. when the maximum temperature in the core has not yet exceeded 1800 °C, and if
the ﬂow of water is sufﬁcient (more than 2 g/s per rod). Despite all the experimental
results obtained, models are unable at present to reliably predict whether or not
reﬂooding a damaged core can cause runaway oxidation of the cladding and aggravatethe situation.
To ﬁnd out more about this phenomenon, since 2010 IRSN has been conducting
test programs with the PRELUDE123 and PEARL facilities at its laboratories in Cadarache, involving reﬂooding experiments on debris beds consisting of steel balls heatedby induction inside a cylindrical cavity (diameter 11 to 29 cm in the case of PRELUDE
and diameter 50 cm in the case of PEARL). The PEARL tests performed in a largediameter section were used to observe the two-dimensional ﬂows. An experimental
and theoretical research program on the reﬂooding of solid debris beds in more
complex conﬁgurations, in particular with the observation of oxidation phenomena, isplanned for 2018.
If core reﬂooding is ineffective, the accident continues and the molten materials ﬂow
to the lower head of the reactor vessel, where they accumulate, as seen in the case of theTMI-2 accident. If these materials come into contact with water at the lower head, vastquantities of steam and hydrogen are produced. Despite the many examinations carriedout after the TMI-2 reactor accident, particularly during the OECD/NEA TMI-VIP
program, and the tests of dropping corium into water performed on a large scale in
the FARO facility at the European Ispra Joint Research Centre in the 1990s, the ﬂow
mechanisms of the corium at the lower head and its mixture with the water present are
still not clearly understood. This difﬁculty is overcome by using pessimistic hypotheses
for the quantity of materials transferred, their fragmentation on contact with water, theoxidation reactions, etc. when evaluating the progression of a core melt accident.
Nevertheless, all the experts agree that a steam explosion violent enough to destroy
the reactor vessel could not occur under these conditions (OECD/NEA project entitled
SERENA124, described further on).
If the corium debris at the lower head is not cooled effectively, it will melt again,
forming a pool in contact with the vessel walls, threatening the vessel’s integrity. Various
test programs have been undertaken to study this conﬁguration. These test programs
either use simulator materials, like the BALI tests conducted by CEA in Grenoble, at fullscale in 2D geometry with water, to study natural convection in the pool, or they use
representative materials (UO2, zirconium alloy, steel), like the OECD/NEA test programs
RASPLAV (1994–2000), MASCA (2000–2003) and MASCA 2 (2003–2006) conducted in
Russia. During the latter test programs, corium (a mixture of UO2-ZrO2-Zr and steel) ofvarious compositions and in different atmospheres (inert or oxidizing) was brought to
123.
124.

Preliminary Study on the Experimental Reﬂooding of a Debris Bed.
Steam Explosion REsolution for Nuclear Applications.

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melting point and kept between 2200 °C and 2650 °C. In RASPLAV, a 200 kg mass ofcorium, with mass composition of 76.6% UO2, 14.3% Zr and 9.1% ZrO2, heated bytungsten walls forming a thin semi-circular section of a reactor vessel, was used. In thesmaller scale tests using corium in crucibles cooled on the outside, heating was achievedby electrical induction and the more varied corium compositions studied meant that
coriums with a high metal content could be used and in some cases simulators of FPs (astable species of the radioelements of interest in order to study their distribution
between the various metallurgical phases of the corium). In parallel, smaller scale tests
were conducted with coriums of identical composition, allowing the physical propertiesof the material mixtures to be determined. Tests using pools of molten salt, eutectic or
binary mixtures, completed the studies of the effect that the material properties have on
the interface temperatures and their impact in correlations for predicting the distributionof ﬂows of heat extracted from a pool to the walls of the vessel containing it. TheRASPLAV tests conﬁrmed that using traditional correlations for the distribution of heatfrom a pool at its edges (established using various simulator liquids that are easier tomanage) with the physical properties of the corium was entirely justiﬁed.
The experiments performed in MASCA and MASCA 2 reveled complex phenomena ofinteraction between materials and of stratiﬁcation between the metal and oxide phasesin the presence of steel. Under certain conditions (particularly where the pools initiallycontain a signiﬁcant amount – around 50% – of unoxidized zirconium), a three-layer
conﬁguration was observed with a high density metal layer containing uranium, topped
by a layer of refractory oxides (mainly uranium and zirconium oxides), itself topped by a
low density metal layer (iron, chromium, nickel, zirconium, etc.). In the stratiﬁed pool
conﬁgurations where there is no reﬂooding, a very intense ﬂow of heat can be transmitted to the walls around the upper metal pool (heat ﬂow concentration or focusing
effect), if its thickness is below a threshold of around 50 cm. The tests also revealed thetransient nature of the stratiﬁcation phenomena where steam is present, since the metal
layers oxidize and gradually mix with the oxide pool.
These phenomena are modeled in the ASTEC simulation code but many uncertainties
remain concerning the transient phenomena, particularly oxidation of the uppermost
layer of metal and the transfer of oxygen between the layers when steam is present. To
reduce these uncertainties, an experimental program known as CORDEB (Corium-Debris)
was launched in 2012 by the Aleksandrov Scientiﬁc Research Technological Institute(NITI) in St Petersburg. It should run for at least seven years in the MASCA facility, inpartnership with EDF, AREVA, CEA, IRSN. Examining the corium pools produced during
the Fukushima Daiichi accident could shed further light, of great relevance, on thesubject.

9.2. Reactor vessel melt-through and basemat erosion
by the molten corium
The OECD/NEA TMI-VP project to examine in detail the damaged TMI-2 reactor
vessel and its contents, mentioned above, established that the internal face of the lower
head, lined with stainless steel, had locally reached a temperature of 1100 °C forAnglais_Book

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30 minutes, while the pressure inside the vessel was 10 MPa. The cooling of the corium
once water circulation had been restored in the reactor coolant system prevented reactor
vessel melt-through, though in 2016 the reasons for this are still not clear. Once of the
hypotheses advanced to explain why the reactor vessel remained intact was that directcontact between the corium and the lower head was limited by the presence of water at
the lower head when the corium ﬂowed in.
The OECD/NEA experimental project known as OLHF125, conducted by the Sandia
National Laboratories in the United States with the involvement of CEA and IRSN,
obtained experimental results on a 1/5 scale model for the creep-induced failure
mechanisms of a PWR lower head at a representative range of temperature/pressure
parameters. However, there are uncertainties involved in applying the results to a fullscale reactor, since a number of welded tubes pass through the lower head. A test
conducted with these tubes showed a signiﬁcant reduction in the rupture elongation of
the vessel.
Following vessel melt-through, the molten corium runs into the reactor pit and, if
there is no proper cooling, causes the chemical decomposition due to heat, and therefore
erosion, of the concrete reactor pit walls and basemat. This could ultimately put the
integrity of the containment at risk and lead to massive radioactive gaseous and liquid
releases into the environment. The interaction between the corium and the concrete also
generates large amounts of gas (steam, which is reduced to hydrogen and carbon
monoxide and dioxide due to oxidation of the rebars or the metal phases present in the
corium), which can increase the pressure inside the containment and also threaten its
integrity.
During the erosion of the concrete by the corium, which can last for several days, thecomposition of the pool changes. It contains some dense oxides (mainly U-Zr-Fe-Ox),
low-density oxides (SiO2, CaO) and non-oxidized metals (Fe, Cr, Ni, Zr). These are
gradually oxidized by the steam or carbon dioxide whereas they can only reduce much
more marginally other less stable oxides. The pool solidiﬁes on contact with the cooler
concrete, temporarily forming crusts that are mechanically unstable since the underlying
concrete wall is breaking down. Stratiﬁcation between the metal and oxide phases canalso occur, at least temporarily, changing the distribution of the heat ﬂows. Whether thisphenomenon occurs depends on the conﬁguration of the pool within the vessel when thevessel fails (if it is stratiﬁed or evenly mixed) and the position of the break (e.g. if it is levelwith a metal layer). In the longer term, it depends on how the composition of the metal
phase changes, the density ratio between metal and oxide phases, the ﬂow rate of thegases generated by decomposition of the concrete, and the residual quantity ofunoxidized metal; all these parameters changing during the interaction. Incorporationof the metals from molten rebars in the concrete can contribute to the formation of ametal layer.

125.

OECD Lower Head Failure.

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In France, these phenomena were ﬁrst studied experimentally in small-scale analytical tests using simulator materials with low melting points. The tests were conducted by
CEA on behalf of IPSN (then IRSN) and EDF:
– the BALISE program studied liquid entrainment at the interface between two nonmiscible liquids with different densities and viscosities (water and liquid parafﬁn,silicon oils, etc.), percolated through a nitrogen ﬂow (1999–2000);– the ABI program measured the heat transfer coefﬁcients at the interface between
two non-miscible liquids with very different densities (Wood metal and gallium tosimulate the corium, water and oils of different viscosities to simulate theconcrete), percolated through a nitrogen and argon ﬂow (2007–2008);– the ARTEMIS program simulated the erosion phenomena in one-dimensional then
two-dimensional geometry using a salt (BaCl2) to simulate the corium and abinary eutectic mixture (LiCl-BaCl2) to simulate the concrete (2003–2008);– the CLARA program measured heat transfers in a pool of water in the presence ofa ﬂow of gas injected through the walls using water and additives to vary theviscosity (2008–2012).
Given the importance of the combination of convection and thermal and physicochemical phenomena, tests with real materials were necessary. At Cadarache CEA built
the VULCANO126 facility, commissioned in 1997127, which uses a rotary plasma arc
furnace to melt prototypic corium mixtures (a UO2 and ZrO2 mixture, also incorporatingoxides from the decomposition of the concrete to study the later phases of theinteraction) with a mass of around 40 kg, which are then poured into an experimental
section. For the programs conducted from 2003 with the participation of EDF and IRSN,
to study interaction between the corium and concrete (CCI128), corium pools werepoured into cylindrical concrete crucibles of varying compositions. High frequency
induction heating was used to reproduce the residual heat released in the fuel throughout
the interaction. The instrumentation in place was used to determine the axial and lateralprogression of the concrete ablation front.
One speciﬁc characteristic of the facility is that it can also be used to study theerosion caused by metal/oxide pools and the impact of stratiﬁcation. A mass of steel ofapproximately 25 kg representing the molten structural materials was melted in three
separate induction furnaces then poured into the experimental section containing themolten corium. In the test section, a system of magnetic screens meant that it waspossible to heat only at the level of the oxide phase of a stratiﬁed mixture.

126.
127.

128.

Versatile UO2 Laboratory for COrium ANalysis and Observation.
The facility was initially used to study corium spreading in the core catcher installed under the EPR
reactor vessel. The tests performed there were also used to validate the CROCO corium spreading
simulation code developed by IRSN.
Corium-Concrete Interaction.

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The following tests were devoted to studying the interaction between corium andconcrete:– nine tests with all-oxide corium (VB-U series),– four tests with corium and steel (VBS-U series),– four tests that were more analytical (VBES-U series).
The results revealed the importance of the type of concrete (siliceous or limestone) tothe concrete's ablation proﬁle, conﬁrming the results of tests carried out by theAmericans (see further on). They also contributed to the development of models forevaluating a conservative approach taking account of the uncertainties that still exist
concerning the time to penetration of the concrete containment basemats of French
reactors if a corium/concrete interaction were to occur. It has therefore been possible todemonstrate that basemat penetration should not occur within 24 hours of the start ofan accident129, allowing time for measures to be taken to protect the public.
In Germany, tests were performed in the 1990s at the BETA facility in the Karlsruhe
research center. The corium was simulated by a molten mixture of 400 kg of alumina andiron obtained through an exothermic oxidation/reduction reaction, to which several
dozen kilograms of zirconium were added. This mixture was poured into a concrete
crucible.
In the United States, three research programs were conducted at facilities of the
Argonne National Laboratory: the MACE (Melt Attack and Coolability Experiments,
1989–2000) programs on behalf of EPRI (Electric Power Research Institute) and theOECD/NEA MCCI130 programs (2006–2010) in which France (IRSN, EDF and CEA) was apartner. The objective of these two programs was to conduct large-scale and smallerscale integral tests using representative materials, to study:– concrete erosion phenomena in a two-dimensional geometry in the absence ofany cooling;– the various cooling mechanisms identiﬁed during the MACE experiments whenwater was poured on the surface of the corium;– new technological concepts for cooling the corium and halting the concrete
erosion.
Two experimental devices were used, SSWICS (Small Scale Water Ingression and
Crust Strength) and CCI (Corium Concrete Interaction). In the ﬁrst, a mass of around
75 kg of UO2 and ZrO2, mixed with siliceous or limestone concrete, brought to atemperature of more than 2000 °C by an exothermic oxidation/reduction reaction, wasmelted in a cylinder of diameter 30 cm with inert walls made from magnesium oxide.
Water was poured on the corium to study the cooling mechanisms and their effectiveness, in particular the mechanism of water penetration into a top crust that has cracked

129.
130.

In the case of Fessenheim, after basemat thickening.
Molten Core-Concrete Interaction.

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due to thermomechanical stresses. Gas could also be injected through the bottom tosimulate the gases released during the corium-concrete interaction.
In the second device, a mixture of 400 to 590 kg of UO2 and ZrO2 was melted using
the same method in a larger cavity with a square cross-section of 50 cm by 50 cm (onetest was performed in a 50 cm by 70 cm cavity with a single ablatable side wall to study
any effects of scale). Tungsten electrodes were arranged on two opposite sides of thecavity so that an electrical current could be applied to the corium pool to generate arelease of heat representative of the residual heat. The other two sides and the bottom
were made from concrete; the type of concrete (siliceous or limestone) was one of theparameters of the study. On this test device, the "Joule effect" heating technique made it
impossible to study stratiﬁed pools with two separate phases, oxide and metal, because
the stratiﬁcation of the metal would cause an electrical short circuit between theelectrodes. Following a dry corium-concrete interaction phase to study the concrete
ablation mechanisms, water was added at the top to study the cooling mechanisms.
In all, 21 tests were carried out during the two OECD/NEA MCCI programs. They
provided important results that signiﬁcantly improved understanding and modeling of
concrete ablation and corium pool cooling mechanisms in a corium-concrete interaction.
As far as concrete erosion kinetics in the absence of cooling are concerned, it wasobserved that the ratio between the ﬁnal erosion thicknesses in the axial and lateral
directions depends on the type of concrete. Whereas the erosion is almost uniform in thecase of limestone concrete, it can be three times greater laterally in the case of siliceous
concrete. As regards cooling mechanisms, it was shown that cracking of the upper crust
and water penetration is much more effective when the cooling above the pool starts
early, i.e. while the concrete content inside the pool is still low and the crust is morefragile. Eruption and fragmentation of the corium ejected into the water through thecrust covering the pool were also observed. They help to cool the pool by turning acompact mass into a bed of debris several millimeters in diameter. This is even more
effective when the concrete is a limestone type (more gas is released ejecting morecorium through the cracks in the crust). At the end of phase 2 of the MCCI program, a testwith siliceous concrete but with cooling at the start of the corium-concrete interaction
meant that it was possible to observe ejections with this type of concrete for the ﬁrst
time. This encouraging result prompted EDF to continue funding tests (CCI), as part of its
work program on extending the operating life of France's nuclear power plants, to study
cooling mechanisms where water is injected early on in the corium-concrete interaction
at the surface of the pool, with support from IRSN, the U.S.NRC, NRA and CEA.
Experiments have also been conducted in Germany to test a different corium coolingconcept (COMET) by injecting water through a special device consisting of several porous
concrete layers. This device could prove more effective than pouring water on the corium
because then the liquid and steam are not circulating against the ﬂow. These experiments
were conducted at the Karlsruhe center, using up to 1300 kg of alumina and iron melted
using an exothermic oxidation-reduction reaction and kept at temperature by induction
heating. Following the erosion of a sacriﬁcial layer of concrete, the pool is cooled bywater injection via a layer of highly porous concrete through porous concrete blocks, thesection of which was optimized to maximize cooling efﬁciency while avoiding a too fastAnglais_Book

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production of steam, which would increase pressure inside the containment. The layer ofhighly porous concrete through which the water was injected on to the blocks was linked
to a tank on a higher level, so water was ﬂowing by gravity only. During the OECD/NEA
MCCI-2 test program and speciﬁcally a VULCANO VWU131-1 test, cooling with a device
following the same principle was also studied using prototypic materials. The results led
to the conclusion that this type of cooling system was effective, meaning that it could be
implemented in new reactor concepts.
Knowledge of concrete erosion and corium cooling phenomena has advanced
signiﬁcantly thanks to all the experimental results produced in the last 20 years.
More robust models based on a large database of experimental data have been
integrated into simulation codes, particularly the ASTEC code. However, it is still notpossible to model in any detail the phenomena occurring at the interfaces between the
corium and concrete and therefore to understand from a mechanistic point of view thedifferences between erosion kinetics due to the concrete type. IRSN, in partnership with
universities and the CNRS, has tried to use direct simulation techniques to advance thisarea of knowledge, e.g. by studying the material ﬂows at the corium-concrete interface or
the consequences of bubbles passing through an interface between two non-miscible
liquids, in terms of the mass of liquid entrained and the effects on heat transfer throughconvection. Without any local information for validation purposes, these efforts have notdelivered any very signiﬁcant advances for the erosion models and so the current models
are still very empirical.
Moreover, without sufﬁcient experimental data, uncertainties remain about thepossible effects of a stratiﬁed pool. Two mechanisms could limit the life of such apool: oxidation of the metal phase and mixture of the materials due to the gases released
by decomposition of the concrete. Further tests are therefore necessary to study these
phenomena and reduce the uncertainty. The impact of the rebars on the erosion
mechanisms is also unresolved because results from the MOCKA tests conducted at
Karlsruhe in the last few years, with simulator materials at high temperatures, are fairly
contradictory and to date no equivalent tests have been performed with prototypic
materials.
Nationally, research is also continuing to study the possibility of cooling andsolidifying the corium by covering it with water or injecting water into a COMETtype device, with a view to designing systems that would avoid basemat penetration inexisting reactors. They are being carried out in France as part of the MIT3BAR132 project in
response to the call for RSNR projects launched in 2012 by ANR in the wake of the
Fukushima accident, and are coordinated by CEA in partnership with EDF and AREVA.
Examination of Unit 1 of the Fukushima Daiichi power plant could also help toimprove knowledge of the interaction between corium and concrete and the effectiveness of cooling mechanisms, by providing information about a situation potentiallyinvolving large corium masses at reactor scale. Indeed, accident calculations performed
131.
132.

VULCANO Water-Uranium.
Assessment and Mitigation of the Risk of Third Containment Barrier Breakthrough at Nuclear
Power Plants.

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using core melt accident simulation codes like ASTEC, MAAP and MELCOR suggest thatthe core of Unit 1 was seriously degraded, that the vessel failed and that large quantities
of molten corium ﬂowed on to the concrete under the vessel, eroding a signiﬁcant
thickness of it. To understand more about the corium-concrete interaction phase in thisreactor, it is necessary to work backwards and analyze the in-vessel degradation phasewhich, depending on the code used, produces some very different results in terms ofcomposition, temperatures and the physico-chemical properties of the debris andcorium [9].

► Possibility of in-vessel corium retention
One countermeasure envisaged by some licensees to try to avoid vessel meltthrough in a reactor where the core has melted is to ﬂood the reactor pit in order to cool
the pit by organizing a two-phase ﬂow of water around it. The effectiveness of this action
depends on many different factors, including the reactor's size and power, the moment
when the materials relocated to the lower head, the fraction and composition of thecorium in the lower head, which determines the distribution of the ﬂow of heat to thelower head internal wall, and the geometry of the vessel pit and characteristics of thethermal insulation around the vessel. The presence (or not) of speciﬁc geometric features(discontinuous thickness between the vessel shell and the lower head, penetrations
through the lower head for instrumentation), which determine the ﬂow of heat that canbe evacuated by the outer face in the water, should not be ignored.
In a deterministic approach, a robust demonstration of the possibility of keeping the
corium in the vessel, known as in-vessel retention (IVR), requires checking that the ﬂow
of heat through the vessel can be evacuated at any point on the vessel’s external surface.
This heat evacuation must be performed without drying out the wall, particularly in areas
where the ﬂow of heat is greatest, and this applies to all feasible conﬁgurations(stabilized or transient) of the debris bed133, which are generally inhomogeneous.
This depends on the distribution of heat on the vessel wall’s internal surface and on
the distribution of the critical ﬂow at each point on the vessel’s external surface, which
depends on a number of parameters as explained above. The distribution of heat on thevessel wall’s internal surface is the result of natural convection mechanisms within the
different layers or piles of corium at the vessel’s lower head, whereas the absolute values
are dependent on the length of time before relocation to the lower head and the change
in mass of the relocated corium, i.e. ultimately the total residual heat to be evacuatedthrough the lower head wall. The physico-chemical balance of the materials thatcompose the corium will change and the situation can gradually evolve so that a
stratiﬁed pool develops in which a non-miscible metal layer ﬂoats at the surface of a poolof oxides, which may itself be composed of two phases with different densities134. This
133.

134.

"Debris bed" is the term used to refer to the corium from the damaged core that has relocated to
the lower head. It may contain solidiﬁed clumps, and pieces of the structures at the bottom of fuelassemblies (grids, etc.) that have fallen into the corium.
The behavior of the corium at the lower head is explained in brief here; it is described in moredetail in Section 5.1.2 of the publication “Nuclear power reactor core melt accidents – State ofknowledge” – Science and Technology Series – IRSN/EDP Sciences – 2013.

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Figure 9.4 Heat ﬂow distribution and risk of dryout © Georges Goué/IRSN.

conﬁguration may be temporary when there is steam in the vessel because the gradualoxidation of the metal phases can produce a more favorable situation in terms of heat
distribution, of a homogeneous pool.
If water has not been injected into the vessel – a likely situation in the ﬁrst few hoursof a core melt accident, which may in turn be due to a loss of coolant – the transfer
through radiation from the metal layer ﬂoating on the oxide pool to the very overheatedupper structures of the core is reﬂected in the fact that the surface temperature of themetal layer is higher than the melting point of steel (the interface temperature between
the metal layer and the internal face of the vessel). The metal layer therefore transfers
more heat to the vessel due to the imbalance between the interface temperatures. This
results in a heat ﬂow concentration, known as a "focusing effect", which causes a heat
ﬂow peak in the vessel at the location of the metal layer (Figure 9.4). The thinner theﬂoating metal layer and the larger the mass of oxides under it (from which the residual
heat is being released), the greater this peak is. If water is present, the formation of ametal crust on the pool’s surface will reduce the interface temperature imbalance andlimit this concentration.
Furthermore, given the big thickness of the vessel wall and the intensity of the heatﬂows to be evacuated through it, the vessel will partially melt starting from its internalface, and its thickness will be reduced to a few centimeters at drying ﬂows of around
1.5 MW/m2. The corium crust may then become unstable on contact with the liquid steelﬁlm, causing the transfer of the underlying molten steel to the corium pool and causingtransient heat ﬂows that are locally very intense when the liquid corium once again
comes into contact with the wall135. From a mechanical point of view, a few centimetersof residual steel will be enough to keep the vessel intact once external cooling is provided
and the vessel has been depressurized.
135.

This phenomenon has been observed experimentally in tests conducted in the SCARABEE reactor
on the total instantaneous blockage (TIB) of fuel assemblies in fast neutron reactors.

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Although in-vessel retention appears possible for small, low-power reactors, thehighest power level at which it would still be possible without being in danger of causing
even more serious consequences is still the subject of uncertainty. If vessel failure should
occur when the reactor pit is completely ﬂooded, the ejection of the molten corium in thelower head could cause a steam explosion on contact with the water, which could
damage the reactor containment (see Section 9.3.1 further on) and the components
contributing to its leaktightness.
This is the reason why France is playing an active part in the European In-Vessel Melt
Retention (IVMR – H2020) project, coordinated by IRSN with the participation of AREVA,
CEA, EDF and 19 other European study and research bodies, with the aim of assessing
whether a strategy of this kind could be successfully applied to Europe's existing
1000 MWe reactors (pressurized water reactors, boiling water reactors, VVER reactors).
A common methodology will be deﬁned for the project, which is expected to ﬁnish in
2019, for analyzing in-vessel retention capabilities and to complete experimental data inorder to reduce uncertainty related to phenomena about which still little is known.
In the context of IVR, the effectiveness of core reﬂooding at a particular moment
determines the mass of corium that will relocate to the lower head and therefore theabsolute value of the residual heat to be evacuated from the lower head. IRSN has begun
some research on this theme, speciﬁcally concerning:– modeling of the geometric characteristics of the different possible debris bed
conﬁgurations (porosity, exchange surfaces, permeability, etc.);– modeling of the thermohydraulic phenomena likely to occur during reﬂooding forthese different conﬁgurations;– modeling of the oxidation phenomena of the metal-oxide mixtures in corium andof hydrogen production;– impact studies of how all these phenomena interact.
The development and validation of the models rely in particular on the PRELUDE tests
and the tests currently being performed by IRSN on a larger scale at the PEARL facility as
part of IRSN's analytical experimental program on debris bed reﬂooding (PROGRES136)(see Section 9.1).
Concerning focusing effect risks, IRSN is working to identify the conditions in which
the layer of metallic materials can ﬂoat for a long period of time on the oxide layer, andfor this it is studying the kinetics of transition between these layers, taking particular
account of crust formation at the interfaces and the oxidation of the metal phases. The
experiments on this subject are being run as part of the CORDEB program mentioned in
Section 9.1 and will continue at the facility used for this program once it is ﬁnished (2016–2019), as part of the European IVMR program coordinated by IRSN.

136.

Analytical Axperimental Program on Debris Bed Reﬂooding.

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Finally, concerning interaction between the corium and water, research activity
has been combined within the ANR RSNR ICE137 project, which will be discussed in
Section 9.3.1 further on.
An in-vessel retention strategy was adopted in Finland in 1995 when the power of the
VVER-440/213 reactors at the Loviisa nuclear power plant was increased to 510 MWe, by
installing a system of mobile insulation to create a cooling space around the vessel and bymodifying the reactor pit ventilation system to create a cooling system. Experimental
programs (COPO138 in Finland and ULPU139 in the United States) and numericalsimulations were used to check that the cooling system performance was sufﬁcient,taking into account bounding heat ﬂow distributions at the lower head.
As far as new reactors are concerned, the in-vessel retention strategy has been
adopted for the AP600. Given the power level of this reactor (600 MWe), the margins
were considered adequate, even in the absence of water injection into the vessel, tovalidate the demonstration, allowing the U.S.NRC to certify the reactor in 1999. The
same approach was used for the AP1000, which was certiﬁed by the U.S.NRC in 2005. Itshould be noted that there are no penetrations through the lower head for instrumentation in the AP600 and AP1000 reactors.
Three types of information were taken into account by the U.S.NRC:– deterministic data;– assessments giving sufﬁcient conﬁdence that if the vessel were to fail – particularly due to the focusing effect – there would be no cliff edge effect, or more
speciﬁcally the reactor pit and the basemat could then provide their conﬁnement
function, withstanding any overpressure that might be caused by interaction
between the molten materials and water140;– probabilistic data, integrating the risk of vessel melt-through and early loss ofcontainment141.
Interestingly, IVR is also the strategy chosen by the designer of the VVER 1000, whichstill provided an external core catcher (TianWan reactor in China and Kudankulam
reactor in India).
137.
138.
139.
140.

141.

Corium-Water Interaction.
Corium Pool Facility.
An IVR-related full-scale boiling heat transfer facility at University of California, Santa Barbara.
For the AP1000, this was the subject of an analysis presented in the document Analysis of InVessel Retention and Ex-Vessel Fuel Coolant Interaction for AP1000 – NUREG/CR-6849, ERI/NRC04-201 – August 2004. For the AP600, refer to the report AP600 Final Safety Evaluation Report
Related to the Certiﬁcation of the AP600 Design – NUREG-1512, 1998.
According to US regulations, probabilistic assessments must show that the frequency of massive
early releases (less than 24 hours after an accident) is less than 10–6/year.reactor. Given theuncertainties surrounding the models for corium progression to the lower head, for the AP1000
the designer assumed a conditional probability of vessel failure of between 4% and 30%, andpostulated a systematic failure of the containment with a massive release into the environment.
Since the calculated frequency of containment failure is 1.9 × 10–8/year.reactor, with 38% of
failures leading to early releases, the designer concluded that its design was well below the limit.

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9.3. Dynamic loading of the containment by a sudden
increase in internal pressureThis kind of phenomenon is likely to cause a relatively early loss of integrity of thecontainment, followed by a massive radioactive release into the environment. In line withthe classiﬁcation proposed by Norman C. Rasmussen in the early 1970s (Figure 9.5), we
will distinguish between steam explosions (failure mode "α") and hydrogen explosions(failure mode "γ"); direct heating of the containment gases (Direct Containment Heating[DCH]) should also be taken into account.

9.3.1.

Steam explosions

Steam explosions can occur when two ﬂuids come into contact, where one (in thiscase the liquid corium from the core melting) is at a higher temperature than the boilingpoint of the other (the coolant). It is a thermodynamic reaction. When it enters the water,
the corium jet fragments and causes intense vaporization of the water. This vaporization
can create a shock wave that fragments the corium even more ﬁnely, further increasing
the contact area between the two ﬂuids and, as a result, the heat transfer. The process
can propagate in the corium-water mixture, triggering an explosion. Simple contact

Figure 9.5 Schematic diagram of the possible containment failure modes according to the Rasmussen
report [1].

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between the two ﬂuids is not enough to cause such an explosion. If the hot ﬂuid
fragments into quite large pieces (measuring millimeters or centimeters), the heattransfer will not be fast enough or intense enough, and will merely cause steam
production characterized by a slow pressure increase in the system. If the hot ﬂuid
stays liquid on contact with the cold ﬂuid due to the formation of a steam ﬁlm andsigniﬁcant conﬁnement restricting steam production in the ﬁrst few moments, this is
likely to trigger a very forceful explosion.
Fine fragmentation and explosion phenomena were initially studied in the 1970s as
part of analyses of core melt accidents likely to occur in fast neutron reactors withsodium as coolant. From the late 1980s, this research was extended to PWRs. In France,
analytical test programs were conducted by CEA on behalf of IPSN (then IRSN) and led tothe development of heat transfer models.
The BILLEAU tests (which ended in 1997) were used to characterize heat transfer on
the surface of solid metallic spheres heated to 2200 °C and plunged into water. The
TREPAM program (which ended in 2003) consisted of measuring the heat ﬂow on thesurface of a tungsten ﬁlament plunged into water, for different ﬁlament diameter values(10 to 250 µm), ﬁlament temperatures (1080 °C to 2630 °C), water pressure values (0.1
to 21 MPa), water undercooling values (0 to 80 °C below the water saturation
temperature) and plunge speeds into the water (0.2 to 46 m/s).
Abroad, tests have been conducted with more representative materials, sometimes
on a larger scale. These include the following programs:– CCM142, run by the Argonne National Laboratory (United States) in the early
1990s, which studied the fragmentation of molten mixtures containing 2 to 12 kgof UO2, ZrO2 and steel at 2800 °C;– SUW143 and WUMT, run at Winfrith (United Kingdom) at the same time, with auranium-based thermite144;– FITS145, run at the Sandia National Laboratories (United States) with an aluminum-based thermite;– ALPHA, conducted by the Japan Atomic Energy Research Institute (JAERI), Japan,
with the same material;– FARO, run by the European Commission's Joint Research Centre at Ispra (Italy),
which ended in 1999 and studied, in the TERMOS vessel, the fragmentation ofmolten mixtures of 18 to 176 kg of UO2 and ZrO2 and the steam explosion
resulting from dropping this into water, under pressures of 0.2 to 5 MPa;

142.
143.
144.

145.

Cold Crucible Melting.
Scale-Urania-Water.
The corium is represented by a mixture consisting of an oxide and a molten metal obtained bymeans of an exothermic oxidation-reduction reaction like the reaction between iron oxide, Fe2O3,
and aluminum (aluminothermic).
Fully Instrumented Test Series.

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101– KROTOS, originally run by the same research laboratory, then by CEA after 2004
in a shock tube (one-dimensional geometry), to study fragmentation on contact
with water of a liter of various molten materials (tin at 1000 °C, alumina from2300 °C to 2800 °C and a UO2-ZrO2 mixture at 2800 °C), and the steam
explosions that occur as a result;– PREMIX, run by the research center at Karlsruhe (Germany), which studied thefragmentation of 16 to 60 kg of aluminum-based thermite producing molten
alumina brought to 2300 °C;– ECO146, conducted at the same laboratory, which studied the explosion resulting
from contact between 6 to 18 kg of molten alumina brought to a temperature of2300 °C and one to two meters of water.
The results of these various research programs provided a large amount of experimental data. However, the results were fairly diverse: some tests produced explosions
but others did not, for no apparent reason except the physical properties of the materials
used (the molten alumina had a greater tendency to cause explosive interactions than the
UO2–ZrO2 mixture, the thermal conductivity of which is lower).
On the basis of the analytical tests, IRSN and CEA together developed the MC3D147
steam explosion simulation code. The code uses an eulerian description with a mixedﬁnite volume (mass and energy balances) and ﬁnite difference (momentum balance)
method with a three-dimensional (cartesian or cylindrical) structured mesh. It simulates
both the ﬁne fragmentation and the explosion phases. The complexity of the phenomena
to be described during fragmentation requires very detailed modeling with precise
numerical schemes. The fuel behavior is described using a three-ﬁeld model:– a continuous liquid ﬁeld,– a ﬁeld for droplets from fragmentation of the continuous ﬁeld,– a ﬁne solid fragments ﬁeld.
The Rayleigh-Taylor instabilities at the interface between the corium and the water
are also modeled.
To advance the modeling of the phenomena, the OECD/NEA launched the international SERENA project (2001–2005) [10], to which France (IRSN and CEA) contributed
alongside Germany, Japan, Korea, the United States and Russia. Its objective was todetermine the state of the art of knowledge about steam explosions, to compare
simulation codes with one another, to assess their ability to predict explosions andto identify the areas where uncertainty needs to be reduced.
The OECD/NEA's international SERENA-2 program [11], which succeeded it (2005–2012) included 12 tests conducted by CEA at the KROTOS facility, which was moved from
Ispra to Cadarache at the end of the FARO program, and by the Korea Atomic Energy
146.
147.

Experiments on energy COnversion during a steam explosion.
Multiphasic and thermohydraulic 3D simulation code.

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Research Institute (KAERI) at the TROI148 facility in Daejon, South Korea. The objective
was to gain a better understanding of the mechanisms that trigger steam explosions andthe effect of the type of materials used to represent corium, and to check the models'capacity to reproduce effects related to the geometry of the interaction zone.
The tests run at the KROTOS facility consisted of injecting 5 kg of molten corium(UO2–ZrO2 mixture but also more realistic compositions including the metallic phases oriron oxide) in a tube (one-dimensional geometry). An X-ray cinematography system wasused to ﬁlm the fragmentation of the jet in real time and the vaporization of the wateraround the corium fragments. In the TROI facility, larger corium masses of 20 kg(identical composition to those at KROTOS in the frame of the SERENA program) were
injected into a container of similar height and width (two-dimensional geometry
representative of a PWR reactor pit). The tests were deﬁned after preliminary calculations had been made with the various simulation codes. The analysis of the results was
discussed by international experts. The program also included code benchmarking
exercises and reactor-scale applications.
The SERENA and SERENA-2 programs advanced understanding of the ﬁne fragmentation of a corium jet in water and the explosions that can be caused by this. They
revealed the inﬂuence of corium fragment solidiﬁcation kinetics on the magnitude of the
explosion (rapid solidiﬁcation restricts the ﬁne fragmentation process, and therefore heattransfers and water vaporization). In general, the simulation codes developed globallygive a good overall prediction of the mechanical energy evaluated in tests using molten
alumina and overestimate the mechanical energy evaluated during tests with molten
corium. An analysis of the images taken during tests at the KROTOS facility shows that
instabilities of the steam ﬁlm surrounding the corium jet are correctly reproduced by the
MC3D code's models. The code also correctly predicts the attenuation of the shock wave
when the explosion occurs in a two-dimensional geometry like those studied in the TROI
facility.
However, the effects related to corium oxidation during the interaction are still notproperly understood. The research, particularly on fragmentation of the jet and oxidationof the corium, is continuing with theoretical and experimental analytical studies as part of
the ICE project (2013–2017), launched in response to ANR's call for RSNR projects
mentioned above and coordinated by IRSN in partnership with CEA, EDF and Nancy
University.

9.3.2.

Hydrogen-related risks

Hydrogen, which is produced mainly by oxidation of the zirconium in the cladding
during the core degradation phase, but also by the oxidation of the other metals in the
corium pool or in the basemat during the interaction phase between the corium and theconcrete, can accumulate in the containment. The amount of movement of theatmosphere in the containment has an impact on the hydrogen distribution, which
may therefore not be uniform. Locally, the hydrogen can reach high concentrations that
148.

Test for Real cOrium Interaction with water.

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exceed the threshold for the gaseous mixture to be ﬂammable; hydrogen detonation in acontainment is an event to be avoided.
The hydrogen distribution in the containment is the result of thermohydraulicphenomena which are complex to study due to the number of ﬂuids in the containment
atmosphere to be taken into account and the containment's compartmentalized
geometry: convection caused by temperature differences between the atmosphere
and walls and the possible use of spraying, condensation of steam on the containment
walls with the water droplets produced by the spraying, diffusion, turbulence, the effectof the hydrogen recombiners, etc. [12, 13].
To assess the risk of hydrogen combustion in the containment, IRSN uses its own
validated simulation codes (ASTEC, TONUS and the FLUENT code pending the implementation of P2REMICS) and criteria from its own research programs, enabling it todetermine which conﬁgurations are potentially dangerous. These tools enable it topredict hydrogen distribution and combustion, taking account of the effect of mitigatingmeans of core melt such as recombiners, spraying and the venting system.
To predict hydrogen distribution inside the containment, IRSN uses two approaches:modeling based on several compartments in which ﬂuid concentrations and temperatures are uniform, and more detailed multi-dimensional modeling with a large number ofmeshes. The ﬁrst approach, which is economical in terms of calculation time, is used in
the ASTEC code. The second is used in the TONUS code developed by IRSN incollaboration with CEA149, and in FLUENT. The two approaches are complementary:
the ﬁrst approach is used to identify scenarios requiring detailed analysis using the second
approach.
The validation of these simulation codes is reliant upon a number of experimental
programs being conducted in France or abroad, at well-instrumented facilities simulating
the geometry of a reactor containment on different scales:– the PANDA facility (Figure 9.6a), at the Paul Scherrer Institute (PSI) in Villigen,
Switzerland, consisting of four compartments with a total volume of 460 m3,
allowing the shape of ﬂuid ﬂows to be studied among other things;– the THAI150 facility (Figure 9.6b), run by Becker Technologies, in Eschborn,
Germany, consisting of several compartments with a total volume of 60 m3,
with thermally insulated walls, designed for studying the distribution andcombustion of hydrogen, and the operation of catalytic hydrogen recombiners;– the MISTRA facility (Figure 9.6c) run by CEA in Saclay, with a cylindrical internal
volume of 100 m3, suited to the study of steam condensation on the walls andspray droplets;– the TOSQAN facility (Figure 9.6d) run by IRSN in Saclay, with a cylindrical internal
volume of 7 m3, dedicated to the study of condensation and spraying, and ofthermal and mass exchanges with the sump water.
149.
150.

EDF is developing the SATURNE code.
Thermal-hydraulics, Hydrogen, Aerosols and Iodine.

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Figure 9.6a Diagram of the PANDA facility. © PSI.

Except in the THAI facility, for safety reasons hydrogen is simulated by helium, a gas withvery similar physical properties but with the advantage of not being combustible. Tests
carried out at the THAI facility have conﬁrmed the similarity of hydrogen and helium
distributions in the test volume.

Figure 9.6b Diagram of the THAI facility [14].

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Figure 9.6c View of the MISTRA facility. © A. Gonin/CEA.

Many national and international projects and programs have collected a large
amount of experimental data, enabling the scope of validation of the simulation codes
to be extended. In particular these include the international OECD/NEA programs SETH(SESAR THermalhydraulics, 2000–2004) at the PANDA facility, SETH-2 (2007–2010)
using both the PANDA and MISTRA facilities, THAI (2007–2009) and THAI-2 (2011–2014)
at the THAI facility. The SETH-2 program studied the phenomena that cause thedestabilization of stratiﬁed conﬁgurations, due to the effect of heat released by thehydrogen recombiners and spraying; because of its low density, the hydrogen is initiallyconcentrated in the upper part of the containment.
Effects of scale have also been studied as part of the ERCOSAM151-SAMARA program(2010–2014), supported by the European Commission and by ROSATOM152 (Russia), in
151.
152.

Containment thermal-hydraulics of current and future LWRs for Severe Accident Management.
Federal Agency on Atomic Energy.

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Figure 9.6d Diagram of the TOSQAN facility. © Stéphane Jungers/IRSN-source IRSN.

particular including tests at the PANDA, MISTRA and TOSQAN facilities alreadymentioned, as well as at the SPOT facility (60 m3) in Russia. The extrapolation of
the results to large scales was performed on the basis of preliminary calculations. It could
be conﬁrmed in the future by means of tests at the 2000 m3 Russian KMS facility(currently at the design stage).
Concerning combustion, it should be noted that a ternary mixture of air, water andhydrogen (Figure 9.7), considered to be ﬂammable, can lead to different combustion
regimes, which depend mainly on the maximum hydrogen concentration reached and the
distribution of this gas between the compartments:– laminar deﬂagration,– rapid deﬂagration,– the deﬂagration-detonation transition (DDT).
– detonation – a phenomenon that can have disastrous consequences and thatshould therefore be avoided.
In a mixture known to be ﬂammable, combustion may be triggered by an energy source ofa few millijoules. Consequently, in the presence of electrical power sources or hot points, it
seems probable that ignition would occur rapidly once the gas mixture enters the ﬂammability domain. In contrast, a much more powerful energy source (at least 100 kJ) is required totrigger a stable detonation. This explains why direct detonation of the gaseous mixture in theAnglais_Book

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Figure 9.7 Shapiro diagram for hydrogen, air and steam mixtures (please note that the detonability
limits shown have been called into question by later studies). @ DR.

containment can be ruled out; the only mechanism considered likely to cause detonation isﬂame acceleration and the deﬂagration to detonation transition.
Two types of criteria have been deﬁned and adopted:– the criterion "σ" related to ﬂame acceleration. The quantity σ is the mixture's
expansion factor, a ratio of fresh and burnt gas densities at constant pressure. It is
an intrinsic property of the mixture in question; the critical value σ∗ above which
ﬂame acceleration is possible depends on the initial temperature of the gases andthe stability of the ﬂame; it has been determined using the results of many
experiments at different scales and in different geometries;– the criterion "λ": similarly, prerequisite conditions have been deﬁned for assessing
the possibility of a deﬂagration–detonation transition (DDT). They are based oncomparison of a characteristic dimension of the space being studied with thedetonation cell size (denoted λ), which characterizes the mixture's reactivity.
Flame propagation after ignition, in an air, steam and hydrogen mixture, has been astudy theme of many experimental programs [15]. The purpose of these tests is twofold:to study the transition between combustion modes and to produce a database ofexperimental data for validating simulation codes.
On this theme, IRSN is engaged in a long-term collaboration with the CNRS Institutefor Combustion, Aerothermal Engineering, Reactivity and Environment (ICARE) inAnglais_Book

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Orléans. Besides determining the ﬂammability limits of the mixtures, work done as partof this collaboration has enabled the ﬂame acceleration criterion "σ" to be reﬁned.
An analytical test program has therefore been run since 2004 at ICARE's ENACCEF153
facility (Figure 9.8). This facility was designed to represent, on a 1/24 scale, a steam
generator bunker and the containment dome of a 900 MWe PWR. ENACCEF consists of adome with a volume adjustable between 781 and 958 liters (representing the free volumeof the containment above the steam generator), connected to a cylindrical tube that is
3.2 m high, has an internal diameter of 0.17 m (representing the steam generator bunker)
and can contain obstacles, in which ﬂame acceleration can be studied. The instrumentation it contains includes photomultiplier and pressure sensors to measure the progressionof the ﬂame front and the pressure generated along the tube. In addition, thecomposition of the gaseous mixture is analyzed at several places in the test section;laser Doppler velocimetry (LDV) and particle image velocimetry (PIV) measurement
techniques are used to measure the velocity ﬁeld of the ﬂow before the ﬂame arrives.
The tests have been used to reﬁne the expression of the critical values of expansion
factor "σ" for homogeneous mixtures and extend it partially to heterogeneous mixtures.
Tests on a larger scale had also been conducted previously in the 62 m-long RUT
facility at the Kurchatov Institute in Moscow, Russia. This facility consists of a straightcanyon connected to a cavity, which is itself connected to a second, shorter canyon
curved at the end. All three zones are rectangular in cross-section and may be obstructed

Figure 9.8 Diagram of the ENACCEF facility [16].

153.

Flame Acceleration Enclosure.

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by obstacles. This geometry can be used to study both monodirectional ﬂame acceleration in the canyons and more complex 3D effects or interactions in the cavity. The
mixtures used contained hydrogen, air and sometimes a diluent (steam). A test program
run on behalf of IPSN in partnership with German researchers at the Karlsruhe center(FzK) studied the different combustion regimes and obtained values for the criteria "σ"(ﬂame acceleration) and "λ" (detonation cell size), which are used for predicting different
combustion regimes in the ASTEC and TONUS simulation codes154.
This program was completed by the HYCOM (integral large scale experiments on
HYdrogen COMbustion for severe accident code validation) program, run from 2000 to2003 with support from the European Commission. Its aim was to study ﬂame
acceleration in mixtures of hydrogen and air (without water), particularly validationof the "σ" criterion. The effect of the expansion of the burned gases (piston effect) andthe effect of compartmentation were studied in the 12 tests conducted in the RUT
facility. The effects of pressure vents, concentration gradients and changes in crosssection were studied in 46 more analytical tests run at the smaller-scale DRIVER and
TORPEDO facilities at the Karlsruhe center in Germany. These facilities consisted of two
cylindrical tubes with diameter 174 mm and 520 mm respectively and length 12 m and
6 m respectively. The program, in which IPSN (then IRSN) played an active part with thesupport of EDF, also included a reactor-scale simulation code benchmark exercise,studying a containment geometry similar to that of EPR.
Research on hydrogen distribution in a reactor containment (effects of spraying andrecombiners – see below), on combustion (deﬂagration to detonation transition on alarge scale) and explosion is continuing as part of the MITHYGENE155 program (2013–2018). The program was launched in response to ANR's call for RSNR projects mentioned
earlier, and is coordinated by IRSN in partnership with CEA, ICARE, AREVA Expansion, theGerman Jülich research center (KfJ), EDF and Air Liquide.
To reduce the hydrogen content of a containment in a core melt accident, passive
catalytic hydrogen recombiners (Figure 9.9) have been installed in all the containmentsof French reactors.
The behavior of recombiners in situations representative of core melt accidents has
been the subject of a number of test programs, conducted primarily by manufacturers(Siemens, AECL, etc.). IPSN ran the H2PAR156 test program (1998 to 2000) in partnershipwith EDF, to study the possible poisoning of catalytic sites in recombiners by ﬁssion
products, particularly the halogens. The facility consisted of an oven that could bringapproximately 10 kg of a UO2–ZrO2 mixture up to a temperature of 2900 °C. Inert
compounds representative of the main ﬁssion products were included in this mixture. The
aerosols and the gases released by the molten mixture spread inside a 7.6 m3
containment made from Terphane, in which a mock-up representing a hydrogen
recombiner used in French reactors was installed. The tests also provided an opportunity
to study the risk of the hydrogen in the recombiner igniting. The atmosphere in the
154.
155.
156.

The equivalent German code, COM3D, uses equivalent criteria.
Hydrogen Mitigation.
Hydrogen Passive Autocatalytic Recombiners.

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Figure 9.9 Diagram of a passive catalytic hydrogen recombiner. © IRSN.

containment contained a mixture of air, hydrogen and steam; the proportions were one ofthe parameters studied. The impact on recombination performance of various geometricparameters (number of catalytic plates, height of stack) and chemical parameters (several
catalytic plates replaced by neutral chemical plates) was also studied.
The KALI H2 test program, run by CEA at the end of the 1990s in partnership with EDF,
also provided an opportunity to study the effects of humidity, exposure to smoke fromcable ﬁres and carbon monoxide on recombiner performance. The results of this program
revealed that humidity has little impact on recombiner performance. They also showedthat recombination was deactivated by carbon monoxide in low-oxygen atmospheres
and that there was a loss of performance of approximately 5% for recombiners exposed
to cable ﬁre smoke. The ﬁre on May 10, 2011 in the containment of reactor 2 at the
Ringhals nuclear power plant in Sweden also revealed the possibility that catalytic
recombiner plates could be perforated by chlorinated products from the combustion ofplastics. Investigations are therefore being carried out by IRSN and the Jülich research
center to study the effect of cable ﬁres on recombiners. The ﬁrst results of this study haverevealed that the start of recombination is delayed in the case of catalytic plates exposedto smoke from cable ﬁres [17].
During the Phebus FPT3 test, specimens from recombiner plates made by various
manufacturers were exposed for several dozen minutes to the radioactive aerosols andsteam in the tank simulating the containment. The low oxygen content of the tank (forsafety reasons) slightly complicated the analysis of the results, which nevertheless notrevealed unexpected phenomenon.
To investigate the effect of a lack of oxygen, of carbon monoxide or of the presence ofdust on the startup and effectiveness of recombiners, IRSN is working with the Jülich
research center in Germany, which has the REKO experimental platform with a numberof facilities on different scales (Figure 9.10). On the basis of the experimental results,
IRSN is developing the SPARK simulation code, which can be used to take account of all
the physico-chemical phenomena governing recombination. This collaboration has

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Figure 9.10 View of the REKO platform. © Forschungszentrum Jülich.

already enabled limits to be deﬁned for hydrogen ignition by recombiners. These results
were conﬁrmed by the test program conducted as part of the OECD THAI and THAI-2
projects. The ﬁndings of these various tests allow the improvement of the models
integrated into the simulation codes used at IRSN (including ASTEC) and as a result to
reﬁne the assessments of hydrogen explosion risk in the annulus space between dual
containments, which IRSN is required to produce for 1300 MWe and 1450 MWe reactors,
and in fuel storage buildings (and the ITER facility).

9.3.3.

Direct containment heating

Direct containment heating (DCH) was originally studied experimentally and analytically by the Americans in the 1990s for the PWRs at the Zion, Surry and Calvert Cliffs
sites. If a reactor vessel fails under pressure, corium could be ejected and fragmented in
the reactor pit and the fragments could be carried into the containment. The metallic
components of corium (such as zirconium and iron) would oxidize on contact with thesteam and air in the containment, releasing heat and producing a quantity of hydrogenAnglais_Book

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inside the containment which, depending on the mass of corium ejected and hydrogenproduced, would threaten its integrity.
The American studies concluded that the corium dispersal and the amplitude of theresulting overpressure were very dependent on the geometry of the reactor pit.
Analytical tests in which the corium was simulated by water and the geometry studied(reactor pit and communication channels with the free volume of the containment) wasthat of a 900 MWe reactor, were conducted at KAERI in South Korea on behalf of IPSN in
the late 1990s. The results enabled an empirical model to be created that was integrated
into the ASTEC simulation code, giving the fraction of corium ejected from the reactor pit
as a function of the geometric parameters and ejection conditions.
In the early 2000s, IRSN launched a new research program in partnership withGerman researchers in the DISCO157 facility at the Karlsruhe research center, consistingof two experimental devices. The ﬁrst was a plexiglass mock-up enabling the ﬂows to
be seen, with the corium simulated by water, gallium or Wood metal158, an alloy witha low melting point. The second was used for more representative experiments, where
the corium was represented by a mixture of 10 kg of molten alumina and iron obtainedusing an exothermic oxidation-reduction reaction, brought to approximately 2700 °C.
Several geometries were studied in these devices: those of an EPR, a P'4-type reactor(1300 MWe), a VVER-1000 and a German KONVOI PWR. The program, which was
funded by IRSN159, consisted of 60 analytical tests at low temperature and 12 integral
tests at high temperature on a 1/16 scale model of the containment of a P'4 reactor
representing the geometry of the reactor pit, the compartments in communication withthe reactor pit and the dome.
The experimental results were analyzed using the MC3D simulation code
(Section 9.3.1). The analysis established a correlation giving the corium fraction dispersed
axially in the reactor pit as a function of the pressure in the vessel at the moment when it
fails. Another ﬁnding is that the pressure increase in the containment is mainly due to
combustion of the hydrogen produced during dispersal of the corium and, to a lesser
extent, of the hydrogen already in the containment when the vessel failed. The rate ofhydrogen production after vessel failure is correlated with the fraction of corium
dispersed axially in the dome.
These results have signiﬁcantly improved the direct containment heating model in
the ASTEC simulation code, enabling it to be applied with greater conﬁdence to thegeometry of French PWRs.

157.
158.
159.

DISpersion of simulated COrium.
Wood metal is an eutectic alloy of bismuth, lead, tin and cadmium. Its melting point isapproximately 70 °C.
Except in the case of the KONVOI reactors.

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9.4. Radioactive releases
The amplitude and nature of radioactive releases depend on three main processes:– the emission of radioactive products by the fuel when it is damaged,– the transporting of these radioactive products into the reactor's circuits,– their behavior in the containment.
The volatility of the radioactive products depends on their physical state, which in turn isdetermined by their chemical nature. Chemical reactions therefore play a key role, bearing
in mind that they occur at a very wide range of temperatures (from 50 °C to 3000 °C) andinvolve a large number of elements: ﬁssion products, the fuel and the core's structural
materials (zirconium, steel, materials that absorb neutrons such as AIC and/or boron
carbide, etc.). These materials release a considerable mass of steam and aerosols, which canbe much higher than the mass of all the ﬁssion products put together. The nature of thematerials used in the walls of the circuits and containment also plays a part.
The research (analytical experiments) conducted speciﬁcally on the above three
processes is discussed in more detail in the sections that follow. The ﬁndings concerning
each of these processes from the more general (integral) tests in the Phebus-FP program
will then be explained.
The reference information [18] provides a summary of the knowledge acquired about
radioactive releases in the event of a core melt accident.

9.4.1.

Emission of radioactive products by fuelThe emission of ﬁssion products by the fuel while it is being heated has been studied
in France as part of the HEVA160 program (eight tests in the 1980s) and VERCORS161
program (17 tests from 1989 to 2002) run by CEA at its Grenoble center on behalf of IPSN
and EDF. At around the same time, similar programs were being run abroad:– the HI162 tests (six tests) and VI tests (seven tests) at the Oak Ridge National
Laboratory (ORNL), United States,– the VEGA163 tests (10 tests) by the Japan Atomic Energy Research Institute(JAERI), Japan,– the AECL-CRI tests (more than 300 tests) by Atomic Energy of Canada Limited(AECL), at the Chalk River Laboratories, Canada.
All these tests consisted of heating, in an oven, specimens of fuel that had been in areactor, to identify the ﬁssion products released and measure the quantities, using in-line
gamma-ray spectrometry and other radiochemical analysis techniques on samples. The
main parameters studied were the type of fuel, its burnup rate, whether or not there was
160.
161.
162.
163.

Helium, Steam.
Realistic Veriﬁcation of the Containment of Reactors.
Horizontal Induction.
Veriﬁcation Experiments of radionuclides Gas/Aerosol release.

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cladding present, the maximum temperature reached during the test and the composition of the gas ﬂowing around the fuel: a mixture of steam and hydrogen in variable
proportions, inert gases, air.
In France, the ﬁrst six tests in the VERCORS program (VERCORS-1 to 6) wereconducted between 1989 and 1994 on samples of UO2 fuel heated to 2300 °C. Then, two
test series, VERCORS-HT (three tests) and RT164 (eight tests), at higher temperatures up
to the melting of the fuel, followed between 1996 and 2002. These tests used UO2 and
MOX fuels in the form of fuel rod segments or a debris bed. The device used for the HT
tests included a cooled, instrumented channel downstream of the fuel, allowing the
behavior of the ﬁssion products emitted during their transfer in a reactor coolant system
to be studied, and in particular their deposition on the walls.
The fuel specimens generally consisted of a fuel rod segment containing three fuelpellets and their cladding, sampled from fuel rods that had spent several years in an EDF
reactor. In most of the tests, the segment was re-irradiated for several days in a research
reactor (the SILOE reactor in Grenoble, then the OSIRIS reactor at the Saclay center once
the SILOE reactor had been shut down) in order to restore the inventory of short-lived
ﬁssion products such as iodine-131 that would have disappeared after the rod was
unloaded from the core of the origin reactor.
The tests enabled a database of very precise experimental data to be obtained on theemission kinetics of ﬁssion products and actinides in different conﬁgurations. This
database was used to develop and validate empirical models integrated into the ASTEC
simulation code. A more sophisticated simulation tool, MFPR165, based on a moredetailed description of the physical and physico-chemical phenomena, was also developed by researchers at the Nuclear Safety Institute of the Russian Academy of Science(IBRAE) in Moscow in collaboration with IRSN; it is used to analyze the results of thetests.
It emerged from all this research that it is possible to distinguish broadly speaking
between the following:– the volatile ﬁssion products such as the isotopes of krypton (Kr), xenon (Xe),iodine (I), cesium (Cs), rubidium (Rb), tellurium (Te), and antimony (Sb), which arealmost entirely released by the fuel when it reaches 2300 °C;– the semi-volatile ﬁssion products, such as the isotopes of molybdenum (Mo),barium (Ba), yttrium (Y), rhodium (Rh), palladium (Pd), and technetium (Tc), theemission rate of which lies between 10 and 100% and depends heavily on theoxygen potential of the ﬂuid ﬂowing along the fuel being studied166;– the low-volatile ﬁssion products such as the isotopes of strontium (Sr), niobium(Nb), ruthenium (Ru), lanthanum (La), cerium (Ce), europium (Eu), neptunium (Np)
164.
165.
166.

Release of Transuranics.
Module for Fission Product Release.
The molybdenum oxides are much more volatile than the metal molybdenum, which explains why
greater quantities are emitted in steam; the opposite is the case with barium, which is emitted
more readily in its metal form in hydrogen.

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and some actinides such as uranium (U), of which 10% maximum is generally
emitted, with the notable exception of ruthenium, which is more volatile when thefuel is exposed to air rather than steam;– the non-volatile ﬁssion products and the actinides such as the isotopes ofzirconium (Zr), neodymium (Nd) and plutonium (Pu), of which no emissions
were really detected during the different tests.
The VERCORS facility was dismantled and a new facility that performed the samefunctions, VERDON, was built by CEA on the Cadarache site. The facility is used to study
not only the release, but also the transporting and deposition of the ﬁssion products inthermal-gradient tubes. Tests with MOX and UO2 fuel with high burnup rates, in a
mixture of hydrogen and steam, and a test in the presence of air, were performed in thisnew facility as part of the International Source Term Program (ISTP, 2005–2013).
The emission of ﬁssion products during core reﬂooding was studied experimentally in
the LOFT-FP tests already mentioned in Section 9.1. They revealed signiﬁcant releases ofvolatile species (iodine and cesium isotopes) during runaway oxidation of the cladding.
Tests were also carried out in the early 2000s as part of the ISTC 1648 research program
run by the Scientiﬁc Research Institute of Atomic Reactors (NIIAR) in Russia, on sectionsof irradiated fuel rods from a VVER reactor heated to 1700 °C.
The emission of steam and of aerosols released by control rods containing a silver,indium and cadmium alloy (AIC) has been modeled by means of measurements made in
the 1990s during speciﬁc tests (EMIS167 program) conducted by CEA in Grenoble on
behalf of IPSN on material specimens brought to melting point. AIC specimens were also
vaporized during certain VERCORS tests (EMAIC168 tests) to study their inﬂuence on thechemistry of the ﬁssion products released.
Once emitted by the core, the radioactive products and elements from the degradationof the core structures undergo rapid cooling. The vapors condense, forming aerosols(homogeneous nucleation), or condense on the aerosols already formed (heterogeneous
nucleation) and on the reactor coolant system walls. The vapor of certain ﬁssion products
such as cesium and iodine can also react chemically with the materials in the RCS walls (Ni,Cr, Fe) and be ﬁxed, to a greater or lesser extent reversibly. The heating of the wall due tothe heat released by the ﬁssion products or the partial pressure reduction of these elementsin the ﬂuid in the RCS can cause some of the deposited ﬁssion products to be re-emitted.

9.4.2.

Transporting of radioactive products into thereactor's circuits

The mechanisms governing the transporting – and deposition – of aerosols in a circuit
has been the subject of many experimental and theoretical studies, often conducted
outside the nuclear ﬁeld. This work has allowed basic models to be established that have

167.
168.

Fission Product Emission.
Silver, Indium and Cadmium Emission.

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been integrated into severe accident simulation codes such as ASTEC. It involves theagglomeration of the aerosols contributing to an increase in their size, thermophoresisand diffusiophoresis169, phenomena responsible for deposition in the presence, respectively, of a temperature gradient between the carrier ﬂuid and the wall or of condensation of the carrier ﬂuid on the wall, gravitational settling or impaction on a wall if thecarrier ﬂuid changes direction, Brownian or turbulent diffusion, etc.
Analytical test programs have been used to check these models, particularly the
TUBA and TRANSAT programs run by IPSN in the 1990s to study thermophoresis anddiffusiophoresis, and the DEVAP170 program run at the same time by CEA on IPSN's
behalf at the Grenoble research center, consisting of measuring deposits by condensationof cesium iodide and hydroxide injected into a tube with a controlled thermal gradient.
The temperature proﬁle of the wall of this kind of tube decreases between the inlet andthe outlet, which is representative of the thermal conditions in the reactor coolant
system (RCS) of a PWR in accidental conditions. The FALCON program in the United
Kingdom in the early 1990s also studied experimentally the deposition of simulated
ﬁssion products (I, Cs) in controlled thermal-gradient tubes and the interaction between
vapors and aerosols.
The aerosols deposited can go back into suspension if there is a sudden increase in
ﬂow rate of the ﬂuid in the RCS, e.g. following an attempt at core reﬂooding. The
phenomena involved were studied experimentally during the 1990s in the STORM facility
at the Ispra Joint Research Centre in Italy.
Tests were also performed in the ARTIST171 facility at PSI, Switzerland, as part of aninternational program, to study the retention of aerosols in the secondary system of asteam generator. The results obtained made it possible to evaluate more precisely thereleases in the event of steam generator tubes rupture during reactor operation.
Given the uncertainties concerning the behavior of iodine in the reactor coolant
system, conﬁrmed during the Phebus-FP tests (see further on), IRSN built an experimental device named CHIP at its laboratories in Cadarache, to study the behavior ofdifferent chemical species in a reactor coolant system. The device consists of:– several generators of chemical species in vapor or gas form: mixture of hydrogenand steam (H2/H20), simulators of highly volatile, reactive ﬁssion products such as
diiodine (I2), cesium (Cs), molybdenum (Mo), reagents present in large quantities
during core melt accidents, such as boron (B), silver (Ag), indium (In), cadmium(Cd) from the reactor's control rods melting;– a heating system to bring the various reagents to a temperature between 1500 °C
and 1600 °C;

169.
170.
171.

Movement of the aerosols caused by diffusion in a gaseous environment of water molecules to acold wall on which they condense.
Volatile Fission Product Deposits in the Vapour Phase on Reactor System Surfaces.
Aerosol Trapping in a Steam-Generator.

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117– a controlled thermal gradient circuit made from the same metal as the reactor
coolant system, into which the reagents are injected; this circuit includes hightemperature (500 °C to 900 °C) sampling devices, and devices for sampling at
lower temperatures (approximately 150 °C) for separating the aerosol andgaseous species of iodine at both temperature levels;– a ﬁlter for recovering at the circuit outlet the aerosols formed during the reagent
cooling process so that they can be analyzed.
Tests were conducted by IRSN as part of the International Source Term Program(ISTP, 2005–2015) in order to quantify the fractions of volatile iodine produced in thereactor coolant system for reaction systems representative of the conditions of a coremelt accident in a water reactor. An analysis of the chemical species that make up theaerosols collected in the ﬁlter also provides more information about the chemical species
of iodine that are transported in a condensed form. Analysis of these results provides abetter understanding of the chemical reactions involved in the production of volatile
iodine and their kinetics, and also improves the iodine transport models in the ASTEC
simulation code, so that potential radioactive releases in the event of a core melt
accident can also be characterized.

9.4.3.

Radioactive product behavior in the containment

The aerosols emitted within the containment are also subject to the phenomena of
agglomeration, deposition and possibly re-suspension. The basic physical phenomena are
the same as those at work in the reactor coolant system. Some components of theaerosols may be hygroscopic and absorb the steam in the containment, which increases
their mass and accelerates their deposition by settling. The main deposition phenomenaof aerosols in the containment are settling and diffusiophoresis. Spraying, which could be
used to cool the atmosphere in the containment and reduce the pressure there, can also
accelerate the deposition of aerosols.
The phenomenon of aerosol spraydown in particular has been studied at IRSN, during
an analytical test program at its CARAIDAS facility in Saclay. This program was used to
study and model the condensation of steam on water droplets and its evaporation,
aerosol collection by the droplets when they fall and the absorption of gaseous iodine by
the droplets. Experiments on the effects of spraying have also been conducted on a larger
scale at the TOSQAN facility already mentioned. The results were used to develop andvalidate an aerosol deposition model integrated into the ASTEC simulation code.
A large amount of the radioactive products released by the core ultimately ends up in
the sump water following their spraydown, leaching of the containment walls by thesteam condensates or their deposition by settling. Consequently, the sumps are intensely
radioactive. Moreover, with the notable exception of silver iodide (AgI), the majority ofthe metal iodides (CsI, RbI, CdI2, InI) are water soluble and form I – ions. Due to the actionof the radiolysis products of the water and a large number of radiolytic and thermally
activated chemical reactions, the I – iodide ions can be oxidized to become volatile I2
iodine, which is then released into the containment atmosphere. The formation ofvolatile iodine depends on many parameters, the most important being water pH. At aAnglais_Book

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to implement in the event of an accident (addition of soda). The water in the sumps also
contains organic materials originating mainly from the immersed paintwork. The iodine
reactions with the organic radicals can produce volatile organic iodides such as methyl
iodide CH3I, which are also released into the containment atmosphere.
These different chemical reactions have been studied in France at IRSN's laboratories
during analytical tests on specimens using a γ irradiator (IODE program in the late 1980s–
early 1990s).
Because the iodine emanations from the sump also mean that mass is transferredfrom the water to the containment atmosphere, more global tests were then performedin the 1990s at CEA's CAIMAN facility on behalf of IPSN in partnership with EDF. Thisconsisted of a 300 liters containment with a 25-liters sump in the ﬁeld of a cobalt-60
source delivering a γ dose rate of 10 kGy/h. Plates coated with the same type of paint as
used in French PWRs could be placed in the sump water and in the containment
atmosphere. The water in the sump could be brought to 90 °C and the atmosphereto 130 °C. The iodine placed in the water in soluble form was mixed with a tracer(iodine-131). A system of selective ﬁlters, known as Maypack172, was used to identify thedifferent iodine species produced (I2 and CH3I) and their physical forms (aerosols orgases) and to measure the quantities using a gamma-ray spectrometer. The SISYPHE
program, conducted at IPSN in support of the Phebus-FP program in a similar 10 m3
containment, also contributed signiﬁcantly to the development of models for iodine
compound mass transfers in sump evaporation conditions.
The different programs provided a better understanding of the phenomena governing
the production of volatile iodine species in sumps and enabled the modeling of thosephenomena to be improved. They also identiﬁed the role of paints, the iodine absorption
rate of which increases signiﬁcantly with temperature. However, the reaction is not
irreversible and some of the absorbed iodine revolatilizes in an organic form. With thisresearch, data were gathered on the phenomenon and enabled initial correlations to beestablished for use in the simulation codes used to evaluate potential radioactive releases
in the event of a core melt accident.
In Canada, a practically identical facility to the CAIMAN facility, the Radioiodine Test
Facility (RTF) was used by the AECL at the Chalk River laboratories to run many
experimental programs on volatile iodine production in sumps. In particular we will
mention the Phebus-RTF tests conducted as part of the international Phebus-FP program.
They were used to quantify the effects associated with the presence of insoluble silver in

172.

Measuring device invented by F.G. May in the 1960s, in which a sampled gas circulates insuccession through a membrane trapping aerosols then two ﬁlters: the ﬁrst consisting of a knit
mesh, which is a silver-coated braid that traps molecular iodine, the second consisting of active
charcoal (impregnated with potassium), which absorbs the other forms of gaseous iodine,
including the organic iodine. Measurement of the concentrations deposited at the different
stages gives the desired result. Improvements were subsequently made by various IRSN
researchers, particularly by replacing the active charcoal with zeolite and measuring the deposits
in real time by γ scanning when the iodine is radioactive.

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119sump water, observed during the ﬁrst Phebus-FP tests. When silver is signiﬁcantly inexcess that reduces the concentration of I– ions and thus volatile iodine production in thesumps, even when the pH is acid.
Analytical tests were performed in the context of two international programs run by
CNL173 (formerly AECL) under the guidance of the OECD/NEA, in which France took part:the Behavior of Iodine Project (BIP, 2007–2011) and the Behavior of Iodine Project –Phase 2 (BIP, 2011–2014). Their objective was to study organic iodine production in moredetail and establish more accurate models than those obtained to date, so that results
obtained in the laboratory could be extrapolated with greater conﬁdence to the scale of areactor.
Given the uncertainties over iodine behavior, IRSN built a new facility, EPICUR, at its
laboratories in Cadarache to study in more detail the physico-chemical processes ofvolatile iodine production; the tests began in 2005. The facility consists of a 5-liters tankthat can be brought to a temperature of 120 °C and placed in the ﬁeld of six cobalt-60
sources emitting an average γ radiation dose rate of 2.5 kGy/h. The iodine is traced by theiodine-131 and a Maypack scanned in real time by a gamma-ray spectrometer allows
accurately measuring the production kinetics of the volatile iodine and whether it ismolecular or organic iodine. A series of 30 tests was conducted as part of the ISTP
international program (2005–2013), allowing the formation of organic iodine frompainted surfaces, whether immersed or not, to be studied; the tests revealed theimportance of organic iodine production from painted surfaces when they are in thetank atmosphere.
The tests were also used to study the formation and stability of the iodine oxides(IOx) formed under radiation in the containment atmosphere. Indeed, the iodine in thecontainment atmosphere in its molecular and organic forms can react with the radiolysis
products of the air, ozone and nitrogen oxides, to form iodine oxides and nitroxides,
which are not volatile. These aerosols are likely to be deposited in the containment or inﬁlters if the containment is vented. Preliminary tests at a laboratory in Germany (PARIS
program), to help with understanding the results from the Phebus-FP program, hadalready showed the importance of these reactions, even for low molecular iodine
concentrations. The IOx formation and decomposition mechanisms are still being
researched as part of the OECD/NEA STEM174 and STEM-2 programs being run by
IRSN in order to develop the associated models.
The EPICUR facility is also being used, particularly as part of the OECD/NEA STEM
program (2011–2015), to study the stability of iodine aerosols subject to radiation at
different humidity levels, the effect of aging of paints on their ability to adsorb iodine andrelease volatile organic iodides, and lastly the mechanisms for long-term revolatilizationof iodine deposited on the walls or in ﬁltration media. The analysis of the results should
reduce the uncertainties about predicting potential iodine releases in the event of areactor core melt accident.

173.
174.

Canadian Nuclear Laboratories.
Source Term Evaluation and Mitigation.

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In the early 2000s, tests conducted by IPSN at the laboratory in Saclay (RECI175 tests)
showed that heating the air loaded with aerosols passing through hydrogen recombiners
can lead to decomposition of the metal iodides (typically CsI, AgI, InI and CdI2) depositedon these aerosols by forming volatile iodine compounds such as I2, HOI and HI. Tests on alarger scale conducted under more realistic conditions at the THAI facility mentioned
earlier (THAI program) conﬁrmed this phenomenon and enabled decomposition rates ofaround a few percent to be measured.
In parallel with all these analytical research programs, programs were run using largescale research facilities. As we have seen, chemistry plays a key role in the behavior of the
ﬁssion products emitted during core melt and evaluating the resulting releases. For thisreason it was considered essential to check, by means of "integral" experiments, i.e.
experiments that are as close as possible to reality, that the phenomena and theirpotential interactions are properly accounted for in the models used in computer codes
for calculating releases. In the 1980s and 1990s, the Americans and Canadians had
already conducted experiments in a reactor using irradiated fuel (PBF tests, LOFT-FP tests
and FLHT tests mentioned earlier), in which appropriate instrumentation was used to
identify and quantify the ﬁssion products emitted by the fuel. However, none of theexperiments modeled the reactor containment. Moreover, the information obtained
about the chemical forms in which the main ﬁssion products (Cs, I) were emitted wherevery imprecise.

9.4.4.

Contribution of the Phebus-FP program on the various
processes involved in radioactive releases

To ﬁll these gaps, IPSN (then IRSN) ran the Phebus-FP program (1988–2012) at the
PHEBUS experimental facility, already described in detail in Section 9.1.
The program's main ﬁndings concerning ﬁssion products and/or the validation of theassociated models are as follows [7]:– the models for the emission of ﬁssion products and actinides when fuel degrades,
deduced from the VERCORS tests, are conﬁrmed except for barium which isreleased in small quantities, probably as a result of the formation with zirconium
of a less volatile chemical compound;– the models for the transporting and deposition of aerosols allows their observed
behavior to be reproduced overall, both in the reactor coolant system and in thecontainment;– cesium is transported in the form of cesium molybdate (CsMoO3), a less volatile
species than the cesium hydroxide (CsOH) commonly assumed by experts to be
the form and used until then in all the tests of separate effects to study the
behavior of cesium;

175.

Recombiner and Iodine.

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121– a fraction of the iodine released into the containment at a break in the reactor
coolant system is gaseous, whereas the equilibrium thermodynamics calculations
predict that all the iodine emitted by the fuel should at 150 °C be in a condensed
form, cesium iodide (CsI); the reactor coolant system therefore constitutes afurther source of gaseous iodine in the containment, which needs to be taken into
account when calculating releases;– this fraction of a few percent in tests performed with a silver, indium and cadmium(AIC) alloy as the control rod material, reaches nearly 100% in tests where the AIC
is replaced by boron carbide;– even when the sump water is acidic, there is very little iodine gas emanation fromthe sump in tests run with AIC since silver, as studied subsequently in more
analytical tests, ﬁxes the I – ions.
– after rapidly decreasing, the volatile iodine concentrations in the containment
stabilize after a few days at a value that is approximately identical regardless ofthe test. This reﬂects an equilibrium between the various chemical reactions that
produce and destroy volatile forms of iodine in the containment: reactions in thecontainment atmosphere with the products of radiolysis of the air (ozone,
nitrogen oxides), transfers between the atmosphere and paints on the walls,
transfers between the atmosphere and the sump water, etc.
Iodine's unexpected behavior observed during the Phebus-FP tests (reactor coolant
system as the main source of gaseous iodine in the short-term, concentration notchanging in the containment in the long-term) led to a reorientation of the research
programs studying the elementary phenomena (ISTP, STEM, BIP, BIP2, THAI, THAI-2),
which until then had focused on studying radiochemical and chemical reactions in the
aqueous phase (IODE, CAIMAN, RTF, etc.), because at that time the sump water wasconsidered to be the only source of volatile iodine production. Very analytical andtheoretical research was also begun as part of the process of setting up a laboratory"without walls", C3R176, involving IRSN, the Laboratory for the Physical Chemistry of
Combustion Processes and the Atmosphere (PC2A) in Lille and CNRS.

9.4.5.

Aspects to be exploredA) Iodine-related and ruthenium-related aspects
In some accident conﬁgurations (after the vessel has failed, during an accident in areactor that is shut down with the vessel open for reloading fuel, or during a spent fuelpool uncovery accident), fuel could degrade in the presence of a steam and air mixture.
The Canadian AECL-CRL tests already mentioned revealed that substantial amounts ofruthenium were released under these conditions whereas, with a steam and hydrogenmixture, very little ruthenium was released as we saw earlier. Ruthenium is highly
radiotoxic, and particularly its two isotopes ruthenium-103 and ruthenium-106, whichhave half lives of 39 days and 369 days respectively. The VERDON-2 test conducted in
176.

Laboratory of Chemical Kinetics, Combustion and Reactivity.

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2012 at CEA's VERDON facility as part of the ISTP program, on a MOX fuel irradiated at
approximately 60 GWd/t, conﬁrmed this phenomenon and enabled it to be better
quantiﬁed. When it is transported in the circuits, ruthenium changes into less volatile
species. However, as shown by analytical experiments performed in Hungary and Finland
in the early 2000s, ruthenium tetroxide (RuO4) in gaseous form, despite being unstableat room temperature, could continue to exist in small quantities and could accumulate in
the containment. Tests conducted at the EPICUR facility as part of the ISTP program
studied the behavior under radiation of this oxide in the containment and enabled it to bemodeled.
Studies of the complex behavior of iodine and ruthenium in the circuits was pursued
until 2015 as part of the OECD/NEA STEM program and continue beyond that date with
the MIRE177 and STEM-2 programs. If necessary, other tests could be performed beyond
that at the CHIP facility to study and model in more detail the effect of the control rod
materials on the chemical forms in which iodine is transported in the reactor coolant
system.

B) Filtration of atmospheric releases
The course of the Fukushima Daiichi accident attracted attention to the value ofﬁtting containment depressurization systems with very efﬁcient ﬁlters, in particular thatwill capture all the volatile forms of iodine (diiodine [I2], organic iodides such as methyl
iodide [CH3I]) even under core melt accident conditions178 and for periods of several days.
Research carried out recently at the PSI in Switzerland tended to show that theeffectiveness of trapping all the volatile forms of iodine by aqueous solutions could
be considerably increased by using an alkaline thiosulfate and quaternary amine
combination. Moreover, in response to the call for RSNR projects launched by ANR
in 2012 in the wake of the Fukushima Daiichi accident, IRSN is coordinating the MIRE
research program (2013–2017) in partnership with EDF, AREVA and various universities(Lille, Marseille, Nancy, Nantes) on improving the retention of all forms of iodine in
ﬁltration media179. Lastly, the Passive and Active Systems on Severe Accident source
term Mitigation (PASSAM, 2013–2016) research and development project coordinated by
IRSN and supported by the European Commission brings together eight partners
including EDF to explore possible improvements to the ﬁltration systems in service
and to study innovative devices that would be much more efﬁcient.

C) A "water-borne release" counter-measure to be explored
IRSN is conducting studies of a geotechnical barrier as a "waterborne-release countermeasure" in a core melt accident situation. If the event of a core melt accident in apressurized water reactor that does not have a core catcher, the molten materials can enter
into direct contact with the reactor building's concrete basemat. As seen earlier, interaction
177.
178.
179.

Mitigation of Releases into the Environment in the Event of a Nuclear Accident.
For example, where the electrical systems for heating the ﬁlters for removing the steam saturating
the iodine trapping sites on certain materials such as active charcoal, are unavailable.
The containment depressurization systems of French nuclear power plant reactors are ﬁtted withsand ﬁlters.

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between corium and concrete is a subject of research aimed at establishing the precise
conditions of erosion and the time until basemat melt-through. There is a risk – which has
not been completely eliminated though it has been greatly reduced since EDF thickened thebasemats of the two Fessenheim reactors seen as critical – that the molten materials andsump water will go into the soil, the ground water, then the nearest river or the sea.
In 2009, IRSN began a study of the feasibility and effectiveness of a verticalgeotechnical barrier in the ground under the foundations of nuclear power plants,
extending downwards until it reaches a relatively impermeable geological formation, toact as a retention system to prevent ground water from ﬂowing to the site boundaries. A
geotechnical barrier like this would offer a grace period of several months following acore melt accident, and also a practical way of limiting pollution outside the site, throughthe pumping out, storage and treatment of the contaminated water inside this enclosure.
Creating a geotechnical barrier of this kind would not be easy but would generally bepossible with existing reactors. However, it would mean that there would have to be alayer of sufﬁciently watertight materials (clay, compact chalk, rock with few fractures,etc.) at an accessible depth in the ground in order to create a closed volume.
The investigations carried out by IRSN suggest that the feasibility of geotechnicalbarriers – using existing geotechnical barriers formed for the purpose of installation
construction – has been demonstrated for all French nuclear power plant sites except
Fessenheim, for which further studies are required, and hydrogeological studies have led
to the conclusion that these systems are likely to be effective at most sites, though the
Belleville and Cruas sites would still require speciﬁc study.
This matter is the subject of discussions with EDF in the context of ASN's formal
requests. It could ultimately lead to research and development. For its part, IRSN plans tolook into the design (from a safety and radiation protection point of view) of a system forpumping out the water collected in a geotechnical barrier during a core melt accident.

References[1] N. Rasmussen et al., Reactor Safety Study. An Assessment of Accident Risks in US
Commercial Nuclear Power Plants, WASH-1400 (NUREG-75/014), Washington DC,
US Nuclear Regulatory Commission, 1975.
[2] Special Issue on ASTEC, Nuclear Engineering and Design, Vol. 272, 119–298, 2014.
[3] Major Achievements after 4.5 years of SARNET, Progress in Nuclear Energy, Vol. 52,
1, 1–157, 2010.
[4] ERMSAR 2013 Conference of the SARNET Network, Annals of Nuclear Energy, Vol.
74, 1–223, 2014.
[5] See the OECD/NEA website containing the NEA/CSNI reports on core melt
accidents: http://www.oecd-nea.org/nsd/docs/.
[6] See the U.S.NRC website containing the NUREG reports on core melt accidents:
http://www.nrc.gov/reading-rm/doc-collections/nuregs/.
[7] Special Issue. Phebus FP Final Seminar, Annals of Nuclear Energy, Vol. 61, November
2013.

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Current state of research on pressurized water reactor safety[8] Ability of Current Advanced Codes to Predict In-Vessel Core Melt Progression and
Degraded Core Coolability – Benchmark Exercise on Three Mile Island-2 Plant – Finalreport, NEA/CSNI/R(2015)3.
[9] Benchmark Study of the Accident of Fukushima Dai-ichi Nuclear Power Plant (BASF
Project) – Phase 1 Summary Report, NEA/CSNI/R(2015)18.
[10] OECD Research Program on Fuel Coolant Interaction. Steam explosion resolution
for nuclear applications – SERENA, Final Report, NEA/CSNI/R(2007)11, 7 September
2007.
[11] OECD/SERENA Project Report. Summary and conclusions, NEA/CSNI/R(2014)15.
February 2015.
[12] State-Of-the-Art Report on containment thermalhydraulics and hydrogen distribution, NEA/CSNI/R(1999)16.
[13] A. Bentaib, H. J. Allelein et al., International Standard Problem ISP-47 on containment thermalhydraulics, Final Report, NEA/CSNI/R(2007)10.
[14] OECD/NEA THAI project ﬁnal report – Hydrogen and ﬁssion product issues relevant
for containment safety assessment under severe accident conditions, rapport NEA/CSNI/R(2010/3).
[15] State-Of-the-Art Report on Flame Acceleration and Deﬂagration-to-Detonation
Transition in Nuclear Safety, NEA/CSNI/R(2000)7.
[16] H. Cheikhravat, « Étude expérimentale de la combustion de l’hydrogène dans une
atmosphère inﬂammable en présence de gouttes d’eau ». Thesis from the Universityof Orléans, September 2009.
[17] A. Bentaib, N. Meynet, & A. Blayer, Overview on hydrogen risk R&D activities
methodology and open issues, Nuclear Engineering and Technology, Vol. 47, 1, 26–
32, 2015.
[18] L.E. Herrantz, T. Haste & T. Kärkelä , Recent advances in the source term area within
the SARNET European severe accident research network (SARNET2), Nuclear
Engineering and Design, Vol. 288, 56–74, 2015.

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Chapter 10
Research on the Behavior ofComponents Important to the Safety of
NPPs and More Particularly the Aging of
Such Components

This chapter provides an overview of the most signiﬁcant R&D efforts on metallic
components, civil engineering structures, and other components that play a role inensuring the safe operation of pressurized water reactors in France's NPPs and which, inparticular, play a conﬁnement role ("barrier").
The R&D efforts in which IRSN is particularly involved are related to:– the behavior of components during accident conditions (earthquake, core melt, etc.);– aging in a broad sense, i.e. the effects of various mechanisms of damage – orpathologies – that may adversely affect components (metallic structures, civilengineering structures, etc.) over time and subsequent to their use (under normal
operating conditions).
A gradual damaging of components can be generated by operating loads and theirsurrounding environment, such as pressure, temperature, thermal transients, vibrations,
irradiation, and the chemical composition of the surrounding medium. Operating
experience, and the results of periodic tests and inspections in particular, shows thatcrippling degradations can occur even though mechanisms of damage—at least thosethat are known—are taken into account in the design, sizing, manufacturing, andAnglais_Book

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operation of components. Cracks on vessel head adapters and tube bundles of steam
generators are just some examples of such degradations. In other cases, the kinetics ofdegradations were faster than expected (thermal fatigue of pipes).
In the case of components studied in terms of loading during accident conditions, thecompliance of "end-of-life" components with safety requirements must of course be
veriﬁed by factoring in the effects of aging during normal operation—or their possible
replacement by new ones.
Research on the behavior of metallic components subjected to loading during
accident conditions focuses on understanding the behavior of complex structures
such as overhead cranes. Regarding civil engineering structures, this research focuses
on developing laws and models of thermomechanical behavior for each component
element of these structures (concrete, rebar, etc.). These laws and models are thenintegrated into simulation codes such as Cast3M, developed by the CEA and which is usedby IRSN as parts of its research in support of assessments or before key deadlines—suchas the ten-yearly safety reviews of France's reactors and the possibility of operating thesereactors beyond their 40-year design lives ("operating life" project [DDF]).
Aging management is based on two key principles—proactiveness and monitoring. It
relates particularly to the components involved in the second and third conﬁnement"barriers"—the reactor-coolant pressure boundary and the containment itself. Indeed, it
is vital that these components retain their intended design characteristics (compliance)
throughout their service lives and up to and including their disassembly during facility
dismantling operations. It is worth reminding that a fundamental objective of metallic
components that play a containment barrier role is to maintain a sufﬁcient degree ofductility throughout their service lives. This requirement particularly relates to the PWR
vessels, which gradually weaken under the effects of irradiation (increase in the ductileto-brittle transition temperature).
The purpose of R&D on aging is to improve knowledge on the mechanisms of damage,or pathologies, that can adversely affect components (Figure 10.1). In France, thisresearch took on a new dimension after EDF announced its intention to continue
operating its NPPs well beyond their 40-year design lives although some of theircomponents are either irreplaceable or difﬁcult to replace (vessel, containment).
The R&D efforts discussed below—speciﬁcally those led by IRSN—thus relate to:– the metallic components of the reactor coolant system,– the concrete containments,– the polymers used as insulation around electrical cables, the inner linings of somecontainments, and seals.
Generally speaking, IRSN's research is intended to highlight phenomena not taken
into account by licensees and improve its understanding of phenomena that are not well
documented but nevertheless are important to safety. Therefore, and along with R&D
efforts led by other bodies (either alone or jointly), it contributes to the development ofstandard practices in terms of the design, fabrication, and in-service monitoring ofAnglais_Book

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Figure 10.1 Areas of aging in PWRs. Areas 1 and 2 are highly critical, for they contain non-replaceable
equipment.

metallic or civil engineering structures, such as those of the RCC-M180 which applies tothe metallic structures in France's NPPs.

10.1. R&D on metallic components
R&D efforts on metallic components involving the key players in France's nuclear
power industry primarily relate to the following phenomena:– irradiation-induced embrittlement of vessel steel. As stated above, this embrittlement alters the mechanical properties of the steel (and increases the ductileto-brittle transition temperature in particular) and adversely affects its behavior in
the case of thermal shocks. This embrittlement can be assessed using empiricalprediction formulas adjusted to data derived from analysis of test specimens fromthe irradiation monitoring program181 and supplemented by irradiation programs
180.

181.

Rules on the Design and Construction of Metallic Equipment (design and construction rules
derived and adapted, in the 1970s, from those issued by the ASME [American Society of
Mechanical Engineers]).
The irradiation monitoring program (PSI) involves testing, for each reactor vessels in France's ﬂeet,
samples representative of the vessel steel placed inside capsules along the circumference of thereactor core. These capsules also contain dosimeters to measure the neutron ﬂux density received
by the samples. Due to their locations, the capsules are exposed to a higher neutron ﬂux than thatreceived by the vessel walls. This makes it possible to foresee the behavior of the materials after
10, 20, 30, and 40 or more years of operation.

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conducted in research reactors. The European PERFORM182-60 (2009–2013)
project led by EDF and associating AREVA and CEA made it possible to develop
a ﬁrst generation of tools for simulating the microstructural effects of steel
irradiation. Research in this ﬁeld is continuing;– thermal embrittlement of cast austenitic-ferritic steels used to fabricate a numberof components (such as the elbows of the RCS in some French reactors currently
in operation). In the early 1980s, it was found that these cast items weresusceptible to thermal aging when the RCS was kept at its service temperaturefor extended periods. The resulting degradation of their mechanical properties(hardening and gradual embrittlement, loss of toughness) was attributed tounmixing of the Fe-Cr-based solid solution into Fe-rich α and Cr-rich α´ domains
via a spinodal mechanisms and precipitation in the ferrite of a nickel and siliconrich intermetallic phase. The potential risk considered was that of a change in the
failure mode of the components, i.e. from a ductile mode to a brittle one requiring
little energy, the suddenness of which was particularly worrisome for pressure
vessels. With CEA, IPSN studied and tested model materials to (i) understand thestructural changes that take place when they are maintained at high temperatures
for extended periods and assess the consequences of such changes, (ii) study thescale effects (tests on large and small specimens), and (iii) identify the inﬂuence ofthe ferrite content and the aging duration. These studies and tests made it
possible to identify that although the failure of aged cast items is macroscopicallybrittle, it is actually caused by a ductile failure mechanism that is less dangerous
because it requires an input of energy. Numerical simulation tools have beendeveloped to predict the aging of cast elbows on RCSs throughout their entire
service life and assess their behavior in the presence of fabrication defects (casting
defects, such as shrinkage cavities) or taken into consideration in safety studies(cracks). EDF is also conducting R&D on this issue to justify whether cast elbows
deemed sensitive should be left in place or replaced;– stress corrosion of stainless steels. Conditions that aggravate this damage
mechanism (water chemistry, choice of materials and manufacturing conditions)have been identiﬁed. Experimental studies have been conducted on this phenomenon as part of international projects, such as the OECD/NEA Halden reactor
project (1995–2008) and the EPRI CIR183 project (1995–2009). Both of theseprojects yielded signiﬁcant ﬁndings about the initiation and propagation of cracks;– irradiation-assisted stress corrosion of stainless steels (affecting the vesselinternals), which signiﬁcantly decreases their ductility and leads to the initiationof cracks. This phenomenon has been studied as part of international projects,
such as the European PERFECT project (2004–2008), which made use of a multiscale approach to predicting damage;

182.
183.

Prediction of the Effects of Radiation FOR Pressure Vessel and in-core Materials using multi-scale
Modelling – 60 years foreseen plant lifetime.
Cooperative Irradiation-assisted stress corrosion cracking Research.

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– irradiation-induced swelling of stainless steels, which can adversely affecttheir operability. This phenomenon is being studied by the GONDOLE program(2006–2016) led by CEA at the OSIRIS reactor in Saclay, France, as part of aninternational partnership (EDF, AREVA, SUEZ-GDF184, EPRI);– stress corrosion of nickel-based alloys used to fabricate components such as
steam generator tubes and bimetal bonds such as in the vessel head adapters andvessel bottom head penetrations. The effects associated with the chemical
conditions of the water in the primary and secondary systems studied as partof the CIRCE and CIRCE-2 programs implemented by EDF and EPRI. IRSN has
undertaken research in this ﬁeld (thesis) as well;– thermal fatigue, particularly following the 1998 incident at Civaux. This subject
will be discussed hereafter;– wear of steam generator tubes due to their supports. This phenomenon is themain mechanism of damage affecting tube bundles on the latest generation of
steam generators in France's NPPs. IRSN, working with CEA, has set up an
experimental program to study this wear and determine changes in the coefﬁcientof friction of these tubes with their surrounding conditions (chemistry, steam,etc.).
The implementation of probabilistic approaches to better take the uncertainties andvariabilities of the parameters involved into account, particularly in terms of assessing
vessel behavior, is also being considered in relation to the possibility of operating France's
NPPs beyond their 40-year design lives. EDF is conducting development efforts incollaboration with CEA on this subject. However, IRSN has reservations about thepossibility and relevance of lending such approaches to safety demonstrations.
IRSN, working with CEA, is also leading research on the seismic behavior of metallic
components such as overhead cranes and precast ﬂoors.
Research is also being conducted in the ﬁeld of non-destructive testing of metallic
components.
Both of these subjects will also be discussed further hereafter.
In 2008, EDF created the Materials Ageing Institute (MAI). Led by EDF, the institute is
co-ﬁnanced by NPP licensees including EPRI (Electric Power Research Institute, which
represents all nuclear reactor licensees in the United Sates) and KEPCO (Kansai ElectricPower Company, in Japan). The MAI brings together the skills of these organizations topredict aging in NPPs and increase the durability of the materials, components, andstructures used in them. Knowledge about this phenomenon will be shared with IRSN.
Some of the aforementioned research subjects are discussed more in detail hereafter.

184.

French gas utility.

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10.1.1.

Research on thermal fatigue

Thermal fatigue is a mechanism of damage that EDF, among others, was forced todeal with when, in May 1998, a 30 m3/h water leak occurred on an RHRS185 pipe inreactor 1 at the Civaux NPP (N4 series) while the reactor had been shut down formaintenance. This leak was caused by a crazing-like pattern of cracks—known as elephantskin fracture—emerging at the extrados of a pipe elbow (Figure 10.2).

Figure 10.2 The inside of the pipe elbow that caused the leak that occurred at the Civaux NPP in 1998.
Crazing-like fractures are visible on either side of the weld bead. © EDF (left) and IRSN–Source: EDF (right).

Thermal fatigue can be summed up as follows: materials expand or contract whensubjected to temperature variations. But stresses occur if they cannot expand or contract
so freely. Damage from thermal fatigue can result if these variations are too frequent186.
The crazing-like fracture area of the RHRS pipe (Figure 10.3) of Civaux reactor 1,
which was made of austenitic stainless steel, was subjected to signiﬁcant temperatureﬂuctuations downstream of a mixture of streams (one at 180 °C and the other at 20 °C)
at a pressure of 27 bar. The leak occurred after only 1500 hours of operation.
The phenomenon of thermal fatigue in the mixing areas had not been foreseen during
the design of the PWRs. The incident that occurred in 1998 could neither be foreseen nor
explained by conventional methods and criteria of mechanical fatigue analysis, such as
those encoded in the RCC-M and based on the assessment of fatigue "usage factor187." Itis worth noting that water leaks induced by thermal fatigue had occurred on the RCS
before the incident at Civaux. However, they were caused by an altogether different
185.
186.
187.

Residual Heat Removal System.
Known as high-cycle fatigue, as opposed to oligocyclic fatigue.
This usage factor corresponds to the ratio between the number of loads applied to a given
component and the maximum number of loads indicated by the mechanical fatigue curve of thecomponent's material.

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Figure 10.3 Crazing-like fracture on a pipe elbow of the RHRS at Civaux in 1998. © Antoine Dagan/
Spéciﬁque/IRSN–Source: IRSN (above) and IRSN–Source: EDF (below).

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Figure 10.3 (afterpart).

phenomenon: dead leg (Farley 2 in 1987, Tihange 1 in 1988, Dampierre 2 in 1992, andDampierre 1 in 1996).
Starting in 1999, ultrasonic examinations were conducted at every NPP in France.
These inspections revealed that the problem was generic (not speciﬁc to Civaux) and that
all of the examined pipes showed signs of cracking. This prompted EDF to replace themixing areas of the RHRS circuits in the entire ﬂeet and make improvements to reduce
the susceptibility to thermal fatigue (particularly by leveling out the welds).
This ﬁnding marked the starting point of more than 10 years of studies and research
to understand the cause of the phenomenon and ﬁnd adequate solutions. These efforts
were conducted by EDF and AREVA, and also by IRSN with CEA in particular. Academic
laboratories were associated.
EDF's work focused on identifying and assessing the risks of fatigue. Systematic
analysis of dismantled pipework, supplemented by mock-up tests (BVS, DUPLEX,
FATHER, etc.) and modeling, revealed key factors in the onset of cracks. The primarycause is temperature differences greater than 50 °C between hot ﬂuids and cold ﬂuids
and extended periods of repeated pipe loading with this high temperature difference.
IRSN, working with CEA and academic laboratories, implemented a two-pronged
research and study program:– thermal-hydraulic simulations of the ﬂow and mixing of water streams at
different temperatures in pipes to better understand the mechanisms at workand key parameters that affect mechanical loads applied to pipes. These simulations were carried out using the Cast3M code in particular;Anglais_Book

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– mock-up tests (FABIME, SPLASH and FAT3D) to understand the initiation andpropagation conditions of thermal fatigue cracks. The FAT3D testing system is
illustrated in Figure 10.4 below. Cold water is periodically injected along the innersurface of the tube (test specimen). This water describes a parabola on the innersurface of the tube such as to obtain a temperature gradient in three directions.
A few key teachings were derived from these studies and tests:– temperature ﬂuctuations at low frequencies (approx. 1 Hz) are responsible for the
rapid propagation of cracks through the thickness of components;– they are caused by turbulence and the speciﬁc ﬂow geometry, which are
inﬂuenced by the conﬁguration of the circuit upstream of the stream mixing
area (elbows, straight sections, etc.);– although thermal fatigue damage is accelerated by welded joints, it can also occur
along continuous weld beads (FAT3D test 6). This important teaching invalidated
the position previously defended by EDF, prompting it to expand inspections of its
facilities to areas not comprising welds;

Figure 10.4 The FAT3D test system (left) and the results obtained from tests on specimens (A) without aweld bead (top right) and with (B) a weld bead (bottom right). © IRSN.

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Current state of research on pressurized water reactor safety– the biaxial nature of thermo-mechanical loads and "environmental" effects may
account for the deﬁciency of conventional methods and criteria used to designand size materials subject to fatigue. These aspects are still being studied (seefurther on);– studies conducted on the nozzle area of the chemical and volume control system(CVCS) of the RCS — an area subjected to temperature differences much greater thanthose occurring in the elbows of the RHRS system (and reaching as high as 280 °C) —led researchers to consider that a priori the role of the ﬂow type is more important
than that of the temperature difference. Indeed, investigations have shown that the
CVCS nozzles in the ﬂeet suffered very little damage from thermal fatigue. Thisﬁnding has not been invalidated by in-service inspections conducted since.
Based on the results of thermal fatigue studies and research, EDF deﬁned a policy forthe operation, in-service monitoring and replacement of mixing areas within all of its
reactors. Starting in 2000, the areas of the RHRS circuits subjected to high temperature
differences were ultrasonically examined after every 450 hours of operation (following,in this regard, a recommendation of IRSN), and maximum periods of operation at hightemperature differences were deﬁned for all susceptible areas.
Nonetheless, research on fatigue is continuing, for much work remains ahead tounderstand the mechanisms involved and the conditions that cause damage to appear. In
2013, IRSN embarked on the EVA188 project with INSA in Lyon. The aim of the project isto speciﬁcally explore the aforementioned environment effects under the conditions
typically encountered in PWRs (pressure, temperature, water chemistry) on the fatigue
life of austenitic (and even austeno-ferritic) steels. These effects may explain the lack ofconservatism of the in-air fatigue curves of the ASME and transposed in the RCC-M. The
testbed will consist of a fatigue loading machine and an autoclave. IRSN has also been apartner of the European INCEFA189 project, which explores the same subject, since 2014.

10.1.2.

R&D on non-destructive testing

Inspections performed during the fabrication of components intended to be used in
the construction and operation (in-service inspections) of nuclear facilities are a vital partof defense in depth. For the French ﬂeet of NPPs, these practices are codiﬁed in a rulebook on in-service inspection for mechanical components (RSE-M190). Nonetheless, anumber of questions and difﬁculties arise, such as the adequacy of the type andperformance levels of inspection equipment used by manufacturers and licensees (ortheir contractors) to areas to be inspected that have complex geometries (see Figures10.5 and 10.6) or materials with speciﬁc metallurgical structures. This prompted IPSN, inthe early 1990s, to begin R&D efforts to develop sensor prototypes able to adjust tothese complex shapes. This research in the ﬁeld of non-destructive testing (or nondestructive examination [NDE]) deals with ultrasonic waves and eddy currents as well as

188.
189.
190.

Study of the Ageing of Steel.
INcreasing Safety in NPPs by Covering gaps in Environmental Fatigue Assessment.
Rules for the Monitoring of Mechanical Equipment in Operation.

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Figure 10.5 Examples of complex geometries: manual ultrasonic examination of a weld bead andultrasonic examination of an intricately shaped nozzle. © IRSN.

simulations of radiographic examinations. Its aim is to motivate licensees to look for anduse the best possible techniques for inspecting the components in their facilities.
Research in this ﬁeld, conducted primarily by IRSN and CEA, pertains to all nuclear
facilities that may be affected by aging. As mentioned above, it deals with thedevelopment of innovative prototypes of transducers or probes and examination
simulation models. The transducers or probes developed through this research are

Figure 10.6 Another example of complex geometry: the transition region of the roll-expanded portion of
the steam generator tubes and the191 U-bend areas of the tubes. © IRSN.

191.

The purpose of roll expansion of a tube into a plate is to create a radially expansion by mechanicalmean or by internal hydraulic overpressure and cause the tube to plastically deform. The residual
stresses cause contact pressure between the tube and the plate, resulting in a strong mechanicalbond between the tube and the plate. The tools used must not create sharp sloped marks on themetal, tear off any metal, or initiate cracks (Figure 10.6).

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intended to be used to check for damage, such as cracks, that may adversely affect safety.
These innovative devices, which are usually patented, relate especially to components
important to the safe operation of PWRs, some of which are considered failsafe192 in thesafety analysis but for which degradation mechanisms are nevertheless postulated.
Signiﬁcant examination difﬁculties still remain for some components, such a reactorcoolant pipes made of coarse-grained heterogeneous materials, in which the ultrasonicwaves propagate with difﬁculty, or components with complex surfaces, or the U-bend
portions of steam-generator tubes (Figure 10.6). The difﬁculties encountered with thesecomponents should be overcome by the higher performance obtained with theseadaptive sensors, which make it easier for ultrasonic waves and signals to penetrate
into materials. Signal processing techniques are also being considered for coarse-grainedmaterials.
Simulation models have been developed for the most commonly used inspection
techniques, i.e. as ultrasonic examination, radiographic examination, and eddy current
examination. These models have been integrated into the CIVA simulation platform
developed by CEA and are now accessible to all the relevant users. They are usedroutinely by IRSN and have become essential to assessing the performance of
examinations conducted by manufacturers, licensees or their contractors, in orderto substantiate its technical opinions to safety authorities. Most of these developments have been engineered in collaboration with CEA. More speciﬁc studies on theexamination of coarse-grained materials are being conducted with the U.S.NRC inassociation with PNNL, and studies on the examination of steam generator tubes arebeing conducted with ANL.

A) Development of transducers and probes► Development of a "conformable" ultrasonic transducer for parts withcomplex geometries
A conformable ultrasonic transducer—i.e. one that conforms to the shapes of parts—
was developed by IPSN in collaboration with CEA and patented in 2003.
The main objective of this transducer (Figure 10.7) is to enable or improve detection
and geometric characterization of defects in components with complex geometries(small elbows, small nozzles, non-uniform surfaces, etc.) on which acoustic coupling of
conventional rigid sensors can no longer be suitably accomplished. It was necessary to
improve these examinations, particularly for components most important to safety andfor which cracking damage is a possibility. The aforementioned multi-element ﬂexible
transducer adjusts to the shape of parts.

192.

Components for which design, fabrication, and in-service inspection provisions make it possible toconsider that their failure is highly unlikely. Their failure is not addressed by speciﬁc consequencemitigation provisions.

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Figure 10.7 Prototype of a conformable transducer – IRSN/CEA patent. © IRSN/CEA.

As seen in Figure 10.8, which shows a study conducted by Laborelec and CEA for theexamination of welds on large nozzles, this type of transducer is now industriallyapplicable.

Figure 10.8 Examination of a nozzle with a conformable transducer (2014 Cofrend Conference – Strategy for
robotized examination of nuclear components: from examination design to experimental results). © CEA.

► Development of eddy current sensors for examining steam
generator tubes
Examination of steam generator tubes is an important safety concern due to the riskof tube failure and release into the environment of radioactive substances (reactor
containment bypass). A steam generator contains anywhere between 3000 and5000 tubes.

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These tubes are examined using eddy current sensors that are generally rigid and well
suited to detecting defects in uniform surfaces, such as the straight sections of the tubes.
However, conventional rigid sensors can quickly reach their limitations with complex
shapes encountered in the transition regions of the roll-expanded portions of steam
generators or in small tube bends that may contain deformations (ovality, crushing, etc.).
IRSN and CEA jointly developed a ﬂexible sensor to overcome such difﬁculties andadequately conduct examinations in these areas potentially subject to damage.
Flexible sensor technology (Figures 10.9 and 10.10) is based on the use of winding
etched on a Kapton substrate193 and giant magnetoresistance (GMR) receivers that arehighly sensitive to magnetic ﬁelds. This technology improves the detection of circumferential defects in areas that are hard to examine due to the presence of longitudinalcracks. CEA and IRSN ﬁled a patent for this sensor in 2009.

Figure 10.9 Flexible eddy current sensor for examining SG tubes. © DR.

Figure 10.10 Prototype of a ﬂexible eddy current sensor for examining SG tube bends. © DR.

► Development of sensors for castings and heterogeneous structures
Prototype ultrasonic transducers and a prototype of a low-frequency eddy current
sensor have also been developed to detect and size planar defects in castings, which havea very coarse grain structure.
Surface-breaking defects along the inner walls of components (pipework, etc.) are
generally adequately detected by ultrasonic waves, for the echoes result from anultrasonic concentration of energy obtained by a "corner effect194." However, thesize of the planar defects is obtained by measuring the diffraction echoes obtained at
the tops of the defects. These echoes actually have very low amplitudes and are

193.

194.

Kapton is in insulating polyamide ﬁlm with mechanical properties (ﬂexibility, thinness, temperature resistance, ﬁlm etching, etc.) that allow for many uses in electrical and electronic
applications.
The corner effect is a phenomenon of overintensity of the ultrasonic echo due to multiple
reﬂections on the corner formed at the intersection of the inner surface of the material withplanar crack, whether surface-breaking or near-surface, directed generally perpendicular to thesurface.

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accompanied by a more or less loud noise due to the coarse-grained metallurgicalstructure that leads to a low signal-to-noise ratio. The progress is signiﬁcant sinceextension defects at depths of 10 to 15 mm are sized (characterized) in structures known
to be difﬁcult. That said, more progress remains to be made to further improve andconﬁrm performance through tests on a wider range of materials and by improving signal
processing and transducer technology.
IRSN and CEA are jointly developing low-frequency eddy current sensors to detect
defects located below the surface of thick components. Encouraging results have beenobtained with the low-frequency sensor, which also includes a GMR receiver. The
prototype detects planar defects that vertically extend between 5 and 15 mm below
the surface and produce signals having an amplitude related to the vertical extension of
the defect. This eddy current technology should be able to supplement and, in somecases, even replace ultrasonic examinations where measuring the vertical depth of planar
defects is not possible, particularly in components made of coarse-grained stainless steels
and other materials that strongly disrupt the propagation of ultrasonic waves. Eddy
current technology nevertheless has the drawback of requiring access to the inner surfaceof the piping or component to be examined.
Research to improve the detectability of defects for both ultrasonic examination andeddy current examination will be conducted through a partnership between IRSN and the
U.S.NRC.

B) Development of simulation models► Simulation of ultrasonic examinations of materials with a homogeneous
structure
In the late 1990s, IPSN and CEA began developing ultrasonic examination simulation
models to provide support tools for the expertises. These models were integrated in the
CIVA platform. The studies made it possible to address the case of homogeneousmaterials in which ultrasonic waves propagate without any particular difﬁculty for partsof simple or complex shape. These models can be applied to the ultrasonic contact
sensors (i.e. in direct contact with components) and contactless immersion sensors195 on
the in-service inspection equipment used to examine welds on the reactor vessels used in
France's NPPs. This research was subsequently continued to take into account defects ofmore complex shape that may be disoriented, in order to be able to adequately simulate
defects most similar to those actually encountered in the components. An example of astudy is shown in Figure 10.11.
Engineered for homogeneous materials, these models are operational and frequently
used in expertises conducted by IRSN for thenuclear safety authorities.

195.

Other models for simulating examinations using the time of ﬂight diffraction (TOFD) techniquehave been developed by CEA and EDF and are available in CIVA.

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Figure 10.11 Ultrasonic response of a multifaceted defect during contact examination using 45°
transverse waves. a) examination conﬁguration; b) "simulated" defect image; c) "experimental" defectimage. © DR.

► Simulation of eddy current examinations
CIVA's functionalities for eddy current examinations make it possible to simulate
examinations of steam generator tubes with an axial probe (Figure 10.12), rotating probes(STL196, STT197, +Point), and multi-element probes. The ﬁrst phase of development ofthese models was intentded for a simple conﬁguration with a tube and an isolated defect.
Now models make it possible to take into consideration a more realistic geometry oftubes and their environment (geometric transition region between the roll-expanded
portion of tubed in tubes plates and their straight portions, areas located under the
spacer plates, U-bend areas). It is also possible to model complex defects or a network ofdefects. Validation of CIVA is accomplished by comparing it with experimental data andby conducting test cases (benchmarks). The resulting conﬁdence in the use of simulation
tools makes it possible to use them to assess examination techniques.
Furthermore, CEA has developed CIVA functionalities that simulate the X-Probe, an
eddy current multi-element probe most commonly used outside France, particularly theUnited States. The functionalities will enable IRSN to better understand the performanceand limits of this probe, which now numbers among the examination methods usable
by EDF.

196.
197.

Long Rotating Probe.
Transverse Rotating Probe.

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Figure 10.12 Simulation of the response of the axial eddy current probe for an outer notch. © IRSN.

► Simulation of radiographic examinations
The development by IRSN, in collaboration with CEA, of tools for simulating
radiographic examinations began much later than research on ultrasonic and eddy
current examinations. As a ﬁrst step, the conﬁgurations of the radiographs mostcommonly used for the components of nuclear facilities—which can have various
dimensions, thicknesses, shapes, access conditions—were identiﬁed. This made it possible
to develop models that reproduced the most common operating conditions and took into
account iridium and cobalt radiation sources and silver-image ﬁlms in use in the facilities.
Attention was then turned to the case of welds of dissimilar materials (e.g. stainless steel
coatings deposited on ferritic steel, bimetallic welds used to bond stainless steel pipes toferritic steel components). This was accomplished, for example, by using computer-aided
design (CAD) software to describe the part to be radiographed and the various areas ofmaterial on the part and to which the properties of each material in question is assigned.
A phased approach was also applied to defects for which detection is to be simulated;these defects shifted from very simple shapes at the start of the studies to much more
complex shapes that can be described using 3D CAD software.
Starting in 2010, emphasis was especially placed on experimental validation ofsimulation models using mock-ups representative of components and with calibrated
defects. Such validation was conducted on a few of the radiograph conﬁgurations studied
previously. They made it possible to compare the results obtained through simulation
with those obtained during the experimental radiographs. These model validations areAnglais_Book

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being continued to the works to address the most common operating conditions198 ofradiographic examination and are the subject of publications or benchmarks.
Since 2013, research is being continued to take into account such aspects as otherradiation sources like selenium (less exposure to operators, but penetrate less throughstructures), X-ray for large thicknesses, and digital ﬁlms.
Here, too, the models developed and integrated into CIVA are used regularly at IRSN
for expertises and parameter studies.

► Simulation of ultrasonic examinations of coarse-grained materials
with a heterogeneous structure and improvement inthese examinations
After developing simulation models suited to ultrasonic examinations of components
fabricated from homogeneous materials or homogeneous welds—which generally do notprevent the propagation of ultrasonic waves and make it possible to assess with a highdegree of conﬁdence the performance of many ultrasonic examinations—IRSN considered that it was necessary to continue its studies in order to develop simulation models
more speciﬁcally suited to ultrasonic examination of heterogeneous materials.
The heterogeneous materials in the RCS of PWRs—found, for example, in castportions or welds used to bond stainless steel pipes to large RCS components made offerritic steel (bimetal bonds)—can signiﬁcantly hinder the propagation of ultrasonicwaves. This is due to the special metallurgical structure of the materials, which containcoarse grains (both castings and bimetal bonds) that vary in direction and size according
to the depth. This causes the speed of the waves to change and leads to various dispersion
or attenuation phenomena at the interfaces with the grains. These alterations impair thedetection or sizing of defects. In either case, simulated and real-life examination of these
speciﬁc areas of materials remains a major focus of R&D and there are many difﬁculties
to overcome. Simulating these examinations requires being able to describe the complex
structure of the material (base metal and weld) in the 2D or 3D model and to calculate
the acoustic ﬁeld transmitted through the materials while not forgetting that complex
structures will produce more noise than homogeneous structures. Lastly, it requires being
able to calculate the various interactions in the disrupted ﬁeld during penetration into thevarious structures with the encountered defects.
Despite the limited number (Figure 10.13) of mock-ups representative of this type ofmaterials, it was possible to develop models that predict the propagation of ultrasonicwaves. This was accomplished by macrographic examinations, conducted directly onthese mock-ups, and which made it possible describe simpliﬁed structures that yield thecontours and directions of grains. The results of these models were very promising, but,due to the limited number of cases studied, tests on other mock-ups had to be conductedto conﬁrm and improve the predictions.

198.

Panoramic radiograph with the source centered in the piping and the ﬁlms located outside,
radiographs with the source located outside in contact with the piping, etc.

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Figure 10.13 Coarse-grained metallurgical structures observed on IRSN mock-ups representative of: (a)a cast component fabricated by Manoir; (b) a narrow groove weld; (c) stainless steel bimetal bondfabricated by Cockerill; (d) "N4" vessel bimetal bond. © IRSN.

Another major difﬁculty in the simulation of ultrasonic examination of these complex
structures is that the description of the structure of the materials must be known in orderto make a simpliﬁed description of the structure and its grains. This can be achieved via
shape recognition (Figure 10.14) from macrographic examinations such as those shown in
Figure 10.13 for mock-ups used to develop the models.
Unfortunately, there are no representative samples of such components for most of
the heterogeneous structures and/or welds encountered in France's NPPs or in those ofother countries. Other methods of description than those using macrographic examination must therefore be investigated.
Studies are thus continuing. Firstly, to explore non-destructive methods that will
make it possible to obtain sufﬁcient descriptions of the structures of materials and their

Figure 10.14 Example of grain-shape recognition for cast steel fabricated by Manoir. © IRSN.

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grains and correctly predict the performance of examinations, and secondly to assess andeven develop codes that simulate solidiﬁcation from fabrication parameters for cast andwelded components for bimetal bonds. It would then be necessary, for example, to beable to simulate examinations on the components in the ﬂeet, to have access tofabrication or welding data that could be described by category. Such access could
be made possible by contacting the manufacturers or builders.
Models developed for bimetal bonds and cast coarse-grained materials require
further improvement, for the differences between simulated and experimental results
remain too signiﬁcant and depend on the reliability of the methods used for determining
the characteristics of the materials. IRSN thus considers that simulations for thesecomponents are not sufﬁciently operational. Research being conducted by IRSN and the
U.S.NRC, in association with CEA and PNNL, will make it possible for these bodies toshare their data- and mock-ups representative of components. This collaboration is
expected to continue until 2017 and will also include research on developing prototypesof transducers and eddy current probes better suited to detecting and sizing defects inthese difﬁcult-to-inspect materials.

10.1.3.

Studies and research on the seismic behaviorof overhead cranes

IRSN, in collaboration with CEA, began conducting studies and research on the dynamic
behavior, under seismic loads, of overhead cranes made up of welded elements199. The
behavior under seismic loading of such structures is highly complex, for phenomena such as
multiple impacts (in case of blocking of a wheel) or trolley slippage may occur.
IRSN investigated this issue during assessments of documents relating to seismic
reviews of various nuclear facilities, for the demonstrations by the licensees of theearthquake resistance of overhead cranes and absence of load drops involved methods
having a robustness that called for conﬁrmation (modal analysis method associated withthe use of "reduced spectra" or "reduced weights").
It ﬁrst appeared necessary to improve the understanding of the seismic behavior ofoverhead cranes by deriving the maximum possible beneﬁt from the experimental (shake
tables) and simulation (ﬁnite-element calculation codes) means available.
This research (2005–2013 [1]) consisted of a theoretical part of numerical simulations, and an experimental part with tests performed on the AZALEE shake table at CEA's
Saclay research center200 (Figure 10.15). The tests were conducted on a reduced-scale
199.
200.

I.e. assemblies with welded connections, as opposed to assemblies with bolted connections, whichhave gaps.
The Seismic Mechanic Studies Laboratory (EMSI) operated by CEA’s Saclay research center has
been conducting earthquake-resistance engineering research for more than 40 years. EMSI’s aim
is to use its numerical and experimental tools to further understand the behavior of structures,
equipment, and components under seismic excitation. The laboratory’s experimental equipment
is housed within the TAMARIS (Tables and Means of Analysing Seismic Risks) facility. AZALEE iscurrently the largest three-axis shake table in Europe.

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Figure 10.15 Overhead crane tested on the AZALEE table (left) and the three main modes ofdeformation obtained by numerical simulation. © CEA (left).

mock-up (1/5). Seismic loading was simulated by controlled jacks that moved the structure
in both horizontal directions of the table. Firstly, bibliographic researches made it possible
to identify the parameters that affect the behavior of overhead cranes. The mock-up's
geometry and scale were deﬁned using ﬁnite-element calculations and the objective was toﬁnd, in the mock-up, the main eigenmodes of the crane's actual structure201. More than 100
conﬁgurations were achieved by varying a number of inﬂuencing parameters, in particular
the trolley position, the type of contact at the wheels (blocked wheels or freely rotating
wheels), the "added weight" (load suspended from the trolley) and its position. A simpliﬁed
numerical model able to reproduce the dynamic behavior of this structure with reduced
calculation times was implemented. This model was validated for several conﬁgurations
representative of the operating conditions of the overhead cranes. It may be used by IRSN
for seismic risk studies, in particular to understand the behavior of overhead cranes in caseof earthquake beyond the design basis earthquake (DBE).
The lessons derived from this research may subsequently enable to IRSN to propose
tracks for ultimately establishing rules on analyzing the dynamic interactions between
overhead cranes and civil engineering structures.

10.2. R&D on civil engineering structures
R&D work on the civil engineering structures of PWRs relates to primarily to thecontainment. The containment is essential, for it is the last "barrier" between the reactor
and the natural environment.
We would like to remind the reader that three types of containment are used at the
PWRs in France's NPPs:– the ﬁrst type (900 MWe reactors) consists of a building having a single wall ofprestressed reinforced concrete lined on the inside with steel plate, referred to as astatic containment;
201.

The reference overhead crane is a generic crane used in nuclear facilities.

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Current state of research on pressurized water reactor safety– the second type (1300 and 1450 MWe reactors) are double-wall buildings
comprising an inner wall of prestressed concrete and an outer wall of reinforced
concrete. Dynamic containment is ensured by ventilation and ﬁltration of theannulus space between the two walls, thus supplementing the static containment
achieved by the inner wall. Composite (resin) liners were applied to the intrados of
the inner walls to improve their leaktightness;– the third type (EPR) is a combination of the previous two: leaktightness is achievedby a metallic liner on the intrados of the inner wall and supplemented by the
dynamic containment associated with double-wall containments.
The behavior of these containments under design-basis conditions and during coremelt accidents is described in the IRSN document on the state of knowledge on core melt
accidents in power reactors202.
Furthermore, an important aspect regarding pre-EPR reactors in France's NPPs is thatthe thermo-mechanical loads on the containments during core melt accidents is more
severe than the design-basis loads (around 5 bar absolute203), since that the "U5" device
was subsequently ﬁtted to decompress the containments should the need arise and ventﬁltered releases to the environment (sand-bed ﬁlter).
One of the aims of research on concrete civil engineering structures is to study the
behavior of containments during accident conditions and taking aging effects into
account. Its purpose is to allow the assessment of the mechanical behavior andleaktightness of the walls of the containments, which can be altered by the combined
mechanisms of shrinkage, creep, and "pathologies." Some of these "pathologies," such asalkali-aggregate reactions (AAR) and delayed ettringite formation (DEF), are caused bychemical reactions that occur a few decades following construction. Research is being
conducted to analyze these complex phenomena and control them in order to ensurethat the safety requirements for these civil engineering structures are met.
Before discussing the R&D efforts being conducted more speciﬁcally by IRSN on
nuclear civil engineering structures, three major research projects that EDF has led or iscurrently leading on such structures must be mentioned:– MAI (mentioned in Section 10.1);– the nationwide CEOS.fr project on the behavior and assessment of cracking andshrinkage defects in special structures [2008–2014] and for which the FrenchInstitute for Applied Research and Experimentation in Civil Engineering (IREX)
provided administrative and logistical support. The aims of this project—which
was supported by 41 partners, including IRSN—were to study cracking induced byvarious loads (monotonic or cyclical static loading) in walls and massive components made of reinforced concrete and the early-age behavior—extending from

202.
203.

See Chapters 6.2 and 6.3 of "Nuclear power reactor core melt accidents – State of knowledge" –Science and Technology Series – IRSN/EDP Sciences – 2013.
This value corresponds to the design pressure that encompasses LOCA, and it is used to conduct
containment integrity tests.

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10 days following casting—to the maturity under the various aforementioned
loads, with special attention given to cracks formed in these structures. Recommendations for controlling these cracking phenomena were compiled in a document published in 2015 [2];– the VERCORS project (2013–2021), which we will discuss further on in thisdocument.

10.2.1.

Development of constitutive equations for civilengineering structures

IRSN is leading research on the behavior of containments under seismic loads or andduring core melt accident conditions. It uses the Cast3M numerical simulation code,
which was developed by CEA with IRSN's contribution (for example, it provided CEA witha number of rheological models derived from IRSN-funded PhD work). An important
point that must be emphasized is that these simulations are checked and validated bycomparing them to the containment monitoring activities conducted continuously by
EDF and during testing.
Since the 1980s, IRSN has been working with CEA's specialist laboratories that
produce the developments necessary to establish constitutive equations for the component materials of the containments (concrete, steel), ranging from the linear behavior of
the structures to their non-linear behavior and taking account of concrete cracking andstructural damage. These equations are used in research on the behavior of thecontainments under complex loads. The objective is to be able to simulate the behaviorof the containments from their construction to a potential accident.
The development of these equations is based on tests conducted on specimens thathave been veriﬁed on structures (slabs, beams, gantries) and validated with the results oftests conducted on large-scale mock-ups (a particular example is the tests conducted on
the RCCV204 and PCCV205 mock-ups at Sandia National Laboratories in the UnitedStates)206.
These R&D efforts make it possible for IRSN to assess, using Cast3M and as part ofphysics research conducted in support of the development of level-2 probabilistic safetyassessments (PSA), the behavior of the containments during core melt accidents. Thesestudies began with the 900 MWe reactor containments in the 1990s. A multi-scale
approach developed by IRSN, and based on the computing resources available at thetime, was used to determine with a sufﬁcient degree of accuracy, the susceptible areas ofthese containments (Figure 10.16). These studies showed that the equipment hatch—andmore particularly the closure ﬂanges—was a weak spot in the event of a core melt
204.
205.
206.

Reinforced Concrete Containment Vessel.
Prestressed Concrete Containment Vessel.
Test conducted as part of a collaboration between Sandia National Laboratories (SNL), the U.S.
NRC, and the Nuclear Power Engineering Center of Japan (NUPEC).

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Figure 10.16 Modeling of the containments and the equipment hatch. © Georges Nahas/IRSN.

accident. This prompted EDF to schedule the closure system to be changed during thethird ten-yearly outages of these reactors207. The system now comprises bolts that arelarger in diameter and made of a different grade of steel.
These simulation tools thus offer the possibility of exploring the behavior ofcontainments subjected to other loads during accident conditions (collisions, earthquake,etc.) and even beyond-design-basis combinations of loads during accident conditions.

10.2.2.

R&D on the behavior of civil engineering structures
under seismic loads

As mentioned above, the development of constitutive equations for the various
components of the civil engineering structures and of rheological models makes it
possible, with the Cast3M simulation code, to study the non-linear dynamic behavior ofstructures under seismic loads up to and beyond the design-basis earthquake.
The experimental programs were conducted at the request of IPSN (then IRSN) on
simple and complex structures such as gantries and walls using the shake tables at CEA's
Saclay research center (dynamic loads). At the same time, other tests on identicalstructures and under alternate static and increasing loads up to failure were conducted at
CEBTP208. All these tests made it possible to acquire new knowledge about the dynamic
behavior of reinforced concrete structures during earthquakes.

207.
208.

See Section 4.4.2.2 of "Nuclear power reactor core melt accidents – State of knowledge" – Scienceand Technology Series – IRSN/EDP Sciences – 2013.
French Construction and Public Works Test Center.

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R&D efforts are being conducted in other areas, in particular:– soil-structure interaction,– evaluations of motions transferred from ﬂoors to equipment installed on different
ﬂoors,– the dynamic behavior of reinforced concrete structures.
Another focus of research conducted in collaboration with CEA pertains to seismic
isolation systems for the structures, in particular regarding their application to new
facilities (Jules Horowitz reactor (JHR), International Thermonuclear Experimental
Reactor (ITER), both at Cadarache center).
These areas are discussed in more detail below.

A) Soil-structure interaction
Knowledge of the effects of the soil-structure interaction (SSI) is key to assessing the
dynamic response, under seismic loads, of a building such as a PWR, which weighs around50,000 tonnes. This interaction can be studied using different methods. Two were
explored by IRSN—one developed at the École Centrale de Paris (integral equations
method), and one developed by researchers at CEA's Saclay research center (absorbing
boundary method209 to limit the size of the ﬁeld representing the soil). IRSN initiated acollaboration with the École Centrale de Paris to use its method with the MISS3D210
code. The use of the integral equations method in the ﬁnite-element calculations of thestructures was achieved by sequentially chaining the MISS3D and Cast3M codes. Bothmethods were validated during IRSN's participation, in collaboration with CEA, in theinternational KARISMA211 benchmark set up by the IAEA following the earthquake(magnitude of 6.8 on the Richter scale) that affected the Kashiwasaki Kariwa NPP
and its seven boiling water reactors on July 16, 2007. Preference was given to the method
developed by CEA following numerical simulations conducted on one of the Kashiwasaki
Kariwa BWRs. However, although the comparison of the calculation results obtained withthe on-site measurements conﬁrmed the ability of the model achieved with this method
to reproduce the horizontal displacements, differences due to the inﬂuence of detachment of the basemats during the earthquake were observed for the vertical displacements. Additional research is being conducted to take into account, in the model, theinﬂuence of the non-linearity introduced by detachment.

B) Assessment of displacements transferred to equipment
Advances in simulation capabilities also serve to improve the prediction qualityof displacements transferred by the buildings to equipment by taking into account the
209.

210.
211.

This method makes it possible to conduct seismic soil-structure interaction calculations via ﬁniteelement analysis by limiting the size of the ﬁeld representing the soil. The resolution is made in
the time ﬁeld for the entire soil-structure meshing and makes it possible to model thedetachment and slippage of the basemat (Cast3M code manual).
3D Modelling of Ground-Structure Interaction.
KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment.

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non-linear behavior of the structures (cracking effect of the concrete) in, for example,
assessments of existing margins beyond design-basis loads for the structures andequipment.
Seismic margin assessment is a priority that emerged particularly in the United Statesin the 1980s. The aim was to understand the possible contribution to the overall risk ofcore melt accident by beyond-design-basis earthquakes that could affect NPPs located in
the eastern212 part of the nation. This resulted in different approaches213 referredcollectively as Seismic Margins Assessment (SMA).
Furthermore, after the earthquake that affected the Kashiwasaki Kariwa NPP on July16, 2007, in situ observations showed that some equipment items had withstood loading
beyond their design basis. More recently, an earthquake affected the North Anna NPP in
Mineral, Virginia, in August 2011. The ground accelerations at the two reactors wereassessed to be 0.2 g and 0.3 g, higher than the design-basis earthquake, which was 0.12 g
and 0.18 g, respectively214. The cumulative absolute velocity (CAV) threshold, of 0.16 g.s,
was exceeded during the quake, causing the reactors to automatically shut down. The
generators started up following the resulting electric network failure. The reactors werereturned to operation in December 2011 after investigations (including inspections by the
U.S.NRC) showed that the earthquake had not caused signiﬁcant damage to the NPP's
safety-important equipment.
Needless to say, the issue of seismic margins returned to the forefront following the
Fukushima Daiichi nuclear accident in the case of complementary safety evaluation (ECS)
conducted in France, where the important concept of robustness—favoring design andconstruction solutions that are stablest under multiple loads with margins to coverunexplored ﬁelds—emerged.
It should be said that, in the case of Generation III reactors, France's "Technical
Guidelines for the Design and Construction of the Next Generation of PWR Reactors"(2004) stipulate that "the designer must also specify how it intends to prove the existenceof sufﬁcient design margins that are consistent with the general safety targets (...). Margin
assessments shall be conducted to demonstrate that no cliff-edge effect in terms ofradiological consequences could occur by assuming acceleration values greater than thosespeciﬁc to the site. The corresponding method shall take into account the actual behaviorof representative equipment and of possible simultaneous equipment failures".
The aim of current R&D efforts that IRSN is participating in is to develop simpliﬁed
numerical models able to simulate the non-linear behavior of structures and determine
the seismic forces that act on equipment (transferred spectra). A number of tests
212.
213.

214.

The high seismic area of the California coastline has been excluded indeed from the scope of thestudy.
See the SMA method described in U.S.NRC NUREG/CR-4334 (1985) as well as NUREG/CR-4482
and NUREG/CR-5076; the probabilistic PRA-based SMA method, which is also described in an U.S.
NRC document; and the EPRI SMA method described in document EPRI NP-6041-SL Revision 1(1991).
These reactors had been seismically reassessed for an earthquake with a PGA (Peak GroundAcceleration) of 0.3 g in the 1990s.

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conducted with the AZALEE shake table have thus been used to validate the ability ofthese models to reproduce the measurement results and observations.
Among these tests are those recently conducted as part of the ENISTAT215 program[3], which aimed to understand the robustness of structures built according to theEurocode 8 standard216 and ﬁtted with thermal insulation systems between their verticalwalls and ﬂoors (thermal break elements, which create discontinuity between the verticalwalls and the ﬂoors).
The half-scale, asymmetrical reinforced-concrete mock-up with a total weight ofaround 40 tonnes was designed for a zero-frequency acceleration at 0.3 g.
The mock-up was subjected to horizontal seismic loading ranging from 0.1 g to 0.8 g.
Although a wall failed at 0.8 g, cracks appeared at low loading. The test results made it
possible to improve the simpliﬁed models to better take into account the displacements
observed at the centers of the ﬂoors.
An important point emerged during the tests. When three-directional seismic loads(horizontal and vertical simultaneously) were applied, the mock-up's stability wasaffected by the displacements even at a low acceleration of 0.07 g. The tests were
therefore continued beyond 0.07 g without any vertical quake. This observation led to
IRSN to plan research on the adverse effects of the vertical component of earthquakes.
The simpliﬁed models will be validated in particular by making the most of the results of
new tests conducted as part of the SMART program (Figure 10.17). These new tests havebeen conducted on a 1/4-scale mock-up designed in accordance with ASN Guide 2/01 forthe structures in nuclear facilities.

C) Dynamic behavior of reinforced concrete structures
The Fukushima Daiichi accident brought back to the fore the issue of load combinations, in particular that of a core melt accident following an earthquake or of anaftershock following a core melt accident triggered by the main quake. Determining thetotal effects of both transient loads of different type and duration is complex, moreparticularly if the reinforced concrete structure enters the non-linear ﬁeld or evenbecomes damaged. IRSN is developing and implementing various methods to meet thisneed. In particular, it is implementing simpliﬁed models—such as the lumped-mass stickmodel (Figure 10.18)—that are able to reproduce the overall behavior of the structures.
The local behavior is subsequently determined for the critical times with a ﬁner model ofthe structures by applying the load sets derived from the simpliﬁed analyses. Other
innovative techniques are being developed with CEA and ENS Cachan217.

215.

216.
217.

Experimental and Numerical Investigation of Shear wall reinforced concrete buildings under
Torsional effects using Advanced Techniques, conducted as part of the European SERIES program(2009–2013), led by Middle East Technical University (METU), in Turkey.
"Design and dimensioning of structures for earthquake resistance".
A prestigious French higher National School.

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Figure 10.17 The SMART2011 instrumented mock-up (used in support of a joint research program between
CEA and EDF) is representative of a 1/4-scale "nuclear" building. Weighing around 47 tonnes (including
36 tonnes of additional weights), SMART2011 was instrumented with more than 200 measurement
channels (acceleration, displacement, deformation) and was subjected to a series of seven earthquakes ofincreasing magnitude. © P. Stroppa/CEA.

Figure 10.18 Simpliﬁed lumped-mass stick model of a containment to simulate its behavior underseismic loads: re-creation of the fundamental modes. @ IRSN.

D) Other Areas of R&D in Seismic Loading
IRSN is exploring other areas:– seismic isolation systems. A number of safety issues have been raised during
studies on the seismic isolation systems adopted for the JHR and ITER facilities(damping pads arranged underneath the buildings). Other mixed systems thatAnglais_Book

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combine passive isolation with active (or semi-active) control, or three-directional
isolation, which is being explored by the Japanese, could also prove useful forFrance's NPPs. In 2013, R&D work was begun with CEA to provide knowledge with
an application part of value for the expertise218;– the seismic behavior of nuclear structures without earthquake-resistant provisions, more particularly the ﬂoors made up of preslabs. Such ﬂoors are found inparticular in the fuel cycle facilities (laboratories and plants) built between 1960
and 1990. IRSN, working with CEA, has embarked on an experimental program on
this issue to study the ultimate behavior of these ﬂoors and identify their modesof collapse as a function of loading. This program assesses the relevance of checks
proposed by technical regulations based on loads obtained by an analysis
conducted by assuming the linear behavior. The full-scale mock-ups were thus
developed by making sure replicate, as closely as possible, the building practices—in terms of concrete characteristics and implementation methods—that were in
use when the facilities were built (1960–1990). These mock-ups were subjected tostatic tests by CEBTP and dynamic tests by CEA on the AZALEE shake table. The
experimental program ended in 2013, and the results have been analyzed [4, 5]. A
ﬁnite-element numerical model simulating the mechanical behavior of thisassembly was used to analyze the behavior of these complex structures. The
comparison of the test results with those of the simulations with ﬁnite-element
models, which presupposes the monolithism and mechanical continuity of thestructures, conﬁrmed the monolithism of the behavior this complex, for theseismic loads explored, despite the absence of seismic protection provisions.

10.2.3.

R&D on the behavior of civil engineering structures incases of collision

Improving methods for assessing the impact vulnerability of nuclear civil engineering
structures (following a load drop or due to a projectile) has been the focus of R&D since
the beginning of France's NPPs program. The risk taken into consideration was anaccidental crash by a Cessna business jet. The aircraft was assimilated with arigid projectile colliding with the wall of a nuclear building such as a containment.
This R&D work was extended to military aircraft in the late 1980s. Since the attacks of
September 11, 2001, this research was further extended to assess the risks in case of acommercial wide-body aircraft impact.
These efforts can currently be split among three areas of R&D:
– an experimental area, with medium-speed collision tests (commercial aircraft) on
reinforced concrete slabs as part of the IMPACT experimental program conducted
at VTT in Finland. This program, which began in 2005, is being carried out incollaboration with the U.S.NRC, HSE, STUK219, ENSI220 and GRS;
218.
219.
220.

A PhD thesis was started in 2014.
Finnish Radiation and Nuclear Safety Authority.
Swiss Federal Nuclear Safety Inspectorate (Eidgenössisches Nuklearsicherheitsinspektorat).

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Current state of research on pressurized water reactor safety– numerical simulations conducted with fast-dynamics codes (LS-DYNA and
RADIOSS). The international benchmark IRIS 2010221 was launched and led by
IRSN, under the leadership of the OECD. Its aim was to assess the ability of various
codes and teams to predict the behavior of structures impacted by hard and soft
projectiles. This benchmark brought together 28 teams from 11 countries.
Following the ﬁndings and recommendations from this ﬁrst exercise, two new
benchmarks have been scheduled with OECD's partners:

the ﬁrst took place in 2012 to calibrate the ﬁrst simulations and present simpliﬁed
models—by providing the partners with the test results and test conditions (actual
boundary conditions), as well as the characteristics of the materials used,
the other began in 2014 and is being conducted to study vibration propagation in a
structure. A mock-up is being built to form the experimental basis of this benchmark.

The knowledge obtained from this area of R&D has already enabled IRSN to draw
up an initial draft of recommendations for the fast-dynamics analyses conductedusing numerical simulations, documented in a report by the OECD;– characterization of impacted material, modeling, and validation. The aim of thisarea is to identify inﬂuencing parameters and improve the rheological model ofthe behavior of concrete subjected to impacts. This aspect was researched222 with
Joseph-Fourier University and CNRS and culminated, in 2013, in the conﬁrmationof the inﬂuence of the moisture content of concrete and the rate of deformationon the dynamic behavior of impacted structures.

10.2.4.

R&D on the behavior of containments during core melt
accidents

A) Assessment of air and steam leaks through an idealized crack
In 1989, IPSN commissioned the Mechanical and Thermal Engineering Department(DMT) at CEA's Saclay research center to conduct an R&D program on local cracking inorder to quantify leakage through a containment wall during emergency situations (air andsteam) compared to dry-air leaks quantiﬁed during in-service testing. Known as SIMIBE, thisprogram comprised tests conducted on two glass plates simulating local cracks. It resulted
in the development of a numerical model integrated in the Cast3M code and able tosimulate the two-phase behavior of the air-steam mixture in such cracks. SIMIBE was used
to interpret the MAEVA tests described below and, in 2000, allowed IPSN to consolidate its
position in relation to the air and air-steam transposition factor, i.e. use a conservativetransposition factor of 1. The transposition of glass to concrete required additional researchto take into account the permeability of cracked concrete and the communication between
local cracks. This additional research, based on the double-porosity theory whereby
exchanges occur between the two media (local cracks and the porous medium), wasimplemented in the ECOBA223 project described below.
221.
222.
223.

Improving Robustness assessment of structures Impacted by missileS.
Doctoral thesis.
Study of Reinforced Concrete Containment Structures.

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B) Assessment of air and steam leaks through a crack under conditions
representative of a containment
The assessment of air and steam leaks through a crack under conditions representative of a containment during emergency situations required using large-scale tests
representative of such conditions. It has been conducted as part of experimental
programs since the early 1980s, which include:– the RCCV tests conducted in 1984 at Sandia National Laboratories in the UnitedStates (tests on a complete containment—made of non-prestressed concrete—ofa Westinghouse 900 MWe reactor, at 1/6 scale, with a metallic liner;– the MAEVA tests conducted in France between 1994 and 2002;– the PCCV tests conducted at Sandia National Laboratories in 2000 on a mock-up ofa Japanese PWR containment, at 1/4 scale (prestressed concrete, with steel liner);– the VK2/2 tests conducted in 2001 at the University of Karlruhe in Germany (tests
conducted on a 1.2 meter-thick pre-cracked slab extended by jacks);– the tests in the ECOBA program (ANR project);– the nationwide CEOS.fr project mentioned above;– the VERCORS, also implemented by EDF.
These experimental programs were accompanied by benchmarks and workshops.
As a reminder, in single-wall containments, conﬁnement is ensured by the steel liner(case of the RCCV and PCCV mock-ups). The goal of research programs on double-wallcontainments (MAEVA, VK2/2 and ECOBA) is to assess the leakage of the inner wall ofprestressed reinforced concrete.
The key difference between these two types of mock-ups is the damage mechanism—a failure mode for the ﬁrst and leak mode for the other.
Some of the aforementioned tests are discussed more in detail below.

► Tests conducted on the MAEVA mock-up
In 1994, EDF decided to build a containment mock-up in order to study its mechanicalstrength and better assess the leakage of the double-wall containments of 1300 MWe
reactors. EDF's aim was to conduct an experimental study of the thermomechanicalbehavior of the inner prestressed concrete wall for design-basis and beyond design-basis
situations (core melt) by subjecting the mock-up to sequences of increased pressure andtemperature. These aims were as follows:– assess the air and steam leakage rates under accidental conditions compared tothose measured with dry air during in-service testing;– study the behavior of the composite liners placed on the intrados of the inner wall
for the in-service tests and the various accidental scenarios as well as validate
their conditions of industrial implementation.

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IPSN participated in the MAEVA tests, in particular by helping to deﬁne twosequences: one involving an increase in air pressure and steam (to simulate a coremelt accident) up to the design-basis pressure initially intended for the EPR containment
without steel liner (6.5 bar absolute), the other up to a greater pressure. IPSN also made anumber of measurements (helium tracing of leakage rates, quantiﬁcation of steam titers
in the water-steam mixture in the core-melt simulation sequence).
The MAEVA224 mock-up (steam and air containment mock-up) represented a straight
portion of the inner wall of the containment, at 1/3 scale for the diameter and 1/1 scale
for the wall thickness. The annulus space between the two walls was also represented,
but the outer concrete wall on the mock-up was replaced by a steel wall (Figure 10.19).
The concrete wall thus had a diameter of 16 m, a thickness of 1.2 m, and a height of 5 m.
The mock-up was built at the Civaux NPP using concrete of the same characteristics as
those of the concrete used for reactor No. 2 of Civaux's NPP (high-performance
concrete). The top slab was supported by four prestressed concrete columns arrangedat each quadrant of the surface area. The inner wall was divided into quadrants, two of
which were lined with a composite material similar to that used for repairing reactors inoperation.

Figure 10.19 Diagram of the MAEVA mock-up. © Georges Nahas/IRSN.

224.

Steam Air Exchange Model.

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The prestress had been calculated to obtain, as for the containments of the 1300 MWe
reactors in operation, a mean residual compression of 1 MPa for an internal pressure of
6.5 bar absolute.
Seven test sequences using air and an air-steam mixture were conducted with the
mock-up. For the sequence simulating a core melt accident with air and steam, thepressure was increased in three steps (2.6 bar absolute, 5.3 bar absolute, and 6.5 bar
absolute). For the air sequence conducted to study the behavior of the containments
beyond their design-basis pressure, the pressure was raised to 9.8 bar absolute.
During each test, measurements were taken to determine the leaks in the annulus
space—divided into four sealed quadrants (each quadrant being referred to as achamber—for dry air and an air-steam mixture—, various temperatures and pressures,and displacements of the inner wall of the mock-up. During the air and steam tests, theextrados of the mock-up's wall was heated to 60 °C in the chambers to help vaporize
the water exiting the cracks and reduce the temperature gradient in the thickness.
The cracks were identiﬁed, the composite liners were auscultated visually, and pull-off
strength tests of the liners were conducted. The pull-off strength test made it
possible to quantify the adhesion of the liners to the concrete wall after the test
sequences.
The analysis of results also made it possible to determine:– the transposition factor to be used between the leakage rate for dry air and theleakage rate an air-steam mixture (factor of 1). This factor is then used todetermine the leakage rates of the actual containments during accidental situations from the leakage rates measured during the tests;– changes in cracking and measured leaks for a number of accidental scenarios.
These last results on the changes in cracking and measured leaks were applied tovalidate the leakage quantiﬁcation method (nationwide CEOS.fr project and the ECOBA
project).
Furthermore, the measures taken during the air and steam sequences showed thatthe leakage rates collected in the chambers were quite high and greater than predicted by
EDF, which believed that the steam would condense in the cracks and remain in thereinforced concrete wall. This assumption was therefore invalidated.
These observations accounted for much in considerations on the robustness of the
EPR. Realizing that composite liners would not ensure prolonged integrity during coremelt conditions—a design basis of the EPR—EDF decided to install a steel liner on theintrados of the containment of the Flamanville 3 reactor.

► ECOBA Project
The ECOBA research project to study the conﬁnement properties of reinforced
concrete structures (2010–2014) was conducted by the French National Research Agency(ANR) as part of its "white" program (interdisciplinary projects in all ﬁelds of research).
ECOBA brought together IRSN and three academic laboratories with complementaryAnglais_Book

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Current state of research on pressurized water reactor safetyskills: GeM225, the ENS Cachan, and the University of Pau and Pays de l’Adour (UPPA).
Using a full-scale mock-up representative of an inner containment wall of a 1300 MWe (P4series) reactor, its aim was to study the various mechanisms of concrete cracking in order toestablish relations between damage, cracking, and permeability. It was based on tests
conducted at the GeM laboratory of the École Centrale de Nantes on two mock-ups
representative of a straight portion, referred to as the "effective area" (1.50 m wide, 1.20 m
high and 0.90 m thick) of an inner wall of the containment. Each mock-up, which measured
3.90 m wide and 2.40 m high (overall dimensions) and weighed 20 tonnes, was built at theexperimental laboratory of the École Centrale de Nantes. Both were instrumented withvibrating-wire strain gauges identical to those used in the actual containments andsupplemented by various measurement devices to validate the results and obtain arobustness of the measurements. The mock-ups were subjected to direct tensile stress—four jacks driven load and displacement—to create cracks representative of those that
could appear in a containment during accidental loading. A metallic box for injecting either
air or air and steam was ﬁtted upstream of the areas of cracking. The leakage rates throughthe cracks were measured using a gas tracer dilution technique developed by IRSN andwhich made it possible to identify the surface openings of the cracks.
The ﬁrst round of tests made it possible to observe the development of cracks in theeffective area—which were similar to those seen on actual structures subjected to these typesof loads—and quantify the leakage rates in the sequences with air injection. The ﬁrst set ofresults conﬁrmed the project's relevance to progress in characterizing the parameters that
affect the integrity of NPP containments and to consolidate the assessments conducted on
the conﬁnement capability of the containments in the event of core melt accidents.

► VERCORS ProjectIn 2013, EDF launched a far-reaching project of studies and R&D on reactor
containments. Known as VERCORS (for Vériﬁcation réaliste du conﬁnement des réacteurs), its aim is to acquire sufﬁcient knowledge in order to demonstrate that NPPs can besafely operated for a period of 60 years. VERCORS is expected to continue until 2021. Its
goals are as follows:– experimentally demonstrate the strength of the containments in core melt
situations (under simultaneous and maintained pressure and temperature loads);– predict by experience changes in containment leaktightness;– improve knowledge about leakage and the models used to predict their changes;– ﬁnd new tools of leak detection and quantiﬁcation.
A number of tests will be conducted on a large-scale and thoroughly instrumented
containment mock-up. In light of the feedback obtained from tests conducted on largescale mock-ups, such as MAEVA or in Sandia National Laboratories, EDF decided to build
225.

French Civil and Mechanical Engineering Research Institute. A joint research unit brings together
the French National Center for Scientiﬁc Research (CNRS), the École Centrale de Nantes, and the
University of Nantes.

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a mock-up of a double-wall containment (P'4 series 1300 MWe reactors) at the same
scale for the large dimensions and thickness (1/3). This mock-up will be built on hard soil
and feature the same discontinuities (such as the ledge used for the crane runway),
penetrations (in particular the equipment hatch), prestressing cables and rebars. The
1/3 scale will make it possible to accelerate some physical phenomena inherent in
concrete, in particular:– shrinkage/curing of a factor of 9,– creep of a factor of 3,
this gives an equivalent aging factor of 7 on average.
At this scale, around nine years will sufﬁce to simulate 60 years for the actual
containments.
The ten-yearly tests will thus be conducted on the mock-up once every 14 months.
The core melt accident will be simulated after the ﬁrst nine years.
Because the delayed strains (shrinkage and creep) in the concrete that will be used aremoderate, the extrapolation of the results will have to be examined closely in the case of
the susceptible containments, for which delayed strains are greater.

C) Assessment of non-localized leaks in a concrete wall
The results of the experimental programs on the large-scale mock-ups together withthe observations and the measurements of the leakage rates during the ten-yearly testsof the containments reveal the existence of non-local leaks—referred to as non-localized
leaks—that contribute signiﬁcantly to total leaks.
Each PWR containment in the France's NPPs has an inner surface measuring around
10,000 m2. As a result, non-localized leaks through these containments as a function of
the applied load are likely to become signiﬁcant. Research, which ended in 2011, wasconducted with the École Centrale de Nantes on the gas (nitrogen) and steam
permeability of concrete under compressive mechanical loads. The experimental setup
made it possible to measure the air-steam mixture permeability of 11 × 22 cm hollow
concrete specimens. The concrete formulation was representative of that of the concreteof the containments themselves. The results revealed a difference in ﬂow between steam
and nitrogen and made it possible to identify a difference in the ﬂow times of the gas andsteam. Nitrogen ﬂows through the accessible porous network in a matter of minutes,
whereas steam takes much longer (around 10 to 25 hours depending on the injection
pressure).
This observation partially explains the time difference between the air and air-steam
leaks observed during the MAEVA tests.

D) Study on the early-age behavior of concrete
Early-age behavior of concrete is an important topic. During this complex phase,
which spans from the setting of the concrete to around one month thereafter,
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weak areas in the concrete's structure. Once concrete shrinks (Le Châtelier effect), thecombined changes in the temperature gradients throughout the structure and themechanical characteristics of the concrete (Young's modulus, tensile strength limit,
compressive strength limit, etc.) cause the concrete to crack. The extent of these cracks
depends on parameters such as the type of formwork, the outdoor temperature, thegeometry of the structure, and the amount of rebars.
These cracks will affect both the durability and mechanical behavior of the structurein the event of accidental loading with propagation of these cracks. In addition, thephysical and chemical changes may, under certain conditions, lead to subsequent
pathologies such as delayed ettringite formation (DEF – see below).
IRSN began researching this aspect in 2007 in order to study the behavior of civilengineering structures from their construction (concrete pouring) onward. Concrete
hydration reactions, with the resulting temperature gradients, water exchanges and theincompatibility of stresses of the constituents as well as the restraining caused byresumption of concreting, which causes concrete to crack, were identiﬁed as areas ofresearch to be developed.
The ﬁrst area looked at was the effect of early-age cracking of concrete on its air andgas permeability. This effect was analyzed by taking into account the rate of cooling, therebars, and resumption of concreting. This research was begun in 2007 (PhD thesis) with
the ENS Cachan. The experimental part of this thesis comprised tests on a 10 × 10 cm
concrete ring representative of that in the containments and heated to a temperature
representative of that measured at the core of the early-age concrete. It yielded
important data on the inﬂuence of construction methods on the creation of the ﬁrst
structural cracks. IRSN used the results obtained to assess the method used by EDF toresume concreting of the Flamanville EPR containment.
This research was followed by another thesis (2011–2014) on the modeling of theaforementioned phenomena supplemented by shrinkage and creep, which also occur
during the early-age phase. This research is currently being applied in the numericalsimulation benchmark of the VERCORS mock-up.
In addition, the quantiﬁcation of early-age exothermic phenomena also makes it
possible to assess the temperatures reached during pouring of fresh concrete and the riskof onset of pathologies, in particular DEF.

E) Development of mesoscopic simulation models
Analysis of the results of the observations and measurements obtained from theexperimental programs and ten-yearly tests of the containments show the importance ofcharacterizing the geometry of local cracks and porosity (corresponding to non-localized
leaks) as a prerequisite to realistic assessments of leakage rates. This characterization
requires using mesoscopic models—i.e. models at the scale of cracking—to be able tosimulate crack openings of less than 1 mm in concrete. High-performance simulation
codes are required to model at the mesoscopic scale. Research conducted by IRSN,
working with the University of Pau and Pays de l’Adour, have made it possible to identify
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creation of cracks in concrete during accident conditions, communication between
microcracks and local cracks, and residual opening of cracks after unloading. These
data may be applied to the macroscopic models used by IRSN for its studies in support ofthe level-2 PSAs on the behavior of double-wall containments of France's NPPs, aparticularly important point for the safety reviews associated with the third ten-yearly
outages of 1300 MWe reactors.

10.2.5.

R&D on containment aging

Aging is the process of changes over time to reinforced concrete structures such asreactor containments. Aging is caused by two phenomena: a natural phenomenon driven
by delayed strains in concrete (shrinkage and creep) and a phenomenon caused by theonset of pathologies.

A) Delayed strains in concrete
In the case of containments, this aging process decreases structural strains, causing
the prestressing cables to lose tension and thus decreasing the compression in thecontainment walls. Because this compression makes it possible ensure the necessary levelof conﬁnement required for these walls, its changes must be controlled throughout thelife of the structure.
R&D on this topic are thus conducted to assess the delayed behavior of the reactors
containments and its consequences on the ability to conﬁne radioactive material in theevent of an accident. This research comprises theoretical parts and experimental parts,
with numerical simulations. Tests are performed on concrete specimens representative of
the concrete of the containments. The main purpose of this work is to obtain thescientiﬁc knowledge and skills needed to make decisions on the risks caused by changes
to delayed strains in concrete. Research is primarily conducted on the following topics:– Tensile creep of concrete and its effects on containment leaktightness:
research was started in 2005 with IFSTTAR and the ENS Cachan. Tests were
performed on cylindrical hollow specimens to study the effects of creep on
concrete leaktightness. The tensile-loading time was limited to 36 hours (inservice testing time of the containments). A numerical model was developed tosimulate the mechanical behavior of the containments. The data acquired during
this research made it possible to better understand the changes in cracks observed
during the ten-yearly tests and the progression of cracking during consecutive
pressure-hold steps.
– Multiaxial creep of concrete and its effect on leaktightness: the containments
are biaxially prestressed structures. To take into account the effects of delayedstrain (concrete creep and shrinkage), these structures were designed in accordance with the RCC-G226 and using formulae in French standards and codes.
However, feedback has shown that these formulae underestimate the effect ofthese strains for multiaxial loads, a fact borne out by numerical simulations
226.

Rules on the Design and Construction of Civil Engineering Structures.

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conducted by IRSN. The reason is that these formulae were established fromobservations and tests on primarily uniaxially loaded structures. Furthermore,monitoring activities conducted by EDF on the containments of its NPPs are notreadily usable in the calculation models because the measured strains are due to a
combination of phenomena (basic creep, drying creep, drying shrinkage, autogenous shrinkage). Research to quantify these phenomena was begun in October
2011 with the ENS Cachan. This quantiﬁcation will make it possible to better
simulate the delayed behavior of the containments, assess the solutions proposedby EDF to reduce the kinetic of these strains (such as wetting of the containment
facings), and predict containment behavior to extend the operation life ofreactors. Once data has been acquired from the other research, the ﬁnal
application will be to simulate a containment from its construction lift by liftwith the prestressing phase, delayed behavior, ten-yearly tests, and end ofoperation. This model will make it possible to simulate the changes over timein the conﬁnement capability of the containment, a decisive factor in decisions onwhether to extend the service life of reactors.

B) Pathologies of reinforced concrete structures
In order to take into account all the phenomena that may alter the safety functions of
nuclear facilities, and of containments in particular, IRSN deemed it essential to initiate
research programs to investigate concrete and rebar pathologies. One reason inparticular for this research is that EDF is seeking to extend the service life of France's
NPPs to 60 years.
The development of pathologies in reinforced concrete structures can lead to damage
that could affect the mechanical properties of the concrete and even undermine theconﬁnement capability required from some structures. Swelling reactions—such as alkaliaggregate reaction (AAR) and delayed ettringite formation (DEF) in particular—are two
examples of potentially harmful pathologies.
EDF conducted research on AAR (a chemical reaction between some types ofaggregate and the cement matrix) and its change over time into an expansive gel
and pop-outs around reactive aggregate. EDF has classiﬁed all of the containments in its
ﬂeet into ﬁve categories (from 0 to 4 and ranging from zero risk to very high risk)
according to the characteristics of the concrete used and the environment of thestructures.
DEF—which is related to the type of concrete, its environment, and its coretemperature during concreting—is identiﬁed as a risk for nuclear structures. DEF canbe triggered by temperatures above 65 °C during construction of structures as well as bycertain moisture conditions. In 2009, IRSN alerted EDF (with ASN) about the importanceof such a risk of DEF. This alert was based on feedback from research programs conductedby reconstituted concrete and led by IRSN in collaboration with IFSTTAR (Figure 10.20).
The temperature of concrete play an even greater role in the development of DEF, inparticular in the case of structures which may, during their operation, be taken totemperatures of around 80 °C followed by a cooling cycle. Recent research conducted by

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Figure 10.20 S-curve of the change in DEF proposed by X. Bruneteaud in 2005 [7]. © Georges Goué/
IRSN–Source: Xavier Bruneteaud.

the scientiﬁc community and veriﬁed by IRSN (EDF-funded tests performed on specimens at IFSTTAR [6]) have conﬁrmed that delayed DEF can occur under some conditions
even though the concrete has not been subjected to temperatures above 65 °C when it
was made. Such a conﬁguration could occur in the containments at the reactor vessel andthe steam-tube penetrations, some fuel-assembly storage facilities, or after a controlled
ﬁre. IRSN is assessing the opportunity of conducting a research program on this subject.
Another parameter, related to the type of cement, affecting the development of DEF
is the use of limestone ﬁller227 as additives. Research started in 2009 (PhD thesis) incollaboration with IFSTTAR and the ENS Cachan made it possible to study the effects ofthese additives as well as other parameters that inﬂuence the development of thispathology, such as the moisture content of the outdoor environment, concrete permeability, and their effects on reaction kinetics.
EDF has classiﬁed all of the containments in its ﬂeet according to the characteristics of theconcrete used and the environment of the structures. The ranking of these ﬁve categories,
which range from zero DEF risk to very high DEF risk, is similar to that used for the AAR.
An analysis of both maps of the potential risk of containment pathology, derived fromthe classiﬁcations, reveals cases where this risk is high for both pathologies. These casesmean that R&D work must be conducted until feedback or experimental results areavailable. In 2014, IRSN—working with IFSTTAR and the ENS Cachan—began research(PhD thesis) on the accelerated aging of concrete affected by both pathologies together
and separately.
In the current state of knowledge, there are no curative solutions for these swelling
pathologies. The only recommendations are to monitor structures and, in the case ofstructures with high and very high risks (see, for example [8] and [9]), limit the ingress ofwater from external sources.

227.

Dust from the cutting of limestone that is added to concrete and other building materials.

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Current state of research on pressurized water reactor safetyC) Predicting the effects of aging: the ODOBA project
In light of the importance pathologies have on the longevity of reinforced concrete
structures and the plans to extend the service life of France's NPPs, IRSN decided the time
was right to create an observatory on the longevity of nuclear civil engineering structures.
This observatory will be shared with the scientiﬁc community and will yield data on the
aging of these structures and the defects that could affect them.
IRSN therefore created the Observatory of the Durability of Reinforced Concrete
Structures (ODOBA228) in 2014. ODOBA's purpose is to study pathologies found innuclear civil engineering structures, such as rebar corrosion (due to chloride ions orcarbonation), concrete swelling (DEF, AAR) and leaching, and their effects on the safety
requirements for these structures.
Knowledge will be acquired concomitantly with the assessment of plans to extend
the service life of France's NPPs beyond 40 years. The experimental part comprises theconstruction, at IRSN's Cadarache research center, of concrete structures representativeof the concrete of the containments at France's NPPs. These structures—60 blocks
measuring 1 m thick and several meters in height and width—will be subjected to either
an accelerated aging process or the natural aging process in order to determine theequivalent durations of accelerated aging. A ﬁrst difﬁcult task was ﬁnding the quarries
and cement used to build a few containments classiﬁed by EDF as having potentially highor very high risks of swelling reactions. This work was conducted with the assistance of
IFSTTAR and consisted in investigating the archives of the French Central Laboratory for
Bridges and Roads (LCPC). This investigation made it possible to locate nine quarries andtheir cement for 13 susceptible NPPs. The ODOBA project will use equivalent genericconcrete for the other susceptible NPPs.
The scientiﬁc part of ODOBA is led by a scientiﬁc committee comprising IRSN, the
ENS Cachan, IFSTTAR, the Construction Materials and Durability Laboratory (LMDC) in
Toulouse and the Mechanics and Acoustics Laboratory (LMA) in Aix-en-Provence. The
purpose of this committee is to monitor the project's scientiﬁc progress and verify thescientiﬁc relevance of the choices made as part of it.
ODOBA is divided into a number of stages that will make it possible to obtain
intermediate results aligned with IRSN's assessment needs, such as analysis of the tenyear safety reviews conducted on France's NPPs.
In 2015, IRSN signed a cooperation agreement with the U.S.NRC to share the resultsof experiments on the pathologies of the concrete used in nuclear reactor containments.
Other organizations are interested in ODOBA and discussions are under way to deﬁne theconditions of their involvement.
The ﬁrst concrete blocks have been cast in 2016.

228.

Monitoring Center for the Durability of Reinforced Concrete Structures.

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D) Non-destructive testing of structures
Non-destructive methods of testing for defects in structures are useful tools forassessing the health of susceptible structures such as reactor containments andstructures were sampling is very limited and controlled. IPSN then IRSN conducted
R&D efforts with CEA's Saclay research center to study the possibilities offered byultrasonic waves. These inspection methods have proven their effectiveness for metallic
structures and make it possible to detect defects. A forward-looking development project
conducted with CEA's Saclay research center consisted in adapting these methods toconcrete by studying the various types of probes and the signal frequency to be emitted,
and by adapting the parameters of the CIVA simulation platform and its simulation codefor predicting noise and attenuation phenomena. Currently, the operational monitoring
capability of this method is limited to a thickness of 40 cm and a very low amount ofrebars. The results obtained by this feasibility study have shown that this method cannot
be used on the containments.
In the frame of the ODOBA project, the need for non-destructive testing methods
meant that other techniques and methods had to be found. These include the use of ﬁber
optics and novel non-destructive testing techniques studied by the LMA, the LMDC andIFSTTAR. These new techniques are being developed to detect the onset of defects andpathologies in the concrete of structures. They will be validated by collecting coresamples from the blocks.

10.3. Research on polymers
The main purpose of research on polymer aging (polymers are used for electricalwires, seals, and coatings) is to study the effects of radiation and temperature on the
degradation of polymer properties.
For licensees, electrical wiring, which is located inside the reactor building andprovides a safety function, is not easily replaceable. It must therefore perform its function
throughout the life of a NPP, including in the event of an accident at the end of the
facility's service life.
The insulation and jackets of wiring are made of polymers. Temperatures andirradiation can, depending on their intensity and synergies, break up the polymer chainsor lead to crosslinking (creation of bonds between the chains), oxidation, or even a loss ofplasticizers due to hydrochloric acid migration. Understanding these mechanisms isnecessary in order to assess the relevance of the conditions of accelerated aging usedduring qualiﬁcation or during accelerated aging simulations. This is because the electricalproperties directly related to the functionality of the wires and cables generally do notvary much before signiﬁcant degradation of the polymers occurs. Indicators of aging
must therefore be looked for among the mechanical and physical and chemical propertiesof polymers.
The service life of wires and cables can then be predicted using experimental data andextrapolation models.

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The purpose of a ﬁrst study initiated by IRSN was to assess the aging resistance of
EVA polymer cables used in the N4 series reactors. Sections of cable were aged thermallythen by irradiation at dose rates between 3 and 1000 Gy/h. They were then subjected tothe accident conditions that would result from rupture of a RCS pipe. The ﬁrst results
showed that EVA is highly stable, virtually immune to temperatures and the dose rate
provided that the antioxidants in the material have not been completely consumed.
Furthermore, in order to determine the representativeness of the conditions ofaccelerated aging, new PVC and EPR/Hypalon229 cables were laboratory aged using an
accelerated method. To deﬁne these tests, IRSN assessed the acceleration of aging so as
not to alter the aging mechanism foreseeable under actual conditions; this assessment
was conducted using a similitude of the experimentally determined activation energies ofthe mechanisms of degradation of the materials. The mechanical properties of these aged
cables were compared to those of identical cables collected from the Cruas site after
seven years of operation. This comparison revealed that the properties of the cables
collected from Cruas and the laboratory-aged cables were similar.
IRSN thus demonstrated that it is possible, after a study of the degradation
phenomena of materials, to establish accelerated aging conditions that are representative of actual aging.
Since 2014, IRSN has been conducting a research program to study the impact of an
accident on aged silicone seals of the equipment hatch. The IRMA irradiator and the
EPICUR facility in particular are being used for this research. Experiments conducted on
mock-ups to assess the temperature (40 °C to 150 °C) and irradiation (25 kGy to 75 kGy)
effects show that the properties of the silicone are greatly affected by these exposure
conditions. All the tests under LOCA conditions, after ageing, have been conducted todate; the results will be synthetized in 2017.
Accelerated aging tests of 3–4 mm coatings made of epoxy-matrix composites wereconducted by CIS-bio international (CEA–Labra230) for IRSN between 2003 and 2008.
Two epoxy resins (manufacturers: Max Perles and Chrysor) similar to those used by EDF
to line the intrados of the containments of its 1300 MWe and 1450 MWe reactors wereused for these tests. A variety of dose rates (1 to 100 Gy/h), doses (40 and 320 kGy) andtemperatures (40 °C to 70 °C) were applied. The results showed that the epoxy resins
were highly susceptible to oxidation. The adhesion of all of the coatings dropped
signiﬁcantly after the four postulated LOCA tests. A critical analysis of this program
in relation to the tests presented by EDF in 2014 is under way. A new R&D program will be
proposed in 2017 to deepen understanding on the degradation mechanisms identiﬁed
and their impact on integrity.
New research on polymer aging (cable insulation, paint, resins, and anti-seismic
damping pads) is being scheduled.

229.
230.

Insulation made of cross-linked ethylene/propylene with a Hypalon jacket.
Laboratory for Applied Radiation.

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References[1] C. Feau, et al., Experimental and numerical investigation of the earthquake responseof crane bridges, Engineering Structure, 84, 89–101, 2015.
[2] CEOS.fr: « Comportement et évaluation des ouvrages spéciaux/Fissuration–Retrait/
Recommandations pour la maîtrise des phénomènes de ﬁssuration », Presse des Ponts,
2015.
[3] B. Richard, N. Ile, A. Frau, A. Ma, O. Loiseau, O. Gitry, F. Ragueneau, Experimental andnumerical study of a half-scaled reinforced concrete building equipped with thermal
break components subjected to seismic loading up to severer damage state,Engineering Structure, 92, 29–45, 2015.
[4] Seismic behavior of composite concrete ﬂoor without earthquake design provisions –Part 1: Experimental program, SMIRT 23, UK, 10/14/2015.
[5] Seismic behavior of composite concrete ﬂoor without earthquake design provisions –Part 2: Numerical analysis, SMIRT 23, UK, 10/14/2015.
[6] B. Kchakech, « Étude de l’inﬂuence de l’échauffement subi par un béton sur le risqued’expansions associés à la réaction sulfatique interne », thesis of the University ofParis Est, IFSTTAR, 2015.
[7] X. Bruneteaud, « Étude de l’inﬂuence de différents paramètres et de leurs interactions
sur la cinétique et l’amplitude de la réaction sulfatique interne », thesis of the École
centrale Paris, ECP, 2005.
[8] IRSN position available to the public n°2014-00314 du 6 août 2014.
[9] ASN Position statement CODEP-DCN-2014-014235 du 4 juin 2014.

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Chapter 11
Research in the Field of Human andOrganizational Factors and, MoreBroadly, in Human and Social Sciences

11.1. From the consideration of human factors in safety
to studies in human and social sciences
Human and Organizational factors (HOF) is a relatively recent discipline born withthe technological developments of the 20th century. These factors have played a central
role in most nuclear accidents. Interest in HOF is growing in the nuclear power industry,and some consider that it can increase safety more than any other discipline.
While the measures implemented in the wake of the Three Mile Island accident (TMI)
focused in particular on the ergonomic and cognitive aspects of workstations, theChernobyl accident raised questions of another kind—organizational factors. The development of a safety culture in nuclear facilities has generally been considered to be theappropriate response. This concept was described in the INSAG-4 report, published in1991 under the aegis of IAEA, just ﬁve years after the Chernobyl accident.
More recently, the 2011 Fukushima Daiichi accident revealed the importance ofsocietal factors in the risks governance in general.
Initial research, conducted in the 1970s, attempted to better understand human"functioning" and its impact on the performance of NPP operators and technicians notjust in the control room, but also during all tasks and activities carried out (tests,

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maintenance, in-service inspections, etc.) outside the control room and which could
affect safety. The scope of this research was subsequently extended to understanding
and assessing organizational factors, then societal factors. IRSN's research also extends
to methodological aspects in order to build and improve baselines and approaches for
conducting its assessments.
The documents of the OECD/NEA referred to herein [1] to [10] testify to theinternational community's continued interest in human and organizational factors in
the wake of the TMI accident.
While the very ﬁrst research on this aspect in nuclear reactors began in 1977 at IPSN's
Department of Nuclear Safety (inﬂuence of human parameters on safety, incident
analysis methodology, human reliability, control room ergonomics), it was mainly after
the TMI accident that human factors were taken into account in facility safety assessments. In the early 1980s, IPSN and EDF equipped themselves with greater resources bycreating dedicated structures (French Laboratory for the Study of Human Factors (LEFH)
at IPSN, Human Factors Group (GFH) at EDF). The topics of interest at the time included:– incident analysis,– operating staff training,– operating procedures,– human-machine interfaces and control-room ergonomics,– organization of control-room staff,– communication between control-room staff,– the use of expert systems,– teleoperation.
In 1989, IPSN also focused on other topics, such as:– maintenance of France's NPPs during outages,– operation via computerized procedures,– ﬁeldwork in maintenance.
These three topics were researched in collaboration with EDF or CNRS. The adoption
by EDF of a new organization of the operation of its NPPs prompted IPSN to research this
topic further.
Starting in 2003, IRSN's research increased in the frame of theses and began
publishing, on topics such as outsourcing, equipment or organizational modiﬁcations
in facilities in terms of human and organizational factors, etc. Outsourcing appeared to bea particular subject of concern in terms of human and organizational factors, as EDF
called on many contractors to carry out work during unit outages.
In 2012, with the creation of the Human and Social Sciences Laboratory (LSHS) andthe post-Fukushima period, IRSN extended its research to social aspects around theAnglais_Book

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governance of nuclear risks seen from the viewpoint of organizational as well as cultural
factors.
Moreover, following a number of radiation therapy accidents that occurred in France,
particularly in Épinal and Toulouse, IRSN extended its research to this issue, focusing on
the appropriation of new technologies and conformity management231.
Generally speaking, a particular feature of HOF research conducted by IRSN is that itis largely based on analyses in the ﬁeld (interviews, observations of projects and of work
situations, such as during "sensitive" activities, etc.)—which, when these analyses are
conducted as part of safety expertise, are set out in protocols established between IRSN
and the nuclear installation operator. The very nature of the topics researched and thespeciﬁc conditions of the assessments carried out thus provide IRSN with a crossdisciplinary and hands-on view of the maintenance, organizational and management
practices at NPPs. Research may also draw on simulations, as we will see further on with
the Halden Reactor Project.
Research is conducted with partners from academia, research organizations (such as
CNRS aforementioned), potentially with industrial companies, and/or as part of national
or international projects (such as the Halden project–see "Focus" further on). IRSN, in
collaboration with AREVA and French naval contractor DCNS232, created the Chair ofSafety, Organization, and Human Research (RESOH233) at the École des Mines de Nantes
in 2012. This ﬁve-year chair is dedicated to the management of safety in high-risk
industries, particularly the nuclear industry. Moreover, in 2013, ANR decided to fund the
AGORAS234 project led by the École des Mines in Paris and in Nantes and the project led
by the Center for the Sociology of Organizations of Sciences-Po on risk governance andemergency management.
A few of the most signiﬁcant projects involving235 IPSN (then IRSN), both past andcurrent, are presented below. As we will see, this research is basically intended tounderstand the mechanisms behind the various activities that involve people andorganizations in order to bring out, as much as possible, relevant levers to the safetyof nuclear facilities.

231.
232.
233.
234.
235.

In collaboration with Paris 8 University for the ﬁrst topic and La Pitié-Salpêtrière hospital for thesecond topic.
French hi-tech company specializing in defense naval systems.
Research on Safety, Organization and Humans.
Improving Governance of Organizations and Networks involved in Nuclear Safety.
Including thesis work.

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11.2. Studies and research on reactor control
room design
11.2.1.

The post-TMI periodA) Background
The 1970s were marked by the implementation of centralized instrumentation andcontrol systems in industries referred to as "continuous process": iron and steel
production, ﬁne chemicals, food processing, etc. France's nuclear program beneﬁted
from these technological advances. During this decade, IPSN was confronted with theissue of analyzing the safety not only of the control rooms in France's PWRs (900 and1300 MWe reactors), but also those in the SUPERPHENIX fast neutron reactor and the La
Hague plants.
The design of the control rooms raised questions about the work of operating crews.
Industrial companies thus wondered about what monitoring in the control room entailed,
how information was displayed on the control room consoles, how information was
passed between roundsmen and the operators in the control room, and many otheraspects. Researchers in ergonomics (specialists in human-machine interfaces and humansystem interactions), especially in Europe, rallied to answer these questions236. Theirresearch inﬂuenced and contributed to IPSN's research.
It was in this scientiﬁc and technical context that the TMI accident occurred in 1979.
TMI showed that the design of reactor control rooms plays a central role in the control ofaccidents, in terms of their prevention, the mitigation of their consequences and therecovery of the accident. The TMI accident was caused by a pressurizer relief valve that
had remained stuck open after automatically opening to reduce a pressure peak in the
RCS. This valve received a closure order, but did not close completely. However, thecontrol interface indicated that the valve was closed because the information used by it
was the order to close the valve, not its actual position—which could have been
determined by a position sensor. As a result, there was confusion in the control
room about the facility's actual status. Furthermore, the emergency shutdown of thereactor and the problems affecting the secondary system activated many alarms, causing
the control console to light up like a Christmas tree. Because there was no way to rank
the information displayed, the operating crew was quickly overwhelmed and unable todetect and extract the relevant information.
Following the TMI accident, EDF took a number of actions to improve the control
rooms in its 900 and 1300 MWe reactors. These improvements were made to the control
panels. The controls for functionally related equipment were grouped into color-coded
areas, the main areas corresponding to overall functions were deﬁned, and secondary
areas related to sub-functions were deﬁned within these main areas. A safety-parameter
display system for controlling the reactor in emergency situations was also added to the236.

Especially V. De Keyser in Belgium, L. Bainbridge in the U.K., J. Rasmussen in Denmark, and
A. Wisner and F. Daniellou in France.

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control room. Other improvements were made to the organization of the operating crew.
The position of "safety and radiation protection engineer" was created to independently
monitor the situation from a viewpoint located away from the operating crew. The safety
and radiation protection engineer makes a diagnosis of the state of the reactor (stateoriented diagnosis) whiled the operating crew uses procedures aimed at diagnosing
events. The organizational logic of the operating procedures to be applied during
emergency situations was changed as well (see Section Section 11.2.2 A further on).
In 1983, IPSN conducted a two-pronged study to analyze the improvement plan
implemented by EDF. First, it compiled a summary of the results of research conducted
on continuous processes and transposed them to the nuclear industry. Then, it conducteda study based on tests carried out on an operating simulator as part of a four-party
agreement involving CEA, EDF, Framatome, and Westinghouse. IRSN used the results ofthese studies to issue an opinion in the improvement plan implemented by EDF.

B) Main findings from research on the operation of continuous processes
In the early 1980s, a number of studies were conducted on the activity of operators
working in "continuous process" control rooms (reﬁneries, cement works, iron and steel
works). These studies focused primarily on operation under normal conditions. In its
initial (internal) reports established in 1983, IPSN summarized the key results of thesestudies in order to raise awareness amongengineers in charge of conducting safetyassessments or expertise on human-machine interactions at NPPs.
These studies revealed the central role played by anticipation in control roommonitoring. For example, in the continuous-process industries studied, it was found that
operators do not wait for an alarm or fault to appear in the control room. Instead, theyseek to anticipate the physical and chemical changes of processes in order to act before
problems occur. These operator interventions help to smooth out changes in the process,which is beneﬁcial for production quality. They also make it possible to anticipate sometransients that, although not risky in themselves, may put the operating crew in a difﬁcult
position or even put challenges of varying intensity on the facility.
These studies also showed that monitoring is an activity that is focused as well as
all-encompassing. It is focused because operators have to concentrate all their attention
on the actions that they are doing to avoid making mistakes. They have to direct theirattention on certain parts of the facilities to be able to clearly assess various phenomena.
At the same time, however, they have to maintain a bird’s eye view so as not to overlook
malfunctions or in order to be able to notice overall changes in the process.
They also showed that monitoring is an activity that is training and experienceoriented. It is widely acknowledged that operators will prefer to monitor parameters
related to functions important to production quality and operating safety. However, thestudy results enabled to qualify this statement. Indeed, operators integrate theiroperating experience and tend to prefer to monitor systems that are proned tomalfunction or have adjustment issues. Equally, they tend to interpret changes inobserved parameters ﬁrst by referring to changes to parameters of the same typethat they have previously encountered.

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Current state of research on pressurized water reactor safetyC) Four-party study program involving CEA, EDF , Framatome, andWestinghouse Operator action
Initially, the goal of this four-party program was to study the response time ofoperators during accident simulations in order to provide data for the probabilistic safetyassessments (PSA). However, in 1981, EDF decided to supplement this quantitative
approach with a qualitative analysis of the difﬁculties encountered by operators during
simulation tests. Various baseline accident scenarios for PWRs (large RCS break, steamline break inside the containment, steam generator tube break, etc.) were thus "acted
out" by operating crews on the simulator at the Bugey training center in April and May
1982.
IPSN conducted its own qualitative analysis of the data collected during thesesimulation tests. This analysis revealed that the recovery phase of accident is the mostproblematic for operating crews. The simultaneous management of several concurrent or
conﬂicting objectives was identiﬁed as a source of difﬁculty. The crews that best succeedin overcoming these difﬁculties are those who have a good ability to anticipate changes inphysical phenomena. In particular, it enables them to not focus on instantaneous changes
in parameters. Crews who lack this ability tend to gather information from multiple
sources and perform multiple checks, leading them to work "in ﬁts and starts" and have afragmented view of the facility. Team cohesion and proper coordination of the activitiesof each member also appear to be factors that can promote the development of foresightcapacity.
Furthermore, at the end of the analysis of the simulation tests, EDF and IPSN
considered that more speciﬁc behavioral data needed to be collected in order to better
identify operator strategies and reasonings. Although collecting oral data during theinterviews conducted at the end of the tests was envisaged, IPSN considered that it was
important to couple the statement with objective data. This led to a feasibility study on
the use of electrooculography237 as a means of studying visual-exploration strategies
during work in the control room.

11.2.2.

The N4 series: the first computerized control roomA) New interfaces and new questions
Following the TMI accident, EDF decided to install a computerized control room in its
future 1450 MWe reactor (N4 series). EDF wanted to beneﬁt from advancements incomputing that made it possible to ﬁlter and rank information and structure its
presentation in control rooms. For example, there were now alarm processing andranking systems, automatic operating sequences, and screens for displaying views offacilities by system, by function or by other logics.
EDF also decided to modify the design of the emergency operating procedures of its
reactors by adopting the state-oriented approach (SOA)238. The purpose of SOA is to
237.
238.

A technique for following eye movements.
See "Elements of Nuclear Safety", J. Libmann, chapter 14–IPSN/Les éditions de la physique – 1996.

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characterize the operational state of the reactor239 and gradually bring it back within asafe operating range regardless of the events that caused the degraded conditions. An
SOA procedure is a ﬂow chart consisting primarily of questions that the operator must
answer by "yes" or "no" and which provides the operations to be carried out depending on
the answer. In the case of the N4 series reactors, these procedures are computerized.
Using a trackball and the arrow keys of his/her keyboard, the operator follows the "steps"in the image of the ﬂow chart. Information that the operator needs to follow theprocedure (parameter values, excerpts of circuit diagrams, etc.) is displayed on the image.
The key operating parameters are displayed in a "dashboard" on a second screen. On athird screen, the operator can display more detailed views of circuits, systems, andparameters (e.g. pressure/temperature diagram) and make the necessary adjustments. A
number of steps in the ﬂow chart are controlled by the computerized procedure. The
procedure compares the operator's answers with those calculated by the computerized
control system using data provided by sensors. If a deviation is detected, the relationshipwith the previous step in the ﬂow chart is displayed in red. These are retroactive checks
that the operator can either accept or refuse to take into account. This is referred to as"overriding the step." In addition, most of the procedures are monitored by the computer
system, which periodically checks whether the answers given by the operator as he/she
proceeds through the procedure are in line with the changes in the parameters. The
system alerts the operator when an answer is no longer valid so that he/she can directly
go to the relevant step.
The impact of these new operating arrangements on the activity of plant operators
raised new questions. To validate its design choices, EDF equipped itself with a full-scale
simulator comprising a mock-up of the future control room coupled with a "process"simulator to allow the crews to "experiment" with operation under various scenarios.
Dozens of tests were conducted during campaigns held between 1987 and 1996. These
consecutive campaigns made it possible to assess the control room's suitability to therequirements of the operating crew's activities. Operation under normal conditions wasassessed during the ﬁrst two test campaigns. Emergency operation was tested between
1994 and 1995, with the development of the computerized operating procedures.
Although the knowledge acquired by IPSN during the aforementioned researchenabled to expertise the NPP operation under normal conditions, the computerizationof the emergency operating procedures raised new questions, prompting IPSN to conduct
new research.

B) Study on operator guidance during emergency operating procedures
In 1994–1995, EDF conducted a series of tests with a simulator to validate the use of
the computerized procedures. The purpose of these tests was to ensure that the means
provided to operators effectively enabled them to operate a nuclear facility in acceptablesafety conditions, in particular during emergency situations. They consisted in closely
observing how operators did their job, i.e. the suitability of their operations in the control
room. At the end of the tests, IPSN considered that, in order to better understand how
239.

The term "process" is also commonly used.

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the operators performed, it was necessary to study their cognitive activities (perceptions,interpretations, reasoning, etc.):– On what basis does an operator identify deviations?
– How do operators construct their vision of changes in reactor operation state?
– How do they understand how the procedure constructs its "own vision" of thesechanges?
– How do operators call on the rest of the crew to conﬁrm their opinions?
Working with researchers at CNRS, IPSN analyzed data that it had collected during
several tests that it had followed. These data consisted of:– the facility parameters recorded by the simulator;– audio recordings of the tests (alarms, information provided orally or by telephone,orders, keylogging (to follow operators as they proceeded through each step inthe computerized procedures), private discussions in order to identify areas ofconfusion or hesitations, etc.);– video recordings;– comments provided by two control room experts: one with extensive experiencein non-computerized control rooms and emergency operation using simulation
technology, the other with solid technical expertise in computerized procedures.
This research made it possible to better understand the impact of step-by-stepguidance of operators by the procedures [11]. Considering the possibility that operators
could carry out unacceptable actions, it can appear appropriate to develop a form of
guidance that would tell operators what actions to take. The computerized SOA
procedures appear to have been engineered with this view. Indeed, they were designed
to break down facility operation to a maximum, both in terms of diagnostics and theoperations themselves. The horizon of each action is greatly circumscribed by the steps inthe procedure, forcing operators to adopt a step-by-step approach to control.
Close analysis of the activities performed by operators to respond to the steps in theprocedures revealed that they use a multitude of skills that are not made explicit by theprocedures. The use of these skills assumes that the operators are trained. However,there are so many skills that it is unrealistic to imagine that operators will ever know
everything. In addition, some skills are speciﬁcally tied to emergency situations and thus
will only be used very rarely. To overcome these difﬁculties, the form of guidance must
help operators to enlist their skills when following procedures.
The study showed that the visualization of the overall structure of procedures helps
operators to understand the logic behind the current operation and helps them to enlist
the necessary skills. Indeed, interpreting the previous steps in a procedure is almost
always necessary to interpret a subsequent step. Equally, the relevance of the step's
interpretation will be reinforced if operators can see the consequences of their choices.
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operators to take initiative when necessary. Operators who passively follow procedures
are not able to effectively enlist their skills. These observations concur with the results of
two studies conducted in other sectors of industry. A 1987 study [12] on the interaction
with an expert system for diagnostic in maintenance activities showed that thisinteraction resulted in inadequate results if operators passively follow the system's
recommendations without constructing in parallel its own diagnostic. Likewise, a 1996
book [13] cites an experimental study on the interaction with a ﬂight planning system.
The study shows that "the pilots allowed themselves to be inﬂuenced by the system andlost a certain level of critical thinking. They accepted illogical choices that they neverwould have considered without the system." In order for guidance to be efﬁcient, we mustget better at taking into account the need for active involvement of the operator incontrol operations. Guidance thus must allow operators to take a certain distance fromrecommendations given in procedures.
This research also makes it possible to better understand operators' positions withregard to guidance that they ﬁnd unsuitable.
These differences between the operator's viewpoint of the procedure's "viewpoint"
occur in various situations. For instance, an operator may take changes trend in thereactor's state into account, whereas the step in the procedure reﬂects the reactor's
instantaneous state. The operator may ﬁnd it hard to follow actions that are recommended by the procedure but appear not to be optimized. The operator may ﬁnd it hard
to follow the recommended order of multiple actions whereas one of these actions isperceived as urgent. These differences in viewpoint reveal the operator's active position in
relation to the recommendations in the procedure. They show his/her ability to resist thelock-in effects caused by the step-by-step guidance of the procedures discussed above.
They can be beneﬁcial to operators, as they provide opportunities for operators to checkthe validity of their interpretation of changes in reactor state.
Two outcomes are possible when an operator notices a difference between his/herpoint of view on changes in the process and that of the procedure: the operator either
overcomes this difference and continues to follow the procedure or he/she stops
following the guidance and risks becoming completely disoriented.
An operator moves past a difference when he/she factors in additional information
allowing him/her to understand the reason for the difference. In this case, this difference
becomes understandable and the operator can accept it—at least for a certain time. Take
the example of an operator who wants to carry out an action not prompted by theprocedure. This difference in viewpoint can be overcome if the operator sees that theaction in question is dealt with by the procedure a few steps later in the ﬂow chart. The
operator can also overcome this difference, without understanding the logic behind it,
provided he/she deems that it is acceptable to continue following the procedure. This
acceptability will, in part, depend on the conﬂict that may appear between therecommended action and the operator's knowledge. But the fact that the procedurecauses the operator to act against his/her knowledge—to put it aside—can lead to a lossof credibility in the procedure or encourage the operator to adopt a passive attitude
toward the procedure and result in the negative effects presented above.

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The EPR project: the move toward increased automation

A) Design principles
As for the N4 series, EDF decided to ﬁt the EPR control room with a digital
instrumentation and control system driven by a computerized man-machine interface.
A goal of the project from the very outset was to increase the level of automation inparticular to help to reduce the number of actions to be performed by the operating crew,
lighten their workload, and establish lines of defense regarding inappropriate actions
which could be taken by the crew.
This automation applied to the start-up and shutdown of equipment (e.g. automaticconnection of the residual heat removal system [RHRS] during normal operation), theintroduction of new regulations (e.g. for the auxiliary feedwater supply of the steam
generators in shutdown states, the supply of borated water) and of new operator support
functions intended to introduce automatic corrective actions to avoid activating thereactor protection system. In addition, an automatic reactor state diagnostics system wasintroduced to aid in emergency operation. Using the main parameters for the reactor
state and the primary and secondary systems, this facility-monitoring system suggests
the operating strategy to be implemented by the operating crew during accidental
situations.
In light of these trends, IPSN began conducting a literature review in 1995 in order tomake a status of research on human-automation interaction. This work made it possible
to characterize two approaches to automation. In what is referred to as the techniciancentric approach, automation is a solution of choice for reducing operator intervention
and thus reduces the risks of human error. Operators thus carry out a set of residualfunctions that are often imprecise in deﬁnition and can be detrimental to the overall
effectiveness of the human-machine system. An alternative approach consists in viewing
automation as a tool for enhancing operator performance. This means that automatedsystems must therefore be designed to be compatible with the characteristics of humanactivity (perception, reasoning, cooperation, interaction, etc.). This activity-centric
approach calls into question systematic automation based on the analysis of its negative
impacts on this activity.
This literature review also made it possible to identify areas to be examined moreclosely—the impact of automation on the control of the operation of a process by
operating crews, the ability to acquire knowledge through operating experience, modesof cooperation between operating-crew members, the level of trust in automation, thedegree of independence, and operator involvement in reactor operation.

B) Study of the impact of automation on operator performance
Between 1996 and 1998, IPSN led a research project and the Halden Man-Machine
Laboratory (HAMMLAB) in Norway—which is afﬁliated with the Halden Reactor Project(see "Focus" below) — to study the impact of automation on operator performance(Figure 11.1). This research was conducted on the laboratory's NOkia Research Simulator(NORS) (PWR version of the VVER 440).

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Figure 11.1 The facility at HAMMLAB during IPSN's experimental work. © DR.

The purpose of the research project was to understand how operators integrate theeffects of automated actions when they manage a situation and must anticipate its
progression. An experiment taking account the type of automation (extended vs. limited)
and type of operations activity (diagnostic or operating sequence) was conducted at
HAMMLAB to answer this question. Diagnostics situations were characterized by theirgradual deterioration, with operators having a certain latitude and having to foresee
future changes. The operating sequences had to be performed either manually or
automatically in order to achieve a change of state. Six scenarios lasting around one
hour were acted out by six operating crews.
Several independent variables were measured during the tests for quantitative
analysis purposes. These included:– measuring the overall performance (crew & automated systems) based on theachievement of some speciﬁc reactor parameter values;– measuring operator performance during time windows predeﬁned for eachscenario and which ranged from the occurrence of a failure signal to thecompletion of a response action by the operating crew;– measuring situation awareness through the operators' responses to a questionnaire about the values of certain parameters. The questionnaire was given several
times during a test while the simulator was in pause mode;– measuring trust in the automated systems via a questionnaire at the end of each
test.
This analysis yielded a number of ﬁndings, such as:– during diagnostics situations, operators tend to place more trust in automatedsystems when they have a limited scope of action;

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performance and reduces operator workload;– human intervention remains preferable in complex diagnostic situations, for in thiscase extensive automation does not lead to good overall performance.

#FOCUS...........................................................................................................................................................................................................................

The Halden Reactor Project
The Halden Reactor Project was created in 1958 under the joint leadership of theOECD and the NEA (www.oecd-nea.org/jointproj/halden.html). It is being carriedout at the Institute for Energy Technology (IFE), in Norway, and is supported by
19 countries, each which funds research in areas such as nuclear fuel, the behavior ofmaterials in nuclear environments, organizational and human factors, humanmachine interfaces, etc. Some of this work is conducted by making direct use ofa small 20 MW experimental reactor (the Halden reactor is a boiling heavy water
reactor—see Figure 11.2) that regularly contains more than 30 experimental devices
at the same time. Research on organizational and human factors is based on
simulations conducted within the Halden Man-Machine Laboratory (HAMMLAB).

Figure 11.2 View of the Halden reactor hall. © IFE (Institutt for energiteknikk).
......................................................................................................................................................................................................................................................................................

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11.2.4.

Contributions and outlook

The knowledge acquired through this research constituted as many "points ofattention" for IPSN's expertises. It was therefore considered that EDF's decision tocreate a "safety and radiation protection engineer" position would make it possible tolimit the focus effects, and even lock-in effects, created by unexpected situations and theimportance of which has been highlighted by the studies. In another example, IPSN
considered that including coordination points in the accident procedures should offset
the risk of the operating crew being split up during unexpected high-stakes situations.
Likewise, the results of the study on use of computerized procedures were used by IRSN
in its assessment of the EPR control room, in particular to emphasize the importance ofproviding operators with an overview of the current procedure in order to allow them tounderstand the operating strategy.
Given the constantly evolving technologies that provide operating crews with new
functionalities, this research on the operational activities of reactors shall be continued.
One example is the development of embedded systems that will allow ﬁeld workers to
locally obtain multiple data on the status of systems and provide control room operatorswith much more accurate information about real ﬁeld conditions.
They must also be continued to derive lessons from the Fukushima Daiichi accident.
IRSN thus conducted an in-depth investigation of all of the ofﬁcial reports and accounts
that were published about the Fukushima accident. It published its ﬁndings in a report
titled "A Human and Organizational Factors Perspective on the Fukushima Nuclear
Accident" [14]. This report highlights a number of consequences of the total loss of power
to the control room: "At 3:37 p.m., the control room [of reactors 1 and 2] lost all electrical
power and was suddenly plunged into silence and darkness, leaving the operators to use
ﬂashlights to read the emergency procedures. These procedures were of no assistance inmanaging the nuclear reactor, however, as the indicators used to monitor its operation
were out of action. It became impossible to check the parameters essential in cooling thereactor: the water level and the vessel and containment pressures." This study has also
made it possible to identify issues that deserve closer examination. For example, whatknowledge should be used to collect and interpret information on the state of thesystems when indicators and procedures are no longer operational following total loss of
power to the control room? How can real cooperation be maintained between thecontrol room and emergency-response center once means of communication are nolonger operational?

11.3. Studies and research on the organization andmanagement of safety at EDF's NPPs
It was not until the early 1990s that IPSN conducted its ﬁrst study on the organizationand operation of reactors (human factors and organization of unit outages240, see

240.

Term used to designate reactor shutdowns during which the reactor is reloaded with fuelassemblies, maintenance is carried out, and other operations are performed.

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Section 11.3.1 below). Previous studies had only focused on operating crews and control roomdesign. Although these organizational studies were prompted by the emergence of issues on
the safety of maintenance operations, they also beneﬁted from research conducted by U.S.
sociologists on organizational reliability in the wake of the TMI accident.
In 1984, a book with the provocative title of "Normal Accidents: Living with High-Risk
Technologies" [15] was published. Providing an in-depth analysis of several industrial
accidents, including TMI, it uses the term "normal accident" to emphasize accidents thatare related to the very nature of high-risk systems. Adding protections to the systems infact increases their complexity, which increases their possibility of failure, reduces theability of operators to understand their overall operation, and in turn increases theirvulnerability. Furthermore, the components of these complex organizations are tightly
coupled, meaning that what affects one component may affect its related components as
well.
Following the publication of Normal Accidents, living with high risk technology, someresearchers considered that the viewpoint presented in it was not borne out by actual
accidents affecting high-risk systems, which were extremely rare. They thus sought tounderstand how high-risk industries were able to maintain a high level of reliability. The
movement known as High Reliability Organizations (HRO) was created by a group of
researchers at the University of California, Berkeley241. Their work explains the organizational measures and methods implemented daily by HROs to maintain a high degree of
reliability. It emphasizes the positive role of organizational ﬂexibility, i.e. rallying several
levels of operation in the organization depending on the situation, the redundancy ofchecks by people, strong consistency between the organization's goals and individual
goals, recognition of the importance of the skills of people whatever their level in thehierarchy, and the continuous organization of training and refresher training courses, theco-existence of centralized decision-making and decentralized operational decisionmaking.
This research remained relatively unknown in France in the 1980s. It was not until a1999 thesis titled Le nucléaire à l’épreuve de l’organisation [16], that information aboutthe HRO paradigm spread. This thesis also emphasized the paradigm's limits andproposed new developments.
Another work of interest is Diane Vaughan's "The Challenger Launch Decision—Risky
Technology, Culture, & Deviance at NASA" [17]. Published in 1996, it explores the reasons
behind the Challenger accident and analyzes several aspects of the organization at
various time spans. It thus looks at the interactions between engineers in the hours
preceding the launch of the shuttle as well as the relationship dynamics between NASAand the U.S. Government over several years. By focusing on the micro-decisions made at
the time, Vaughan shows that what in hindsight appears to be a series of clearly
identiﬁable errors is actually a series of decisions and interpretations that are perfectly
understandable in the context in which they were made, but which are in fact slight241.

Todd LaPorte, Karlene Roberts, and Gene Rochlin, joined by Paul Schulman and Karl Weick inparticular.

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deviations from normal limits and lead imperceptibly to "normalization of deviance". Thisis one of the key contributions of this book.
Also worth mentioning is the PhD thesis "When designers anticipate the organization
to control risk: two installation modiﬁcation projects on two SEVESO 2 classiﬁed sites"
completed in 2008 [18]. This thesis, which contains many sociological lessons that cannot
all be listed herein, brought to light or conﬁrmed a few of the "workings" that form thebasis of organizational reliability:– the involvement, right from the design of facility retroﬁts, of operators (in all the
relevant ﬁelds) and contractors (such as maintenance subcontractors) who will all
ultimately participate in the operation of the retroﬁtted facilities, is a factor of
success;– corrections or additions to the initial design of the facilities is a factor of risks incomplex systems (as is a nuclear reactor and its operation).
All of this work will inﬂuence the research carried out by IPSN (then IRSN) detailed
hereafter.

11.3.1.

Organization of maintenance activities

In the summer of 1989, three incidents occurred while maintenance was being
conducted on reactors in France's NPPs. These incidents highlighted the fact that safety
could be compromised by a forgotten or incorrectly basic operation (gaseous releases
beyond the threshold, a latent failure that could have caused an engineered safeguards
system to fail during an accident sequence). EDF reinforced its organization starting in
1990. It created new positions (checkers and contract managers), reinforced site
engineering and preparation, provided contractors with training on quality and safety,
and introduced the performance of a risk analysis prior to all maintenance activities.
Although these arrangements may, in principle, reinforce the reliability of maintenanceactivities, IPSN and EDF agree that a deeper understanding of the conditions of unit
outages was necessary to be able to better assess the relevance of these measures.

A) "Human factors and the organization of unit outages: safety
challenges"
In 1991, IPSN undertook a study to understand the maintenance work in order to beable to assess the organizational changes proposed by the nuclear reactors operator. An
ergonomic analysis of the activities of various workers during two unit outages—in July
1992 and July 1993—on the same site was implemented. In concrete terms, more than
20 projects (essentially related to mechanical work) and the activities of more than
50 workers with various positions were closely tracked. Two requaliﬁcation tests, among
the most complex, of the safety-injection system items were observed. An ergonomist
took part in the daily outage meetings during which progress on the maintenance work
was checked and rescheduled where necessary. This ergonomic analysis revealed thespeciﬁc aspects of each worker's job, its speciﬁc constraints, and the associated risks, anddescribed how the workers coped with contingencies in their established organization.

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Both of these aspects made it possible to identify possibilities of organizational failureand failure in individual work, as well as the implicit strategies and skills developed byworkers to mitigate these risks of failure.
The study revealed that the conditions under which maintenance was performed
were insufﬁciently taken into account during the maintenance-preparation activities
and when drawing up operating experience feedback on unit outages. The little
interaction between contractors and preparers was identiﬁed as a contributing factor
to this situation. Information about the difﬁculties in carrying out maintenance work did
not rise, despite the fact that these difﬁculties led to unexpected events, lost time, andworkers making decisions themselves. All this had a human "cost" and could compromisesafety.
The management of unexpected events is a primary concern in the life of workers.
Nothing happens exactly as planned and unforeseen events can crop up. When theyoccur, they must be detected and maintenance work must be adjusted accordingly. Their
impact on the safety of this readjusted maintenance work must be assessed in real time.
The time and resources needed to do this are often underestimated.
The study also revealed that jobs and skills of diverse nature are involved in themaintenance activities performed during a unit outage. The comparison of viewpoints inreal time, when organized, helps to ensure maintenance activities are carried out
smoothly. With this in mind, skills management and time constraints, attention tocollaboration among various maintenance workers, and the creation of structures wherelimits encountered during maintenance can be openly discussed are elements importantto safety.

B) "The organization of unit outages in a period of reform"
In 1994, a study by the Center for the Sociology of Organizations (CSO) was carriedout in French NPPs as part of a PhD thesis being written on the organization of work inhigh-risk systems. Titled "The organization of unit outages in a period of reform," thethesis subsequently found an echo in the 1999 thesis [16] referred to herein. It provided acomparative analysis of the preparation and completion of four unit outages at two NPPs
in the United States and two NPPs in France.
The main objective of the thesis was to assess changes initiated by EDF, starting in
1991, to reinforce the organization of its maintenance activities during unit outages. It
sought to identify the difﬁculties encountered during outages. However, it exceeded andrenewed the scope of analysis of the previous study (see Section A). On the one hand, it
attempted to understand how a number of outage constraints occur during the lead-up
to the outage itself. On the other, it sought to understand how some characteristics of
the intervention situations are related to overall constraints that burden nuclear sites. In
terms of the methodology, the thesis made use of strategic analysis, developed in thesociology of organizations (see, for example, the work referred to herein [19]). It useddata collected from 68 interviews to trace back to the "games" that take place between
the members of an organization and the strategies that they use to perform theiractivities and achieve their goals.

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The thesis revealed that setting up a project organization ("permanent structure of aunit outage") dedicated to the completion of unit outages facilitates the preparation ofmaintenance operations. However, if the project team is formed around six months
before an outage, each department gradually assigns its representatives and somepeople, such as inspectors, are assigned late in the process. Furthermore, the people
assigned to the project team are chosen by their respective departments. Outage project
managers therefore do not choose their "troops." The thesis also showed that the"ramping up" of the outage structure was accompanied by a withdrawal of thedepartments in the preparation of outages. Broadly speaking, the departments did
not appear concerned with outage projects, considering that they were solely a matterfor outage workers.
Another prominent change observed on the site was how work is awarded tocontractors. Buyers participate in drawing up work orders and business aspects (thedeﬁnition of technical requirements, competitive tendering, etc.) play a larger role. The
level of detail required in speciﬁcations is a source of difﬁculty for EDF staff, especially for
those who lack experience or deal with operations for which it is difﬁcult to anticipate theworking conditions. Furthermore, although competitive tendering can help contain
maintenance costs, it creates contractor instability and the possibility of having towork with new contractors whose technical expertise remains to be seen over time.
The study also revealed the complexity in the scheduling of outage work. A ﬁrst
draft of the master outage schedule is established based on aspects such as thepreventive maintenance programs, the retroﬁts to be made, and constraints related
to the provision of tools. The coordinators then deﬁne the work packages and send thecorresponding orders to contractors. This preparation essentially takes place for eachindividual package with relatively little overall coordination. The result is that crossdisciplinary activities, such as the procurement of spare parts or routine radiation
protection equipment, are insufﬁciently taken into account.

C) Study on the implementation of EDF's outage control centers
Since 2012, EDF has been implementing COPAT (Centres opérationnels de pilotage
des arrêts de tranche), its own version of the outage control centers (OCC) in use in North
America. IRSN conducted a benchmark study in June 2011 to assess the impact of thisnew organization of safety and radiation protection. It did so by consulting licensees in
the United States and Canada as well as the Canadian Nuclear Safety Commission(CNSC). Prior to this, EDF had conducted its own study in 2007.
EDF's COPATs are designed to coordinate outage activities 24 hours a day to ensurethat planned outage schedules are better adhered to. Sensitive phases and sequences ofcritical activities, which are identiﬁed beforehand, are closely monitored. The COPAT isalerted of probable or conﬁrmed deviations of 30 minutes in the critical path in order totake the appropriate arrangements without affecting outages. This assistance makes it
possible to solve difﬁculties encountered during outages more quickly and share realtime information among teams. This organization requires having quick response teams.
Its aim is to be able to anticipate potential difﬁculties and imagine scenarios for managing
them by foreseeing the sufﬁcient necessary resources.

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IRSN's study made it possible to capitalize the facts, observations, and lessons,particularly regarding the key factors of success, the limits, and the challenges ofmanaging unit outages via outage control centers. Its objective was more to identify
key issues than collect organizational solutions. It thus helped to establish a method forassessing EDF's proposal for implementing COPATs. Although the study validated anumber of topics for in-depth study that had been identiﬁed previously by IRSN
(complexity of the management of interfaces between hundreds of people and withinprojects comprising thousands of activities, forecasting of human resources in a period ofmassive retirement, integration of hundreds of contractors, operating experience feedback, change management, etc.), it revealed other important topics to be investigated indetail during IRSN's assessment of the measures adopted by EDF.
The interviews conducted during the study revealed that OCCs were seen as
structures that were added to solve problems of coordination and cooperation between
operating crews and maintenance crews and bridge cultural gaps. It was thus deemedwise to investigate how COPATs could increase the complexity of unit outage
management.
The study also showed the variety of project structures in North America that adhere
to OCC principles, have different stafﬁng levels (10 to 20 people), and place various levelsof importance of operation and maintenance. The question was thus raised about
whether COPATs should be tailored to each French NPP.
It was noted that North American facilities placed importance on preparation (longer
by a few months and standardized) and controlling the forecast volume of maintenance.
Licensees in North America tolerate only 10% additional unexpected242 maintenanceover the initial volume determined six months prior to the start of an outage. In the caseof France's NPPs, this additional maintenance can rise to 100% (50% during the six
months prior the outage and 50 % during the outage). This observation raised a numberof questions:– Was there a lack of foresight detrimental to the organization of the preparation ofmaintenance activities in France's NPPs?
– Was the restriction of unexpected maintenance by North American licensees
conducive to organizational success but detrimental to equipement reliability andsafety?
The study found that OCC staff had a great amount of expertise (three years as an
assistant in an OCC, then eight years in that position) that enabled them to calmly deal
with situations, particularly with unexpected events. The situation in Canada and France
was vastly different. It was found that COPATs could not be training centers, which
related to issues of human resource management.

242.

The term unexpected maintenance work encompasses both that which may be overlooked during
the inventory made six months before the outage and that which is caused by anomalies foundduring the six months before or during the outage.

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Research in the Field of Human and Organizational Factors and, More Broadly, in Human and Social Sciences 187

The study also showed that there was a relationship between foresight andresponsiveness at licensees in North America. There was a high priority on foreseeing
potential failures and scheduling slippages and on being prepared to manage potential
unexpected events (capitalization of past unexpected events managed, prepared crews to
face similar events). This aspect had to be looked at in the case of EDF's COPATs.
Lastly, the study revealed that North American licensees (30%) signiﬁcantly outsourced fewer maintenance activities than EDF (80%), pooling of resources among sites,
and greater internalization of logistical means (ensuring that maintenance activities arecarried out under the right conditions).

11.3.2.

Contractor management

The studies conducted between 1992 and 1994 (Sections 11.3.1 A and B) chieﬂyfocused on the organization and performance of maintenance activities during unit
outages. In the early 2000s, IPSN decided to conduct research on the use of contractors.
There were two reasons for this: contractors are increasingly being used, in particular for
maintenance activities, and outsourcing in the nuclear industry is a subject of intense
debate that, in the end, has little basis on extensive studies.

A) Outsourcing relationships and their consequences on security
and safety
Some research emphasizes the difﬁcult working conditions of some employees ofcontractors and directly link these difﬁculties to their subcontractor status. The use ofcontractors is thus seen as a factor that deteriorates working conditions and even thevery quality of work and thus safety. However, some arguments cast doubt on thisnegative effect of outsourcing. For example, some types of sensitive equipment (pumps,
valves, etc.) have always been maintained by the employees of their manufacturers eversince the ﬁrst NPPs were commissioned. This situation is widely viewed as an assurance ofquality. Likewise, some maintenance tasks are extremely difﬁcult regardless of which
employee is performing them. In order to more clearly identify the effects of outsourcing,
the research focused on the outsourcing relationship, how this relationship is built
through the joint action of the project owner and the contractor, and how thisrelationship affects not just the work of contractors' employees, but that of the project
owner's employees as well.
Such research was conducted as part of a sociology PhD thesis in collaboration with
the French rail operator SNCF and the French Gas, Network management, Distribution
Company GrDF [20]. This partnership made it possible to access maintenance projects to
replace railroad tracks and track ballast (SNCF) and to replace gas-distribution networks(GrDF). These projects were carried out mainly by the employees of contractors. Morethan 50 interviews were conducted with the local, regional, and national employees of
SNCF and GrDF. These employees were in charge of preparing and conducting themaintenance projects and providing feedback on them. These interviews made it possible
to collect data on the meaning each employee ascribed to their work practices, how they
saw their work, and the ties linking them to the contractors. Several projects were alsoAnglais_Book

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observed as they took place. For example, a project to replace track ballast and railway
ties in Northern France (November 2007 to January 2008) was followed, making it
possible to attend initial project planning and scheduling meetings, follow an employee incharge of bringing in and clearing out the work trains on the railtracks, and more.
Although the ﬁndings yielded by this research are too numerous to mention here, it
provided some important observations and important lessons.
First of all, this research made it possible to provide a typology of the various types ofoutsourcing activities243, including in the case of insourcing by a contractor. Aspects of"cascading" outsourcing were also discussed in this research.
The research also revealed that there are many types of relationships between project
owners and contractors and that they depend in particular on the extent of outsourcing,
how much technical skills are shared between project owners and contractors, theirdegree of mutual dependence, and the duration of outsourcing agreements. The 2003
collapse of the gangway leading to the Queen Mary 2 liner at the Saint-Nazaire shipyard
is cited as an example highlighting a number of aspects of the relationships between the
project owner (Les Chantiers de l’Atlantique) and its contractor (SAS Endel) thatjeopardized safety at the shipyard and ultimately led to the accident244.
The thesis showed how much the premise that "the market' quality, cost, andturnaround time results—always better than those obtained internally—make it easy tooverlook the transaction costs required to remain in control of facilities". Indeed,outsourcing often leads to "virtualization" of safety or security where hands-on monitoring may be severely decreased (or even prohibited) in favor of supervision based on
the validation of technical studies and the veriﬁcation of data used by contractors with
reduced hands-on checks ("paper-based" safety or security). "Cascading" outsourcing cancompound this situation and cause “risk to migrate to the weakest links in the outsourcing
chain.”The thesis furthermore emphasized that outsourcing affects the project owner's
internal organization, thus challenging the "idea that outsourcing is simply a "transfer" ofoperations to the contractor without any deep-rooted changes to the organization in which
the change occurs."
Another aspect highlighted is that "it is impossible to study the link between
outsourcing and safety or security with a rationalistic vision consisting in identifying
factors that would directly affect safety or security and lead to the conclusion thatmanagerial intervention on such-and-such factor would, with such-and-such percentage ofodds, guarantee safety or security". The thesis shows the oversimpliﬁcation of theoutsourcing "approach" from the sole perspective of the "sustainability of the skills"of the project owner and "contractor monitoring." Outsourcing must be considered andanalyzed as a relationship or partnership.

243.
244.

Technology-driven or capacity-driven insourcing, technology-driven or capacity-driven external
subcontracting, and outsourcing.
These factors were clearly established by the criminal investigation into the accident.

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Research in the Field of Human and Organizational Factors and, More Broadly, in Human and Social Sciences 189

The thesis also pointed out that "established" contractors—thanks to their acquired(and well-earned) reputation—can become elevated to favored status: "contractors who
know that they are well positioned create a lot of leeway for themselves".
These ﬁndings and lessons are points of attention in IRSN's assessments of outsourcing practices used by licensees, in particular those conducted on EDF in 2015. They
also opened the way to other tracks of research, such as that carried out as part of RESOH
research chair.

B) Outsourcing relationships—the RESOH research chair
Inter-organizational relationships, such as outsourcing, were seen as a source ofadvances in safety following the progress made in human and organizational factors ofsafety since the 1980s, a fact attested as much by the conclusions of the complementary
safety evaluation (ECS) conducted by IRSN and ASN following the Fukushima accident.
The Chair of Safety, Organization, and Human Research (RESOH) is dedicated to studying
organizational and human factors of safety at high-risk industrial facilities throughout
their life cycle (design to dismantling and waste management). It focuses in particular onthese inter-organizational relationships.
RESOH was created in March 2012 by its four founding partners—the École des minesde Nantes, IRSN, AREVA, and DCNS—for a ﬁve years duration.
The RESOH research chair focuses on two areas of research:– inter-organizational relationships, in particular via the development of outsourcing and co-contracting relationships245;– the integration of safety in all the constraints and management systems related toresearch on industrial competitiveness.
Safety is a collective "construct" that brings into play the liability and individual
activity of not just each contractor, but of teams and institutions as a whole as well(project owners, contractors and co-contractors, inspectors, etc.). RESOH's objective is toanalyze the construction of safety at the scale of the system formed by the complex
relationships between all of these parties and by taking their environment (economic,
legal, social, etc.) into account.
Its aim is to provide answers to these issues by identifying the vulnerabilities,
robustness and resilience246 of these contracting and co-contracting networks and by
studying the managerial practices and management systems that can enhance safety
within them.
RESOH uses two research methods:– qualitative ﬁeld studies that call on techniques used in the sociology of work andorganizational ethnography;
245.
246.

This term describes the pooling of contractors (consortium) for the purposes of providing service
for a licensee.
Adaptability to unexpected situations.

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Current state of research on pressurized water reactor safety– an observatory on outsourcing practices, for quantitative and cross-disciplinary
purposes.
RESOH aspires to bring in additional partners and thus explore other areas ofindustry.
Its research on the management of complex projects and the use of outsourcing
covers two aspects:– the contribution of unit outage schedules to improving time management andcoordination between parties;– the role agreements play in mutual commitments between project owners andcontractors.
This research is grounded in ﬁeld studies conducted with AREVA (La Hague) and
DCNS (Cherbourg).

11.4. Research on human and social sciences: theAGORAS project
In his introductory message to ofﬁcial report by the Fukushima Nuclear Accident
Independent Investigation Commission commissioned by the National Diet of Japan,
Chairman Kiyoshi Kurokawa mentioned a number of societal factors "that could andshould have been foreseen and prevented". He wrote of "mindset that supported thenegligence behind this disaster" (...) "made in Japan" Chairman Kurokawa's reﬂections
have a general scope and should be of concern to any party involved in the "governance"of risks of any type. It was in this ﬁeld of human and social sciences that IRSN decided, in
2012, to conduct new research, in particular as part of the AGORAS247 project. Rather
than investigate the human and organizational factors that contribute to the operationalsafety of nuclear facilities, it focused on the general functioning of all stakeholders
involved in ensuring safety.
In France, the complementary safety evaluation (ECS) conducted in wake of the
Fukushima accident pointed up the importance of studying the link between safety andinter-organizational relationships. It was this as yet little-explored topic that became thefocus of the AGORAS project to improve governance of nuclear safety organizations andnetworks. AGORAS was selected in late 2013 as part of a call for nuclear safety andradiation protection projects launched by the French National Research Agency (ANR)
and spurred by the highest levels of France's Government. Its goal is to understand how
institutional equilibriums among licensees, subcontractors, partners, and also safety
regulators and public assessment bodies are built and evolve in a world where civil society
is playing an increasingly greater role. Emphasis will be placed on the dialog that isestablished between stakeholders and which partially forms the basis of safety in the
nuclear industry as a whole.

247.

Amélioration de la gouvernance des organisations et des réseaux d'acteurs pour la sûreté nucléaire.

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Research in the Field of Human and Organizational Factors and, More Broadly, in Human and Social Sciences 191

Expected to last six years, AGORAS is structured around two aspects—accident
prevention and emergency management. The goal of the ﬁrst aspect is to analyze theimpact of the Fukushima accident on the safety "approach" for nuclear facilities and therelationships among those involved in the governance of nuclear risks. The goal of thesecond aspect is to analyze how this accident has helped to change the perception of
nuclear accidents and the methods used to manage the preparation of accident and postaccident situations. It will look at changes in emergency-response organizations anddoctrines since the TMI accident by analyzing feedback obtained for emergency exercises
and actual emergencies (the ﬂooding of the Blayais NPP in late 1999 is a possible
example). The expected results should reveal vulnerability factors related to the growing
complexity among emergency-response teams and institutions, tracks for improvement
and opportunities making simulation exercises more realistic.
IRSN is coordinating two efforts as part of the project. The ﬁrst (2014–2018) will
identify organizational and cultural conditions that inﬂuenced decisions about the
Fukushima NPP (technical choices, design options, etc.) and which ultimately proved
to be inappropriate. This effort comprises two areas of research:– the ﬁrst, the study of assessment processes and technical decisions, consists inunderstanding, as part of a socio-historical analysis developed to analyze anumber of major accidents (such as the Challenger Shuttle disaster), the dynamics
and factors that may explain certain concepts, tools, and data that are poorly
known, ignored, or inadequately taken into account by all stakeholders involved in
the governance of nuclear risks. It is based on an analysis of French cases;– the second consists in analyzing the preparation and implementation of riskcontrol documents, such as the "extreme cold weather baseline" or the "ﬂoodguide" (see Section 8.2 herein). The aim is to analyze the dynamics andrepresentations at work during their preparation and how they are able to either
strengthen or weaken the legitimacy of the documents. How are some proposed
tools, data, recommendations, or approaches interpreted and used? Whichdeviations or amendments in relation to the initial intentions of the designersof these documents are ultimately accepted by regulators? What are the limits,checks and impact assessments?
The second action (2013–2018) relates to technical dialog during the post-accident
complementary safety evaluation (ECS) to understand how the occurrence of a major
accident can lead to a reassessment of previous safety practices. This involves identifying
the conditions that both favor and impede reassessment. Complementary safety
evaluation (ECS) has effectively opened a period of intense technical dialog among
licensees, ASN and IRSN. In concrete terms, the goal will be to answer the following
questions: What new inter-organizational dynamics will be brought to light by the lines ofreasoning and potential controversies among the stakeholders? Which "paradigm breaks"(the unforeseen accident has happened) or, on the contrary, what lines of "defense" will
underpin these lines of reasoning? How will media coverage of some debates and theparticularity of extending the service life of NPPs affect the positions and arguments ofkey players of nuclear safety?

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References[1] Proceedings CSNI specialist meeting on operator training and qualiﬁcations (12–15
October 1981, Charlotte, N.C., USA), rapport NEA/CSNI-63, Vol.1, Vol. 2, 1982.
[2] Identifying signiﬁcant human actions in reactor accidents, NEA/CSNI-89, 1984.
[3] Analysis of incidents involving cognitive error and erroneous human actions, NEA/CSNI-180, 1990.
[4] Proceedings of the Specialist Meeting on Operator Aids for Severe Accidents
Management and Training (1993, Halden, Norway), NEA/CSNI/R(1993)9, 1993.
[5] Conclusions of the Specialist Meeting on Operator Aids for Severe AccidentManagement and training (Samoa) (1993, Halden, Norway), NEA/CSNI/R(1994)
13a, 1994.
[6] Joint OECD/NEA-IAEA Symposium on Human Factors and Organization in NPP
maintenance outages: impact on safety (1995, Stockholm, Sweden), NEA/CSNI/R(1995)27, 1995.
[7] Identiﬁcation and Assessment of Organizational Factors Related to the Safety of
NPPs: State-Of-the-Art Report (SOAR), NEA/CSNI/R(1998)17, Vol.1, Vol.2, 1998.
[8] Report on the CSNI Workshop on Nuclear Power Plant Transition from Operation
into Decommissioning: Human Factors and Organization Considerations, NEA/CSNI/R(1999)17, Rome, Italy, May 17–18, 1999.
[9] Identiﬁcation and Assessment of Organizational Factors Related to the Safety of
NPPs, State-Of-the-Art Report (SOAR), NEA/CSNI/R(1999)21, Vol.1, Vol. 2, 1999.
[10] CSNI Technical Opinion Paper No. 12, Research on Human Factors in New Nuclear
Plant Technology – also referenced as: NEA no. 6844, NEA/CSNI/R(2009)7, 2009.
[11] F. Jeffroy et al., « De l’évaluation de la sûreté à la recherche dans le domaine desfacteurs humains : le cas de l’activité de conduite avec procédures informatisées »,p. 237–249. In: sécurité et cognition, Ganascia J-G. Éditeur. Hermes, Paris, 1999.
[12] E.M. Roth, K.B. Bennett and D.D. Wood , Human interaction with an "intelligent"machine. International Journal of Man-Machine studies, 27:479–525, 1987.
[13] R. Amalberti, La conduite des systèmes à risques, Presses universitaires de France,
Paris, 1996.
[14] A Human and Organizational Factors Perspective on the Fukushima Nuclear
Accident, IRSN report PSN-SRDS/SFHOREX No. 2015-01, 2015.
[15] Charles Perrow, Normal accident, living with high risk technology, Princeton
University Press, new edition, 1999.
[16] Mathilde Bourrier, Le nucléaire à l’épreuve de l’organisation, Presses universitaires
de France, 1999.
[17] Diane Vaughan, “The Challenger launch decision, risky technology, culture anddeviance at NASA”, The University of Chicago Press, 1996.
[18] Cynthia Colmellere, « Quand les concepteurs anticipent l’organisation pour maîtriser
les risques : deux projets de modiﬁcations d’installations sur deux sites classés
SEVESO 2 », Thesis of sociology, University of Compiègne, 2008.
[19] M. Crozier, E. Friedberg, L'Acteur et le système, Éditions du Seuil, 1977, 1981.
[20] Marie Ponnet, « Les relations de sous-traitance et leurs effets sur la sûreté et lasécurité dans deux entreprises : SNCF et GrDF », Thesis of sociology, University of
Nantes, 2011.

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Chapter 12
Other Research and Tracks of New
Research

IRSN also conducts research of varying scope in many other technical areas related to
PWR safety, including:– the reliability of critical codes in the code-based digital instrumentation andcontrol systems used in N4 series and EPR reactors;– the behavior of materials subjected to seismic loads with the proprietary
development, via numerical simulation, of material "fragility" curves that canbe implemented in probabilistic assessments of seismic hazards. Determination offragility curves via numerical simulation is an alternative to the American method(EPRI curves based on a lognormal model and predetermined parameter sets) andcould also make it possible to determine in a most suitable way the parameters of
EPRI's lognormal model. To achieve this, IRSN developed Interactive Seismic
Analysis of Fragilities of Equipment and Structures (ISAFES), a simulation codethat makes it possible to conduct dynamic time simulations using representation
objects of structural components referred to in France as MARC (Weight, Damper,
Stiffness, Criterion). This development thus addresses the ﬁelds of dynamic
numerical simulation of equipment with uncertainties as well as soil-structureinteraction with seismic excitation;– in-vessel mixing phenomena in PWRs, which could occur, for example, if borated
water is diluted with clear water, a situation that can lead to reactivity accidents;

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Current state of research on pressurized water reactor safety– airborne contamination. This research, apart from that on ﬁltration systems, ismore speciﬁcally geared to the issues encountered in fuel cycle facilities anddismantling operations.
Furthermore, the designers of the Generation III248 and III+ reactors, as well as ofprojects for integral or modular reactors (SMR249), are increasingly emphasizing passive
systems250. There has been renewed interest in passive systems since the Fukushima
Daiichi accident. Needless to say, the accident's consequences led researchers to look at
prolonged loss of electrical power and LOCA. Passive systems may be solutions forhandling such situations. However, the increasing use of passive systems reveals that
more needs to be known about their use in deterministic safety demonstrations and
PSAs. Indeed, questions remain to be answered about their reliability, their effectiveness,the possibility of "recovery" in case of failure, and the veriﬁcation of their performancefrom reactor design to reactor operation.
IPSN (and after IRSN) addressed the subject of passive systems during assessments
conducted in the 1980s and 1990s on the PHENIX and SUPERPHENIX fast neutron
reactors and, more recently, for the ASTRID (Advanced Sodium Technological Reactor for
Industrial Demonstration) project. These systems related speciﬁcally to the possibility ofremoving the residual heat from the reactor via natural convection of the sodium in thereactor vessel—primary circuit—and in the secondary circuits, which also contained
sodium. At the same time, IRSN was involved in the European Thermal-Hydraulics ofInnovative Nuclear Systems project (THINS), particularly for the numerical simulation ofnatural convection tests conducted at PHENIX in the late 1990s (as part of its "end-of-life
tests") and for the qualiﬁcation of the CESAR module of the ASTEC simulation code.
However, proposals for research projects (extending beyond sodium-cooled reactors andspanning prestandardization research, testing of components, numerical simulation, etc.)
are being prepared following the European Commission's H2020 call for projects. These
projects include REPASS251 and NUSMOR252, in which IRSN may participate withpartners such as ENEA253, KIT and GRS.

248.

249.
250.

251.
252.
253.

The use of passive systems in an EPR is limited to gravity drop of the control rods, pressurized
accumulators to reﬂood the core, the possibility of thermosiphon cooling of the reactor following
failure or voluntary shutdown of the primary pumps—to be checked in the reactor itself as well as
in the case of reactors in operation—, hydrogen recombiners, and the core catcher in the event ofa core melt accident.
Small Modular Reactors.
The concept of a passive system is broad: spanning systems that make limited used ofcomponents requiring mechanical movement to perform the systems' functions, the supplyof outside power or support functions, human intervention (to activate and operate the system
for the appropriate period), and systems that use natural phenomena (gravity, heat transfer byconduction, natural convection, or radiation, pressure differences, etc.).
Reliability Evaluation of Passive Safety Systems.
NUgenia Small Modular Reactor with passive safety features.
Italian National Agency for New Technologies, Energy and Sustainable Economic Development.

Science and Technology Series

Jean Couturier, Michel Schwarz
Current State of Research
on Pressurized Water Reactor Safety
Jean Couturier, Michel Schwarz

IRSN is a public industrial and commercial body undertaking research and consultancy
in the field of nuclear safety and radiation protection. It provides the public authorities
with technical support. It also carries out various public service missions entrusted toit under national regulations. In particular, these include radiological monitoring of
the environment and workers in France, managing emergency situations, and keeping
the public informed. IRSN makes its expertise available to partners and customers
both in France and worldwide.

Head office31, avenue de la Division Leclerc
92260 Fontenay-aux-Roses
RCS Nanterre B 440 546 018
Telephone +33 (0)1 58 35 88 88
Postal Address
B.P. 17 - 92262 Fontenay-aux-Roses Cedex
FranceWebsite www.irsn.fr

ISBN : 978-2-7598-2164-8

9 782759 821648

Current State of Research on Pressurized Water Reactor Safety

This publication reflects IRSN’s desire to keep an enduring record of its results andto share its knowledge.

Enhancing nuclear safety

Science and Technology Series

For more than 40 years, IPSN then IRSN has conducted research and developmenton nuclear safety, specifically concerning pressurized water reactors, which are
the reactor type used in France. This publication reports on the progress of thisresearch and development in each area of study – loss-of-coolant accidents, coremelt accidents, fires and external hazards, component aging, etc. –, the remaining
uncertainties and, in some cases, new measures that should be developed toconsolidate the safety of today’s reactors and also those of tomorrow. A chapterof this report is also devoted to research into human and organizational factors,
and the human and social sciences more generally. All of the work is reviewed
in the light of the safety issues raised by feedback from major accidents such as
Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments
conducted, for example, as part of the ten-year reviews of safety at French nuclear
reactors. Finally, through the subjects it discusses, this report illustrates the many
partnerships and exchanges forged by IRSN with public, industrial and academic
bodies both within Europe and internationally.

Current State of Research
on Pressurized Water
Reactor Safety

Science and Technology Series