Materials for sustainable nuclear energy : the Strategic Research Agenda (SRA) of the Joint Programme on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA).
Date: 2019-05-22
Institutions: Joint Research Centre
Subjects: energy production, energy technology, nuclear reactor, nuclear research, nuclear safety

MATERIALS FOR 
SUSTAINABLE 
NUCLEAR ENERGY 

The Strategic Research Agenda (SRA) of the Joint 
Programme on Nuclear Materials (JPNM) of the 
European Energy Research Alliance (EERA)

L. Malerba (CIEMAT) 
M. Bertolus (CEA, DEN) 
K.-F. Nilsson (JRC)

Joint Programme on Nuclear Materials of the European Energy Research Alliance

Coordinating sustainable nuclear materials research for a low carbon Europe

EUR 29740 EN

MATERIALS FOR 
SUSTAINABLE 
NUCLEAR 
ENERGY 

The Strategic Research Agenda (SRA) of the Joint 
Programme on Nuclear Materials (JPNM) of the 
European Energy Research Alliance (EERA)

Chair:

L. Malerba (CIEMAT)* 

Main authors:

L. Malerba (CIEMAT)*
M. Bertolus (CEA, DEN)
K.F. Nilsson (JRC)

Contributors:

M. Angiolini (ENEA), M. Cologna (JRC), A. Jianu (KIT), J. Kalivodová (CVR)
C. Pareige (CNRS), M. Serrano (CIEMAT), J. Somers (JRC)

Editing:

J. Marrow (U. Oxford)

Joint Programme on Nuclear Materials of the European Energy Research Alliance

Coordinating sustainable nuclear materials research for a low carbon Europe

* Previous affiliation SCK•CEN

JRC116292
EUR 29740 EN
ISBN 978-92-76-03318-9 ISSN 1831-9424 doi:10.2760/59099

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Summary

1.

INTRODUCTION .......................................................................................................................................... 19

1. 1.  Societal, economic and technical challenges ..................................................................................... 19

1. 2.  The EERA JPNM in the European platform landscape ..................................................................21

1. 3.  Generation IV systems in Europe ..................................................................................................................... 23

1. 4.  Objectives and Grand Challenges of the EERA JPNM .................................................................. 24

1. 5.  Purpose, target and structure of this Strategic Research Agenda ..................................... 25

1. 6.  Sources and links with other roadmaps and SRAs ..........................................................................28
1.6.1.  Materials roadmap enabling low carbon energy technology (MR)  .............................28
1.6.2.  The SET-plan Integrated Roadmap (IR)  ..........................................................................................28
1.6.3.  OECD/NEA Technology Roadmap for Nuclear Energy and 

other OECD/NEA initiatives .....................................................................................................................29

1.6.4.  SNETP Strategic Research and Innovation Agenda 

and Deployment Strategy  ........................................................................................................................29
1.6.5.  GIF Technology Roadmap3 and R&D Outlook  ............................................................................30
1.6.6.  Other roadmaps ...............................................................................................................................................30

2. NUCLEAR MATERIALS FOR GEN IV SYSTEMS AND RESEARCH

APPROACHES ............................................................................................................................................... 33

2.1.  Material degradation processes and relevant properties .......................................................... 33

2.2.  Nuclear materials, components and systems .....................................................................................34
2.2.1.  Structural materials........................................................................................................................................34

2.2.2.  Fuel materials ....................................................................................................................................................38

2.3.  Materials research approaches .......................................................................................................................40
2.3.1.	 Materials	qualification:	design	rules	and	codes,	fuel	performance	codes ..............41

2.3.2.  Advanced materials modelling and characterisation ............................................................44

2.3.3.  Development of advanced material solutions ............................................................................ 47

2.4.  Data Management .......................................................................................................................................................51

3. NUCLEAR MATERIALS RESEARCH AGENDA .........................................................................55

3.1.  Structural materials ...................................................................................................................................................55
3.1.1.	 Pre-normative	research:	qualification,	test	procedures,	design	rules ........................55

3.1.2.  Advanced structural materials modelling and characterization .................................... 61
3.1.3.  Development of advanced structural materials ....................................................................... 66
3.1.4.  Summary of structural material issues ............................................................................................. 75

3.2.  Fuel materials .................................................................................................................................................................. 75
3.2.1.	 Fuel	materials	qualification	and	design	rules .............................................................................. 75

3.2.2.  Advanced fuel materials modelling and characterization ................................................. 80

3.2.3.  Development of advanced fuels ......................................................................................................... 86

4. CROSS-CUTTING ISSUES ....................................................................................................................93

4.1.  Objectives ...........................................................................................................................................................................93

4.2.  Commonalities with nuclear GenII/III reactor materials .............................................................94

4.3.  Commonalities with materials for (V)HTR ...............................................................................................94

4.4.  Commonalities with fusion materials ..........................................................................................................95

4.5.  Commonalities with materials for other energy technologies ............................................... 96

5. WORKFLOW AND TIMELINE ........................................................................................................... 101

6.

INFRASTRUCTURES FOR NUCLEAR MATERIALS R&D ..................................................107

6.1. 

Irradiation Facilities ..................................................................................................................................................107
6.1.1.  Fast Neutron Facilities ................................................................................................................................107

6.1.2.  Material Testing Reactors and associated ‘hot’ cells ........................................................... 108

6.1.3.  Charged particle irradiation facilities .............................................................................................. 109

6.2.  Handling of irradiated materials .................................................................................................................. 109
‘Hot’ Cells and shielded facilities ...................................................................................................... 109

6.2.1. 

6.2.2.  Transportation of samples for dedicated analysis .................................................................110

6.3.  Computational facilities........................................................................................................................................110

6.4.  General considerations on nuclear infrastructures in Europe ..............................................110
6.4.1.  Costs related to crucial infrastructures build and use ........................................................110

6.4.2.  Renewal of nuclear infrastructures ....................................................................................................111

6.4.3.  Sharing and joint programming of nuclear infrastructures ...............................................111

7. EDUCATION, TRAINING AND MOBILITY OF RESEARCHERS .....................................115

8.

INDUSTRY AND REGULATORS INVOLVEMENT ..................................................................121

9.

INTERNATIONAL COOPERATION ................................................................................................. 125

10. RESOURCES .................................................................................................................................................131

11. RISK ASSESSMENT ................................................................................................................................. 135

12. SUMMARY AND RECOMMENDATIONS .................................................................................... 143

List of abbreviations

ADS 

AFA 

Accelerator-Driven System

Alumina Forming Austenitic (steels)

AI/ML	

Artificial	Intelligence	/	Machine	Learning

AMPEA 

Advanced Materials and Processes for Energy Applications (EERA JP)

ATF 

APT	

BU 

C/M 

CSE	

CSP 

CTR	

CVD	

CVI	

Accident Tolerant Fuel

Atom	Probe	Tomography

Burn-Up

Carbide/Metal

Creep-Strength	Enhanced

Concentrated Solar Power

Carbothermal	Reduction

Chemical	Vapour	Deposition

Chemical	Vapour	Infiltration

DBTT 

Ductile-Brittle Transition Temperature

DoW 

dpa 

DS 

EC 

EELS 

EERA	

EII 

Description of Work

Displacements per atom

Deployment Strategy

European Commission

Electron Energy Loss Spectroscopy

European	Energy	Research	Alliance

European Industrial Initiative

ENEN 

European Nuclear Education Network

EOL 

ESNII 

End Of LIfe

European Sustainable Nuclear Industrial Initiative

ESTEP	

European	Steel	Technology	Platform

EMIRI	

Energy	Materials	Industrial	Research	Initiative

EUMAT	

European	Technology	Platform	for	Advanced	Engineering	Materials	and	Technologies

ETIP	

ETP	

European	Technology	and	Innovation	Platform

European	Technology	Platform

E&T(&M) 

Education and Training (and Mobility)

EU 

European Union

EURATOM  European Atomic Energy Community

FCCI	

FCMI	

Fuel-Clad	Chemical	Interaction

Fuel-Clad	Mechanical	Interaction

FeCrAl	

Alumina	forming	F/M	steels	(from	the	main	composing	elements)

6 

FEM 

FIB 

FGR 

Finite Element Models

Focused Ion Beam

Fission Gas Release

FOAK 

First Of A Kind

FPC 

F/M 

Fuel Performance Codes

Ferritic/Martensitic

GenIV 

Generation IV

GFR 

HPC	

HEA	

HIP	

HLM 

Gas-cooled Fast Reactor

High	Performance	Computing

High	Entropy	Alloys

High	Isostatic	Pressing

Heavy Liquid Metal

HRTEM	

High	Resolution	Transmission	Electron	Microscopy

HT	

HTR	

ICME 

ICT	

IAEA 

IEA 

IPR	

IR 

High	Temperature

High	Temperature	Reactor

Integrated Computational Materials Engineering

Information	and	Communication	Technology

International Atomic Energy Agency

International Energy Agency

Intellectual	Property	Rights

Integrated Roadmap (SET-plan)

JPNM 

Joint Programme on Nuclear Materials

JOG	

Joint	Oxyde-Gaine	(French	for	oxide-clad	joint)

KEMS	

Knudsen	Effusion	Mass	Spectrometry

LBE	

LFR 

LME 

LT 

LTO 

LWR	

MA 

MAX 

MOX 

MoU 

MR 

MSR 

MS 

MTR 

M2F	

NC2I 

Lead-bismuth	Eutectic

Lead-cooled Fast Reactor

Liquid Metal Embrittlement

Low Temperature

Long-Term Operation

Light	Water	Reactors

Minor Actinides

(M- transition metal, A- A group element, X- C or N)

Mixed uranium-plutonium OXide

Memorandum of Understanding

Energy Materials Roadmap (SET-plan)

Molten Salt Reactor

Member State

Material Testing Reactor

Expert	Group	on	Multiscale	Modelling	of	Fuels	of	the	WPMM	(OECD/NEA)

Nuclear Cogeneration Industrial Initiative 

Strategic Research Agenda of the EERA JPNMList of abbreviations

7 

NDE 

NEA 

N/M 

NPP 

Non-Destructive Examination

Nuclear Energy Agency

Nitride/Metal

Nuclear Power Plants

NUGENIA  NUclear GENII/III Association

ODS	

Oxide	Dispersion	Strengthened	(or	Strengthening)

OECD 

Organisation for Economic Co-operation and Development

O/M 

PAS	

PIE 

PLD 

PyC 

ROG	

R&I	

R&D	

SANS 

SCW 

Oxide/Metal

Positron	Annihilation	Spectroscopy

Post Irradiation Examination

Pulsed Laser Deposition

Pyrolitic Carbon

Réaction	Oxyde-Gaine	(French	for	oxide-clad	reaction)

Research	and	Innovation

Research	and	Development

Small-Angle Neutron Scattering

SuperCritical Water

SCWR 

SuperCritical Water Reactor

SEM 

Scanning Electron Microscopy

SETIS 

SET-plan Information System

SET-plan	 Strategic	Energy	Technology	plan

SFR 

SL 

Sodium-cooled Fast Reactor

Surface Layer

SNETP	

Sustainable	Nuclear	Energy	Technology	Platform

SPS 

SRA	

SRIA	

TEM 

TMT	

TRL	

TSO	

Spark plasma sintering

Strategic	Research	Agenda

Strategic	Research	and	Innovation	Agenda

Transmission Electron Microscopy

ThermoMechanical	Treatment

Technology	Readiness	Level

Technical	and	Scientific	support	Organizations

(V)HTR	

(Very)	High	Temperature	Reactor

WPMM  Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear 

Systems (OECD/NEA)

XRD	

2DS 

X	Ray	Diffraction

2 Degree Scenario

8 

PREFACE

Nuclear	energy	in	the	European	Union	and	its	associate	countries	is	expe-
riencing	a	difficult	transient.	However	it	is	obvious	that,	worldwide,	nuclear	
energy	will	contribute	substantially	to	the	low-carbon	energy	mix	that	is	
expected	(and	hoped)	to	limit	global	warming	in	the	coming	decades.	In	
contrast,	in	the	EU	only	three	member	states	are	currently	building	new	
nuclear power plants: France (1), Finland (1), and Slovakia (2); and only a few 
more are planning or considering new builds, mainly in Central or Eastern 
Europe,	as	well	as	the	UK.	So,	in	the	period	to	2030	more	nuclear	capac-
ity	will	be	lost,	due	to	the	closure	of	old	reactors,	than	gained	from	new	
ones.	Accordingly,	the	Nuclear	Illustrative	Program	(PINC)	issued	in	2017	
by	Euratom	forecast	a	decline	in	EU	nuclear	capacity	to	2025	followed	by	
(hopefully)	a	levelling	out	to	2050.	Global	energy	in	general,	and	nuclear	
energy	in	particular,	require	a	strategic	and	technical	long-term	vision	for	
a	safe	and	decarbonized	energy	supply,	which	seems	to	have	been	lost	
these	last	years	due	to	political	and	populistic	debates.

The	rapid	and	welcome	development	of	renewable	energy	sources	poses	
new	challenges	for	the	global	energy	system	and	particularly	for	the	
security	of	supply.	The	need	of	flexible	power	capacity	to	complement	
fluctuating	sources	is	one	of	these	major	challenges.	Nuclear	is	able	to	
regulate	fluxes	and	is	one	of	the	few	available	solutions,	as	well	as	fossil	
fuels	during	the	transition	period,	but	nuclear	currently	lacks	market	sig-
nals	to	make	its	production	economically	attractive,	specifically	suffering	
from	adverse	investment	environment.	Here,	not	only	economic,	but	also	
technological	issues	are	at	stake,	largely	connected	with	the	capability	of	
materials	to	withstand	operational	conditions	for	which	the	existing	plants	
were	not	conceived.	Facing	this	problem	requires,	among	other	political	
and	economic	measures,	innovation	in	the	field	of	materials	for	power	
plants,	in	terms	of	wider	performance	and	lower	cost.

Nuclear	energy	in	particular	requires	innovation	in	a	broader	sense.	While	
the	level	of	safety	of	nuclear	installations	in	Europe	is	probably	one	of	the	
highest	in	the	world,	it	can	be	further	improved	with	better	materials.	In	
addition,	there	is	a	need	for	better	flexibility	and	adaptability	in	a	wider	
spectrum	of	applications,	and	to	pursue	the	ultimate	goal	of	sustainability	
by	improving	efficiency,	reducing	waste	volumes	and	hazard	and,	through	

Strategic Research Agenda of the EERA JPNMPreface

9 

fuel	breeding	and	recycling,	making	available	resources	sufficient	to	fuel	
power	plants	for	the	next	centuries	and,	why	not,	millennia.

This	Strategic	Research	Agenda,	elaborated	by	the	EERA	Joint	Programme	
on	Nuclear	Materials	largely	during	my	mandate	as	chairman	of	the	Executive	
Committee	of	EERA,	is	a	small,	but	still	important,	contribution	to	solve	the	
above	challenges.	It	is	therefore	with	satisfaction	that	I	commend	the	work	
done	by	the	management	board	of	this	joint	programme	to	let	this	docu-
ment	see	the	light	and	recommend	its	reading	and	adoption	by	the	involved	
policy	and	decision	makers,	research	managers,	and	researchers	at	large.

Hervé BERNARD

CEA, former EERA Chair

10 

EXECUTIVE 
SUMMARY

With	a	yearly	production	in	excess	of	850	TWhe,	nuclear	energy	is	the	sin-
gle	largest	source	of	low-carbon	electricity	in	the	EU.	Consistent	with	this,	
the	2	degree	IEA	scenario	concludes	that	the	worldwide	nuclear	energy	
production	should	increase	from	11%	in	2015	to	16%/15%	in	2050/2060.	
Thus,	nuclear energy plays an important societal role, together with 
renewables, in the energy transition from	fossil	fuels.	Yet,	three	main	
open issues remain: (1) sustainable and responsible use of resources; (2) 
accident	risk;	and	(3)	long-lived	nuclear	waste.

The sustainability of nuclear energy will be ensured by deploying 
Generation IV (GenIV) systems.	These	can	(i) produce	more	fuel	than	
they	consume,	guaranteeing	low-carbon	energy	production	for	centuries	
through	recycling,	without	additional	mining,	in	a	circular	economy;	(ii)	offer	
~50%	higher	thermal	efficiency	and	increased	standards	of	passive	safety	
than	current	reactors,	thereby	becoming	both	societally	and	economically	
attractive;	and	(iii)	reduce	significantly	the	volume	and	radiotoxicity	(decay	
time	<	1000	years)	of	nuclear	waste.	However,	materials	will	be	exposed	to	
high	levels	of	temperature	and	irradiation,	with	some	in	contact	with	po-
tentially	aggressive	non-aqueous	coolants,	targeting	a	60	year	operation	
reactor	design.	Thus, the development, screening and qualification of 
suitably performing and affordable materials are crucial to make GenIV 
reactors	an	industrial	and	commercial	reality,	with	positive	impacts	on	
economy,	safety,	waste,	and	thus	sustainability	of	nuclear	energy.

This Strategic Research Agenda (SRA) has	been	prepared	by	the	EERA-
JPNM, based on a wide consultation with the scientific and industrial 
community	involved,	to	identify	the research lines to	be	pursued	in	the	
EU	to	ensure	that suitable structural and fuel materials are available 
for the design, licensing, construction and safe long-term operation of 
GenIV nuclear systems.	Emphasis	has	been	put	on	the	fast	neutron	spec-
trum	systems	considered	in	the	European	Sustainable	Nuclear	Industrial	
Initiative	(ESNII),	namely	sodium-cooled	fast	reactor	(SFR),	heavy	liquid	
metal (HLM) cooled systems (accelerator driven systems –ADS- and 

Strategic Research Agenda of the EERA JPNM11 

lead-cooled	fast	reactor	–LFR-),	and	gas-cooled	fast	reactor	(GFR).	However,	
links	with	other	GenIV	systems,	namely	(very)	high	temperature	reactor	((V)
HTR),	supercritical	water	reactor	(SCWR),	and	molten	salt	reactor	(MSR)	have	
been	established.	Effort	has	been	devoted	to	identify	GenIV materials’ re-
search issues that are of interest for other nuclear and also non-nuclear 
energy technologies,	as	well.	This	was	done	with	a	view	to	optimise	the	use	
of	available	facilities,	knowledge	and	human/financial	resources,	whenever	
possible,	by	joining	forces	with	other	research	communities.	Importantly,	the	
content of this SRA is consistent with relevant strategic documents and 
roadmaps that have been compiled by linked organisations and platforms 
(SETPlan	integrated	roadmap,	OECD/NEA	technology	roadmap,	SNETP	
SRIA,	…),	of	which	it	is	an updated extension for what concerns materials 
for sustainable nuclear.

Three Grand Challenges	have	been	identified,	namely:	(i)	Elaborate	design	
correlations,	assessment	and	test	procedures	for	the	structural	and	fuel	
materials	that	have	been	selected	for	the	demonstrators	under	the	service	
conditions	expected;	(ii)	Develop	physical	models	coupled	to	advanced	
microstructural	characterization	to	achieve	high-level	understanding	and	
predictive capability; (iii) Develop innovative materials solutions and fab-
rication	processes	of	industrial	application	to	achieve	superior	materials	
properties,	to	increase	safety	and	improve	efficiency	and	economy.	

Based	on	the	current	status	of	the	research	in	the	field	of	nuclear	material	
science	and	on	the	needs	of	the	industrial	stake-holders,	and	consistent	with	
the	Grand	Challenges,	this	document	identifies	the	categories of materials 
of interest,	the	research approaches, tools and instruments, and espe-
cially	the	specific scientific and technical gaps and objectives, to enable 
the	design, licensing and construction of GenIV system demonstrators, 
as	well	as	to	open	the	way	to	their	longer term commercial deployment.	
The	timeframe	is	defined	by	the	actual	plans	of	construction	of	the	dem-
onstrators,	which	depend	largely	on	political	decisions	and	support,	as	well	
as	on	the	economic	contingency.	The	document	also	addresses	the	neces-
sary	corollaries	to	the	proposed	research	activities,	namely:	infrastructure	
needs,	education	and	training,	mobility	schemes,	industry	and	regulators	
involvement,	importance	of	international	cooperation,	...

The	research	activities	are	organized	in	blocks	that	result	from	the	appli-
cation, for structural and fuel materials, of a well-established materials 
science approach,	which	is	based	on	the	combination	of	three	classes	
of activities: (1) materials testing for full qualification in environment and 
definition	of	design	rules	in	a	pre-normative	spirit;	(2)	development	of	mech-
anistic and physical models	in	support	of	materials	behaviour	correlations	
used	in	design	rules	and	improvement	of	materials	properties;	and	(3)	devel-
opment of advanced materials solutions	through	experimental	screening	
of	solutions,	assisted	by	models	that	are	rooted	in	understanding	of	the	
physical	processes	that	govern	materials	behaviour.	Considerations	of	the	

Executive Summary12 

circular	economy	and	advanced	manufacturing	methods	based	on	au-
tomation of procedures, as well as materials monitoring,	also	enter	this	
third	class	of	activities.	In	all	cases,	data management is a crucial issue to 
ensure	data	collection,	storage,	interoperability,	and	re-usability.

The	materials of interest for all demonstrators,	on	which	the	qualifica-
tion	and	pre-normative	research	effort	is	mainly focused, are austenitic 
stainless steels for structural functions (Ni-based alloys for out-of-core 
applications; in some cases ferritic/martensitic –F/M- steels) and MOX 
as	fuel,	with	ceramics (mainly SiC/SiC) and UO2	as	fuel	for	the	high	tem-
perature	operation	of	e.g.	the	GFR	demonstrator.	Surface protection may 
be	necessary	already	for	demonstrators	to	provide	sufficient	compatibil-
ity	with	coolants.	Improvements	of	safety,	performance	and	economy	in	
future	prototypes	and	then	commercial	reactors	advises	the	exploration	
of improved austenitic steels, advanced F/M steels, refractory alloys 
(V and Mo), oxide dispersion strengthening, advanced surface protec-
tion methods	and,	for	the	farther	future,	prospective	materials	such	as	
high entropy alloys or MAX phases.	Reduction	of	waste	and	increased	
safety	requires	the	development	of	minor actinide bearing fuel from 
multirecycling,	as	well	as	advanced	fuel	forms	such	as	uranium nitrides 
or carbides.

For structural materials,	the	requirement	of	60 years design lifetime 
for	non-replaceable	components	is	in	perspective	the	most	demanding	
requirement,	which	includes	under	its	umbrella	several	issues	related	
with	the	reasonable	prediction	of	long-term	degradation	processes:	high 
temperature processes	(creep,	fatigue,	thermal	ageing),	compatibility 
with –especially- heavy liquid metal and helium coolants,	and	effects	of	
low flux prolonged irradiation,	with	emphasis	on	welded	components	in	
all	cases.	In	terms	of	testing,	there	is	a	need	for	standardization,	especially	
for	sub-size	and	miniature	specimens.	The	modelling,	supported	by	micro-
structural	characterization,	has	as	its	main	objective	the	development	of	
suitable microstructure evolution models to be used as input to models 
for the mechanical behaviour	under	irradiation	and	at	high	temperature,	
eventually	linking	with	fracture mechanics.	Specific	developments	are	
required for coolant compatibility models, as well as for models in sup-
port of the use of charged particle irradiation	for	the	screening	of	new	
materials	solutions,	such	as	those	listed	above.	

Concerning fuel materials,	the	properties	and	processes	that	govern	its	
behaviour	in	pile,	on	which	research	effort	is	focused,	are:	margin to melt-
ing	(establishment	of	phase	diagrams	and	evolution	of	thermal	properties),	
atomic transport properties and ensuing microstructural evolution, fis-
sion product (non-gaseous) and helium	(gas)	behaviour	and	transport,	
mechanical properties	(their	evolution,	subsequent	fragmentation	and	
cracking,	fuel-cladding	mechanical	interaction),	and	compatibility with 
cladding and coolant	(internal	cladding	corrosion,	chemical	interactions	

Strategic Research Agenda of the EERA JPNM13 

especially	in	case	of	severe	accident).	These	are	all	addressed	from	both	
an	experimental	and	a	modelling	perspective.	

Several materials issues	identified	for	GenIV	demonstrators	are	common 
to other nuclear fission technologies, namely GenII/III reactors and (V)
HTR,	especially	concerning	the	integrity	of	structural	materials	(qualification	
of	welds,	procedures	for	small	specimens,	modelling	and	use	of	charged	
particle	irradiation),	fuel	(oxides,	innovative	fuels	and	synthesis	routes,	fuel	
performance codes, modelling, materials for accident tolerant fuel), inno-
vative	materials	solutions	and	manufacturing.	Most	materials	and	issues	
related	to	high	temperature	operation	are	common	between	GFR	and	(V)
HTR.

Important commonalities exist also with fusion materials, namely all ac-
tivities	related	with	F/M	steels	(design	rules,	low	temperature	radiation	
embrittlement,	high	temperature	behaviour,	welding,	compatibility	with	
heavy	liquid	metals,	advanced	F/M	steels),	materials	screening	methods,	
ceramic	coatings,	mechanistic	and	physical	modelling	of	irradiation	effects	
and	compatibility	with	HLM.

More	in	general,	the	development	and	qualification	of	materials	resistant	
to	high	temperature	and	aggressive	environment,	as	well	as	an	approach	
based	on	testing	and	characterization,	modelling	and	development	of	new	
materials solutions is common to several other energy technologies,	such	
as	concentrated	solar	power,	geothermal,	bioenergy,	fuel	cells	and	hydro-
gen.

Several	classes	of	facilities	and	infrastructures	are	necessary	for	the	qualifi-
cation	of	nuclear	materials	in	general,	and	specifically	for	GenIV	systems.	The	
specificity	of	nuclear	energy	makes	irradiation facilities essential for the 
qualification of nuclear materials.	Ideally	GenIV	materials	should	be	qual-
ified	by	exposure	to	fast	neutrons	up	to	high	dose,	however	at	the	moment	
Europe is totally dependent on non-European countries to have access 
to fast neutron facilities.	Working	materials testing reactors	(MTR)	with	
dominantly	thermal	spectrum,	equipped	with	relevant	hot	cell	facilities,	exist	
in Europe,	but	there	are	not many	of	them.	Moreover,	hot cells and shielded 
facilities	where	active	materials	can	be	handled	and	tested	are	costly	in-
frastructures	to	maintain	and,	for	this	reason,	are	also	limited in number 
and often becoming obsolescent,	despite	the	construction	of	new	facilities	
in	some	countries:	this	limits	severely	the	number	of	tests,	measurements	
and	examinations	that	can	be	performed	on	neutron	irradiated	materials.	
Transport of active materials, especially fuel, is also becoming problematic, 
largely	because	of	the	lack of harmonisation of regulations	throughout	
Europe.	For	all	these	reasons,	charged particle irradiation facilities become 
crucial	to	provide	data	that,	although	unsuitable	for	qualification,	can	be	
exploited for screening and modelling purposes.	For	the	latter,	access	to	
suitable computational resources	becomes	also	important,	though	there	

Executive Summary14 

is currently no dedicated resource for fission	material	studies.	Overall,	
a number of actions are needed	in	order	to	make	sure	that	sufficient	
facilities and infrastructures are available in Europe for nuclear materials 
research.	Amongst	them,	a	better	coordinated use and development of 
infrastructures	throughout	Europe	is	desirable,	based	on	a	principle	of	
complementarity,	rather	than	competition,	including	common rules for 
the access to facilities and infrastructures.

Strongly	linked	to	use	and	sharing	of	infrastructures	and	facilities	is	the	
problem	of	education	and	training	to	their	use	for	nuclear	materials	re-
search.	A suitable and inherently attractive education and training (E&T) 
programme is	needed	to	reduce	the	risk	of	a	future	shortage	of	nuclear	
skills	and	ensure	the	maintenance	of	the	acquired	knowledge	and	ex-
pertise.	Nuclear materials can be a way to attract young researchers to 
the nuclear field,	thanks	to	the	inherent	cross-cutting	nature	of	materials	
science	through	several	technologies.	This	fact	should	be	better	exploited.

Finally, a close link of	nuclear	materials	research to the industrial ap-
plication is essential.	In	particular	the	goals	and	needs	of	the	reactor 
designers	must	be	very	clear	in	order	to	support	the	processes	of	licensing	
and	construction	of	advanced	nuclear	systems,	for	which	the	amount	of	
experience	is	limited.	The	involvement	of	technical and scientific support 
organizations (TSOs) and regulators,	to	follow	the	procedures	used	for	
materials	qualification	and	to	possibly	guide	them,	is	also	desirable,	though	
difficult,	and	is	likely	to	further	accelerate	the	licensability	of	nuclear	com-
ponents.	The	connection	with	materials’ manufacturers is essential as 
well	in	the	process	of	development	of	new	materials’	solutions	in	view	
of	industrial	upscaling,	especially	when	innovative	fabrication	routes	are	
explored.	For	these	reasons,	explicit	efforts	are	made	on	the	EERA	JPNM	
side	to	facilitate	industrial	involvement,	e.g.	through	participation	in	task 
forces.

International cooperation is currently not sufficiently encouraged	by	the	
means	that	are	offered.	However,	international	cooperation	is	essential for 
the	optimized/harmonized	use	of	infrastructures,	harmonization	of	testing	
procedures	and	methodologies	for	interlaboratory	data	consistency,	data	
collection	and	sharing	and	also	synergy	on	modelling.	The	role	of	inter-
national organisations	(IAEA,	NEA,	…)	to	facilitate	this	is	key,	in	particular	
to	enable	better	connection	with	the	GIF.

Besides	the	obvious	need	of	adequate	financial	resources	in	order	to	
address	the	research	problems	outlined	in	this	SRA,	as	well	as	the	nec-
essary corollaries, four recommendations	emerge	that	this	document	
is	intended	to	bring	to	the	attention	of	stake-holders,	particularly	deci-
sion-makers:

Strategic Research Agenda of the EERA JPNM15 

R1: Data from materials property measurements after exposure to relevant 
conditions	are	the	essential	ingredient	for	robust	design	curves	and	rules.	
Plenty	of	data	were	produced	in	the	past	that	are	now	de	facto	unusable;	this	
is	either	because	they	are	covered	by	confidentiality	or	because	they	were	
not	properly	archived.	Correct	data	management	to	guarantee	availability	
for	future	re-assessment	is	therefore	essential	and	should	be	encouraged	
and	fostered.	In	particular,	financially supported policies to foster data 
sharing and encourage old data disclosure should	be	implemented.

R2:	Some	infrastructures	are	absolutely	essential	to	enable	the	correct	qual-
ification	of	nuclear	materials,	not	only	irradiation	facilities,	but	also	suitable	
‘hot’	cells	where	active	materials	can	be	safely	handled	and	tested,	nu-
clearized	characterization	techniques,	loops	and	pools	for	compatibility	
experiments,	etc.	They	are	also	crucial	for	education	and	training	of	young	
researchers	and	operators.	These	infrastructures	are	costly	to	build	and	
maintain.	Other	research	facilities	are,	on	the	other	hand,	more	common	
and	 sometimes	 redundant.	A	rational	and	 harmonised, pan-European 
management of infrastructures, based on joint programming, including 
trans-national infrastructure renewal planning and a scheme for facility 
sharing and exploitation,	would	be	highly	desirable	and,	at	the	end	of	the	
day,	beneficial	for	all.	

R3:	International	cooperation	with	non-EU	countries	where	research	on	nu-
clear	materials	is	pursued	can	be	very	valuable	for	Europe.	Quite	clearly,	the	
goals	of	this	cooperation	are	in	the	end	the	same	as	in	the	case	of	internal	
European	cooperation,	namely	coordination	of	activities,	sharing	of	data,	
and	access	to	infrastructures.	Currently,	however,	the instruments availa-
ble in Europe for international cooperation are not sufficiently attractive 
to	motivate	significant	cooperation	with	non-EU	researchers.	Efforts	should	
be	made	to	improve	their	attractiveness	and	ease	of	access.	International	
organisations	such	as	OECD.NEA,	IAEA,	but	also	Euratom	and	JRC	for	the	
connection	with	GIF,	have	here	a	crucial	role.

R4:	The	nuclear materials research community in Europe is currently 
strongly integrated	and	engaged	in	thriving	collaboration,	in	a	bottom-up	
sense.	This	is	in	contrast	with	the	inadequacy	of	the	top-down	instruments	
offered	to	make	this	integration	efficient	and	functional.	This	SRA is largely 
the	result of matching bottom-up research proposals with top-down 
strategies.	The	appropriate	instrument	to	allow	this	community	to	deliver	
according	to	the	SRA	goals	should	provide	the	conditions	to	implement	
the	agreed	research	agenda	and	to	set	up	suitable	E&T&M	schemes	that	
allow	knowledge,	data,	and	facility	sharing.	Since	the	financial	support	of	
Euratom	will	never	be	sufficient,	earmarked	funding	from	the	MS	dedicated	
to	support	integrated	research	on	nuclear	materials	is	crucial.	In	this	sense,	
a co-fund instrument, such as a European Joint Programme, seems to 
be most suitable.

Executive Summary16 

These recommendations are clearly based on the willingness to pursue 
a policy of increased integration rather than of isolation at all levels: 
research organisations, EU Member States, and European Commission.	
Given	the	amount	of	resources	that	can	be	reasonably	allocated	to	cover	
a	need	that	has	been	estimated	to	range	-depending	on	the	ambition	of	
the	goals-	between	15	and	50	M€/yr,	this	requires	the	finding	of	a	difficult	
equilibrium	between	the	need	to	make	the	best	use	possible	of	the	limited	
resources	available,	in	a	framework	of	nuclear	energy	where	support	is	
politically	difficult	to	obtain,	and	the	legitimate	ambition,	in	a	context	of	
healthy	competition,	to	preserve	each	stakeholders	assets.

Strategic Research Agenda of the EERA JPNM1

INTRODUCTION

© SCK-CEN

1.  INTRODUCTION

1. 1.  Societal, economic and 
technical challenges

With a yearly production in excess of 850 TWhe, 
nuclear energy is the single largest source of 
low-carbon electricity in the European Union.1 
This  corresponds  to  more  than  ¼  of  the 
electricity in Europe, and contributes to guaran-
teeing a secure and reliable base-load supply. 
Consistent with this, in the 2 degree scenario 
(2DS) of the International Energy Agency (IEA), 
it is considered that worldwide nuclear energy 
production should increase from 11% in 2015 to 
16%/15% in 2050/2060.2 Thus, nuclear energy 
plays an important societal role, together with 
renewables, in the energy transition. However, 
three main open issues remain: (1) sustainable 
and responsible use of resources; (2) accident 
risk; and (3) long-lived nuclear waste.

1 

In 2014, 876 TWhe were produced by nuclear. This, com-
bined with 406 TWhe of hydroelectricity covering especially 
peaks, corresponds to about 40% of the total electricity and 
represents by far the largest portion of low-carbon electricity 
in Europe. Wind comes third, with 253 TWhe (source: Eurostat  
- http://ec.europa.eu/eurostat/statistics-explained/index.php/ 
Electricity_and_heat_statistics).  However, wind  keeps  grow-
ing and produced, in 2016, almost 300 TWhe (source: Wind-
europe - https://windeurope.org/about-wind/statistics/euro-
pean/wind-in-power-2016/).

2  Source: Energy Technology Perspectives, IEA, 2016/2017.

Current nuclear systems are insufficiently sus-
tainable: just less than 1% of the fuel’s energy 
content is used in present day nuclear power 
plants  (NPPs),  more  than  90%  of which  are 
light water reactors (LWR). Sustainability can 
be greatly enhanced by deploying fourth gen-
eration (GenIV)3 fast neutron reactors, along 
with the facilities that are needed to close the 
fuel cycle. The combination of fast reactor and 
closed fuel cycle allows the energy extracted 
from the available uranium resources to be 
maximised. Fuels irradiated by fast neutrons 
generate as much Pu from the 238U by neutron 
capture as the 235U is consumed by fission. The 
reactor cores can be thus optimized, pushing 
the burnup to high values, i.e. letting the fuel 
remain for longer in the reactor, to produce 
more Pu than they consume (breeder reac-
tors). This can be then extracted and used for 
refuelling, in a circular economy logic.

Increasing the thermal efficiency is another key 
factor to improve sustainability in terms of use 
of resources. The thermal efficiency is the ratio 
between electricity and heat produced and its 
increase means not only a larger amount of 

3  Technology  Roadmap  of  the  Generation  IV  International  
Forum: https://www.gen-4.org/gif/upload/docs/application/ 
pdf/2014-03/gif-tru2014.pdf

20 

electricity produced for a given thermal power, 
but also less waste heat, less environmental 
impact and less need for cooling. Thermal ef-
ficiency depends on the temperature of the 
reactor core and on the performance of the 
conversion system. Present LWRs have ther-
mal efficiencies around 33%, while modern 
coal-fired plants reach approximately 39% and 
combined-cycle gas plants are even 50 to 60% 
efficient. The use of non-aqueous coolants, 
mandatory in fast neutron reactors, will allow 
operation at temperatures well-above those of 
LWRs, thereby pushing the efficiency close to 
(or higher than) coal plants and, in the case of 
very high temperature, combined-cycle sys-
tems. With liquid metals as coolants this can be 
achieved close to atmospheric pressure, with 
a design that is based on an increased use of 
passive safety systems. This provides design-
ers with an important tool to make fast breeder 
reactors both safe and economically attractive. 

Fast reactors offer also an additional virtue, 
they have the ability to transmute the minor ac-
tinides (the elements that present a long term 
source of radiotoxicity and heat) into short lived 

GenIV systems create more 
fuel than they burn and 
operate at high temperature 
using passive safety 
systems, with the possibility 
of burning waste: this 
increases substantially the 
sustainable use of resources 
and guarantees safe energy 
production for centuries, 
significantly reducing waste 
production.

fission products with lower radiotoxicity. In this 
way, provided that high burnups are reached, 
the decay time of the waste can be reduced 
by a factor 1000 to time scales that are below 
1000 years,4 which significantly reduces the 
required capacities of geological repositories 
through an up to ten times reduced volume, 
much shorter-term hazard and less heat from 
the waste. In addition, the adoption of a closed 
fuel cycle requires only a short-term storage 
of the irradiated fuels before their reprocessing 
and reuse.

Thus, GenIV systems composed by fast reac-
tors and close fuel cycle facilities can create 
more usable fuel than they burn and operate 
at high temperature using passive safety sys-
tems, increasing substantially the efficiency 
in the use of resources and guaranteeing safe 
energy production for several centuries, with 
significantly  reduced  production  of  high-
level, long-lived waste. 

In order to reach the above goals, however, 
materials in GenIV systems will be exposed 
to higher temperatures and higher irradiation 
levels than in today’s LWRs. Moreover, in sev-
eral cases the compatibility of materials with 
non-aqueous coolants needs to be demon-
strated. The operating conditions for the fuel 
pins are further complicated by possible large 
temperature gradients and the internal pres-
ence of fission products. All these factors lead 
to substantial changes of materials’ properties, 
which reduce their performance. At the same 
time, the overall cost of these systems must 
be on a par with other low-carbon energy sys-
tems, including current LWRs. Since the capital 
cost of the construction represents the larg-
est part of the investment in the case of NPPs, 
to be economically viable GenIV reactors are 
expected to be designed and licensed for a 
60 year lifetime. This is obviously quite a chal-
lenge, given in particular the harsh conditions 
to which both structural and fuel materials are 
going to be exposed. Thus, the performance 

4  J. Magill, V. Berthou, D. Haas, J. Galy, R. Schenkel, H.-W. 
Wiese, G. Heusener, J. Tommasi, G. Youinou, “Impact limits opf 
partitioning and transmutation scenarios on the radiotoxicity of 
actinides in radioactive waste”, Nuclear Energy 42(2003)263 

Strategic Research Agenda of the EERA JPNM21 

of nuclear materials is an essential point to 
make GenIV reactors a reality. Furthermore, 
the development of materials with superior re-
sistance to high temperature and exposure to 
aggressive coolants can also be beneficial for 
other energy technologies, such as concen-
tratred solar power (CSP), geothermal energy, 
bioenergy and fuel cells and hydrogen.

1. 2.  The EERA JPNM in the 
European platform 
landscape

The EERA JPNM provides 
the R&D for materials 
that is needed for the 
development and 
implementation of GenIV 
systems in Europe

The Joint Programme on Nuclear Materials, 
JPNM  (www.eera-jpnm.eu),  was  launched 
in  2010  and  is  one  of  the  16  current  joint 
programmes  (JPs)  of  the  European  Energy 
Research Alliance, EERA (www.eera-set.eu). 
Altogether, these JPs cover all the low-carbon 
energy technologies and systems. EERA, cre-
ated in 2008, supports the European Strategic 
Energy Technology (SET) Plan of the European 
Commission (EC), which was launched in 2007. 
It does so by coordinating the work of almost 
250 (in 2018) public research organisations, 
towards the development and deployment of 
cost-effective low carbon technologies. The 
goal is to meet the sustainability targets set 
by Europe for 2020 and 2050, to counteract cli-
mate change and guarantee security of energy 
supply and competitiveness. One important 
defining feature of EERA is the focus on rela-
tively low technology readiness levels (TRL5<5). 
EERA deals mainly with research towards inno-
vation, while industrial implementation (TRL>5) 
is characteristic of the technology platforms 
and the industrial initiatives.

nuclear energy technology platform (SNETP7) 
at the same time as the EERA JPNM. The EERA 
JPNM provides the R&D for the structural 
and fuel materials needed for the develop-
ment and implementation of fast reactors in 
Europe, as defined by ESNII. Currently, this is 
the main reason of existence of the JPNM in 
EERA. Figure 1 illustrates the connections be-
tween energy platforms, in particular nuclear 
energy platforms, in Europe.

However, the scope and goals of the JPNM 
go beyond this. By operating mainly at TRL<5, 
the EERA JPNM deals mainly with fundamen-
tal research, albeit projected towards specific 
technological applications, bridging with the 
industrial initiatives via, mainly, pre-normative 
research. The SNETP recognises the impor-
tance  of  basic  technology  developments, 
because, as stated in its Deployment Strategy 
(DS) of 20158 they “open routes for the iden-
tification of common trunks for Gen II, III, IV 
and cogeneration application, notably in areas 
such as: 

The  European  Sustainable  Nuclear  Energy 
Industrial Initiative (ESNII6), which has the task 
of  developing  GenIV  fast  neutron  reactors 
in Europe using different technologies, was 
launched under the umbrella of the sustainable 

 y Material behaviour for structural compo-

nents and fuel

 y Structural integrity of systems and compo-

nents

 y Manufacturing & assembly technology”

5  https://ec.europa.eu/research/participants/data/ref/
h2020/wp/2014_2015/annexes/h2020-wp1415-annex-g-
trl_en.pdf

6  http://www.snetp.eu/esnii/

7  http://www.snetp.eu/

8  http://www.snetp.eu/wp-content/uploads/2016/01/
SNETP-DEPLOYMENT-STRATEGY-2015-WEB.pdf

1. Introduction22 

figure 1: Illustration of the connection between (nuclear) energy platforms in Europe.

Thus, research on structural and fuel mate-
rials’ behaviour belongs to one of the areas 
where commonalities across nuclear reac-
tor generations and types can be actually 
found. For this reason the SNETP explicitly 
mentions9 that “the interface with EERA/JPNM 
should be reinforced for the development of 
new and innovative materials”. The memo-
randum of understanding (MoU) signed in 
December 2016 between EERA JPNM and 
SNETP cements this intention, extending the 
collaboration to enhance synergy not only 
with ESNII, but also with the other SNETP pil-
lars, namely the NUclear GENII/III Association 
(NUGENIA)9 and the Nuclear Cogeneration 
Industrial Initiative (NC2I)10. Moreover, several 
issues faced by materials for fission reactors 
are common to fusion reactors and systems, 
as well: hence it is possible to find cross-cut-
ting topics with this other, longer-term form 
of nuclear energy.

9  http://www.nugenia.org/

10  http://www.snetp.eu/nc2i/

It  is  also  clear  that  materials with  superior 
properties in terms of high temperature and 
corrosion  resistance  may  find  their  way  to 
other energy technologies. In particular, within 
EERA  the JPNM  finds  natural  grounds  for 

Research on structural and 
fuel materials’ behaviour 
belongs to one of the areas 
where commonalities 
through nuclear reactor 
generations and types, as 
well as with other energy 
technologies, can be 
actually found

Strategic Research Agenda of the EERA JPNMcollaboration with other joint programmes, 
targeting high temperature operation in envi-
ronmentally harsh environments, for example 
AMPEA (Advanced Materials and Processes for 
Energy Application), CSP, geothermal energy, 
bioenergy and fuel cells and hydrogen.

1. 3.  Generation IV systems  

in Europe

In the vision of the SNETP the European nu-
clear  industry  can  continue  to  deliver  safe 
low-carbon  nuclear  energy  for  the  present 
and the coming centuries, with a commitment 
towards even higher safety standards and sus-
tainability, in two phases: 

 y Safe extended operation of existing GenII/
III nuclear power plants or long-term oper-
ation (LTO), including new builds; 

 y Parallel deployment of GenIV fission reac-
tors and systems, guaranteeing more sus-
tainable and safe nuclear energy, with the 
potential for nuclear heat generation.11

GenIV reactors are expected to be commer-
cially deployed around, or after, the middle 
of this century, depending on industrial com-
mitment and political support, in particular 
to close the fuel cycle with fast reactors or 
to alternative uses of nuclear energy.12 At the 
moment, four GenIV fast reactor systems, at 
different maturity levels, are being studied in 
Europe within ESNII, namely: sodium-cooled 
fast  reactor  (SFR),  lead-cooled  fast  reactor 
(LFR), gas-cooled fast reactor (GFR), and ac-
celerator-driven system (ADS).

The  SFR  is  the  most  mature  technology 
and LFR is considered the next technology 
in terms of TRL, with advantages in terms of 
passive safety features, as well as potential 

11 

In the long run, the gradual insertion of fusion systems in 
the  energy  production  market,  in  cohabitation  with  fission 
systems, may also occur.

23 

Gen IV fast reactor systems 
considered by industrial 
initiatives in Europe 
include sodium, lead and 
gas cooled, as well as 
accelerator-driven system, 
technologies.  

High temperature thermal 
reactors are also the focus 
of industrial initiatives. 

Other GenIV systems not 
linked to industrial initiatives 
are the supercritical water 
and molten-salt reactors.

modularity, while GFR is a longer term alter-
native that would open the way to even higher 
temperatures and therefore efficiency. In ESNII, 
ASTRID, ALFRED and ALLEGRO are demon-
strators for SFR, LFR and GFR technologies, 
respectively.13 In addition, MYRRHA is a flex-
ible research facility for material testing and 
demonstration of ADS for waste minimization, 
with features strongly related to LFR technol-
ogy. These systems are all in the focus of the 
EERA JPNM research. 

NC2I addresses the design of high tempera-
ture reactors for the supply of heat to industry 
(cogeneration), in order to demonstrate the fea-
sibility and advantage of the coupling between 

12  For  an  analysis  of  the  prospects  fof  GenIV  systems,  in-
cluding  technical  descriptions  and  economics  considera-
tions,  see  Locatelli,  Mancini  and  Todeschini,  Energy  Policy 
61 (2013) 1503.

13  At the time of preparation of the current document the AS-
TRID  project  is  experiencing  profound  revision  and  may  be 
cancelled  as  known  now,  however  it  keeps  being  used  as 
reference SFR demonstrator throughout this document.

1. Introduction 
 
24 

a nuclear reactor and a non-nuclear plant, in a 
framework of CO2 emission reduction in heat 
production.

should generally talk of GenIV systems, and 
not simply GenIV (fast) reactors.

Despite being pushed forward by industrial 
initiatives, the time to construction of these 
European demonstrators is currently difficult to 
assess. It will depend substantially on industrial 
strategy evolution and governmental financial 
support, including in some cases the possibility 
of accessing European structural funds. 

GenIV includes other reactor concepts, with 
specific advantages, namely: the supercrit-
ical water reactor (SCWR) as an advanced 
upgrade of existing LWRs, and so potentially 
easier  to  demonstrate  in  terms  of  technol-
ogy, and the molten salt reactor (MSR) as a 
potentially proliferation-resistant system (in 
the  thorium  cycle) with  advantages  also  in 
terms of fuel recycling and safety. These sys-
tems are not included in the ESNII portfolio, 
nor are there established industrial initiatives 
pushing  them  forward;14 yet work  on  them 
supported  by  national  programmes  is  un-
derway in Europe . The SCWR is included in 
the NUGENIA portfolio as advanced LWR, 
and  the  MSR  is  increasingly  mentioned  as 
a  long  term  interesting  option. SCWR  and 
MSR,  like  the  HTR, are not in the focus of 
the JPNM, but may be addressed in terms 
of materials, especially as part of cross-cut-
ting activities, if there is a sufficient critical 
mass  of  interest within  the JPNM  partners.

Importantly,  GenIV reactors should not be 
decoupled from the relevant fuel cycle facil-
ities. These enable the fabrication of MOX fuel 
and, in the longer run, the fabrication and use 
of fuel that contains minor actinides, including 
advanced fuels (e.g. nitrides), so as to reduce 
to the very minimum the waste, by burning it 
in the reactor, as well as in ADS. These facilities 
include reprocessing and recycling, and to-
gether are needed to guarantee sustainability 
for centuries to come. For these reasons, one 

14  This is consistent with the analysis done in the reference 
cited in note 13, where the feasibility of the different systems 
based on the state-of-the art is classified as medium or low 
for existing SFR and LFR concepts, but uncertain or critical for 
SCWR and MSR, as well as for GFR.

1. 4.  Objectives and Grand 

Challenges of the EERA 
JPNM

The objective of the EERA JPNM is to improve 
safety and sustainability of nuclear energy by 
focusing on materials aspects. This has two 
implications: 

1. 

2. 

 Better knowledge of materials behaviour 
under service conditions, seeking predic-
tive capability, to select the most suited 
materials and define for them safe design 
rules, especially considering radiation and 
temperature effects, and taking care of their 
compatibility with coolants. 
 Development  of  innovative  material 
solutions  to  improve  resistance  to  high 
temperature, irradiation and aggressive en-
vironments, through suitable processing/
protective methods applied to existing ma-
terials or developing new types of materials

The EERA JPNM vision paper15 expresses a 
view  on  nuclear  energy  materials  strongly 
linked with the ESNII systems and identifies 
three Grand Challenges. These are based on 
the three pillars of the EERA JPNM’s research 
approach and strategy (see Figure 2):

 y Materials characterization: Assessment of 
candidate structural and fuel materials’ be-
haviour in operational conditions: screening, 
selection and qualification with a view to de-
veloping design rules;

 y Materials modeling: Development of ad-
vanced  models  to  rationalise  materials’ 
behaviour,  underpin  design  correlations 
and provide a basis for the improvement of 
materials properties by providing predictive 
capability;

 y Innovative  materials:  Development  of  in-
novative structural and fuel material solu-

15  https://www.eera-set.eu/wp-content/uploads/Vision-
Paper-EERA-JP-Nuclear-Materials-February-2015.pdf

Strategic Research Agenda of the EERA JPNM 
tions  and  advanced  fabrication  processes 
for industrial use, with superior capabilities 
in  terms  of  resistance  to  irradiation,  high-
temperatures and aggressive environment. 

Figure 2: The research approach and strategy pillars 
of the EERA JPNM.

The three grand challenges correspondingly 
identified are:

 y Grand Challenge 1: Elaboration of design 
correlations, assessment and test proce-
dures for the structural and fuel materials 
that have been selected for the demonstra-
tors, under the service conditions expected. 
This involves deployment of infrastructures 
for exposure to ageing environments, test-
ing of materials and production of data and 
knowledge.

 y Grand  Challenge  2:  Development  of 
physical  models  coupled  to  advanced 

The objective of the EERA 
JPNM is to improve safety 
and sustainability of nuclear 
energy by focusing on 
materials aspects

25 

microstructural characterization to achieve 
high-level understanding and predictive 
capability. These are essential assets, given 
the scarcity of experimental data and the 
difficulty and cost of obtaining them.

 y Grand Challenge 3: Development of inno-
vative structural and fuel material solutions 
and  fabrication  processes,  in  partnership 
with industry, to achieve superior thermo-
mechanical properties, better compatibility 
with  coolants  and  improved  radiation-re-
sistance,  so  as  to  increase  safety  and  im-
prove efficiency and economy.

Addressing these Grand Challenges requires 
a concerted action at European level involving 
research community and industrial partners.

1. 5.  Purpose, target and 

structure of this Strategic 
Research Agenda

The SNETP Deployment Strategy (DS) of 20158 
provides a possible timeline for the different 
nuclear technologies. While the actual timeline 
will evolve, two important messages can be 
deduced: 

 y Research on materials that impact structural 
integrity, component ageing and advanced 
solutions for components in nuclear systems, 
is a continuous process without a deadline: 
it constitutes the research ‘humus’ on which 
innovation, and consequently better safety 
and efficiency for nuclear systems, can grow;
 y In  order  to  allow  the  licensing  and  cons-
truction  of  GenIV  demonstrators,  taking 
into  account  also  recent  developments, 
the research on materials needs to provide 
sufficient data for qualification and possibly 
codification of design rules, as well as com-
puter simulation, depending on system and 
issue, in a horizon that has a span of 10 to 
20 years.16

16 

It is not possible to establish a priori when the data about 
a given material can be considered sufficient, because it de-
pends  enormously  on  the  system,  the  component  and  the 
targeted operating conditions. This is why a continuous inter-
action between materials scientists and designers is needed.

1. Introduction26 

This horizon, taking into account the specifi-
cities of nuclear energy, is relatively short. To 
be met, if the goal of sustainability is to be 
reached, the deployment is required now of 
significant resources that are devoted to ma-
terials research and development.

In this framework, the present SRA identifies 
scientific and technical gaps that need to be 
addressed, as well as innovative routes that 
may be followed, to provide sufficient data, 
knowledge, simulation tools and experimen-
tal instruments, concerning structural and fuel 
materials, to enable the design, licensing and 
construction of GenIV system demonstrators, 

This SRA identifies scientific 
& technical gaps and 
innovative routes to provide 
sufficient data, knowledge, 
simulation tools and 
experimental instruments, 
concerning structural and 
fuel materials, to enable 
the design, licensing and 
construction of GenIV 
system demonstrators, 
opening the way to 
longer term commercial 
deployment. 

It is expected to be of use 
for scientists, industries and 
research managers and 
funders, decision-makers.

Vision
paper

Strategic Research 
Agenda

Description of Work
(general + 6 DoW of SPs)

Figure 3: Documents describing the research pro-
gramme of the EERA JPNM with increasing level of detail 

from top to bottom.

as well as to open the way to their longer term 
commercial deployment.

The research programme of the EERA JPNM 
is defined by three documents, as shown in 
Figure 3. 

 y The  vision  paper  provides  the  general 
high-level  context  and  is  meant  for  any 
audience;

 y The  strategic  research  agenda  (SRA), 
which is revised periodically,17 describes 
for stake-holders the R&D route to be fol-
lowed (roadmap) to face the three Grand 
Challenges of the EERA JPNM identified 
in the vision paper, extending also to the 
other strategic activities that accompany 
research.

 y The description of work (DoW), regularly 
revised, is the implementation of the SRA: 
it  describes  work  that  is  underway,  inclu-
ding tasks and deliverables at the level of 
research subprogrammes.

The present SRA is thus expected to be of 
use to guide the activities of all nuclear ma-
terials stakeholders: scientists, industries, 

17  This  SRA  may  need  to  be  updated  if:  (a)  materials  con-
sidered have become obsolete; (b) new materials solutions 
appear;  (c)  the  reference  system  conditions  change  sub-
stantially.

Strategic Research Agenda of the EERA JPNM 
27 

Figure 4: Matrix structure of the SRA: the crossing of infratechnologies and materials identifies blocks which form 
the sections of the part 3 of this document.

research  managers  and  decision-mak-
ers.18 It is aimed at researchers active within 
and without the EERA JPNM, but it has the 
ambition to reach other stakeholders also, 
such as members of the SET-plan Working 
Groups, nuclear and non-nuclear technology 
platforms representing both industry and re-
search organizations, as well as managers and 
decision-makers, especially member states 
representatives and European commission 
officers. Its goal will be reached if and when 
the industrial counterpart adopts the results 
of  the  planned work  for  specific  compo-
nent/system design.

This  SRA  adopts  a  matrix  structure  (C. 
Featherstone and E. O’Sullivan19), as schemat-

ically described in Figure 4, which is a visual 
representation of its “fabric”: the horizontal 
‘warp’ is given by the cross-cutting research 
strategies, or infratechnologies, while the ma-
terials addressed, structural and fuel, provide 
the vertical ‘weft’20. Within each matrix ele-
ment, or block, different issues and relevant 
goals are identified. 

The three infratechnologies that are identified 
are obviously interdependent, as exempli-
fied in the virtuous circle of Figure 2, which 
illustrates the JPNM research strategy. Pre-
normative research on existing materials and 
screening of new material solutions clearly 
overlap and feed each other in terms of meth-
odology and the necessary infrastructures, 

18  http://ec.europa.eu/research/industrial_technologies/
pdf/research-road-mapping-in-materials_en.pdf

vation, Institute for Manufacturing, University of Cambridge, 
March 2014.

19  Featherstone  and  O’Sullivan,  A  review  of  international 
public sector strategies and roadmaps: a case study in ad-
vanced materials, Centre for Science Technology and Inno-

20  Research  Road-Mapping  in  Materials, A.  F.  de  Baas  (Ed.), 
European Commission, DG Research, EUR 24210 EN (2010), 
doi: 10.2777/87000.

1. Introduction28 

while physical models and simulation tools 
describing the behaviour of materials sup-
port and have the potential to accelerate both 
pre-normative research and development of 
new material solutions.

This document presents the research agenda, 
including discussions of all related strategical 
aspects:

 y Actual roadmap in terms of description of 

key issues and relevant activities;

 y Identification of cross-cutting research with 
other nuclear and non-nuclear energy tech-
nologies;

 y Needs for specific facilities and infrastruc-

tures and viable access schemes;

 y Interaction with industry;
 y Benefits of international cooperation;
 y Some aspects related with education and 

training;

 y Risk assessment.

1. 6.  Sources and links with other 

roadmaps and SRAs

To elaborate the present SRA, the priorities of 
ESNII concepts and the objectives of SNETP 
were  taken  into  account,  while  identifying 
commonalities with  other  nuclear  and  also 
non-nuclear  energy  technologies. Towards 
this aim the EERA-JPNM launched, in 2015, a 
wide consultation involving ESNII representa-
tives to report on the most recent design of the 
prototypes, and materials scientists active in 
the research performed within the six sub-pro-
grammes of the JPNM. Subsequently, contacts 
were established with the SNETP pillars21  and 
the fusion community, as well as with other 
joint programmes in EERA22. A MoU was signed 
between EERA and SNETP in December 2016, 
with the purpose of agreeing officially on com-
mon research topics. As a first action for the 

21  More  specific  joint  task  descriptions  with  SNETP  pillars 
than in this SRA can be found in the Technical Annexes of the 
EERA JPNM/SNETP MoU.

implementation of this MoU, technical annexes 
that consensually identify common research 
activities  between  EERA JPNM  and  SNETP 
pillars have been produced. Moreover, other 
roadmaps, SRA and roadmapping initiatives 
have been taken into account. These sources 
are listed in what follows.

1.6.1.  Materials roadmap enabling low 
carbon energy technology (MR) 

This roadmap was issued after a wide consul-
tation among stake-holders that was launched 
by the EC in December 2011.23 It highlighted the 
steps to be followed in the field of materials 
for advanced low-carbon energy technologies, 
defining  a  10-year  European  Research  and 
Innovation (R&I) agenda. For nuclear fission it 
proposed a Research and Development (R&D) 
programme on commercially available materi-
als for the prototypes and demonstrators, and 
on advanced materials for industrial scale sys-
tems. This programme targeted materials for 
cladding, coatings for enhanced corrosion and 
erosion/wear resistance and novel advanced 
materials. Specifically: (i) manufacturing and 
out of pile testing of F/M 9%Cr steels for heat 
exchanger, (ii) manufacturing and out of pile 
(and possibly in-pile) testing of oxide dispersion 
strengthened (ODS) claddings, (iii) manufactur-
ing and out of pile (and possibly in-pile) testing 
of SiCf/SiC composites cladding and (iv) man-
ufacturing and testing of coatings. While not 
all of these targets are still relevant, several 
technical issues mentioned in the MR are also 
in the focus of this SRA, which updates them 
to the current perception of needs.

1.6.2.  The SET-plan Integrated Roadmap 

(IR) 

Nuclear energy is recognized by the SET-
plan  as  a  low-carbon  energy  source. 
Accordingly, in the Integrated Roadmap (IR), 
which was officially presented and launched 
on the occasion of the SET-plan conference in 

22   https://www.eera-set.eu/wp-content/uploads/EERA-JP-
workshop-Materials_for_Energy_report.pdf 

23  https://setis.ec.europa.eu/system/files/Materials_Road-
map_EN.pdf

Strategic Research Agenda of the EERA JPNMNuclear energy is 
recognized by the  
SET-plan as a low-carbon 
energy source. Materials 
have clearly an important 
role to play to reach the 
SET-plan target of nuclear 
safety and its related 
Implementation Plan.

December 2014 in Rome, Heading 5 of Part II 
reads: “Supporting Safe Operation of Nuclear 
Systems  and  Development  of  Sustainable 
Solutions for the Management of Radioactive 
Waste”. Under this Heading, nuclear materi-
als for GenIV reactors are explicitly mentioned: 
“Qualify nuclear materials for operation under 
Gen IV conditions and develop innovative ma-
terials to improve plant safety and efficiency”. 
Indirectly, moreover, research and innovation 
on nuclear materials are implicit in other parts 
of the document under the same Heading, 
where they are necessary for the accomplish-
ment of the objectives mentioned.

Among  the  10  SET-plan  key  actions  from 
the IR, used as a base for discussions with 
Member States and stakeholders on the prior-
itisation of energy research activities in Europe 
in 2017-18, nuclear energy enters as the 10th 
objective, as follows: “Maintaining a high level 
of safety of nuclear reactors and associated 
fuel  cycles  during  operation  and  decom-
missioning, while improving their efficiency”. 
Materials have clearly an important role to 
play in reaching this 10th SET-plan target and 
its related Implementation Plan. The pres-
ent SRA provides essentially an expansion of 
the IR, to allow its practical implementation 
for what concerns research on GenIV nuclear 
materials.

29 

1.6.3.  OECD/NEA Technology Roadmap 

for Nuclear Energy and other OECD/
NEA initiatives

The  OECD/NEA  Technology  Roadmap  for 
Nuclear Energy 201524 recommends that, in the 
timeframe 2015-2030, the “governments [are] 
to recognise the long-term benefits of devel-
oping GenIV systems […]”, this , as mentioned, 
involving crucially the selection and qualifi-
cation of materials. The present SRA is thus a 
timely document, to ensure efficient use of the 
funding to be provided by governments.

At  the  time  of  preparation  of  this  SRA,  the 
OECD/NEA is engaged in the elaboration of a 
Nuclear Innovation Initiative (in the low-carbon 
perspective) with a horizon to 2050. Via the 
MoU signed between EERA JPNM and OECD/
NEA a strong link was created between the 
present SRA and the priorities and projects 
identified in this NEA initiative.

1.6.4.  SNETP Strategic Research and 

Innovation Agenda and Deployment 
Strategy 

The  SNETP  released  a  Strategic  Research 
and Innovation Agenda (SRIA)25 in February 
2013 and the corresponding DS in June 2015. 
Together, these documents define clearly the 
three pillars for nuclear energy research and 
demonstration in the following decades, as 
follows:

 y Support the fully safe operation of present 
and newly built LWR, so-called GenII/III 
reactors, allowing the development of sus-
tainable solutions for the management of 
radioactive wastes;

 y Prepare the development and demonstra-
tion of advanced fast neutron GenIV reactor 
technologies associated with a closed fuel 
cycle to enhance the sustainability of nu-
clear energy;

24  https://www.oecd-nea.org/pub/techroadmap/

25  http://www.snetp.eu/wp-content/uploads/2014/05/
sria2013_web.pdf

1. Introduction30 

Materials behaviour for 
structural components and 
fuel and more generally 
the structural integrity of 
systems and components 
are common trunks through 
GenII/III and IV, cogeneration, 
and also fusion. 

Consistently, in the present 
SRA, significant effort is 
made to identify cross-
cutting issues with other 
energy technologies.

 y Promote the use of nuclear energy beyond 
electricity  generation,  namely  in  cogene-
ration  of  heat  or  hydrogen  production  or 
water desalination. 

The  importance  of  transversality  between 
SNETP pillars and with other energy platforms 
is a key message of the SNETP DS. In particu-
lar “the identification of R&D project clusters 
for Gen II, III, IV and cogeneration applications 
for basic technology developments, e.g. per-
formance and ageing of NPPs for long-term 
operation  and  high  reliability  components 
for structure and fuel, could valuably build 
a bridge between the different nuclear sys-
tem  developments  and with  other  ETPs  as 
well”. In other words, materials behaviour for 
structural components and fuel and more 
generally the structural integrity of systems 
and components are common trunks through 
GenII/III and IV, and cogeneration as well. 
Consistently, in the present SRA, significant 
effort is made to identify cross-cutting issues 
between materials for GenIV and for GenII/III, 

as well as for high temperature. (Similar effort is 
made for fusion energy and other energy tech-
nologies). The EERA JPNM operates at low TRL 
(<5), while industrial initiatives, involving util-
ities, necessarily deal especially with higher 
TRL solutions and approaches: this is a criterion 
used to identify overlaps and set boundaries. 

1.6.5.  GIF Technology Roadmap3 and R&D 

Outlook 

The Generation IV International Forum (GIF) 
was founded in 2001 with the objective to es-
tablish international co-operation in R&D for six 
future nuclear systems (SFR, LFR, GFR, SCWR, 
VHTR, MSR). 

The Technology Roadmap was published in 
2002 and updated in 2013; a R&D outlook was 
published in 200926. The Roadmap and R&D 
Outlook provide an overview of the major R&D 
objectives and milestones for the coming dec-
ade, aiming to achieve the Generation IV goals 
of sustainability, safety and reliability, economic 
competitiveness, proliferation resistance and 
physical protection. The emphasis is on the sys-
tems for which materials is an integrated part. 

1.6.6. Other roadmaps

Other documents connected with the present 
SRA in terms of issues, materials and goals, 
that were consulted during its preparation, are: 
(i) the SRA of EuMaT, the European Technology 
Platform for Advanced Engineering Materials 
and Technologies27; (ii) the roadmap of Materials 
Modelling of the European Materials Modelling 
Council28;  the  Roadmap  of  the  Metallurgy 
Europe – EUREKA cluster29; the Roadmap to 
Fusion Electricity (2012 version)30.

26  https://www.gen-4.org/gif/jcms/c_42188/publications

27  http://eumat.eu/filehandler.ashx?file=11580

28  https://emmc.info/wp-content/uploads/2015/04/
EMMC_Roadmap_V3.0.2.pdf

29  http://metallurgy-europe.eu/

30  https://www.euro-fusion.org/fileadmin/user_upload/
EUROfusion/Documents/Roadmap.pdf

Strategic Research Agenda of the EERA JPNM 
2

NUCLEAR 
MATERIALS 
FOR GEN IV 
SYSTEMS AND 
RESEARCH 
APPROACHES

© CVR

2.  NUCLEAR  

MATERIALS FOR  
GEN IV SYSTEMS  
AND RESEARCH 
APPROACHES

2.1.  Material degradation 

processes and relevant 
properties

The GenIV concepts will involve very harsh 
conditions in terms of temperature, irradiation 
levels and interaction with the coolant for all 
materials. Here we focus on structural and fuel 
materials. 

Liquid-metal cooled system operation tem-
peratures  are  expected  to  be  between 
400-550°C for demonstrators, with off-normal 
excursions up to 600°C (unless new materials 
allow to go higher), but ideally targeting 600-
700°C or even beyond for commercial plants, 
while operation temperatures in gas-cooled 
systems target 850°C or beyond. In the latter, 
this will lead to temperatures around 2200°C 
at the centre of the fuel in normal conditions, 
and temperatures may exceed 1000°C in struc-
tural materials in off-normal conditions. These 
temperatures, coupled to tremendous temper-
ature gradients (up to 500°C/mm), will inflict 
severe thermal and mechanical stresses on 
the fuel and plant components that will act to 

cause the degradation of both structural and 
functional materials through several simul-
taneously occurring mechanisms. Moreover, 
cooling fluids represent invariably a chemi-
cally hostile environment, the exposure to 
which also affects severely the performance 
of materials in contact with them. These effects 
are unavoidably exacerbated by high temper-
ature. Thus, what often limits the operating 
temperature is not only the degradation of 
the properties of the materials per se, but the 
effect on the materials of the aggression by 
heat transfer fluids above a certain temper-
ature. Finally, nuclear materials are exposed to 
varying levels of irradiation: for example, the 
dose measured in number of displacements 
per atom (dpa) will reach 1 dpa in the fuel in 
less than 1 day in reactor, will exceed 100 dpa 
in the cladding during its stay in the reactor, 
and may be less than 2 dpa in the vessel over 
the whole reactor life. Invariably, these levels 
of irradiation will negatively affect the per-
formance of the materials.

The mechanical degradation mechanisms 
that are known to operate at high temperature 
are creep, thermal and mechanical fatigue, 

34 

creep-fatigue, ratcheting and thermal ageing. 
Furthermore, corrosion (oxidation), dissolution 
and  erosion  are  thermochemical  mecha-
nisms  that  are  due  to  interaction  between 
fuel, structural materials and cooling fluids. 
In some cases, the exposure to fluids may 
affect the mechanical properties (e.g. liquid 
metal embrittlement). Irradiation also affects 
the materials by changing substantially and 
progressively their chemical composition, mi-
crostructure and microchemistry. Importantly, 
these  mechanisms  act  simultaneously, 
thereby introducing high complexity in terms 
of synergistic effects that cannot necessarily 
be simply “linearly superposed”.

For  fuel,  one  of  the  main  parameters  that 
needs to be maintained under irradiation is 
the temperature at the centre of the pellet. 
This is governed by the thermal conductivity 
and needs to be less than the fuel melting 
temperature, itself strongly affected by the 
irradiation-induced modifications of the fuel 
within  the  reactor  (creation  of  defects  and 
fission products, restructuring). Secondly, the 
integrity of the fuel pin, which is affected by 
the increasing mechanical load on the clad-
ding as burn-up increases and by the possible 
chemical interaction between fission products 
and cladding elements, must be maintained 
in order to prevent fission product release.

For materials with a structural function the 
properties that need to be preserved are: (i) 
strength and resistance to creep deformation, 
with rupture beyond the design stresses and 
lifetime; (ii) toughness and resistance to crack 
initiation and propagation under static and 
also cyclic loading (fatigue); and (iii) especially 
for cladding materials, high thermal conduc-
tivity and limited thermal expansion, as well 
as hermeticity against gaseous and volatile 
fission  products.  In  addition,  the  materials 
need to maintain their thickness by resisting 
the possible corrosive effect of the coolant 
environment.

Only a few classes of materials have the poten-
tial to meet these requirements, with different 
levels of quality in their response: these are 
described in the next section.

2.2.  Nuclear materials, 

components and systems

2.2.1.  Structural materials

Structural materials are those used to fabri-
cate a component that bears load or stress, 
whatever its origin (mechanical, thermal, vi-
brations…). 

In nuclear reactors it is important to distinguish 
between two types of structural components, 
replaceable and non-replaceable:

 y Replaceable  components  are  designed 
to be (relatively) easily extracted from the 
reactor,  or  their  replacement  is  a  normal 
part  of  the  reactor  operation.  The  most 
obvious examples of this type are the fuel 
elements, which are periodically reshuffled 
or removed once the burnup limit allowed 
by  neutronics  and  materials  has  been  re-
ached. In general, these elements are the 
hollow tubes that contain the fuel pellets, 
the  bundles  that  hold  the  tubes  together 
and  the  structures  that  support  these.  A 
significant deviation from this general des-
cription of the fuel element may be found in 
HTR or GFR, where the high temperatures 
may  require  the  use  of  ceramic  materials. 

Replaceability  generally  goes  hand  in 
hand with the expectation of more severe 
degradation  and therefore shorter life-
time. This is the case for the fuel cladding, 
which is exposed to the highest irradiation 
dose, experiences the highest temperatures 
and also high temperature gradients, being 
in contact with the coolant and the fuel.

 y Non-replaceable  components  constitute 
the  main  structure  of  the  reactor.  Major 
examples of this type are the vessel or, in 
pool reactors, the upper cover of the con-
tainment.  These  components  are  charac-
terized by the fact that their replacement, 
though  theoretically  not  impossible,  es-
sentially  corresponds  to  building  a  new 
unit, i.e. it is so costly and complex that it is 

Strategic Research Agenda of the EERA JPNMeconomically not affordable.31 These com-
ponents  need  to  be  designed  for  the  full 
lifetime of the reactor or, correspondingly, 
their lifetime defines the lifetime of the re-
actor itself.

Non-replaceable components need to be 
designed in such a way that the degrading 
agents are mitigated as much as possible, 
so that the materials ageing is sufficiently 
slow to guarantee that the component re-
mains fit for purpose until the end of the 
life of the plant. Typically, the vessel will be 
sufficiently distant from the fuel elements 
to be subject only to marginal or negligible 
irradiation and the system will be designed 
so  that  the  vessel  temperatures  will  be 
lower  than  close  to  the  fuel  element.  In 
addition, it is sometimes possible to apply 
protection the vessel from the effect of the 
coolant, although the problem of prolonged 
exposure to the coolant may in fact be the 
main lifetime limiting factor for irreplaceable 
components.

The  choice  of  the  most  suitable  materials 
for a given component and application obvi-
ously comes, to a large extent, from previous 
experience. The current design of GenIV dem-
onstrators planned in Europe envisages the 
use of austenitic steels as the main class 
of in-core structural materials, specifically 
316L(N) for most components, including the 
vessel, and 15-15 Ti for the cladding and other 
fuel element parts32. Depending on the spe-
cific demonstrator, other materials may also 
enter, e.g. ferritic/martensitic (F/M) steels 
in the SFR core, while for the GFR core ce-
ramic materials similar to those used in HTR, 
or perhaps refractory metallic alloys, must 
be considered. Other materials foreseen for 
use in out-of-core components are Ni-based 

31  In  the  history  of  nuclear  energy,  however,  some  compo-
nents initially meant to be irreplaceable have been eventual-
ly replaced, because the high cost was counterbalanced by 
the expected gain from the exploitation of the plant.

32  These are AISI and customarily used names. In Europe the 
formally correct way to denote them is X2CrNiMo 17 12 2 con-
trolled nitrogen content (for 316L(N)) and X10CrNiMoTiB 15 15 
(for  15-15 Ti);  the  latter, when  of  German  production,  is  also 
denoted as 1.4970.

35 

alloys, especially for the intermediate heat ex-
changer, the steam generator, turbine blades, 
coaxial pipes or hot gas headers of GFR and 
HTR. However, austenitic steels are clearly 
the current dominant choice across system 
demonstrators. The reason is that they are a 
very good compromise between several re-
quirements, even without excelling specifically 
in any one of them. The critical reason for this 
choice is the return of experience from their 
use in the fast reactors built and operated in 
the past, such as Phénix and Superphénix in 
France. There exists, therefore, a wealth of ex-
perimental data on them, on the basis of which 
design rules have been already established 
and introduced in standard codes. This is a 
significant advantage towards design and li-
censing. Nonetheless, still several aspects 
need to be qualified in austenitic steels and 
require intensive research effort, especially 
concerning  compatibility  with  coolants. 
Therefore, the largest effort of qualification 
foreseen in this SRA is dedicated to auste-
nitic steels (section 3.1.1). Research activities 
also include paths to improve the properties 
of austenitic steels (section 3.1.3): (1) improve 
swelling  resistance;  (2)  improve  corrosion 
resistance,  especially  against  HLM.  Higher 
swelling resistance is pursued via microstruc-
ture stabilisation of the steels with suitable 
alloying elements (e.g. AIM1 and its improved 
version AIM2). Better corrosion resistance is 
pursued with the creation of corrosion barriers, 
either by changing the surface composition 
with the implantation of aluminium to form a 
protective alumina layer, or by directly apply-
ing alumina (or other ceramics) coatings on 
the metallic substrate. Another possibility is to 
add aluminium directly to the alloy, to allow the 
formation of a self-healing protective alumina 
layer when in contact with oxygen-containing 
fluids: these are the so-called alumina form-
ing austenitic (AFA) steels, which would also 
exhibit inherently good creep properties due 
to NiAl precipitation. None of these solutions, 
however, is currently included in design codes.

F/M steels (e.g. T91, EM10, HT9, reduced acti-
vation F/M like EUROFER or F82H for fusion, …) 
potentially offer a number of desirable prop-
erties as cladding and core materials that are 

2. Nuclear materials for GEN IV systems and research approaches36 

superior to those of austenitic steels, namely 
better thermal conductivity, lower expansion 
coefficient  and  better  resistance  to  radia-
tion-induced void swelling. The third point is 
key: while austenitic steel cladding will not sus-
tain more than 100 dpa of irradiation dose,33 
F/M steels are expected to reach 200 dpa 
without experiencing any swelling, thereby 
offering a promising solution to significantly in-
crease the burnup. However, F/M steels can 
currently operate only below 550°C, suffer from 
radiation-induced embrittlement below 350-
400°C and are especially prone to liquid metal 
embrittlement (LME), when in intimate contact 
(wetting) with, in particular, heavy liquid met-
als (HLM)34. Reliable correlations and models 
for the identification of design rules allowing 
for all these phenomena are needed to make 
F/M steels usable in GenIV systems; alterna-
tively, the resistance to high temperature and 
embrittlement of F/M steels, as well as the 
compatibility with HLM, need to be improved. 

Two paths are pursued to improve the creep 
properties  of  F/M  steels:  (1)  production 
of  oxide-dispersion  strengthened  (ODS) 
steels,  currently  using  powder  metallurgy 
techniques; (2) tuning of the composition to-
gether with appropriate thermal-mechanical 
treatments (TMT) in a conventional metallurgy 
framework (creep-strength enhanced, CSE). 
The use of ODS F/M cladding would allow 
the upper  limit  of the temperature window 
to be increased beyond 700°C, improving ra-
diation resistance as well. However, the ODS 
steel fabrication process is costly and there 
is a lack of established industrial production. 
The second way to improve the properties of 
F/M steels, via composition tuning and ther-
mal-mechanical treatments (TMT) to stabilise 
the microstructure, is very attractive because 
the  manufacturing  process  remains within 
conventional metallurgy and has been used 
successfully outside the nuclear field (650°C 

33  Currently, 15-15 Ti is supposed to reach 90 dpa, better ver-
sions of it (AIM2) might allow this limit to increase up to 110 dpa.

34  Liquid lead and its alloys, such as Lead-Bismuth Eutectic 
(LBE) or Pb-Li, are collectively denoted as heavy liquid me-
tals (HLM). Recently it has been observed that also austenitic 
steels  in  liquid  sodium  may  experience  LME,  although  the 
conditions for this to happen are not easy to comply with.

operation limit). Similar techniques may also 
help to limit radiation embrittlement below 
350°C. The methods to improve corrosion re-
sistance are the same as in austenitic steels: 
either surface modifications/coatings or alloy 
composition tuning with aluminium addition: 
FeCrAl alloys may also be ODS, giving rise to 
a potentially very promising material that is re-
sistant to irradiation, creep and corrosion. 

For temperatures above 800°C, such as those 
targeted in GFR and HTR, no steel is able to 
maintain its fitness for purpose, with the possi-
ble exception of high Ni austenitic steels such 
as alloy 800. In this case, Ni-based alloys (e.g. 
Inconel 617, Haynes 230 and Hastelloy XR) are 
a possibility for components outside the core, 
particularly heat exchangers (not only for GFR 
and HTR). However, these alloys suffer from 
quite  severe  irradiation  embrittlement  and 
also swelling, so their use in the core can be 
critical. Other heat-resistant materials need 
therefore to be considered for in-core applica-
tions and the spectrum is quite wide, ranging 
from refractory metals (e.g. molybdenum or 
vanadium) to ceramics. The latter generally 
offer very attractive properties in terms both 
of stability to high temperature and resistance 
against wear and corrosion/erosion. However, 
both are penalized by brittleness. Thus, re-
fractory materials for structural components 
are likely to be mainly composites, enabled by 
fibres or other reinforcements to exhibit some 
type of pseudo-ductile mechanical behaviour. 
The most intensively studied ceramic com-
posite is SiCf/SiC, i.e. silicon carbide fibres in 
silicon carbide matrix, and SiCf/SiC is the main 
candidate material for GFR cladding. It is also a 
material of use for the HTR and it is a suitable 
candidate for LFR and SFR cladding, as well. 
However, this material is very expensive and 
design rules for it are not developed. 

Moreover, the design of the high temperature 
gas cooled reactor concept (GFR, V/HTR), im-
poses the need for materials solutions to be 
devised for the purposes of thermal shield-
ing and insulation, as well as control rods and 
seals. These materials must be gas-tight, cor-
rosion resistant, and exhibit high fracture and 
creep strength, while being inexpensive and 

Strategic Research Agenda of the EERA JPNM37 

Table 1: Main structural materials expected to be used in the different ESNII GenIV systems, distinguishing different 

phases that define the timeline, from demonstrator to prototype (FOAK), and finally commercially deployed plants.

PHASES →

ESNII demonstrator

SYSTEMS ↓ 

As licensed 
(phase I)

Evolving  
(phase II)

FOAK 
(prototype)

Commercial 
deployment

Periodically 
Replaced 
Components

Permanent 
Structural 
Components

Periodically 
Replaced 
Components

Permanent 
Structural 
Components

Periodically 
Replaced 
Components

Permanent 
Structural 
Components

Periodically 
Replaced 
Components

SFR (ASTRID)

ADS (MYRRHA) 

LFR (ALFRED)

GFR 
(ALLEGRO) / 
(V)HTR 

AuSS: 15-15Ti – 
AIM1 (cladding)
F/M: EM10 
(wrapper)

AIM2 or F/M 
ODS (cladding)

TMT F/M or 
F/M
ODS

TMT F/M, 
F/M ODS, 
perhaps SiC/
SiC

AuSS: AISI316L(N); 800SPH

AuSS: AISI316L(N); TMT F/M

Coated 15-15Ti 
(FeAl, FeCrSi, 
FeTa, MAX 
phases, …) or AFA

N/A

Cladding: 
1.4790; 
structures: 
316L(N)

316L(N)

Cladding and 
structures: 
(Al2O3 coated) 
15-15Ti (AIM1)

Cladding and 
structures: Al203 
Coated 15-15Ti 
or AFA

Cladding: AFA 
or FeCrAl ODS
Structures: AFA

AFA or FeCrAl 
ODS, or 
(coated) Mo-
ODS, or SiCf/
SIC ,

316L(N)

GFR: T<550°C: 
15-15Ti 
(cladding) – 
AFA?
HTR: TRISO 
(SiC)

AFA or ferritic steel lined with 
AFA

GFR: T> 850°C: 
SiC/SiC 
(cladding) / HTR 
TRISO (SiC)

SiC/SiC, perhaps Mo-ODS/ HTR 
TRISO (SiC)

Permanent 
Structural 
Components

GFR : T<550°C: 
316L(N) – AFA ?
HTR: graphite

GFR: 
550<T<850°C: 
AFA, FeCrAl ?
HTR: graphite

GFR: AFA or FeCrAl, perhaps , 
Mo or V alloys HTR: graphite

preferably fabricated in a net-shape design, 
with the ability to be joined. At the moment, 
in addition to SiC, also other materials such as 
carbon composites, mullite, Al2O3, TiO2, ZrC, 
ZrN, B4C, graphite or graphene are being con-
sidered for this function.

Perspective ceramic or metallic refractory 
materials,  including  innovative  ones,  that 
can be considered for cladding or other ap-
plications  in  GFR/HTR,  and  also  for  other 
systems like LFR to mitigate corrosion effects, 

are: ODS-Mo, high entropy alloys (HEA), and 
MAX phases.35 For all these materials, includ-
ing  SiCf/SiC,  there  are  issues  concerning 
fabrication  processes,  optimal  choice  of 
composition/architecture of the component, 
joining,  standardization  of  testing,  and  the 
usual problems of radiation resistance in con-
tact with coolants. 

35  Most  likely  suitable  as  a  corrosion  protective  coating  at 
not-too-high temperatures.

2. Nuclear materials for GEN IV systems and research approaches 
38 

Based on this excursus on structural mate-
rials for GenIV systems, Table 1  attempts a 
classification and a prediction of the main 
structural materials that are expected to be 
used in the different GenIV systems that are 
currently part of ESNII, distinguishing differ-
ent time phases from demonstrator (phase 
I and II) to prototype first-of-a-kind (FOAK) 
and, finally, commercially deployed plants. 
This table is important because it allows 
the different materials and therefore the 
research  and  development  devoted  to 
them  to  be  set  in  the  correct  timeframe 
and in connection with a specific system. 
The actual timing for each system is not the 
same and will crucially depend on the pri-
orities  that  are  politically  associated with 
one  system  or  another  and  the  readiness 
with which funding is allocated: none of this 
can  be  easily  predicted.  However,  from  a 
conceptual viewpoint the phases schema-
tized in the table remain valid and therefore, 
through this table, it is possible to estab-
lish priorities and perspectives. Table 1 is 
not  expected  to  be  fully  comprehensive, 
but simply highlights which classes of ma-
terials the EERA JPNM research focuses on 
and where they fit, in order to immediately 
understand the application and context. This 
table  is  based  on  current  knowledge  and 
perceptions  about  the  future  and will  re-
quire updates. Phases I/II of the prototypes 
are not necessarily defined in the ESNII plans 
and can be of very different nature, but es-
sentially correspond to switching to different 
materials, particularly cladding and fuel ele-
ment materials, for better performance. For 
instance, the GFR considers a first demon-
strator working at low temperature (~500°C) 
to be later upgraded for high temperature 
operation (~850°C), when the qualification 
of the necessary materials is sufficiently ad-
vanced; a similar strategy is envisaged for 
the LFR demonstrator (380520°C). Finally, 
concerning  FOAK  and  commercially  de-
ployed reactors, the materials mentioned in 
the table are obviously an educated guess. 
The message is that materials that are not 
currently envisaged for the demonstrators 
remain a target of the EERA JPNM research, 
because of their long term potential. 

2.2.2.  Fuel materials

Nuclear fuel is a consumable and removable 
component at the heart of the reactor. It is the 
seat of the fissions of nuclei, which produce 
the energy to be ultimately used to produce 
electricity. It remains several years in the reactor 
until the maximum burnup admissible has been 
reached. It is exposed to the harshest irradi-
ation and temperature conditions. Together 
with the cladding, however, it is the only com-
ponent whose performance can be significantly 
improved during the lifetime of a reactor. 

The fuel materials are combined with cladding 
in the fuel elements. By conception, while en-
suring dimensional stability within the design 
margins, nuclear fuels and fuel elements must 
in particular:

 y  Provide the power expected during their 

whole stay in reactor;

 y  Use the fissile elements as best as possible 
to reduce the cost of energy production;
 y  Confine the fission products inside the fuel 
elements  in  all  operating  and  accidental 
conditions.

Nuclear fuels and fuel elements differ widely 
from reactor to reactor, in geometrical config-
uration, fuel composition and cladding. Fissile 
atoms used in nuclear reactors, however, are 
mainly uranium and plutonium. Fuel materials 
are thus compounds of these elements, either 
refractory ceramics such as oxides, carbides, 
nitrides or silicides, or stable metallic alloys. 
More complex systems include ceramic/ce-
ramic or ceramic/metal composites, or molten 
salt fuels. Different geometries of fuel rods, 
fuel plates and pellets have been developed. 
Actinide oxides (U and mixed U-Pu oxides, 
MOX) are the most industrially used fuel ma-
terials and have been used in power reactors 
since the 1960s. 

MOX will be the fuel for the first cores of ALL 
ESNII  fast  neutron  reactor  prototypes,  al-
though UO2 with <20% U-235 is planned to be 
used in the first core of ALLEGRO. Variations of 
MOX were used in the previous European fast 
reactor programmes: Phénix, Superphénix, and 

Strategic Research Agenda of the EERA JPNMDounreay fast reactors. MOX fuel is Europe’s 
only major knowledge and competence base 
for fast reactors, even if all European fast reac-
tors are now closed. It is thus natural that this 
fuel is the first choice for the future. 

MOX fuel crystallises in the cubic fluorite struc-
ture, so isotropic behaviour can be expected. It 
has high melting point, though only moderate 
thermal conductivity, resulting in typical op-
erating temperatures at 80% of the melting 
point. Its manufacturing technology is known 
and proven, although it needs new validation 
when a new plant is deployed.

Though significant knowledge is available 
from the past, it is not usable directly and 
must be leveraged to enable the licensing 
of future MOX fuels for first ESNII cores and 
later for Gen IV reactors. First, the MOX product 
can bear important intrinsic hallmarks that are 
linked to the fabrication methods used, e.g. po-
rosity distribution, grain size, and impurity levels, 
all of which have a big impact on its perfor-
mance. Furthermore, reactor core designs have 
evolved, with particular effort based on ther-
mal hydraulics and neutron physics to reduce 
safety relevant parameters as void coefficients, 
so pellet geometry has evolved from the clas-
sical one to thicker pellets with an annulus to 
limit centre line temperatures. Furthermore, the 
GEN IV cores, in addition to their differing reactor 
coolants, will not always be operated under the 
same power rating as in the past, necessitat-
ing further reactor specific investigations. The 
accelerator-driven systems also present spe-
cific issues that may affect the fuel pellets, for 
example due to the multiple beam trips, which 
will affect power and might create temperature 
oscillations, so fatigue effects could be substan-
tially more serious than in a critical reactor.

The introduction of multi-recycling of Pu will 
also perturb the Pu concentration in the fuel, 
but more importantly higher concentrations 
of  241Pu in the Pu isotopic vector will lead to 
higher contents of  241Am, should significant 
delays between separation and insertion in 
the reactor occur. This will lead to increased 
demands on the fuel fabrication technology 
(remote and shielded handling) and also on the 

39 

in pile performance as helium will be produced 
in the fuel in greater quantities than for today’s 
typical values.

In the longer term, the reduction of the long 
term toxicity of the waste can be dramatically 
improved by the introduction of advanced 
nuclear fuel cycles within which the minor 
actinides (MA) – americium, neptunium and 
curium – are extracted from the spent fuel 
and  introduced  in  the  fuel  cycle  for  their 
transmutation in fast reactors. 

Two types of concepts are envisaged for trans-
mutation.

The first concept, known as homogeneous 
mode,  involves  diluting  minor  actinides  in 
standard fast-reactor fuel. The advantage is 
the strong similarity between the structures of 
the various actinide oxides (fluorite-type cubic 
structure) and their mutual solubility. To min-
imize the impact of the introduction of minor 
actinides on reactor safety parameters, fuel mi-
nor actinide content is kept relatively low (a few 
% of heavy atoms). In such conditions, only a 
slight evolution is anticipated in fuel behaviour 
and performance, the properties remaining 
quite close to those of standard fuel, which 
facilitates the qualification of these fuels.

In the second concept, the heterogeneous 
mode, the minor actinides are concentrated in 
specific assemblies (also called minor actinide 
bearing blankets) located at the periphery of 
the reactor core. This results in a limited per-
turbation of the core behaviour. In addition, the 
MA bearing assemblies are manufactured in 
dedicated plants, separately from standard 
fuel, which enables limited quantities of MA-
bearing fuels to be handled. The low neutron 
flux level experienced at the periphery of the 
core, however, slows down the transmutation 
process and this is compensated by increasing 
the MA fraction up to 15-20%. A large R&D ef-
fort is required for the design of these specific 
objects and to ensure their qualification.

Because of the high neutron emission, ther-
mal power and toxicity of the minor actinides, 
the fabrication of MA-bearing fuels requires 

2. Nuclear materials for GEN IV systems and research approaches40 

Table 2: Main fuel materials expected to be used in the different ESNII GenIV systems, distinguishing different 

phases that define the timeline, from prototype/demonstrator to FOAK and finally commercially deployed plants.

PHASES →

ESNII demonstrator

SYSTEMS↓ ↓

SFR (ASTRID)

As licensed (phase I)

Evolving (phase II)

FOAK 
(prototype)

Commercial 
deployment

MOX 20-25% Pu from 
UOx recycling

MOX 20-25% 
Pu from MOX 
recycling, MA 
bearing fuels?

MOX 20-25% Pu from MOX 
multirecycling, MA bearing oxide 
fuels

ADS (MYRRHA)

MOX with up to 30-35% 
Pu

MOX, MA bearing 
fuels 

N/A

LFR (ALFRED)

MOX with up to 30% Pu

GFR (ALLEGRO) / (V)
HTR

UO2 with <20% U-235

MOX, advanced 
MX

MOX, advanced 
MX

Advanced MX, MA bearing fuels

Advanced MX, MA bearing fuels

heavy shielding. A  down  selection  of  opti-
mal fabrication routes must be achieved and 
remote handling for fuel fabrication and as-
sembly production engineered and qualified. 
Then, the assessment of the irradiation perfor-
mance of these fuel types, performed in the 
frame of several European projects since the 
90’s, needs to be completed.

In a further long term step, optimising core per-
formance in terms of breeding and increased 
margins to melt would necessitate the adop-
tion of mixed uranium and plutonium carbides 
and nitrides (denoted MX = MC and MN). These 
fuels have less moderation, and lead to harder 
neutron spectra and shorter doubling times. In 
addition, they have high melting points and sig-
nificantly higher thermal conductivity than MOX 
fuels, so that they operate at about 40-50% of 
their melting point, which provides an appeal-
ing safety aspect. Their fabrication is not trivial, 
though, if high purities are to be achieved. There 
are questions about their volatility at tempera-
tures below the melting point; this is a matter 
that increases in significance for Pu multi recy-
cling, as the built in Am component could be 
(but it is not fully proven) even more volatile than 
the U and Pu constituents.

Similarly  to Table  1  for  structural  materials, 
Table 2 lists the fuels that are envisaged to 
be used in the different ESNII systems in the 

different time phases. The same caveats apply 
as for Table 1.

2.3.  Materials research 

approaches

One of the main duties of materials engineers 
is to be able to foresee the time when, due 
to degradation, the material is no longer fit for 
its purpose under any kind of external influ-
ence, thereby defining the lifetime of systems 
and components. This implies knowing what 
happens to the material and how its proper-
ties change while in operation, as well as in 
off-normal conditions that may lead to acci-
dents, i.e. knowing the degrading agents and 
processes and their effects. This knowledge, 
which can be defined as qualification of the 
material, is generally obtained by testing the 
material  or  the  component  under  relevant 
conditions, as well as by monitoring it during 
operation. This knowledge is also essential to 
identify the limits of materials, often leading 
to a redefinition of the system and component 
design, to ensure that the operation conditions 
are such that the materials remain fit for pur-
pose for a sufficiently long time.

The acquired knowledge of the behaviour of a 
material is then transferred to models, that can 

Strategic Research Agenda of the EERA JPNMbe more or less detailed, specific or general, em-
pirical or theoretical, or a mixture of these. Tests 
and monitoring are necessary to verify up to 
what extent models are followed in each specific 
case. In turn, models are needed to interpo-
late or extrapolate to conditions under which 
tests and monitoring cannot be performed, or 
to reduce the need for testing and monitoring, 
knowing that in the event the materials in op-
eration can, in some cases and within limits, be 
monitored using non-destructive techniques, 
but cannot be continuously tested. Models un-
derlie the design rules and codes and provide 
a guide to plan the maintenance, inspection 
and, if needed, replacement of components, 
based on the assessment of their lifetime or 
probability of failure. Modelling is therefore of 
crucial importance, although the modern trend 
is to enhance the materials monitoring ca-
pabilities, in particular through digitalisation 
(big data analysis, use of artificial intelligence, 
…),  in  order  to  reduce  the  need  for  testing 
and  to  complement  the  model  indications.

As an alternative to changing design condi-
tions, which in some cases may penalise the 
efficiency or the economy of the system, the 
knowledge of the behaviour of materials leads 
naturally to the identification of ways to im-
prove their response under given conditions. 
This can be done by protecting them, or im-
proving their properties through appropriate 
processing, or by identifying or even develop-
ing in a targeted way entirely new materials. 
The development of new material solutions 
allows  design  constraints  to  be  relaxed, 
thereby improving, sometimes substantially, 
the efficiency or the economy of the system.

These three aspects of materials research, 
namely  qualification,  modelling  and  new 
materials solutions, define the approaches 
(infratechnologies) of interest for this SRA, i.e. 
methods and techniques that are common to 
structural and fuel materials and appear as 
the warp of the research agenda of Section 3, 
as illustrated in Figure 4. They are also linked 
and constitute a virtuous circle, as illustrated 
in Figure 2. They are described in this section 
as applied to nuclear materials, in particular for 
GenIV systems.

41 

2.3.1.  Materials qualification: design rules 
and codes, fuel performance codes

GenIV system demonstrators will necessarily 
be constructed using structural and fuel ma-
terials on which some return of experience 
exists from previously operated reactors, and 
which are possibly already included in design 
codes. However, innovative design such as for 
the ESNII systems can generally rely only on 
limited return-of experience.36 Thus, safety 
and other requirements need to be demon-
strated by qualifying the materials selected 
for the operating conditions expected in the 
demonstrators,37 to which they need to be 
proven to resist for sufficient time. These in-
clude high temperature, high irradiation levels, 
and compatibility with coolants, for instance 
between HLM and structural components.38 In 
turn, the actual operating conditions of dem-
onstrators will be influenced and determined 
by the properties of these materials, after ap-
plying suitable safety margins, to avoid their 
degradation beyond acceptable limits.

Off-normal conditions are especially impor-
tant in the nuclear field in the aftermath of 
the Fukushima accident, as a consequence 
of which stricter safety requirements are now 
imposed and more effort to demonstrate struc-
tural integrity in accident scenarios is needed.37

36  SFRs have been built and operated in the past and the De-
sign  Code  RCC-MRx  was  specifically  developed  to  support 
the development of these fast reactors. Thus ASTRID or any 
advanced SFR demonstrator can to a large extent be based on 
the experience from the French SFR reactors: Rapsodie, Phé-
nix and Superphénix. For the lead- and gas-cooled reactors, 
however, there is very limited, if any, return of experience.

37  In design codes, e.g. RCC-MRx, operating conditions can 
be  defined  as:  (1)  conditions  to  which  component  may  be 
subjected in the course of normal operation, including nor-
mal  operating  incidents,  start-up  and  shutdown;  (2)  emer-
gency  conditions  corresponding  to  very  low  probability  of 
occurrence, but which must nonetheless be considered, and 
which  imply  shut  down  and  appropriate  inspection  of  the 
component  or  of  the  plant;  (3)  highly  improbable  operating 
conditions, whose consequences on components are stud-
ied  among  others  for  safety  reasons.  Severe  accident  con-
ditions are not necessarily part of design codes, although In 
the aftermath of Fukushima this is being discussed (“Design 
Extended Conditions”).

38  HLM compatibility is not included in any code today and 
it  is  quite  urgent  to  develop  codes  and  standards. The  two 
key HLM related degradation mechanisms are corrosion (by 
oxidation and dissolution) and liquid-metal embrittlement.

2. Nuclear materials for GEN IV systems and research approaches42 

Such accident scenarios generally translate 
into load excursions with high temperatures 
and  load  rates. The  temperature  range  for 
mechanical  properties  assessment  needs 
therefore to be extended and strain rate prop-
erties  are  necessary  in  the  relevant  range. 
When such data do not exist, specific research 
programmes are required. 

In this context materials qualification means 
generation and maintenance of evidence to 
ensure that material or equipment will op-
erate  on  demand,  under  specified  service 
conditions, meeting system performance re-
quirements.39

Materials qualification is the pre-requisite for 
the establishment of robust design rules for 
structural components and fuel performance 
codes and is therefore defined as pre-norma-
tive research, where pre-normative refers to 
research aimed at establishing standards or 
common procedures to ensure that materials 
and components in nuclear reactors are de-
signed and operated in accordance with the 
best available engineering practices and the 
current scientific knowledge.

Materials qualification requires dedicated 
experiments both out of pile and in pile,40 to 
collect comprehensive and reliable data of the 
material properties relevant for the operational 
conditions.

Even for out-of-pile experiments, suitable in-
frastructures are needed where, for example, 
candidate materials are exposed for increasing 
times to a wide range of temperatures, in con-
tact or not with specific coolants, which may 
be stagnant or flowing. After or during expo-
sure, these materials need to be suitably tested 
and examined to verify the level of degradation 
they exhibit, in terms of changes of properties 
of engineering interest. 

Test standards are developed for this purpose 
by organizations such as ASTM, CEN and ISO 
that prescribe how tests should be conducted 
and how the data should be analysed to assure 
consistent material properties, irrespective of 
where the tests are performed.

Since exposure times are necessarily limited, 
accelerated tests are often a requirement. 
In some cases of interest for GenIV systems, 
the  characterisation  may  necessitate  the 
identification of new and bespoke standard 
procedures  to  execute  the  exposure  and, 
especially, the tests. The procedures are par-
ticularly  critical  in  the  case  of  accelerated 
tests, because their relevance to real operat-
ing conditions needs to be proven.

In-pile  experiments,  in  principle,  imply  re-
peating the same type of exposures under 
irradiation,  and  preferably  in fast  neutron 
spectra for the case of GenIV systems, up to 
the dose expected in service. In the absence 
of suitable facilities, Material Testing Reactors 
(MTRs), with predominantly thermal neutron 
spectrum, can be used, although irradiation is 
limited to lower dose, so that the possibility 
of safely extrapolating to different spectra and 
higher doses is essential.41 

Eventually, the data gathered for each candi-
date material through this expensive procedure 
need to be rationally translated into robust 
design rules for components and laws and 
models for fuel performance codes. 

For structural materials, data from standard 
tests are processed from a design standpoint; 
for instance a lower envelope is obtained from 
scattered material data for fatigue curves or 
accelerated  laboratory  creep  data  are  ex-
trapolated to the operational conditions that 
correspond  to  lower  stresses  and  temper-
atures.  Design  rules  eventually  comprise 
closed-form equations and strict criteria, to-
gether with basic material properties. They are 

39  Since  the  requirements  for  nuclear  applications  are  very 
strict  and  the  operational  conditions  harsh,  special  nuclear 
grades  of  materials  are  often  developed, AISI  316L(N)  is  an 
example. 

40  Pile  is  the  core  of  a  reactor,  where  the  materials  are  ex-
posed to neutron irradiation.

41  No neutron irradiation facility currently exists anywhere in 
the world  that  allows  doses  on  the  order  of  100  dpa  to  be 
reached  within  reasonable  times  and  at  affordable  costs. 
Such facilities did exist in the past, though.

Strategic Research Agenda of the EERA JPNMdeveloped to be conservative and simple to 
apply at the design stage. An important con-
cept is “allowable stresses”; these depend on 
the material and the temperature and should 
ensure that deformations are in the elastic re-
gime. Eventually, design rules are collected in 
design codes,42 together with materials spec-
ifications & design data for different materials 
and components, by making reference to test 
standards and qualification procedures. 

The  RCC-MRx  design  code  has  been  de-
veloped  specifically  to  support  the  SFR 
technology and has been identified as the 
most appropriate design code for all ESNII 
reactors.  It  consists  of  a  single  document 
that covers in a consistent manner the de-
sign and construction of components for high 
temperature and research reactors and the 
associated auxiliaries, examination and han-
dling  mechanisms  and  irradiation  devices. 
The design rules were developed to cover the 
mechanical resistance of structures close to 
neutron sources that can, depending on the 
situation, also operate in significant thermal 
creep conditions. It also includes materials 
manufacturing specifications.43 

RCC-MRx  does  not  contain  any  specific 
rules for environmental effects except thin-
ning (by corrosion), nor does it cover, yet, the 

42  The  two  main  design  codes  are  RCC-MRx  developed  by 
AFCEN and ASME Boiler and Pressure Vessel Code 2015 Edi-
tion. The R5 for high-temperature and R6 for defect assess-
ment  developed  in  the  UK  are  the  most  used  assessment 
procedure codes in Europe.

43  RCC-MRx is divided in three main sections: Section I con-
tains general provisions common to the entire code; Section 
II  gives  additional  requirements  for  the  alternative  use  of 
rule sets applicable for non-nuclear classified components 
and  special  instructions  for  component  subject  to  specific 
regulation; Section III is a set of applicable rules organized 
in 6 Tomes.

 y Tome  1  addresses  design  and  construction.  Subsection 
Z with  20  technical  appendices  contains  for  instance  ba-
sic design material properties, welded joints, elasto-visco 
plastic analysis and defect assessment.

 y Tomes 2 to 5 contain the rules for various technical areas: 
procurement  of  metal  products;  destructive  and  non-
destructive  examination  methods;  welding  requirements; 
other  fabrication  operations  such  as  cutting,  forming  and 
surface treatments.

 y Tome 6 contains a collection of probationary phase rules 
which do not yet have sufficient feedback in the standard 
code. This corresponds to ASME’s “Code Cases”.

43 

high temperature area for GFR and (V)HTR; 
moreover, material property curves and de-
sign rules are based on 40 years operational 
life. RCC-MRx is updated every three years. 
Other  codes  include  the  R-codes  devel-
oped to support the plant life management 
of the UK reactors; R5 is for high-tempera-
ture problems and R6 is primarily for defect 
assessment. There are some commonalities 
between  the  RCC-MRx  appendices  (note 
44) and the R-codes. It should be noted that 
Design Codes such as RCC-MRx and ASME 
BVPH only address structural materials and 
components. Hence fuel claddings are not 
included in these codes.

In  the  case  of  the fuel element, which  in-
cludes the cladding and other parts, the main 
design  tools  are fuel  performance  codes 
(FPC) that enable the simulation of the ther-
mal and mechanical behaviour and evolution 
of the fuel element in reactor as a function 
of the irradiation and thermal parameters in 
normal operational, incidental and accidental 
conditions.44 To this aim, these codes solve 
coupled partial differential equations govern-
ing heat transfer, stresses and strains in the 
fuel element, the evolution of isotopes and 
the behaviour of various fission products in 
the fuel rod with boundary conditions defined 
by the reactor operational conditions. These 
equations involve material properties for the 
fuel and the cladding, which evolve with the 
residence time in reactor. Models are there-
fore needed to describe the very complex 
relationships between the evolution of these 
properties and the relevant parameters, es-
pecially the temperature, composition and 
microstructure. 

Such modelling is essential to understand 
and interpret the measurements carried out in 
reactors and the results of post-irradiation ex-
aminations, to predict the behaviour of specific 

44  Fuel  assemblies  generally  need  full  qualification  in-pile 
and  are  not  included  in  design  codes.  There  is  an  AFCEN 
code for fuel assemblies, RCC-C, “Design and Construction 
rules  for  Fuel  Assemblies  of  PWR  Nuclear  Power  Plants”, 
but it is restricted to PWR and it is essentially a synthesis of 
best practices of the French industry for fuel fabrication, with, 
however, very few quantitative criteria.

2. Nuclear materials for GEN IV systems and research approaches44 

fuels and/or in specific operating conditions 
and to demonstrate the satisfactory behaviour 
of fuels in all operational situations to support 
safety reports. 

The development of FPC started quite em-
pirically, relying on simple laws derived from 
experimental data. Progressively, more and 
more irradiation data were capitalized in the 
codes and more physics was introduced in 
them. It is now universally recognized that a 
better understanding of the underlying phe-
nomena of fuel behaviour is the prerequisite 
for a significant improvement of the codes.

The main European codes for GenIV reactors 
are TRANSURANUS45 developed by JRC in 
collaboration with various institutions through 
Europe and the GERMINAL code46, which is 
part of the PLEIADES platform47, co-devel-
oped by CEA and EDF. 

Figure 5 schematically illustrates the process 
of materials qualification and generation of 
design rules and laws that should enter codes. 
The relevant infratechnology described here 
corresponds to the first line in the warp of 
Figure 4: “Exposure, testing and measuring, 
data correlation, elaboration/&standardiza-
tion of test procedures”, concisely referred to 
as materials qualification. 

2.3.2.  Advanced materials modelling and 

characterisation

Exposing, testing and measuring to collect 
data of engineering relevance is essential, 

45  Lassmann, TRANSURANUS: a fuel rod analysis code ready 
for use, J. Nucl. Mater. 188 (1992) 295; Di Marcello et al., Exten-
sion  of  the TRANSURANUS  plutonium  redistribution  model 
for fast reactor performance analysis, Nucl. Eng. Design, 248 
(2012) 149.

46  Lainet  et  al.,  Current  status  and  progression  of  GERMI-
NAL  fuel  performance  code  for  SFR  oxide  fuel  pins,  IAEA-
CN-245/222, FR17, Yekaterinburg, Russia,  2017; Lainet  et  al., 
Recent  modelling  improvements  in  fuel  performance  code 
GERMINAL for SFR oxide fuel pins, IAEA-CN-199/241, FR13, 
Paris, France, 2013.

47  Marelle, Validation of PLEIADES/ALCYONE 2.0 fuel perfor-
mance code, Water Reactor Fuel Performance Meeting, Jeju, 
Korea, 2017.

but may not be sufficient or affordable. The 
exposure to real conditions in the laboratory 
is challenging in terms of infrastructure, time, 
costs, and also know-how and the data col-
lected cannot cover all possible conditions. In 
particular, exposure times comparable with 
the lifetime of the reactor,48 or high irradia-
tion levels such as those reached by cladding 
and fuel, are hardly accessible in a labora-
tory, or the dose-rate will be much higher. 
Also,  the  combination  of  effects  and  their 
synergy is difficult to reproduce in a labora-
tory. So, extrapolation of data is unavoidable, 
but purely empirical extrapolations may be 
difficult or even unreliable. Nevertheless, in 
the past materials have been largely quali-
fied in this manner, describable as a paradigm 
of “observe and qualify”: the observation of 
the materials performance under a variety of 
conditions was the main ingredient in their 
qualification and licensing. This practice is 
still used today and will continue to be used, 
but  is  gradually  undergoing  a  paradigm 
shift, whereby the materials are subjected to 
“design and control”. This paradigm shift is 
expected to mitigate costs and reduce lead 
times to deployment. 

At the heart of the design and control par-
adigm lies a greater reliance on advanced 
modelling and simulation, partly generated 
by improved theory, but also by the vastly 
increased computational power of the last 
decades, crucially coupled with advanced 
microstructural  and  micromechanical 
characterization, using ever more powerful 
techniques  for  materials  examination  and 
testing  at  all  scales. This  approach,  lately 
denoted as integrated computational mate-
rials engineering (ICME),49 bears the promise 
of  providing  increased  robustness  in  the 
long term, initially only underpinning, then 
gradually improving and finally replacing tra-
ditional empirical approaches, such as those 

48  1/3  is  considered  sufficient  for  extrapolation,  but  this  can 
also be a challenging duration/dose.

49  ICME-based  microstructure-based  description  of  the 
deformation of metals: theory and application, in: Helm, Butz, 
Raabe  and  Gumbsch,  Microstructure-Based  Description  of 
the Deformation of Metals: Theory and Application. JOM 63 
(2011) 26. http://www.dierk-raabe.com/icme/

Strategic Research Agenda of the EERA JPNM45 

Figure 5: Schematic representation of the main elements that enter the process of generation of component 

design codes and fuel performance codes.

used in current fuel performance codes or in 
dose-damage correlations for LWR vessels. 

The  ICME  approach  aims  to  reach predic-
tive capability by being able to describe in 
a physical way, possibly with a single tool 
(e.g. a Monte Carlo or a phase field code) fed 
by more fundamental models or calculation 
methods (e.g. density functional theory), the 
evolution in time of both the microstructure 
(redistribution  of  lattice  defects)  and  the 
microchemistry  (redistribution  of  chemical 
species) of materials exposed to irradiation 
and/or high temperature and/or coolants. 
The output of these microscopic evolution 
models should then be the input of models 
that operate at the meso- and macroscopic 
length  scales  and  predict  accordingly  the 
corresponding changes of materials behav-
iour and properties, in a multiscale modelling 

framework and spirit.50 For example, in order 
to bridge from the microscopic description to 
higher scales to inform continuum mechanics, 
models that describe dislocations (involved in 
all plastic deformation processes, from tensile 

50  As an example, here is an idealised chain of models to pre-
dict radiation effects at the microscopic scale:

 y Models  to  predict  damage  production  in  collision  casca-
des: techniques of application here are electronic structure 
and interatomic potential calculations using molecular sta-
tics and dynamics.

 y Atomistic  models  that  allow  mobility  and  stability  to  be 
studied for all radiation defects involved in the microstruc-
tural and microchemical evolution. Techniques of applica-
tion  are  as  above,  but  extend  to  atomistic  kinetic  Monte 
Carlo models.

 y Parameters and mechanisms determined as above are the 
base  of  microstructural/microchemical  models  that  de-
scribe how radiation damage evolves versus dose. Tech-
niques of application here are object kinetic Monte Carlo, 
rate theory-based cluster dynamics, phase field, … 

2. Nuclear materials for GEN IV systems and research approaches46 

properties to creep) are generally needed.51 
The  modelling  tools  are  generally  costly-
to-run and often use parallelized computer 
codes, so the use of high performance com-
puting (HPC) is expected to be a crucial asset 
in many cases.

All  these  models  require  suitable data  for 
validation and also calibration from mod-
elling-orientated  experiments. These  are 
experiments in which materials are exposed 
to external factors, as for qualification pur-
poses,  but  are  designed  to  understand 
mechanisms, by separating variables and 
effects,  rather  than  to  measure  engineer-
ing properties. Experiments of this type, in 
which key variables such as temperature, ir-
radiation-dose and dose-rate need accurate 
control and variation over sufficiently wide 
ranges,  are  invariably  delicate  to  perform, 
and may be long and costly. As in the case 
of qualification, specific facilities are needed 
for exposure, in particular for irradiation and 
subsequent characterisation. In this context, 
the use of charged particle irradiation (ions, 
protons, electrons, …) is a very valuable and 
affordable tool.

To fully characterise the microstructure of 
the materials exposed, the use of combined 
modern characterization techniques is cru-
cial, e.g. transmission electron microscopy 

51  For example the following combination may be used:

 y Calculation of the interaction of gliding and climbing dis-
locations with both pre-existing (e.g. grain boundaries) and 
radiation induced microstructural features (not only single 
point  defects  but  also  voids,  precipitates,  solute  clusters 
and dislocation loops), as a function of type of feature, type 
of dislocation, temperature and strain rate. Techniques of 
relevance  here  are  atomistic  ones,  especially  molecular 
statics and dynamics with interatomic potentials, suppor-
ted by electronic structure calculations.

 y Transfer the above understanding in the form of local ru-
les, to dislocation dynamics models, applicable at single-
crystal level.

 y Derive  and/or  parameterize,  from  dislocation  dynamics 
and  other  tools,  appropriate  constitutive  laws  for  crystal 
plasticity  models,  applicable  to  aggregates  of  crystals 
(grains), that should contain and reproduce the effects of 
the  microstructural/microchemical  changes  due  to  irra-
diation or thermal ageing.

 y Identify  suitable  homogenization  schemes  to  apply  con-
tinuum plasticity approaches and evaluate the mechanical 
behaviour at component level.

(TEM)  in  all  of  its  multiple  forms,52  atom 
probe tomography (APT), small angle neutron 
scattering (SANS), positron annihilation spec-
troscopy  (PAS),  and  many  others.  Suitable 
mechanical characterization experiments are 
equally crucial, including micromechanical 
characterization from specimens at single 
grain scale (nanoindentation, micro-pillars, 
…) with different crystal lattice orientations. 
Micromechanical characterization techniques 
are often the only possibility for charged par-
ticle irradiated specimens, due to the limited 
penetration  of  these  particles.  In  addition, 
mechanical  tests  addressing uni- vs  mul-
ti-axial load, cyclic load, relaxation, load 
sequence, non-proportional loading, etc. are  
of interest.

Since the requirements of experimental pre-
cision, accuracy and completeness for model 
validation  are  extremely  high,  it  is  impor-
tant to establish, for all the characterization 
techniques used, accepted best practices, 
protocols and possibly standards for their 
application  and  for  the  analysis  of  the  re-
sults. This should allow full inter-laboratory 
comparability and provide high guarantee of 
reliability, to reduce scatter and uncertainty. 
The analysis of characterization results is often 
helped by models and software, often irre-
placeably: these models should be improved, 
enlarging their physical bases, and the codes 
benchmarked, in order to estimate and once 
again reduce scatter and uncertainty.

Advanced computational tools such as artifi-
cial intelligence/machine learning (AI/ML) 
can be very valuable in this modelling frame-
work, either to help bridge through scales, 
whenever  the  complexity  of  a  lower  scale 
(for example chemical complexity or micro-
structural complexity) needs to be reduced 
to a limited number of variables that affect 
the higher scale, or to by-pass model chain-
ing, through training on examples obtained 
from  complex  models. Additionally,  these 

52  E.g. high resolution TEM (HRTEM), or scanning TEM (STEM), 
coupled to energy dispersive X-ray (EDX) or electron energy 
loss spectroscopy (EELS), to identify and analyse defects and 
also measure the local chemical composition.

Strategic Research Agenda of the EERA JPNM47 

Figure 6: Modelling and experimental characterisation techniques at various time and length scales involved in 

the ICME approach, used to elucidate the basic mechanisms of nuclear material behaviour and ageing, and their 

coupling.

methods can provide support to the analy-
sis of experimental data. In particular AI/ML 
is a very powerful technique for the analysis 
of big data, e.g. from extensive experimen-
tal measurements, in order to deduce a logic 
to be used to guide or underpin models and 
design correlations, but also to improve fabri-
cation routes or materials compositions and/
or architectures (see section 2.3.3). Similarly, 
AI/ML  may  assist  online  non-destructive 
monitoring of materials and components to 
help predict failure.

Figure 6 illustrates in a schematic way the idea 
of the ICME approach, as well as the model-
ling and experimental techniques involved. 
The relevant infratechnology here introduced 
corresponds to the second line in the warp of 
Figure 4, defined as “Advanced modelling and 
microstructural characterization for predictive 
capability”, which can be concisely referred to 
as modelling. 

2.3.3.  Development of advanced material 

solutions

For  some  components  or  parts  of  compo-
nents, there may be no design-driven ageing 
mitigation strategy identified that fully guaran-
tees the materials’ fitness for purpose over a 
sufficiently long timespan, thereby penalising 
the efficiency and/or the economy of the sys-
tem. The component may then have reduced 
availability (because of the requirement of 
frequent inspections) or too short lifetime 
(because of materials’ expected rapid deg-
radation), or offer too little safety margin. In 
this case there is a strong push to find new 
material solutions instead of, or in addition 
to, design solutions. A further push towards 
new materials solutions comes from the need 
to reduce the cost of material or component 
fabrication and also to enhance the economy 
of the plant within ample safety margins by 
further reducing the frequency of inspections 

2. Nuclear materials for GEN IV systems and research approaches48 

and therefore shutdowns and increasing the 
lifetime of components.

New material solutions may correspond to: 
(i) select a different existing material of the 
same class with better properties for a specific 
application; (ii) suitably protect the selected 
material with functional materials to reduce its 
degradation; (iii) improve the inherent proper-
ties of existing materials or classes of materials, 
either by changing their composition (e.g. al-
loys) and/or architecture (e.g. composites), by 
applying different post-fabrication treatments 
(e.g. thermal treatments in steels), or else by 
using different fabrication processes; and (iv) 
select from an entirely different class of ma-
terials or, if possible, develop new materials 
in a targeted way. 

This situation applies specifically to commer-
cial GenIV reactors. It is very unlikely that the 
high fuel burnups and operating temperatures 
targeted in these systems, which are in contact 
with non-conventional coolants, can be fully 
achieved with  existing  materials  under  the 
overarching principle of safety. It is well-known 
that GenIV  demonstrators will necessarily 
operate at conditions that are not optimal in 
terms of efficiency and waste minimisation, 
and at smaller scale and power than eventual 
commercial reactors. This is the only way to be 
compatible with the conditions that materials 
can be expected to tolerate, based on the ex-
isting return of experience and the available 
qualification data. Moreover, protection against 
degradation (e.g. against corrosion) through 
materials solutions is likely to be needed for 
the demonstrators. To ensure the commercial 
deployment of truly GenIV systems with a de-
sign lifetime of 60 years and able to sustain 
high fuel burn-ups and transmutation, new 
material solutions are needed.

A  substantial  research  and  development 
and innovation effort is required, leading to 
the identification of effective material solutions 
to mitigate the consequences of harsh service 
conditions, the improvement of the properties 
of existing materials through new fabrication or 
treatment processes and, if possible, the elab-
oration of completely new materials.

The elaboration of new materials solutions 
is,  as  always,  largely  based  on  previous 
experience: often via the adaptation and ap-
plication to a different technology of solutions 
that have already been successfully used in 
other technologies where similar problems 
had to be faced. For instance, the applica-
tion of coatings is a quite obvious solution 
to protect against environmental aggression 
(corrosion); and procedures successfully used 
to improve the creep strength at high tem-
perature of steels used in gas- or coal-fired 
plants may also appear as a good solution 
for components that are subjected to irradia-
tion. However, the transfer of solutions from 
other technologies must be done within the 
constraints of nuclear applications, where 
irradiation represents a very specific type of 
load on materials (either for property degra-
dation or transmutation/activation) and safety 
requirements are particularly strict. 

Each time a new materials solution, as op-
posed to a design solution, is proposed for 
a  nuclear  reactor,  a  long  process  of  full 
material  qualification  and  codification  is 
required. Irradiation in suitable facilities and 
subsequent testing under all those conditions 
that may put safety in question, makes the 
process currently very long, of the order of 
decades. Moreover, before a new material 
or material combination is introduced, effi-
cient procedures for joining pieces made of 
that material need to be developed and also 
tested and qualified. Finally, materials solu-
tions  and  joining  procedures  are  typically 
developed in the laboratory: crucially, before 
the solution is adopted there needs to be an 
industrially upscaled fabrication and joining 
procedure. Clearly, these steps can only be 
taken for a very reduced number of materials, 
which have emerged from a selection based 
on a previous screening.53

53  A  positive  consequence  of  the  strictness  of  the  qualifi-
cation  procedure  that  needs  to  be  undergone  for  nuclear 
application, combined with the especially harsh conditions 
that  materials  face  in  nuclear  reactors,  especially  GenIV 
systems,  is  that  the  materials  solutions  developed  in  this 
framework, where safety is an imperative, are likely to be of 
use also for other technologies where harsh conditions are 
expected.

Strategic Research Agenda of the EERA JPNMThe screening is currently performed essen-
tially in the same way as for existing materials 
that need to be qualified for the demonstra-
tors, i.e. by exposure and testing (the “observe 
and qualify” paradigm). The only difference is 
that the first goal here is not to fully define 
the correlations and rules that will allow the 
design, licensing and construction of a re-
actor, but only to give a first assessment of 
the behaviour of the candidate materials to 
identify the most suitable one(s). It would be 
very desirable for the screening process to be 
done at an accelerated pace. Accelerating 
the screening of materials is very much con-
nected with the shift towards a “design and 
control”  paradigm.  This  implies  address-
ing two aspects: (i) identifying experimental 
methods  to accelerate  the  exposure  and 
subsequent testing and characterisation of 
materials; and (ii) making use of advanced 
modelling and characterization techniques 
to guide the development of new materials, 
by using a targeted quantitative methodology 
rather than a trial-and-error approach that is 
based on the experience or intuition of the 
researchers involved.

In  the  case  of  nuclear  materials, charged 
particle irradiation can be a suitable method 
to accelerate exposure, also in combination 
with other parameters (temperature and en-
vironmental attacks), provided that: (i) suitable 
characterization  techniques,  including  for 
mechanical characterization, are identified, 
developed and, importantly, standardized or 
at least guided by accepted best practices 
and protocols; and (ii) models are developed 
that allow the assessment of the effect of ir-
radiation by particles that are different from 
neutrons (accounting for effects of spectrum, 
dose-rate,  limited  penetration,  injection  of 
species, …etc.). This links very much with the 
support offered by the modelling infratech-
nology  (section  2.3.2),  as  well  as  with  the 
accelerated testing envisaged for pre-nor-
mative research (section 2.3.1). Accelerated 
characterization  would  be  best  achieved 
by  the  creation  and  availability  of  suitable 
integrated  test-beds,  possibly  including 
exposure and characterization in the same 
package.

49 

An  even  stronger  link  with  the  modelling 
comes from the ambition to target the de-
velopment of materials solutions based on 
calculations, for instance, by identifying in 
advance a range of promising material com-
positions  or  architectures  for  the  relevant 
application. Whether  these  are  reliable  or 
not will depend enormously on the level of 
reliability achieved by the ICME approach (de-
scribed in section 2.3.2) as well as, importantly, 
on which properties are of interest, since in 
many cases the prediction by calculation of 
mechanical properties may not be reliable, or 
even not possible. Here, too, AI/ML may be 
of use, depending on whether a sufficiently 
large amount of experimental data becomes 
available, by allowing extrapolation or interpo-
lation to identify the best materials features to 
maximise a certain measurable property. HPC 
is here also an important asset.

The combination, possibly in an automated 
way, of accelerated screening with the use of 
calculations corresponds to the use of high 
throughput approaches for materials devel-
opment, particularly if integrated fabrication 
methods (for instance additive manufacturing) 
are included.54 In practice, this is unlikely to be 
immediately applicable in the case of the de-
velopment of nuclear materials, at least in the 
near future, because of the complexity of the 
combined exposure to irradiation, temperature 
and environmental aggression, the size of the 
components involved and the high reliability 
required for long exposure times. Moreover, 
these  approaches  do  not  eliminate  the  re-
quirements to license a material solution for a 
nuclear installation that are associated with the 
strictness of the safety authorities. It is, howe-
ver, worth considering these new approaches, 
which are emerging in the field of materials 
science at large.

The relevant infratechnology here described 
corresponds to the third line in the warp of 
Figure 4, defined as “design, manufacturing and 
processing of innovative and better performing 

54  E.g.  Chen,  High-Throughput  Computing  for  Accelerated 
Materials Discovery, in: Computational Materials System De-
sign, Shin and Saal Editors, Springer (2018).

2. Nuclear materials for GEN IV systems and research approaches50 

Figure 7: Illustration of the process of development of new materials solutions and its desirable acceleration.

Strategic Research Agenda of the EERA JPNMmaterials”, which can be concisely referred to 
as advanced materials. Figure 7 illustrates the 
procedure for materials development and its 
desirable acceleration.

2.4.  Data Management

Trends in the research sector, such as the adop-
tion of data management policies by funding 
agencies55,56, publishing houses57 and partner 
organizations58, the emergence of data jour-
nals59,60 and frameworks for data citation61, are 
indicative that data management is becoming 
an integral component of the mainstream re-
search process. While the specific reasons for 
this circumstance are varied, the underlying 
motivations are improved science and greater 
opportunities for innovation. Self-evidently, 
membership organizations that rely on data to 
achieve their objectives stand to benefit from 
improved data management practices. This is 
certainly the case for the EERA JPNM, where 
large volumes of materials’ testing, character-
isation and modelling data are generated that 
are of inherently high intellectual and/or com-
mercial value.

In the case of EERA JPNM, the lack of data is a 
significant barrier to the formulation of design 
rules to be put in design codes and in feeding 
fuel performance codes, whether the data are 

55  Data  management  -  H2020  Online  Manual.  Retrieved  29 
May  2018  from  http://ec.europa.eu/research/participants/
docs/h2020-funding-guide/cross-cutting-issues/open-ac-
cess-data-management/data-management_en.htm

56  Funders’ data plan requirements. Retrieved 29 May 2018 
from  http://www.dcc.ac.uk/resources/data-management-
plans/funders-requirements

57  Availability  of  data  &  materials:  authors  &  referees  @  npg. 
Retrieved 29 May 2018 from https://www.nature.com/authors/
policies/availability.html

58  ILL Neutrons for Society - Data Management. Retrieved 29 
May  2018  from  https://www.ill.eu/users/user-guide/after-
your-experiment/data-management

59  Data  in  Brief  - Journal  -  Elsevier.  Retrieved  29  May  2018 
from https://www.journals.elsevier.com/data-in-brief.

60  Scientific  Data.  Retrieved  29  May  2018  from  https://www.
nature.com/sdata

61  Welcome to DataCite. Retrieved 29 May 2018 from https://
datacite.org

51 

needed to trace empirical curves or to validate/
calibrate models. Historically, large amounts of 
test and measurement data have been gener-
ated through national and international research 
programmes, but these data are often not avail-
able or, if available, they are not complete (for 
instance data for time to creep-rupture, but 
not the creep-curve itself). While appropriate 
web-enabled databases have been developed, 
e.g. the materials database MATDB of Online 
Data & Information Network of the European 
Commission Joint  Research  Centre62  or  the 
International Fuel Performance Experiments 
database held by the OECD,63 often data are not 
made available because they are protected by 
confidentiality and therefore cannot be shared, 
or in other cases because they were not prop-
erly stored. Furthermore, in ongoing and future 
research programmes, a very large share of the 
effort, and the cost, will be dedicated to material 
testing and property measurement, including 
novel materials. Since experimental data provide 
the basis for design curves and rules, as well as 
for the development and validation/calibration 
of models, their availability must be guaranteed 
with a view to their re-assessment in the context 
of new models, different operating conditions, 
different regulatory requirements, etc. Given 
that such data are also expensive to generate, 
it should become standard practice for the data 
to be collected, preserved, and made readily 
accessible, with respect to intellectual property 
rights, of course. The latter issue may require 
agreements, e.g. in terms of embargo periods, 
before the data become disclosed. 

With improved data management depending 
on so many (sometimes competing) factors, 
best practices are evolving that are designed 
to support improved data management; these 
include data management plans64,65 and the 

62  https://odin.jrc.ec.europa.eu/odin/index.jsp 

63  https://www.oecd-nea.org/science/wprs/fuel/ifpelst.
html 

64  DMPonline | Digital Curation Centre. Retrieved 29 May 2018 
from http://www.dcc.ac.uk/dmponline

65  EUROPEAN  COMMISSION  Directorate-General  for  Re-
search  &  Innovation,  Guidelines  on  FAIR  Data  Management 
in  Horizon  2020  (2016).  Retrieved  29  May  2018  from  http://
ec.europa.eu/research/participants/data/ref/h2020/
grants_manual/hi/oa_pilot/h2020-hi-oa-data-mgt_en.pdf

2. Nuclear materials for GEN IV systems and research approaches52 

FAIR (Findable, Accessible Interoperable and 
Re-usable)  Guiding  Principles  for  scientific 
data management and stewardship66. The FAIR 
guiding principles are concerned with data 
becoming findable, accessible, interoperable 
and reusable, so probably the most pressing 
need in the engineering materials sector is for 
data interoperability. In this context, standards 
for engineering materials data are needed to 
ensure consistency of the data coming from 
different laboratories and to facilitate transfer 
between different information systems. There is 
therefore a urgent need to develop standard 

formats for engineering materials data and 
to leverage the highly interconnected infor-
mation and communication technology (ICT) 
infrastructures that have emerged in recent 
years, with a view to ensuring that the data 
on which design curves and rules are based 
remain available for the purposes of future 
re-assessment. It should also become estab-
lished practice for projects to maintain a data 
management plan beyond the project duration, 
so that the resources for data management 
can be allocated and concrete deliverables 
identified.

66  Wilkinson  et  al., The  FAIR  Guiding  Principles  for  scientific 
data management and stewardship (2016), Scientific Data, 3, 
doi:10.1038/sdata.2016.18.

Strategic Research Agenda of the EERA JPNM3

NUCLEAR 
MATERIALS 
RESEARCH 
AGENDA

© CEA

3.  NUCLEAR 

MATERIALS 
RESEARCH 
AGENDA

This section describes the materials research 
issues that need to be addressed to provide 
system designers and component manufac-
turers with suitable qualified materials and 
relevant design rules/fuel performance codes 
for the licensing and construction of GenIV re-
actor demonstrators, FOAK prototypes and 
eventually commercially plants. The two main 
subsections concern structural (3.1) and fuel 
(3.2) materials, i.e. the ‘weft’ of the ‘fabric’ of the 
SRA as in Figure 4. In each subsection the issues 
are organised according to the three comple-
mentary approaches illustrated in Figure 2 and 
described in section 2.3. namely: pre-normative 
materials research and qualification, advanced 
materials modelling and characterisation, and 
development  and  screening  of  advanced 
materials solutions. These correspond to the 
‘warp’ of the SRA ‘fabric’ in Figure 4.

3.1.  Structural materials

3.1.1.  Pre-normative research: qualification, 

test procedures, design rules

This  section  concerns  the  development  of 
the assessment tools and the collection of 
the structural materials data in support of the 

design, licensing, and construction of the ESNII 
demonstrators, with a view also towards the 
FOAK prototypes and commercial reactors to 
be deployed later.67 

An especially delicate issue in this context 
is  the extension  of the operational life  of 
non-replaceable  components  from  40  to 
at  least  60 years  as  a  general  Gen  IV  re-
quirement. 68  All  relevant  slow  long-term 
degradation processes need to be accounted 
for, especially high-temperature processes 
(creep,  fatigue,  thermal  ageing),  but  also 
corrosion and low-dose-rate long-term irra-
diation effects. A fundamental issue is how to 
measure material properties representative 
of long-term operation. This is a tremendous 
challenge shared by all Gen IV concepts, for 
which design and assessment methodologies 
need  to  be  developed  and  translated  into 

67  Pre-normative research has been a key objective in the FP7 
projects MATTER and MATISSE (http://www.fp7-matisse.eu/), 
and is the key focus in the H2020 project GEMMA (http://www.
eera-jpnm.eu/gemma/) and in ongoing JPNM Pilot Projects. 
See in particular the summary report from the MATTER project 
for an overview of work done and envisaged: “Deliverable 9.1: 
The relevance of MATTER results for the design of ESNII reac-
tors”, http://www.eera-jpnm.eu/gemma/

68  Technology  Roadmap  Update  for  Generation  IV  Nuclear 
Energy Systems, January 2014, https://www.gen-4.org/gif/
upload/docs/application/pdf/2014-03/gif-tru2014.pdf 

56 

codes and standards. For instance, for creep 
an extrapolation by a time factor 3 is consid-
ered feasible, but this would require tests of 
20 years duration for reliable 60 years design 
data. In a future low-carbon energy mix, the 
nuclear contribution will also need to operate 
in a load-following mode, whereby the reac-
tor components will accumulate more load 
cycles, which should also be taken into ac-
count in a long-term operation perspective. 
Although the 60-years operational life is a re-
quirement for commercial deployment, and is 
thus a long-term need, there is a short-term 
R&D need to start in due time, i.e. now, in view 
of long-term test programmes69. 

Specifically,  for  the  HLM-cooled  systems, 
the most important issue is to demonstrate 
the  structural  integrity  of  components  in 
prolonged contact with a corrosive coolant. 
There are no dedicated design rules, assess-
ment procedures or material data in the codes 
to address the HLM environmental effects. 
Thus,  there  is  an  urgent  need to  perform 
pre-normative research on the compatibility 
between components and HLM coolants and 
develop procedures to be included in codes 
and standards.70 

Importantly, most structural integrity issues 
in metallic components occur at the welds, 
the assessment of which is especially com-
plex. This means that much emphasis in the 
pre-normative  research  should  be  dedi-
cated to welded components. 

Finally, the high investment cost is perhaps 
the  largest  obstacle  for  the  development 
and  deployment  of  innovative  nuclear  re-
actors. Thus, in addition to demonstration 
of safety, the codes and standards should 

69  See Final Report EERA JPNM Task Force 60 years Oper-
ational  Life  Future  Reactors  Review  and  Roadmap  for  Future 
Activities,  available  from  http://www.eera-jpnm.eu/fileshar-
er/documents/ and Technical Area 1a of the MoU between 
EERA and SNETP: Development of a methodology for design 
and  plant  life  management  for  60-years  operational  life  for 
non-replaceable components of Generation IV reactors

70  See  Technical  Area  1b  of  the  MoU  between  EERA  and 
SNETP:  Development  of  Design  Rules,  qualification  and  miti-
gation strategies for reactor components in heavy liquid metal 
coolants.

also ensure cost efficiency. Pre-normative 
research related to reducing cost without re-
ducing safety margins include: reduction of 
undue conservatism in design rules by apply-
ing more advanced assessment procedures; 
assessment of alternative materials and de-
signs; more accurate and less conservative 
descriptions  of  loads;  updated  materials 
curves based on additional tests methods as 
well as extended test conditions. This implies 
that conceptually the development of codes 
and standards should be based on a physi-
cal understanding of materials degradation 
processes.

The  pre-normative  research  includes 
closely integrated experiments and mod-
elling work at the meso- and macroscale. 
The engineering scale is the basis for reactor 
design, but the understanding of the degra-
dation  mechanisms  and  of  how  to  include 
them  in  the  design  rules  and  assessment 
procedures requires the proper length and 
time scale to be addressed for the particular 
degradation mechanism. One example is the 
extrapolation of accelerated data to opera-
tional conditions where the proper relevant 
degradation mechanism must be explicitly 
taken into account. The end product would 
then be engineering tools based on a mul-
ti-scale informed approach. This section thus 
includes engineering modelling approaches, 
while more mechanistic approaches are the 
subject of section 3.2.1. 

In what follows the main pre-normative re-
search issues for the key materials of the ESNII 
demonstrators are listed and summarised. 
More details can be found in the Technical 
Annex to the SRA, in section TA.1.1.1, which 
is a more extended version of this one.71 At 
the  end  of  this  section,  Table  3  (page  58) 
further summarizes the main pre-normative 
research issues. The 60-years design lifetime 
is not mentioned anywhere as a specific is-
sue, because it corresponds to a combination 
of several of the issues listed in what follows 
and  in  the  table. The reference  structural 

71  http://www.eera-jpnm.eu/?q=jpnm&sq=nboard 

Strategic Research Agenda of the EERA JPNMmaterials  are  austenitic  steels,  on which 
most research effort is focused, in particu-
lar 316L and 316L(N), that are planned to be 
used for the core of all prototypes.  In the 
case of the GFR, Ni-base alloys are also ref-
erence materials, in particular the high-Ni 
alloy 800H for out-of-core components. Alloy 
800H is qualified up to 750°C for the steam 
generator; but for the other high tempera-
ture components – turbine blades, hot gas 
header, intermediate heat exchangers – Ni-
base alloys with higher creep rupture strength 
are required, e.g. Inconel 617, Haynes 230, or 
Hastalloy-XR.72 Some emphasis remains on 
9Cr F/M steels which, although not used for 
MYRRHA and ALFRED, and only to a limited 
extent  considered  in ASTRID,  remain can-
didate materials for the phase II of some 
demonstrators. For cladding the reference 
material is the 15-15Ti austenitic steel. 

High temperature behaviour and 
degradation of metals

Creep effects need to be taken into account 
above  the  negligible  creep  temperature, 
which is given in RCC-MRx for codified mate-
rials. Many fast reactor components operate 
in the creep regime, where creep, creep-fa-
tigue and thermal ageing are the life-limiting 
factors.73 In addition to the 60-years design 
life, high-temperatures must also be consid-
ered for accidental conditions, well-above the 
normal operational temperature. Accordingly, 
the following issues need to be addressed, 
focusing in particular on the identification of 
mechanisms and development of relevant 
engineering models:

 y creep and creep-fatigue deformation: collect 
data and develop models that incorporate 

72  There has also been effort to develop special ODS alloys 
with high temperature strength and stability (MA6000).

73  For a general overview of high-temperature assessment 
approaches see Comprehensive Structural Integrity, Fracture 
of  Materials  from  Nano  to  Macro Vol.  4  Cyclic  Loading  and 
Fatigue and Vol. 5 Creep and High-Temperature Failure, 2007, 
Elsevier-Pergamon. For nuclear applications see Structural 
Materials for Generation IV Nuclear Reactors, Edited by Pas-
cal Yvon, Woodhead  Publishing  Series  in  Energy,  Elsevier, 
2017.

57 

the dominant degradation mechanisms (e.g. 
non-isothermal creep-fatigue, especially se-
rious in F/M steels that undergo softening 
under cyclic loading) or microstructural evo-
lution at the proper scale;

 y creep and fatigue damage and crack propa-
gation: develop models that incorporate the 
initiation of damage and crack propagation 
until final failure and integrate creep and fa-
tigue with proper damage criteria;

 y thermal  ageing:  include  thermal  ageing 
effects in design codes and assessment 
procedures predicting the microstructural 
evolution with thermodynamic codes; 
 y interaction creep, creep-fatigue and ther-
mal ageing and upscaling: develop mod-
els that combine effect of creep, fatigue 
and  ageing  for  total  life  assessments  of 
components, upscaling from mechanistic 
to polycrystal continuum models through 
homogenization schemes;

 y representative  data  collection  and  produc-
tion  for  60 years  design  life:  (i)  compilation 
and  assessment  of  ‘historical’  data,  (ii)  me-
chanical testing and microstructural analysis 
of materials from reactor components ope-
rated in the past, (iii) new test programmes.

Environmental compatibility between 
coolant and structural materials

HLM-cooled systems: In these systems the 
main safety issue is to guarantee integrity of 
materials in contact with the coolant for the 
lifetime of the component. It is thus necessary: 
(i) to address corrosion, dissolution, erosion by 
mapping relevant rates as functions of all the 
variables involved; (ii) address liquid metal 
embrittlement (LME) by identifying appropri-
ate steps to be taken towards the definition 
of relevant design rules (either demonstra-
tion of immunity, or identification of mitigation 
strategies, or else complete mapping and use 
of coated or alumina-forming materials). The 
resistance to HLM degradation of austenitic 
steels  needs  further  demonstration:  long 
terms tests are needed as well as accelera-
ted tests in more aggressive environment, 
as a complementary shortcut, supported by 
appropriate  models  and  microstructural 

3. Nuclear materials research agenda58 

Table 3: Main issues concerning pre-normative R&D: materials qualification, design rules, assessment & test 

procedures.

Main issue

Breakdown in sub-
issues

Materials concerned Techniques/Methods

Creep, relaxation and 
cyclic deformation

High temperature 
behaviour and 
degradation of 
metals

Creep and creep-
fatigue damage and 
crack propagation

Austenitic steels 
(316L, alloy 800), 
F/M steels (Grade91), 
Ni-base alloys 
(Inconel 617, Haynes 
230, Hastalloy-XR)

Thermal ageing

Liquid Metal 
Corrosion and 
erosion (LMC)

Austenitic 
steels:316L, 15-15Ti 
and 15-15Ti with 
alumina surface 
protection

Environmental 
compatibility 
between coolant 
and structural 
materials

Liquid Metal 
Embrittlement in 
HLM (LME)

Austenitic steels 
316L and alloy 800, 
Ni-based alloys 
(Inconel 617, Haynes 
230, Hastalloy-XR), 
as well as SiC/SiC 
and AFA steels. 

Austenitic and F/M 
steels (irreplaceable 
& structural 
components)

Compatibility with 
HT He

Low temperature 
embrittlement 
& plastic flow 
localisation

Long-term/low 
dose irradiation in 
environment

High dose irradiation 
swelling and creep

Fuel cladding 
materials: 15-15-Ti 
austenitic steel

Radiation effects

Experiments:

For long-term operation: Mechanical tests 
of in-service material, long-term tests, 
accelerated tests.

Basic tests for model calibration: creep, 
low-cycle fatigue, crack propagation 
tests. Special emphasis on long hold 
times.

Microstructural analysis to link 
mechanism-based models to 
experiments are needed.

Models: 

For creep, relaxation and cyclic 
deformation, emphasis on unified visco-
plastic continuum models, mechanistic 
models for different creep mechanism 
and damage crack propagation fracture 
mechanics 

Models need to be translated into Design 
Rules or Assessment Procedures.

Mechanical tests: slow-strain rate tensile; 
fracture, fatigue, and creep-fatigue in 
flowing and stagnant conditions;

Corrosion tests: Erosion and corrosion 
(oxidation and solution tests) in flowing 
conditions

Qualification tests (mechanical and 
corrosion) for 316SS and welded 
components.

Accelerated tests to map bounding 
conditions.

Emphasis on long-term tests; A very 
careful documented control and 
monitoring of the test conditions (in 
particular oxygen control) is required for 
all tests.

Tests to be complemented by detailed 
microstructural analysis (e.g. SEM, EBSD, 
XRD, TEM);

Engineering related approaches need 
to be developed Coupling with models 
developed in 3.1.2

Exposure to irradiation, also including 
coolant environment.

Standard mechanical test in hot cells of 
neutron irradiated materials (irradiated in 
test reactors or in-service exposed) and 
complementary ion/proton irradiation. 

These tests need to be supported by 
irradiation models from 3.1.2

Transfer the data into reduction factors 
for material properties. 

Strategic Research Agenda of the EERA JPNM59 

Materials concerned Techniques/Methods

Main issue

Breakdown in sub-
issues

Non-isothermal 
thermo-mechanical 
loads

Assessment of 
complex loadings

Complex stress 
distributions 

All

Load transients 
and beyond design 
conditions

Residual stresses

Weld procedures

Degradation 
modes and defect 
assessment

Compatibility with 
HLM

All welds: 

austenitic, F/M; Ni-
based alloys and 
dissimilar metal 
welds

Sub-sized/ 
miniaturised 
specimens for 
mechanical property  Fuel cladding 

material, 15-15Ti

Thin-walled cladding 
tubes

Small Punch test

All

Micro-pillar tests and 
nano-indentation

All

Integrity and 
qualification 
of weldments 
and welded 
components

Sub-size and 
miniature 
specimen test 
standardization

Component and 
material health 
monitoring

- - -

All

Experiments:

Component or specimen tests that 
simulate thermo-mechanical loads (e.g. 
thermal shocks), variable amplitude tests 
for fatigue; tensile high strain rate tests, 
multi-axial tests, e.g. cruciform specimen 
for biaxial loading.

Modelling:

Finite element models of complex load 
cases and simplification to translate into 
design rule load cases.

Experiments: 

Residual stress measurements by 
neutron diffraction and simpler but 
less accurate methods such as X-ray 
diffraction, contour method. 

Standard test for welded specimens 
(tensile, fracture toughness tests, fatigue) 

Characterization of the different regions 
of a weld (small punch, nano and micro 
indentation)

Mock-up test of welded component for 
validation. 

Microstructrual analysis (SEM, TEM, XRD, 
EDX)

Modelling:

Simulation of weld process and post-
weld heat treatment for residual stresses;

Structural integrity assessment (defect 
assessment crack propagation, damage) 
of welded specimens and components 
by FEM. 

Translation of structural analysis 
assessment into Design Rules.

Experiments: 

Various fuel cladding tests (internal 
pressure, ring-compression, small-
punch, cone mandrel) with emphasis on 
hot-cell tests;

Small-punch test for tensile and creep 
properties;

Nano-indentation, micro-pillar tests for 
tensile properties;

Miniature specimen fracture and fatigue 
tests.

Modelling: 

Test to be complemented with finite 
element analyses, and meso-scale 
models (dislocation dynamics and crystal 
plasticity)

Patterns of response of material to NDE 
techniques as part of codification;

Exploration of possibility of lifetime 
estimation based on NDE in view of 
online monitoring

3. Nuclear materials research agenda60 

characterization. LME is a serious issue in, 
especially, F/M steels, that lacks a complete 
understanding: engineering models exist but 
need to account for the microstructure.

Gas-cooled systems: the corrosion behaviour 
depends here to a large extent on the rela-
tive impurity level in the coolant and whether 
oxidation, carburization or de-carburization 
occur, so mapping of rates is needed in this 
case as well. Due to the high temperature of 
operation, the possibility of environment and 
high stress synergism on corrosion and crack 
formation, including coupling with creep and 
creep-fatigue, needs investigation. The large 
number of possible materials for GFR imposes 
a need for wide screening. High Ni alloys re-
quire particular effort of full codification.

Radiation effects

For irradiation effects it is necessary to dis-
tinguish between fuel cladding, necessarily 
exposed to high irradiation dose rates, and 
structural components that are exposed to 
very  low  irradiation  fluxes.  For  structural 
components there are two border lines: the 
negligible irradiation curve that defines con-
ditions for absence of significant irradiation 
effects, and the maximum allowable irradi-
ation that is acceptable but needs irradiation 
effects to be accounted for. For 316L(N) aus-
tenitic steels these values are defined based 
on irradiation hardening and embrittlement. 
The 60-years operational life low dose may 
have significant effects from helium produc-
tion induced by transmutation and potential 
He embrittlement, due to the elevated opera-
tion temperature. For fuel claddings the main 
concern is irradiation creep and swelling, 
the latter also related to helium production, 
which  are  less  problematic  in  F/M  than  in 
austenitic steels. However, in the case of F/M 
steels,  low temperature (<350°C)  radiation 
embrittlement may lead to a significant shift 
of the ductile-brittle transition temperature 
(DBTT), even after fractions of one dpa, and 
potentially to severe loss of elongation due 
to plastic flow localisation, which still needs 
to be properly accounted for in design rules. 

For all materials, the possibility of synergistic 
effects between irradiation and corrosion by 
dissolution and LME need to be quantified.

Assessment of complex loadings

All structural components in a nuclear reactor 
are exposed to complex loadings. Good de-
scriptions of the loads and the computation 
of the stress and strain distributions, which 
are directly related to the constitutive models, 
are necessary to predict safety margins, evo-
lution of damage and remaining lifetime of 
components.

Integrity and qualification of weldments 
and welded components

The operating experience of all nuclear reactor 
types clearly demonstrates that welds are the 
weak spots in metallic components. Testing and 
assessment of welds and welded components 
must thus be an integrated part of pre-norma-
tive research programmes for GenIV conditions: 
(i) defect assessment through inspection can 
reduce costs and increase safety, (ii) post-weld 
heat treatment to minimize residual stresses in 
welds manufactured in accordance with code 
requirements is especially important, (iii) reliable 
methods to measure and calculate residual 
stresses need to be pursued, (iv) mechanistic 
understanding remains essential.

Sub-size and miniature specimen test 
standardization

Tests methods and associated procedures and 
standards for very small volumes of material 
using small and miniature specimens74 be-
come essential for:

 y Neutron irradiated materials that are avail-
able in limited quantity, or in order to mini-
mize the radioactivity;

74  Hyde et al., Requirement for and use of miniature test spe-
cimens to provide mechanical and creep properties off ma-
terials: a review, Int. Materials Reviews 52 (2007) 213.

Strategic Research Agenda of the EERA JPNM y Very thin material layers affected by charged 
particle irradiation (typically a few µm); 
 y Service-exposed materials, i.e. tested in ser-

vice during inspections;
 y Thin-walled cladding tubes;
 y Evaluation of local properties, for instance 
heat  affected  zone  of  a  weld  (mm  scale), 
coatings, single crystals (µm scale), or gra-
ded functional materials. 

There is a need to review in depth the test 
methods in terms of strength, weaknesses 
and restrictions to provide material data. 
This should be followed by the development 
of  common  procedures  and  standardiza-
tion of these tests. The standardization work 
must go through the involvement of national 
and international standard committees with 
participation of main testing laboratories and 
designers. These activities should concern:

 y miniaturised specimens for standard me-
chanical property assessment: these are 
smaller versions of already used and stand-
ardized specimens;

 y thin-walled cladding tube tests: these tests 
require specific equipment and standardi-
zation;

 y micro-  and  nano-mechanical  testing  me-
thods: these are unconventional and/or truly 
miniaturised  specimens/techniques,  e.g.  
nano-indentation,  compression/tension/
bend ing tests on micropillars, etc.

Component and material health 
monitoring

Non-destructive examination (NDE) of com-
ponents is crucial to verify, during inspections, 
that components are performing according to 
expectations and to detect cracks well ahead 
of time, as they could potentially lead to failure. 
However, it is not obvious how to deduce pre-
cise information on microstructural evolution 
and associated changes in mechanical prop-
erties from NDE. The conventional approach 
is thus to monitor the material degradation 
by analysing surveillance specimen exposed 
to in-service conditions by mechanical tests 
and, if required, microstructural analysis. The 

61 

current  industrial  tendency  is,  in  contrast, 
towards online materials monitoring and life-
time estimation based on NDE, possibly by 
applying artificial intelligence-based big data 
analysis to derive patterns capable of warning 
sufficiently in advance of the possibility that the 
component may fail.

3.1.2.  Advanced structural materials 
modelling and characterization

The issues to be addressed with an ICME ap-
proach (section 2.3.2) are necessarily strongly 
linked to those addressed in pre-normative 
research  (previous  section),  as  they  corre-
spond  to  the  main  degradation  processes 
that affect structural materials for GenIV sys-
tems. Since the early 2000s, several European 
projects have addressed, from a modelling 
point of view, material degradation processes 
connected to irradiation, such as radiation hard-
ening and embrittlement (in bainitic, austenitic 
and ferritic/martensitic steels), as well as, to 
a certain extent, plastic flow localisation and 
irradiation assisted stress corrosion cracking in 
water, namely FP6/PERFECT,75 FP7/GETMAT,76 
FP7/PERFORM60,77 FP7/MatISSE.78 Some of 
the relevant activities currently continue in 
the H2020/SOTERIA79 and the H2020/M4F80 
projects. Issues connected with the model-
ling of the high temperature behaviour of F/M 
steels have also been partially addressed in 
FP7/MATTER81 and FP7/MatISSE,77 while the 

75  FP6 IP PERFECT Project: Prediction of Irradiation Damage 
Effects in Reactor Components, Journal of Nuclear Materials 
– Special Issue, vol. 406, issue 1 (2010). 

76  http://www.iaea.org/inis/collection/NCLCollectionSto-
re/_Public/46/040/46040904.pdf?r=1 

77  http://www.sciencedirect.com/science/article/pii/
S0029549311001804?via%3Dihub; http://proceedings.asmedi-
gitalcollection. asme.org/proceeding.aspx?articleid=1628620; 
https://www.openaire.eu/search/project?projectId=corda__
_____::190daa7e4bf2ff682991a1c9e23ccfd2

78  www.fp7-matisse.eu;  http://www.fp7-matisse.eu/wp-con-
tent/uploads/2015/12/MatISSE-2015-GEN-IV-materials-
Holmstrom.pdf 

79  www.soteria-project.eu 

80  http://www.h2020-m4f.eu/

81  http://www.eera-jpnm.eu/?q=jpnm&sq=nboard;  http://
proceedings.asmedigitalcollection.asme.org/proceeding.
aspx?articleid=1627574 

3. Nuclear materials research agenda62 

modelling of oxidation and dissolution, as well 
as prolonged irradiation, in austenitic steels is 
being partly addressed in H2020/GEMMA82. 
Several modelling activities are also included 
in EERA JPNM pilot projects. All these projects 
provide solid bases on which to progress fur-
ther, both in terms of modelling techniques 
that  have  been  developed  and  results  ob-
tained, as well as in terms of experimental work, 
performed with a view to understanding mech-
anisms and for model validation. However, the 
development of models necessarily proceeds 
from simpler to more complex materials and 
issues, therefore there is not only still a lot of 
room for the improvement of existing models 
and there are still several open issues to be 
addressed, but there are in fact cases where no 
defined computer-simulation based physical 
modelling approach has been identified yet, 
beyond classical continuum approaches. 

The most important issues to be addressed 
are briefly described in what follows and sum-
marised in Table 4: Main issues concerning 
structural materials advanced modelling and 
characterization (page 64).

A more detailed account is provided in section 
TA.1.1.2 of the Technical Annex.71

Microstructural/microchemical evolution

Irradiation  produces  damage  in  materials 
through  processes  of  interaction  between 
energetic subatomic or atomic particles with 
the atoms and molecules that form the tar-
get. Therefore, in the case of nuclear materials, 
which have irradiation as their most specific 
origin of degradation, large effort needs to be 
devoted to the development of models that 
describe the material at the microscopic sub-
microscopic and atomic scale, accompanied 
by relevant modelling-orientated experiments 
(see section 2.3.2). Modelling the microstruc-
tural  and  microchemical  evolution  under 
irradiation and/or while exposed to high tem-
perature is essential in order to understand 

82  http://www.eera-jpnm.eu/gemma/

virtually all degradation processes.83 Although 
significant advances have been made in the 
development  of  microstructural  and  micro-
chemical evolution models in the last couple 
of decades, multiple challenges remain to be 
addressed in order to develop reliable ICME ap-
proaches that provide truly predictive capability. 
These essentially aim at being able to describe, 
with a single tool, the evolution of both the 
microstructure (redistribution of lattice de-
fects) and the microchemistry (redistribution 
of chemical species).

Machine learning schemes based on artifi-
cial intelligence, e.g. artificial neural networks, 
could  be  of  valuable  help  to  address  the 
modelling problems related with high chem-
ical complexity, as well as to bridge between 
scales.84

Mechanical behaviour after and under 
irradiation

Microstructural and microchemical evolution 
models and the corresponding understand-
ing should be used as input to move to larger 
scales and address, for example, the predic-
tion  of  the  subsequent  plastic  behaviour, 
particularly in tensile tests, as a function of ma-
terials type and composition, dose received, 
irradiation  and  test  temperatures,  and  also 
applied strain-rate, as summarily described in 
section 2.3.2. Issues to be addressed are radi-
ation hardening and embrittlement, plastic 
flow localisation and deformation at constant 
load due to irradiation creep. Model develop-
ments are necessary to address these issues, 
as described in section 2.3.2, essentially aiming 
at describing in a single framework the kinetic 
evolution of the irradiation-induced and the 
dislocation microstructures, to deduce consti-
tutive laws for crystal plasticity85 and devise 

83  A summary of key irradiation effects in structural materials 
can be found in, e.g.: Zinkle, Phys. Plasma. 12 (2005) 058101; 
Zinkle and Was, Acta Mater. 61 (2013) 735.

84  Castin et al. Comp.Mater. Sci. 148 (2018) 116.

85  Volegov et al., Physical Mesomechanics 20 (2017) 174 https://
link.springer.com/article/10.1134/S1029959917020072.

Strategic Research Agenda of the EERA JPNMhomogenization  schemes86  for  continuum 
mechanics  approaches  applied  to  compo-
nents. The challenge remains, in particular, 
to introduce chemical effects (e.g. decorated 
dislocation loops or heterogeneously nucle-
ated precipitates). In the case of (irradiation) 
creep, the coupling between microstructural 
and dislocation evolution needs to be effective 
because both microstructures evolve simulta-
neously over time. 

Mechanical behaviour at high 
temperature

The model developments sketched in the 
section 2.3.2 are potentially applicable also in 
the case of high temperature mechanical be-
haviour, mainly describable as thermal creep, 
which may be associated or not with cyclic 
loading, i.e. fatigue, via creep-fatigue interac-
tion. However, in these cases visco-plasticity 
comes into play to describe the fact that ma-
terials at high temperature deform and lose 
strength also under constant load through sev-
eral mechanisms. These models are currently 
not developed to the same level of advance-
ment as microstructure evolution or plasticity 
models. It is especially important to allow for 
changes of creep mechanisms between ac-
celerated laboratory tests and actual operation 
conditions. Moreover, for many engineering 
applications connected with high temperature 
operation in particular, thus affecting crucially 
GenIV systems, finite element models may 
not be usable. This happens when there is no 
sufficient understanding of the underlying pro-
cesses available to guide the development of 
a consistent integrated modelling approach. In 
these cases, microstructural examination to-
gether with the semi-empirical identification of 
the important variables are essential.

Fracture mechanics

All materials contain defects, i.e. microcracks 
from which cracks can be initiated and then 

86  Allen, Comp. Sci. & Technol. 61 (2001) 2223.

63 

propagate leading to failure. In design codes 
the assumption is normally that the compo-
nents are defect free, but for the structural 
integrity assessment of components the ex-
istence of microcracks needs to be accounted 
for, especially for welded components. Existing 
models need to be improved and further val-
idated,  but,  importantly,  non-destructive 
examination techniques are key, because cur-
rently neither plastic nor visco-plastic models 
can explicitly describe crack initiation. Thus, 
developments in this direction are needed, 
assisted by extensive dedicated microstruc-
tural examination. Crack propagation can be 
described under simple loading conditions but 
not in the case of complex loading, such as for 
instance under creep-fatigue. Local modelling 
approaches are to be pursued.

Coolant compatibility models

The  processes  that  govern  the interaction 
between solids and coolants are crucial in 
the context of GenIV reactors and need to be 
appropriately modelled beyond current capa-
bilities.87, 88 

More  detailed  approaches  are  required  to 
understand the origin of the processes ob-
served through microstructural examination. 
A full description of these processes requires 
the identification of new appropriate model-
ling approaches, because the complexity of 
the relevant physical and above all chemical 
processes  challenges  the  possibilities  of-
fered by existing simulation tools. Especially 
challenging is the problem of liquid metal 
embrittlement, which requires modelling of 
the liquid/solid interaction, which is assumed 
to occur by adsorption of embrittling atoms 
from the liquid at stress concentrators along 
interphases, and the induced embrittlement 
and fracture mechanisms. Depending on the 
specific solid/liquid couple, fracture can be 
intergranular or transgranular and although it 

87  Zhang et al., Review - Models of liquid metal corrosion, J. 
Nucl. Mater. 404 (2010) 82.

88  Zhang, A review of steel corrosion by liquid lead and lead-
bismuth, Corrosion Science 51 (2009) 1207.

3. Nuclear materials research agenda64 

Table 4: Main issues concerning advanced structural materials modelling and characterization.

Main issue

Breakdown in sub-issues

Main materials 
concerned

Techniques/Methods

Formation of radiation 
hardening and embrittling 
microstructural features 
(low temperature)

Formation of voids and 
onset of swelling (high 
temperature/dose)

Radiation-induced 
segregation/precipitation 
at extended defects

Microstructural evolution 
under load in relation with 
irradiation creep

Thermal ageing and 
subsequent precipitation/
segregation

Oxide formation/ 
dissolution in ODS during 
fabrication, thermal ageing 
and irradiation

F/M (possibly 
also ODS) 
and austenitic 
steels, ceramics

ODS alloys

Correct interpretation 
of microstructural 
examination technique 
signals and raw data

Any

Radiation hardening: flow 
behaviour

Plastic flow localisation

F/M and 
austenitic steels

Irradiation creep

Cyclic plasticity (softening 
and hardening)

Austenitic and 
F/M steels

Thermal creep

Microstructural 
and 
microchemical 
evolution

Mechanical 
behaviour after 
and under 
irradiation (low 
temperature)

Mechanical 
behaviour at high 
temperature: 

Creep-fatigue interaction

Fracture 
mechanics

Crack initiation

Crack propagation

Austenitic and 
F/M steels

Molecular statics and dynamics using 
either electronic structure techniques 
or interatomic potentials for energy 
calculation, as well as atomistic kinetic 
Monte Carlo methods, as atomistic 
models to study stability and mobility of 
radiation defects, or phase stability.

Object kinetic Monte Carlo, or rate 
theory based cluster dynamics, or to 
some extent phase field to develop 
microstructural/microchemical 
evolution models.

Machine learning / artificial intelligence.

Electron microscopy, atom probe, 
scattering techniques, positron 
annihilation, etc used in a combined 
way to obtain complete description. 

Specific computational techniques to 
simulate the response of experimental 
techniques.

Dislocation dynamics models for 
single crystal behaviour, informed 
to atomistic models, to deduce 
constitutive laws. (Strain gradient) 
crystal plasticity models for aggregates. 
Homogenisation techniques and 
continuum mechanics for the 
component scale. 

Wide range of mechanical testing, from 
micropillars to tensile and impact, for 
model validation/calibration.

Different models at different length-
scales are needed: from dislocation 
based to continuum mechanics. The 
techniques and methods are as in 
the previous case, but their use is 
more challenging because of the 
number and complexity of possible 
mechanisms, involving not dislocations 
and grain boundaries, and complex 
thermo-mechanical loading conditions. 
The up-scaling includes crystal 
plasticity and visco-plastic models.

(Visco)-plastic models including 
damage criteria; classical fracture 
mechanics governed by crack tip 
parameters; local approaches where 
the fracture processes are explicitly 
modelled

Strategic Research Agenda of the EERA JPNM65 

Main materials 
concerned

Techniques/Methods

F/M and 
austenitic steels Atomistic and thermodynamic models 

for corrosion and LME; phase field 
models; extensive microstructural 
characterization, especially electron 
microscopy

Main issue

Breakdown in sub-issues

Compatibility with 
coolants & coolant 
chemistry

High temperature 
oxidation/corrosion

HLM dissolution corrosion 
of steels

Liquid metal embrittlement F/M steels

Properties of 
composite/
ceramic materials 
depending on 
microstructure/ 
architecture

Tomography of composite 
and correlation with their 
macroscopic properties 
(mechanical, thermal, …)

Development of models 
for thermal and mechanical 
composite behaviour

Development 
of methodology 
to perform 
ion/electron 
irradiation 
experiments

Other issues

Design of ion irradiation 
experiments and 
interpretation of 
microstructural data

Use and interpretation 
of micromechanical 
techniques (e.g. 
nanoindentation)

Residual stresses after 
welding

Modes of deformation in 
steels

SiC/SiC, Max 
phases, other

Model materials, 
to extend to all

Model 
materials, to 
extend to all

Austenitic and 
F/M steels

ODS steels

In addition to known atomistic and 
microstructural/microchemical 
evolution models as for steels, larger 
scale models that take into account the 
architecture of these materials need to 
be identified and developed

Essentially the same techniques as 
for microstructural/microchemical 
evolution, but specifically targeted to 
charged particle irradiation issues

Specific simulation tools combining 
atomistic to continuum descriptions 
to be developed for the simulation of 
micromechanical techniques

In principle a full suite of codes and 
methods, through all scales, should be 
deployed to simulate what happens 
in the welding process or under 
deformation in ODS steels. Neither of 
them exists to date.

appears as cleavage some micro-plasticity is 
always involved. The detailed mechanisms are 
still largely unknown, or at least still debated.

Models in support of the use  
of charged particle irradiation

Neutron irradiations are long and expensive 
experiments and the access to the relevant fa-
cilities is generally restricted. Charged particle 
irradiation (ions, protons, electrons, …) is a very 
valuable and affordable tool for the purpose of 
modelling-orientated experiments and mate-
rials screening. However, transferability issues 
exist between charged particle and neutron 
irradiation  environment  (see  also  section 
6.1.3).89 It is thus essential that suitable models 
address the specificities of these modes of ir-
radiation. Difficulties need also to be overcome 

89  Was, J. Mater. Res. 30 (2015) 1158.

concerning the post-irradiation examination, 
particularly the possibility of obtaining a mean-
ingful assessment of mechanical properties 
after charged particle irradiation. Overall, the 
goal is to develop an established and possibly 
standardized methodology, based on mod-
els, to perform charged particle irradiation 
and relevant PIE, including mechanical char-
acterization.

Development of standard methodology 
for model validation

The validation  of  models  is  a  complex  and 
costly  task  that  requires  specific  attention 
also from the methodological point of view, in 
order to reduce uncertainties to the minimum 
and assess those that are associated with the 
model  and  cannot  be  avoided. Modelling-
orientated experiments need to be designed 
on appropriate materials and trying to separate 
variables. Protocols need to be established to 

3. Nuclear materials research agenda66 

apply microstructural characterization tech-
niques and analyse their outcome. Methods 
to simulate the response of materials to ex-
perimental techniques are needed. Protocols 
for the comparison between simulation and 
experiment need to be established, as well.

Note on materials of interest and different 
levels of model development

It is important to emphasise that the level of 
development  of  advanced  models  is  not 
equal for all materials of interest for GenIV 
reactors. For example, the development of 
microstructure evolution physical models for 
austenitic  steels  is  much  less  mature  than 
for  F/M  steels.  In  general  models  for  con-
centrated alloys are less developed than for 
diluted alloys: 

 y The modelling of F/M steels is worth be-
ing  pursued  because  these  materials  re-
main important in the long-term for GenIV 
applications, and especially because they 
represent an important reference case for 
the  methodology;  F/M  steels  represent 
also the starting point to develop models of 
use to simulate ODS and/or alumina-form-
ing alloys (FeCrAl); 

However:

 y The modelling of austenitic steels needs 
to  be  pushed  forward,  “catching  up” with 
the level of maturity that characterises mo-
dels for F/M steels on the low-scale side.

Ceramic materials are often chemically and 
crystallographically  complex  systems  (e.g. 
MAX phases), that require specific develop-
ments.  In  terms  of  atomic-level  modelling, 
SiC and Al2O3 are probably the most studied, 
although the activities are globally very lim-
ited and scattered. Importantly, these materials 
are generally used in the form of composites, 
which opens different types of problems for 
modelling, one of them being the monitoring 
of the behaviour of the architecture, such as via 
3D tomography, and the assessment of the role 
of the interfaces between e.g. fibres and matrix.

3.1.3.  Development of advanced structural 

materials

The  development  and  codification  of  new 
materials solutions for nuclear application is 
a very long process, due to both the overall 
strict requirements to comply with the safety 
constraints of the nuclear industry and the re-
quests of the regulators, and the unavoidable 
need to perform long and costly irradiation 
campaigns. This prevents, for example, the 
use of optimised materials for the short term 
ESNII prototypes. Yet, for further demonstrator 
phases, FOAK prototypes and longer term ap-
plications in commercial reactors it is important 
to pursue the development and codification 
of better performing materials and materials 
solutions.

The main targets are to improve high temper-
ature behaviour, minimise radiation effects 
and mitigate environmental degradation. In 
this section, R&D needs for six material classes 
are discussed: 

 y Improved austenitic steels;
 y Ferritic/martensitic steels;
 y SiCf  /SiC composites;
 y Refractory metallic alloys;
 y Modified  surface  layers  for  protection 

against corrosion;

 y Other perspective materials.

For all these materials there are general issues 
of fabrication processes to be developed/
optimized and/or made less expensive. For 
some of them additive manufacturing may 
be a way forward that is worth pursuing. There 
are also general issues of protection against 
coolant attack for those solutions that do not 
explicitly address this problem. Finally, for all 
of them different solutions need to be con-
sidered and screening procedures applied, 
which are especially costly when irradiation is 
involved: for some, but not for all, ion irradiation 
can be a way to go. In what follows, possible 
development paths for each class of materials 
are briefly listed. At the end, Table 5 (page 68) 
summarizes the main issues related with the 
different classes of advanced structural materi-
als. More details can be found in section TA.1.1.3 

Strategic Research Agenda of the EERA JPNMof the Technical Annex, which is an extended 
version of the present one (next page).71

Improved austenitic steels

Austenitic steels have been used as cladding 
and core component structural materials in 
past fast reactors. Their behaviour is therefore 
relatively well-known as a large database of 
irradiated materials is available. Overall, they 
offer excellent mechanical properties in a wide 
range of temperatures, including satisfactory 
creep resistance up to 700°C. The main short-
coming for the use of austenitic steels as 
core components is the susceptibility to ir-
radiation void swelling, 90 probably both an 
inherent property connected with the fcc crys-
tal structure and also a consequence of the He 
production by transmutation through nuclide 
chains that start with Ni. Swelling leads to un-
acceptable embrittlement above 3% (∆V/V, 
where V is the volume). The path is for develop-
ment of swelling resistant austenitic steels, 
by further stabilising and optimising steels of 
the 15-15 Ti-class, mainly by working on the 
composition, with addition of swelling inhibi-
tors like Ni, Si and P, as well as Nb and V. 

Another issue affecting austenitic steels is that, 
despite their generally higher resistance to cor-
rosion than e.g. F/M steels, they may not offer 
sufficient guarantees of corrosion-resistance 
in HLM-cooled systems,91 especially with a view 
to pushing up the operating temperature for 
higher efficiency. Here one of the most promis-
ing paths currently pursued is the development 
of alumina-forming austenitic (AFA) steels that, 
through the addition of 4-6 %wt Al, exhibit supe-
rior oxidation resistance up to 900°C, due to the 
formation of a protective Al2O3 scale. They also 
offer creep strength comparable with some su-
peralloys that contain a much higher amount of 
nickel, thanks to strengthening via precipitation 
of NiAl particles. Nevertheless their application 
as fuel cladding materials will require several 
compositional changes and qualification under 

90  C. Cawthorne, E.J. Fulton Nature, 216 (1967), pp. 575-576.

91  K.  Lambrinou,  et  al Journal  of  Nuclear  Materials, Volume 
490, 2017, pp. 9-27.

67 

neutron irradiation, before converging to a suit-
able composition. Moreover, the threat of LME 
remains open and will require the identification 
of a proper framework for materials codification 
and licensing.

Ferritic/martensitic steels

Despite their high thermal conductivity and 
excellent dimensional stability under neutron 
irradiation (swelling resistance), the core ap-
plications of F/M steels in fast reactors has 
been so far limited. Their use as cladding or 
core material in other systems is at present 
prevented by:

 y Loss  of  strength  at  high  temperature 
(T>550ºC), including softening under cyclic 
operation;92

 y Neutron irradiation embrittlement al low 
temperature (T<350ºC), including plastic 
flow localisation with subsequent drastic 
reduction of uniform elongation;93, 92

 y Susceptibility  to  liquid  metal  embrittle-

ment in contact with HLM.94

However, in the long term the use of F/M steels 
is very desirable in a context of optimal use of 
resources (high-burn up), given that, at present, 
these are the only available industrial materi-
als that can bear the promise of withstanding 
neutron doses in excess of 150 or even 200 
dpa.  Their  use  in  future  GenIV  commercial 
power reactors will require a number of actions:

 y Improvement  of high  temperature  me-

chanical behaviour;

 y Improvements of ductility after low tem-
perature irradiation, including slip locali-
zation with loss of uniform elongation;
 y Improvement  of  the  compatibility  with 
HLM,  by  investigation  the  mechanisms 
responsible for corrosion and, even more 
importantly, liquid metal embrittlement.

92  Raj  &  Vijayalakshmi,  Ferritic  Steels  and  Advanced  Ferrit-
ic-Martensitic  Steels,  in:  Comprehensive  Nuclear  Materials 
4.03 (2012) 97.

93  Farrell, Byun and Hashimoto, J. Nucl. Mater. 335 (2004) 471.

94  Ersoy et al., J. Nucl. Mater. 472 (2016) 171.

3. Nuclear materials research agenda68 

Table 5: Main issues concerning classes of advanced structural materials.

Main issue

Breakdown in sub-
issues

Materials 
concerned

Techniques/ Methods

Improved 
austenitic steels

Reduction of 
susceptibility to 
swelling

Multi-stabilized 
steels

Improvement of 
compatibility with HLM

AFA steels

Addition of elements (Ti,Si,P, Ni …) that (a) 
in solid solution enhance point-defect 
recombination (b) create nanoprecipitates 
that stabilize the dislocation network 

Addition of aluminium with double benefit 
of alumina-SL formation and NiAl particle 
strengthening

Improvement of high 
temperature behaviour

ODS and CSE F/M 
steels

Dispersion of 
oxide particles 
by powder 
metallurgy 
(target > 700°C) 
or tuning of 
composition and 
TMT

ODS steel fabrication requires 
optimization/standardization and 
industrial upscale of fabrication process, 
and possibly development of alternative 
fabrication procedures.

Ferritic/
martensitic steels

Welding and relevant 
qualification

ODS steels

Development, optimization, standardization 
of welding procedures for ODS steels

Improvement of 
compatibility with HLM

FeCrAl or 
modified surface 
layers

Current F/M as well as ODS or CSE 
F/M can be coated or their composition 
modified with addition of Al (FeCrAl, incl. 
ODS)

Mitigation of low 
temperature irradiation 
embrittlement

All F/M starting 
from baseline 
steel

(Strategy may be identified via modelling, 
see section 3.1.2)

Development and 
qualification of 
SiC/SiC cladding

Degradation under 
neutron irradiation, 
especially thermal 
conductivity

Corrosion by oxidation, 
especially in moist 
oxygen deficient, high 
temperature conditions 
(in He, but also HLM 
Na)

Joining and relevant 
qualification

Standardization of 
mechanical tests 
and analysis for 
tubular structures 
of continuous fiber 
ceramic composites

SiCf/SiC tube 
with metallic 
liner (“sandwich” 
tubes) or ceramic 
layer

Neutron irradiations & PIE by SEM & TEM 
(especially PyC layer), thermal conductivity 
and hermeticity measurements (thermal 
characterisation methods by steady state 
or transient measurements)

Corrosion testing in service environment 
under careful control of the chemistry 
and varying the concentration of the 
contaminating species expected. 

Post-test microstructural characterization

Thermodynamic modelling of the SiC 
environment/coolant system.

Testing after application of protective SL

Screening of materials/processes/designs 
assisted by thermodynamic modelling

Realization of sample joints and prototypes

Mechanical testing and performance 
assessment in the operative conditions

Post-test examination by SEM, TEM, XRD

Standardization of testing procedures

Development of standards & design rules

Strategic Research Agenda of the EERA JPNMMain issue

Breakdown in sub-
issues

Materials 
concerned

Techniques/ Methods

Development of 
processing routes

Alloy design & synthesis, regain processing 
skills. Development of welding procedures.

69 

Protection with coatings and relevant 
qualification.

ODS Mo, V-Cr-Ti 
alloys

Refractory alloys

Environmental issues 
(oxygen corrosion and 
interstitial impurity 
embrittlement)

Characterise and 
improve behaviour 
towards low 
temperature radiation 
embrittlement

Surface layers 
for protection 
against coolant 
attack

Optimisation of SL 
composition and 
deposition method

GESA surface 
modification on 
steels

Evaluation of 
effectiveness of 
protection against 
prolonges exposure to 
HLM, especially high 
speed erosion

Evaluation of 
performance under 
irradiation

Mechanical 
compatibility with the 
substrate & thermal 
aging.

Welding of steels with 
modified SL

High entropy alloys

GESA and 
PLD alumina 
protected 
austenitic steels, 
F/M steels, 
refractory alloys

Prospective 
materials

MAX phases

Improve DBTT shift under irradiation 
through composition and processing 
changes. Verification through neutron 
irradiation at relevant DPA doses, post 
irradiation characterization by TEM, SEM, 
mechanical testing. 

Assessment of the long term behaviour 
of the system by thermal aging tests 
& thermodynamic modelling of the 
coating substrate system, with a view to 
progressive optimization.

Corrosion testing in service and off-normal 
conditions. Post-test microstructural 
characterization (XRD, SEM, TEM, Raman, 
etc.) to identify the compounds formed. 

Ion and neutron irradiations & post-
irradiation microstructural examination.

Mechanical testing under relevant 
conditions: fracture toughness, cyclic loads 
(thermo-mechanical fatigue, high cycle 
fatigue, creep and creep rupture)

Analysis of welded samples under above 
conditions.

Explore different combinations of elements, 
test them from the high-temperature 
mechanical, corrosion resistance and 
irradiation behaviour viewpoint.

Explore different compositions (solid 
solutions) and microstructures to optimize 
the material for the specific application. 

Perform relevant qualification.

Two alternative, though not mutually excluding 
and partly complementary, paths are consid-
ered to improve high temperature strength 
of  F/M  steels,  namely  (i) oxide-dispersion 
strengthening  (ODS)  using  powder  metal-
lurgy  or  alternative  production  processes, 
that  can  potentially  reach operating  tem-
peratures  above  700°C,  offering  also  high 

radiation resistance,95 and (ii) creep-strength 
enhancement (CSE), through composition and 
thermomechanical  treatment  optimization, 
leading to optimised carbide distributions and 
microstructures, including the assessment of 

95  P.  Dubuisson,  Journal  of  Nuclear  Materials,  Volume  428, 
Issues 1–3, September 2012, pp. 6-12.

3. Nuclear materials research agenda70 

the potential cyclic softening, low temperature 
irradiation (slip localization) and compatibility 
with coolants.96 The main issue affecting ODS 
cladding is the high cost associated with the 
multiple fabrication steps needed in a classi-
cal powder metallurgy route, which is also a 
bottleneck for upscaling to industrial produc-
tion, affected by limited reproducibility of the 
quality of the end product. Thus the TRL of 
ODS steels remains low and the priority is to 
improve the production routes, in partnership 
with steel-makers. There is also an open issue 
of appropriate welding procedures. CSE has 
the advantage that it involves conventionally 
produced (i.e. by casting in crucible) F/M steels, 
However, it is unlikely that operating temper-
atures higher than 650°C, as already achieved 
outside nuclear energy, can be targeted. No 
nuclear grade steel of this type currently exists. 
It is considered that appropriate compositional 
tuning and thermomechanical treatments may 
also help minimise the problem of low tem-
perature radiation embrittlement.

The problem of corrosion by dissolution and 
erosion can be faced either by applying suit-
able surface modification (see below), or by 
producing corrosion resistant steels that pro-
duce such a protective layer by themselves, 
e.g.  through  Al  addition  (FeCrAl). 97  These 
steels exist but still need composition optimi-
sation, to minimise side-effects of the addition 
of Al (decrease of ductility despite fully ferritic 
microstructure), while guaranteeing the for-
mation of a stable and continuum oxide layer, 
which requires reducing the content of both Cr 
and Al. It is possible - although it remains to be 
demonstrated - that these alumina-protected 
steels  may  also  provide  protection  against 
LME: the still missing clear understanding of 
the mechanisms behind LME in F/M steels 
and the synergy between LME and irradia-
tion hardening prevents clear conclusions to 
be drawn in this respect. The identification of 
a proper framework for materials codification 
and licensing whenever LME is involved is ex-
pected to be complex.

96  R.L. Klueh, et al, Scripta Mater., 53 (2005), p. 275

97  Jun  Lim  et  al.  Journal  of  Nuclear  Materials,  Volume  441, 
Issues 1–3, October 2013, Pages 650.

SiCf  /SiC composites and other ceramics

Increasing the temperature of operation be-
yond 800°C, such as in the case of GFR and 
VHTR, requires the use of refractory materials 
that allow access to temperatures well beyond 
the  current  limits  of  the  most  heat-resist-
ant super-alloys. Of all the possible material 
classes considered in the past, the main and 
conceptually most advanced candidates are 
those based on SiCf/SiC composites, i.e. SiC 
fibers in a SiC matrix. Sublimating into vapour 
directly from the solid at 2830°C, this is indeed 
one of the most temperature resistant materi-
als available on earth. These composites are 
also proposed in the commercial deployment 
of the SFR for the realization of the hexago-
nal cans and the Na gas heat exchanger, as 
well as for cladding and structural materials 
for high temperature HLM cooled systems, 
due to their good corrosion and erosion re-
sistance,98 while being considered also for 
advanced accident tolerant fuel cladding for 
GenIII+ reactors.99

The main weakness of these composites as 
cladding materials is that, both as produced 
and especially after deformation, they are not 
hermetic  and  therefore  cannot  retain  the 
fission  products.  Suitable  solutions  to  this 
problem have, however, been proposed and 
are being tested.

There are four issues that require attention and 
further work for the full qualification of SiCf/SiC 
composite cladding:

 y behaviour under irradiation;
 y corrosion issues;
 y joining;
 y standardization of testing procedures.

98  A. K. Rivai, M. Takahashi, J. of Pow. and Ener. Sys. 1 (2007) 
134; M. Takahashi, M. Kondo, Prog. in Nucl. Ener., 53 (7) (2011) 
1061;  M.  Kondo,  M.  Takahashi,  J.  Nucl.  Sci.  and  Tech  43(2) 
(2006)  174;  A.K.  Rivai,  M.  Takahashi,  Prog.  Nucl.  Energy  50 
(2008) 560; M. Takahashi, S. Uchida, Y. Kasahara, Prog. Nucl. 
Energy 50 (2008) 197.

99  http://www.westinghousenuclear.com/Portals/0/Tech-
novation%20Stuff/Accident%20Tolerant%20Fuel%20Brochu-
re%20.pdf; http://www.ga.com/accident-tolerant-fuel. 

Strategic Research Agenda of the EERA JPNMSiC is also considered as material for thermal 
shielding, insulation, control rods and seals, 
for  high-temperature  gas-cooled  reactors, 
together with other materials such as carbon 
composites, mullite, Al2O3, TiO2, ZrC, ZrN, B4C, 
graphite or graphene.

SiC behaviour under irradiation

Nuclear-grade  SiCf/SiC  composites  have 
shown to be stable to extremely high irradia-
tion doses. Overall swelling of monolithic SiC 
remains less than 1-2% in the whole range of 
possible irradiation temperatures, therefore the 
main issue to be addressed concerns swelling 
in the composite used for the actual compo-
nent (fuel cladding tube). The loss of thermal 
conductivity of SiCf/SiC under irradiation by 
one order of magnitude or more  poses the 
most  severe  limitations  for  its  application 
as fuel cladding. Ion irradiations can provide 
only limited information because, contrarily 
to metals, covalent compounds are severely 
damaged by electronic excitations, making it 
especially difficult to extrapolate to their be-
haviour under neutron irradiation. 

SiC corrosion Issues

Chemical compatibility in reactor environments 
is highly dependent on the thermodynamic 
stability of SiC in the coolants and in contact 
with the possible impurities contained therein. 
For example helium coolant may contain small 
amounts of gas impurities such as CO2, CO, 
H2O, H2, CH4, O2, as well as solid particles com-
ing from a variety of sources throughout the 
reactor system.100 The key oxidising impurities 
to consider in the He coolant will be oxygen 
and water vapour: the reaction of SiC with O2 
and moisture at elevated temperatures leads 
to three typical oxidation features, passive ox-
idation, active oxidation and bubble formation. 
The use of SiC composites requires extensive 
testing  to  evaluate  the  impact  of  non-pas-
sivating oxidation and corrosion in relevant 
environments. Barrier coating to hinder or limit 
corrosion may have to be envisaged. 

100  Estimation for ALLEGRO from HTR-10 experiments.

71 

SiC Joining

SiCf/SiC tubes are initially fabricated with one 
closed end; the open end must then be her-
metically sealed after loading the fuel pellets 
with an end-cap that should withstand the 
pressure of the fission gasses and the neutron 
radiation field, while being chemically stable 
in the coolant environment. A crucial technol-
ogy gap is the lack of a reliable, reproducible 
technique to join and hermetically seal the 
tubes. Presently there is insufficient information 
pertaining to the compatibility with coolants 
and stability under irradiation of the joints and 
additional studies are needed for the assess-
ment of reliable joining technologies

Standardization  of  testing  procedures  for 
SiCf/SiC

The qualification and eventual codification of 
SiCf/SiC requires a vast effort of pre-norma-
tive research, starting from the problem of the 
standardization of tests as a pre-requisite to 
characterize the behaviour of the material in en-
vironment (tightness against fission products, 
contact with flowing He, irradiation). Standards 
on mechanical tests for nuclear grade SiCf/
SiC are thus necessary to produce accurate 
and reliable data, based on well-defined test 
methods,  detailed  specimen  preparation, 
comprehensive reporting requirements, and 
commonly accepted terminology. For instance, 
the failure behaviour of SiCf/SiC components 
having tubular geometries is anticipated to be 
significantly different from that observed for flat 
two-dimensional architectures.

Other ceramics

The design of the high temperature gas cooled 
reactor concept (GFR, V/HTR), imposes the 
need for materials solutions to be devised 
for the purposes of thermal shielding and 
insulation, as well as control rods and seals. 
These materials must be gas-tight, corrosion 
resistant, and exhibit high fracture and creep 
strength, while being inexpensive and pref-
erably fabricated in a net-shape design, with 
the ability to be joined. Achieving reasonable 
leak tightness is one of the most important 

3. Nuclear materials research agenda72 

issues, since He gas permeates easily through 
most materials. While also some metallic ma-
terials are being explored, (Inconel/Ag) and 
ceramics are especially being investigated 
and tested for O-ring seals. In many cases 
the material chosen also should be resistant 
to shocks related to thermal transients, which 
might occur when the flow of the process fluid 
or the coolant is interrupted. Therefore, next 
to the super alloys, ceramics with improved 
fracture  toughness  are  being  developed, 
i.e. for intermediate heat exchangers. At the 
moment,  in  addition  to  SiC  (deposited  us-
ing CVI), also other materials such as carbon 
composites, mullite, Al2O3, TiO2, ZrC, ZrN, B4C, 
graphite or graphene are being considered for 
this function.

Refractory alloys

The requirements of safe operation at very 
high temperature calls for materials with melt-
ing  temperature  above  1500°C.  Refractory 
metals (V, Nb, Mo, Ta and W) have melting 
points above 2000°C and are thus an obvious 
choice for high temperature structural appli-
cations, as metallic alternative to SiCf/SiC 
and also as competitors to advanced steels 
for in-core applications. However, these alloys 
exhibit high affinity for oxygen, hydrogen, ni-
trogen and carbon, which easily diffuse in the 
bulk causing hardening and embrittlement 
during both the production process and the 
service life. The upper temperature limit of 
applicability of these materials is mostly de-
termined by oxygen corrosion and interstitial 
impurity embrittlement. The low temperature 
limits for their use under neutron irradiation 
are determined by irradiation embrittlement 
and DBTT shift at or above the operational 
temperatures. 

Another key challenge at present is to re-es-
tablish large-scale production capabilities and 
to recapture former expertise on processing, 
joining, brazing, cold work in various shapes, 
irradiation effects, and weldability. 

Several  refractory  metals  and  alloys  have 
been considered in fusion for first wall and 

blanket structural materials applications. Here 
only two alloys of this type are further dis-
cussed as prominent examples, both studied 
to some extent in the past and recently ex-
periencing renewed interest for cladding: V 
and Mo alloys.

Vanadium alloys

High quality manufacture of V-4Cr-4Ti heats 
of high purity, with state-of-the art properties 
required for fusion blanket application, has 
been demonstrated into a variety of engineer-
ing-relevant shapes, including small diameter 
thin wall tubes. Research on V-alloys as clad 
materials for the previous generation SFR was 
initiated in Europe during the 1970s but soon 
discontinued. V-alloys remain, however, on the 
list of potential interests for GenIV applications.

The  essential  condition  to  make V-  alloys 
potentially suitable for core applications is 
to enlarge its reference operating window, 
by expanding both the low and high opera-
tion temperature limits. This involves finding 
the  most  appropriate  approach  (composi-
tion, fabrication process, …) to improve high 
temperature strength and low temperature 
ductility of V-alloys, in a way similarly to F/M 
steels. 

Another element of concern is their high vul-
nerability to oxidizing species and proneness to 
embrittlement by impurities that are unavoida-
ble in fast reactor coolants. The development 
of  adequate  protective  barriers  ensuring 
protection  against  oxidation  and  corrosion 
(see below) is probably mandatory. Attention 
should be focused on suitable fabrication and 
processing technologies that avoid contamina-
tion, are compliant with operative requirements 
and nuclear environment, and are applicable 
to complex surfaces, including tubes and their 
internals. 

Molybdenum alloys

With a melting temperature in excess of 2600°C, 
molybdenum is the most versatile among the 
refractory metals and an excellent structural 
material.  Mo  alloys  have  been  extensively 

Strategic Research Agenda of the EERA JPNMinvestigated for their application as cladding 
of fast spectrum high temperature gas cooled 
reactors for space, as well as for fusion first wall 
applications and, more recently, ATF cladding. 
Sufficient experience on fabricability and joina-
bility exists, so component construction is not a 
major concern. Molybdenum and its alloys have 
also good swelling behaviour under neutron ir-
radiation, less than 3% up to 110 dpa.

The main issues are: 

 y The affinity for oxygen, carbon and nitrogen, 
which easily diffuse in the lattice occupying 
the interstitial sites, causing embrittlement;
 y The  susceptibility  to  oxidation  corrosion 
in oxidizing environments where the low 
melting temperature (795°C) and the for-
mation of volatile oxides may give rise to 
catastrophic oxidation,101 although Mo has 
good corrosion resistance in several liquid 
metals;

 y The limited ductility and irradiation induced 
embrittlement with, depending on the dose, 
DBTT shift up to 800°C for unalloyed Mo;

The improved ductility of the oxide disper-
sion strengthening makes molybdenum an 
interesting option for core applications. ODS 
Mo-La alloys show a DBTT at -100°C in the un-
irradiated state and an outstanding reduction 
in the DBTT shift after irradiation, even at low 
temperatures;  although,  depending  on  the 
mechanisms leading to the observed corro-
sion resistance, their application as cladding 
may require the deposition of a suitable dif-
fusion barrier to prevent oxidation corrosion 
and mass transfer issues for high temperature 
applications. Systematic evaluation of these al-
loys is needed to verify whether they meet the 
challenging requirements for application to the 
core of the HLM cooled systems, via: (a) char-
acterization of the corrosion in HLM of varying 
[O] and temperature; (b) characterization of the 
creep performance; (c) mechanical testing to 

101  Simnad  and  Spilners  JOM  7  (1955)  1011.  https://doi.
org/10.1007/BF03377603; Lai, High-Temperature Corrosion 
and  Materials  Applications,  ASM  Intl.  Materials  Park  Ohio; 
Smolik et al., Oxidation and Volatilization of TZM Alloy in Air, 
9th  International  Conference  on  Fusion  Reactor  Materials 
ICFRM-9 10 - 15 Oct 1999, Colorado Springs, CO, USA.

73 

investigate LME issues; (d) development of 
welding procedures; (e) development of bar-
rier coatings to prevent oxygen gathering; and 
(f) irradiation at dpa doses relevant for GenIV, 
to evaluate swelling and embrittlement issues.

Modified surface layers for protection 
against coolant attack

The adoption of corrosion-resistant materials 
and the mitigation strategies set in place to limit 
the effects of aggressive media and wear may 
not be sufficient or applicable, thereby calling 
for protection via coatings or other surface treat-
ments. The underlying idea is that protective 
surface layers should not alter the mechanical 
properties of codified materials with known 
properties, but would impart the desired cor-
rosion (and/or wear) resistance. This need is 
common to (a) steels in HLM operating above 
400-450°C; (b) SiCf/SiC in oxygen deficient high 
temperature environments; (c) Refractory alloys, 
(d) parts and components subject to fretting 
wear (stellite –Co- replacement issue). 

There exist five main criteria for the develop-
ment of an optimized surface layer (SL):

1.  The SL should form thermodynamically sta-
ble passivating phases by reaction with the 
environment. 

2.  These phases should be slowly growing in 
order to keep SL reservoir depletion rates 
low. 

3.  Interdiffusion between layer and substrate 
should occur as slowly as possible, via in-
troduction  of  an  interdiffusion  barrier  or 
substrate where the diffusion rate of the SL 
species is high.

4.  The values of the coefficients of thermal ex-
pansion of SL and substrate should be very 
similar, to minimize cooling and reheating 
stresses during temperature oscillations. 

4.  If deposited, the SL deposition processes 
should be carried out at low temperature 
to avoid the degradation of the substrate 

3. Nuclear materials research agenda 
74 

performance, in general and especially un-
der neutron irradiation. 

Prospective materials

The last constraint rules out the adoption of 
diffusion coating techniques for the core com-
ponents that imply exposure to temperatures > 
800°C (in steels) for times of the order of hours.

Most work on protective SL for core applica-
tions in ESNII prototypes involves two systems:

 y alumina-forming metallic layers produced 

by GESA;

 y ceramic Al2O3 barrier coatings produced by 

pulsed laser deposition (PLD).

These  two  are  discussed  in  the  Technical 
Annex (section TA.1.1.3) as prominent exam-
ples,71 together with detonation gun spraying 
as an especially straightforward solution, and 
atomic layer deposition (ALD) as an emerg-
ing technique. Other processes are available 
and possible, amongst other coatings based 
on refractory alloys, alumina forming alloys, 
silica forming alloys and MAX phases. FeCrAl 
coatings  produced  by  pack  cementation 
are  considered  for  the  steam  generator  of 
ALFRED.

The qualification of coated materials is es-
pecially delicate because, bearing in mind the 
licensing of the component, it will be neces-
sary to prove that in case of (local) failure of the 
SL protection no major safety-threatening con-
sequences will ensue. Essentially this implies 
foreseeing not only tests aimed at verifying 
the stability of the protective SL and standard 
qualification procedures (See section TA1.1.1 of 
the Annex to this SRA, http://www.eera-jpnm.
eu/?q=jpnm&sq=nboard), but also including in 
any case the qualification of the substrate ma-
terial. This is necessary in order to know which 
corrosion-rate or effects in general should be 
expected in the case of SL failure, as well as, 
crucially, worst case condition testing, accord-
ing to criteria that will necessarily depend on 
the individual case. 

Welding coated components may also pose 
issues because of the likely risk that the pro-
tection is lost in the process of welding.

The materials solutions listed in the previous 
sections correspond to materials that have 
been long considered, but that still require 
significant work in terms of composition, man-
ufacturing processes and property screening 
before a final codifiable material emerges and 
reaches sufficiently high TRL. Here two truly 
new classes of materials with promising fea-
tures, but which for the moment have hardly 
ever been used in any technological applica-
tion, are discussed, namely high entropy alloys 
(HEA) and MAX phases.

High-entropy alloys

High entropy alloys (HEA) are a fundamen-
tally new metallic material concept proposed 
in recent years.102 Elements are combined in 
roughly equimolar concentrations so that, in 
theory, the high entropy of mixing stabilizes 
simple solid-solution phases with relatively 
simple crystal structures. These alloys may be 
compositionally complex, but can be micro-
structurally simple. HEAs exhibit high strength 
due to their compositional complexity (solute 
strengthening), being thus considered prom-
ising for high temperature applications. HEAs 
containing passivating elements, such as Cr, 
Al, etc., have shown equivalent or superior 
resistance to corrosion compared with con-
ventional alloys in aggressive environment.103 
These good mechanical and corrosion prop-
erties make HEAs attractive wherever extreme 
service environments exist, such as in nuclear. 
Relatively little is known, however, about their 
stability under neutron or ion irradiation. The 
key point is that their conception offers the 
possibility to tailor via a suitable alloy design 
the desired thermo-mechanical, corrosion and 
radiation resistance properties, by modifying 
the composition. The field offers therefore 
wide opportunities to explore, discover, and 
develop new classes of alloys for structural 

102  Yeh et al. Adv Eng Mater 6 (2004) 299; Huang et al. Adv Eng 
Mater 6(1-2) (2004) 74; Yeh Ann Chim Sci Mat 31(6) (2006) 633; 
Zhang et al. Adv Eng Mater 10 (6) (2008) 534.

103  Liu et al. Transactions of Nonferrous Metals Society of Chi-
na 25 (4) (2015) 1341.

Strategic Research Agenda of the EERA JPNM75 

and functional applications. Further research 
should be devoted to explore their applica-
bility to the GenIV systems.

3.1.4.  Summary of structural material 

issues

MAX phases

The  MAX  phases  are  layered  solids  with 
hybrid metallic-ceramic behaviour and prop-
erties that depend on stoichiometry, given 
by the general formula Mn+1AXn, where M is 
an early transition metal, A is an A-group el-
ement (Al-S, Ga-Se,Cd-Sb, Tl-Bi)), and X is C 
or N, while n is typically 1, 2 or 3.104 They are 
versatile materials, whose properties can be 
tailored by forming solid solutions on the M, 
A and X sites, that often exhibit better prop-
erties than the ‘parent’.105 In particular, these 
materials are characterized by unusually high 
–for ceramics– damage tolerance. In terms 
of response to irradiation, MAX phases seem 
to have a remarkable capacity for self-an-
nihilation  of  neutron-induced  defects  at 
elevated temperatures.106 They are not, how-
ever, refractory materials, i.e. they will have a 
limitation in terms of operating temperature, 
dictated by their stability in the specific work-
ing environment.107 Because of their excellent 
compatibility with HLM they are promising 
core materials for HLM-cooled systems. As 
with every innovative nuclear material, the 
MAX  phases  need  to  be  optimised  for  the 
envisaged application. Optimisation involves:

 y Selection  of  appropriate  composition  fo-

llowed by microstructural tailoring.

 y Phase purity.

 y Collection  of  statistically-relevant  experi-

mental data.

Because  of  the  large  number  of  structural 
materials classes and types, it is convenient 
to provide a summary of the main issues con-
nected with pre-normative research, modelling 
and advanced materials solutions for each of 
them. This is done in Table 6 (next page).

For  each  of  the  many  possible  routes  to-
wards novel materials solutions that can be 
pursued, there are weaknesses that may act 
as show-stoppers. In principle, if the materi-
als solution fails to pass the test concerning 
this issue, it may be sensible to abandon that 
route. This issue is broadly addressed in the 
Risk Assessment section (section 11). 

3.2.  Fuel materials

3.2.1.  Fuel materials qualification and 

design rules

This section addresses the development of 
materials data and simulation tools, in parti-
cular for fuel performance codes (see section 
2.3.1), in support of the design and licencing 
of fuels for ESNII prototypes and later GenIV 
reactors. 

Although the processes occurring in pile in 
nuclear fuels are interconnected, the fuel ma-
terial properties and processes governing the 
behaviour of nuclear fuel under irradiation 
can be grouped in five main categories of is-
sues, as follows. 

104  Barsoum, MAX Phases: Properties of Machinable Ternary 
Carbides and Nitrides, 2013 Wiley-VCH Verlag GmbH & Co. 
KGaA, Weinheim, Germany.

105  Lapauw et al., Inorganic Chemistry 55 (2016) 5445; Tunca et 
al., Inorganic Chemistry 56 (2017) 3489.

106  Tallman et al. J. Nucl. Mater. 468 (2016) 1; ibidem 484 (2017) 
120; Ang et al. Scripta Materialia 114 (2016) 74; Ang, et al., Jour-
nal of the European Ceramic Society 37 (2017) 2353.

107  Low,  Thermal  Decomposition  of  MAX  phases,  https://
www.azom.com/article.aspx?ArticleID=6711.

Margin to fuel melting

The maximal temperature of the fuel at the 
centre of the pin is a key safety parameter since 
design rules indicate absence of melting as a 
criterion to be respected at all time. The margin 
to melting is defined as the difference between 
the  melting  point  (solidus  temperature)  of 
the fuel in its actual state and the centreline 

3. Nuclear materials research agenda76 

Table 6: Summary of issues relevant to the different structural material classes and types.

Materials

Type of related  
issues 

Pre-normative 
research

Modelling

Advanced materials’ 
solutions

316L(N)
(prototype 
irreplaceable 
components)

Thermal ageing, thermal creep, compatibility 
with heavy liquid metals (HLM)/gas: 
increased database (including welds), 
accelerated testing, models describing 
micro/macro evolution → refinement of 
existing, or elaboration of new, design rules

Improve 
compatibility with 
coolants, apply 
high temperature 
protective barriers

15-15Ti (cladding)

Irradiation creep and swelling, thermal creep, 
compatibility with coolants & fuel: increased 
database, models describing micro/macro 
evolution → refinement of existing, or 
elaboration of new, design rules

Austenitic 
steels

Alumina forming 
austenitic (AFA) 
steels

Exposure needed for 
screening between 
candidates

Thermodynamic 
models for 
composition 
optimisation, 
microstructure 
evolution models

Ni-based 
alloys (heat 
exchangers, 
valves, 
coaxial 
pipes)

Alloy 800 (high Ni 
austenitic steel)

Actual Ni-based 
alloys (eg Inconel 
617, Haynes 230, …)

Design properties 
are available for 
Alloy 800H from 
existing Codes and 
Standards;
Need to validate 
the properties of 
thin-section under 
extreme conditions 
due to strength 
reduction (not 
suitable for HLM 
coolants due to 
corrosion issues).

Exposure to high 
temperature in 
environment needed 
for screening 
between candidates 
and then for 
qualification.
Not suitable for 
irradiation field 
environments due 
to swelling and 
embrittlement. Also 
not suitable for HLM 
coolants due to 
corrosion issues. 

To identify mitigation 
strategies, models 
of radiation-induced 
microstructure 
evolution in 
connection with 
predictions of 
embrittlement 
and swelling (He 
production). 
Creep and creep-
fatigue engineering 
models in support of 
design correlations 
and rules.

Improve swelling 
resistance and 
compatibility with 
coolants (apply 
high temperature 
protective barriers).

Addition of 
Al increases 
compatibility 
with coolants 
(protective alumina 
layer), but causes 
embrittlement at low 
T, although improves 
high T creep strength 
(NiAl precipitates): 
compromise 
searched

Improve 
compatibility 
with alternative 
coolants and high 
temperatures; 
increase strength 
properties

Compatibility with 
coolant at high 
temperatures; 
manufacturing and 
joining

Strategic Research Agenda of the EERA JPNM77 

Materials

Type of related  
issues 

Pre-normative 
research

Modelling

Advanced materials’ 
solutions

9-14 %Cr

Low temperature irradiation embrittlement, 
irradiation creep, thermal ageing/creep, 
creep-fatigue (cyclic operation softening) 
compatibility with coolants, liquid metal 
embrittlement: increase database (including 
welds), models → define design rules and 
develop models in support 

Ferritic / 
Martensitic 
(F/M) steels 
(cladding 
and core)

Oxide dispersion 
strengthened (ODS)

FeCrAl alloys (also 
ODS)

Exposure needed for 
screening between 
candidates, suitable 
treatments for 
recrystallization 
to eliminate 
anisotropy after 
powder metallurgy 
production of 
bars and tubes by 
extrusion.

Oxide formation/
stability, 
microstructure 
evolution, modes of 
deformation

Thermodynamic 
models for 
composition 
optimisation, 
microstructure 
evolution models

Molybdenum alloys 
(including ODS)

Refractory 
metallic 
alloys 
(cladding 
and core)

Vanadium alloys

Exposure needed for screening between 
candidates. Irradiation creep and swelling, 
thermal creep, compatibility with coolants & 
fuel: increase database, models describing 
micro/macro evolution → refinement of 
existing, or elaboration of new, design rules, 
supported by models

Need solution 
to minimize 
embrittlement, 
improve creep 
resistance (e.g. by 
thermomechanical 
treatment) 
and improve 
compatibility with 
coolants

ODS steels (tubes) 
have better creep 
resistance, but 
manufacturing 
and joining are 
issues (optimization 
needed); toughness 
and compatibility are 
also issues

Addition of 
Al increases 
compatibility with 
coolants (protective 
alumina layer), but 
worsens mechanical 
behaviour: 
compromise 
searched

Prospective 
materials, mainly for 
cladding, studied 
also in the past, 
with problems of 
manufacturing, 
compatibility 
with coolant 
and mechanical 
behaviour

High Entropy Alloys

Prospective metallic materials with potentially excellent mechanical 
properties, coolant & radiation resistance, need extensive 
investigation for screening, including understanding of origin of 
properties through modelling, before applications are identified. 

Ceramics 
(cladding 
and 
coating)

SiC/SiC (also 
C/C) composites 
(cladding)

Graphite

Mechanical test 
standardization, 
radiation 
resistance (thermal 
conductivity, 
hermeticity, swelling, 
…) and corrosion 
resistance → define 
design rules

Microstructure 
evolution 
models under 
irradiation, finite 
element models 
for composite 
architectures, 
X-ray tomography 
techniques

Liners to guarantee 
hermeticity of 
cladding, or 
other techniques 
to guarantee 
hermeticity. Limit 
thermal conductivity 
degradation under 
irradiation.

Irradiation effects 
on oxidation 
resistant graphite, 
irradiation creep, 
Codes & Standards 
development

Dependence of 
properties on 
porosity, graphite 
structure dynamics 
(stress states) 

SiC/Graphite 
“composites”

3. Nuclear materials research agenda78 

Materials

Ceramics 
(cladding 
and 
coating)

Type of related  
issues 

Pre-normative 
research

Modelling

Advanced materials’ 
solutions

Non-metallic core 
support structures 
(ad hoc ceramics)

Screening of 
candidates. Test 
standardization 
(mechanical & 
thermophysical 
properties)

Microstructure 
evolution models 
under irradiation

Protection against 
oxidation

Al2O3 coatings

Applied with different techniques on different substrates to protect 
against coolant attack and temperature: exposure for screening and 
qualification

Max phases

Prospective ceramic materials with excellent mechanical properties 
(for ceramics), coolant and radiation resistant, though stability to 
high temperature needs to be verified case by case. Need extensive 
investigation for screening, including understanding of origin of 
properties, before applications are identified. Usable as coatings.

temperature of the fuel. This criterion is all the 
more important for ESNII reactors, which will 
operate at high temperature.

Therefore, the melting point and the actual 
maximal fuel temperature in pile, which are 
both largely dependent on the fuel compo-
sition, chemistry and burn-up (BU), must be 
known with great accuracy. This necessitates 
a comprehensive knowledge of the system 
phase diagrams and of the evolution of the 
thermal conductivity and specific heat as a 
function of the above-mentioned fuel param-
eters. . If the phase diagram of the (U-Pu-O) 
system  is  largely  known108,  further  data  is 
needed at high temperature and in presence 
of minor actinides or fission products, which 
lead to significant changes in fuel composition. 
Data on thermal conductivity and specific heat 
are even scarcer.

Atom transport and microstructural 
evolution

the oxygen diffusion, which governs the lo-
cal oxide/metal (O/M) ratio for oxide fuels. 
Furthermore, solid-state diffusion is involved 
in the drastic restructuring of fast reactors fuels 
taking place in the early stages of irradiation, 
alongside pore migration, a consequence of 
the very high temperature gradient across the 
fuel pellet radius. 

These atom transport properties are strongly 
influenced by the atomic-scale defects cre-
ated by energetic particles produced through 
fissions and radioactive decay, as well as in-
tense ionization.

Reliable data and in depth understanding of 
the  key  controlling  phenomena  are  essen-
tial for accurate and reliable modelling and 
assessment of the consequences of the relo-
cation effects taking place in FR MOX in fuel 
performance code. There is very limited liter-
ature information on diffusion in mixed oxide 
fast reactor fuel published since 1990.

The atom transport properties are at the origin 
of several important phenomena taking place 
during irradiation, for instance the redistribu-
tion of Pu in the fuel, which strongly affects 
the heat and fission product distribution, and 

Fission Product and Helium behaviour

The cladding constitutes the second retention 
barrier against radioactive fission product re-
lease.109 A design criterion is the absence or 

108  C. Guéneau et al. J. Nucl. Mater. 419, 145 (2011).

109  The first retention barrier is the inherent retention capabi-
lity of the fuel. Attempts at improving this exist both for UO2 

Strategic Research Agenda of the EERA JPNMminimisation  of  mechanical  interaction  be-
tween the fuel and the cladding. To calculate 
the mechanical load exerted on the cladding, 
it is necessary to know, inter alia, the quantities 
of fission gases and helium released to the pin 
plenum and the pressure generated. The fuel 
swelling, the quantity of gas retained in the 
fuel and its location (in solution in the lattice, 
at grain boundaries, or in bubbles, intra or inter-
granular), must also be known. Quite extensive 
information behaviour or rare gases is available 
on fresh and irradiated UO2 and much progress 
has been made in the last two decades in the 
understanding of the mechanisms governing 
the behaviour in this compound.110, 111, 112 The 
same progress must now be made on more 
complex compositions.

In addition, the increasing inventory of non-gas-
eous fission products in the fuel with burnup 
and time has a substantial effect on the fuel 
chemistry, including for instance the oxygen 
or carbon potential of the fuel.

Mechanical properties

The  load  on  the  cladding  depends  on  the 
mechanical evolution of the fuel in the reactor 
since it governs the fuel-cladding mechanical 
interaction (FCMI), which becomes significant 
during permanent low power operation or in 
case  of  a  power  increase  after  a  long  low-
power operation. The mechanical properties 
also come into play in gap closure, which has 
a supreme influence on the temperature of 
the  fuel  and  the  corrosion  of  the  cladding, 
but whose mechanisms are largely unknown. 
Very little data is currently available on the me-
chanical properties of MOX. The experimental 
measurement of the plastic and elastic behav-
iour of fuels at high temperature and under 
irradiation (ideally as a function of O/M ratio, 

and  MOX,  but  mainly  for  GenII/III  applications,  for  instan-
ce Chromox: https://tel.archives-ouvertes.fr/tel-01037885/
document

110  J.  Rest  et  al.,  J.  Nucl.  Mater.,  in  press  52018),  https://doi.
org/10.1016/j.jnucmat.2018.08.019 

111  M. Tonks et al., J. Nucl. Mater. 504, 300 (2018). 

112  L. Luzzi et al., Nucl. Eng. Des. 330, 265 (2018).

79 

microstructure, chemistry and burn-up), as well 
a deeper understanding of the basic mecha-
nisms involved in the evolution of mechanical 
properties, are needed for the development of 
better models to predict mechanical evolution 
and integrity.

Compatibility between fuel, cladding and 
coolant

Among the fission products created under ir-
radiation, volatile elements such as caesium, 
tellurium,  iodine,  as  well  as  molybdenum, 
migrate from the centre to the periphery of 
the fuel pellet to form, in the case of oxide 
fuels, the “Joint Oxyde-Gaine (JOG)”, i.e. the 
oxide-clad joint, layer. This layer constitutes a 
second potential interaction risk between the 
fuel and the cladding. The fuel-cladding chem-
ical interaction (FCCI) results eventually in the 
corrosion of the cladding, commonly called the 
“Réaction-Oxyde Gaine (ROG)”. This corrosion is 
another major factor limiting the integrity, and 
therefore the lifetime, of the fuel pin, eventually 
impinging safety and economics of operation. 

In addition, in case of a breach in the cladding 
(pin failure), the primary coolant can enter the 
fuel element and come in contact with the fuel, 
which can lead to a chemical reaction, in parti-
cular between MOX and liquid metal coolants, 
such  as  sodium  lead  or  lead-bismuth. This 
reaction could lead to a further degradation 
of the fuel pin and potentially to the dissemi-
nation of fuel in the coolant. Thermodynamic 
and kinetic aspects of these reactions are still 
unknown, especially at high burn-up.

The properties and mechanisms relative to 
these five issues need to be known with accu-
racy for the improvement of fuel performance 
codes, which play an essential role in the im-
provement of safety margins and performance, 
as well  as  in  the  qualification  of  innovative 
nuclear fuels. These improvements and qual-
ification rely traditionally on integral irradiation 
testing (full length pins and assemblies) repre-
sentative of the conditions of GEN IV reactors 
either in material testing reactors or in the lim-
ited number of fast neutron reactors available 

3. Nuclear materials research agenda80 

today. It then involves the examination of the 
irradiated  fuels,  as  well  as  corresponding 
measurements on fresh fuels for reference. 
Traditional post-irradiation examinations (PIE), 
including the measurement of fuel pin dimen-
sion changes, fission gas release (FGR), volatile 
and non-volatile material migration, micro-
structure evolution, etc., must be performed 
in  specific  hot-labs  enabling  the  handling 
and characterization of irradiated fuels. The 
advances  made  in  measurement  methods 
during the last decades permit greater infor-
mation and knowledge harvesting than in the 
past, with thermal conductivity, fission gas re-
tention, and fuel vaporization behaviour being 
now readily attainable. New PIEs should also 
be performed on “treasure” materials from past 
irradiation that have been held in storage to 
reap their full potential.

As indicated in section 2.2, in the short term, 
ESNII reactor first loads will rely on MOX fuels. 
The qualification and understanding of these 
fuels is thus the focus of this section. 

Most of the models and properties used today 
in the macroscopic scale simulation of nuclear 
fuels are derived from the 1990 Fast reactor 
Data manual, which gathered the recommen-
dations on properties of (U,Pu)O2 fuel made by 
a group of European experts 113. 

The FP7 ESNII+ project114 (2013-2017) has started 
the update of this property catalogue by con-
tinuing the characterization of irradiated fuels 
and reviewing the results obtained since the 
catalogue was published to assess the impact 
of the new results on the recommendations. 115 
Thermal  properties  (thermal  conductivity, 
specific heat, melting temperature and emis-
sivity), structural and mechanical properties 
(lattice parameter, thermal expansion, elastic 

113  Fast Reactor Data Manual, Technical report NT CEA DRN/
DEC/SPU/LPCA 2 (1990)

114  http://www.snetp.eu/esnii/

115  K.  Tucek  et  al.,  New  catalogue  on  (U,  Pu)O2  properties 
for  fast  reactors  and  first  measurements  on  irradiated  and 
non-irradiated fuels within the ESNII+ project, FR17, Interna-
tional Conference on Fast Reactors and Related Fuel Cycles: 
Next Generation Nuclear Systems for Sustainable Develop-
ment, 26–29 June 2017, Yekaterinburg, Russian Federation.

constant, brittle-to-ductile transition tempera-
ture, yield and ultimate stresses, thermal creep) 
and atomic transport properties (diffusion and 
migration of oxygen, uranium, plutonium or mi-
nor actinides, fission gases, as well as of fission 
gases pores, fission gas bubbles) were particu-
larly considered. 

The data to be measured to fill the gaps in 
knowledge that are consistent among others 
with the final recommendations of the ESNII+ 
project, as well as the materials concerned and 
the techniques that can be used are listed in 
Table 7.

3.2.2.  Advanced fuel materials modelling 

and characterization

As stressed in Section 2.3, a combination of 
advanced  modelling  and  separate  effect 
experiments  including  detailed  materials 
characterization  is  used  in  complement  to 
technological research. On fuels, this involves 
basic research investigations on the five issues 
described in the previous section. 

The aim is to use the approach described in 
Section 2.3.2, and combine the results with 
those of the pre-normative research to un-
veil the missing relevant data and elementary 
mechanisms  underpinning  the  fuel  behav-
iour, and also to extend the reliability regime 
of traditionally deduced empirical laws gov-
erning various aspects of nuclear fuel under 
irradiation, which are implemented in fuel per-
formance codes.

This advanced modelling and characterization 
approach has started later on fuel compounds 
than for structural materials, the first atomic 
scale computational studies date to the late 
1990s. Nuclear fuels are usually insulators or 
semi-conductors and have therefore specific 
thermo-mechanical and transport properties. 
The defects generated by irradiation are also 
significantly more complex than in metals, for 
instance because of their electric charge. The 
modelling of nuclear fuels is also particularly 
challenging because of the complex behaviour 
of 5ƒ electrons in actinide compounds. 

Strategic Research Agenda of the EERA JPNMTable 7: Main issues concerning fuel materials qualification and design rules.

Main issue

Breakdown in 
sub-issues

Materials 
concerned

Techniques/ Methods

81 

Evolution of 
margin to fuel 
melting 

Atom 
transport and 
microstructural 
evolution

Fission 
Products and 
Helium

Establishment of 
phase diagrams

Evolution 
of thermal 
properties 

Atom transport

Microstructural 
Evolution

Gas (fission 
gas and He) 
behaviour

Non-gaseous FP 
transport

Irradiated oxide 
fuels and fuel 
pins: MOX (U,Pu)
O2
UO2
(U,Pu,Am)O2
(U,Am)O2
Fresh oxide fuels 
as reference

Determination as a function of O/M, Pu, Am 
content, BU of 

•  Melting points 

•  Heat capacity

•  Vapour pressure

Techniques: laser heating, calorimetry, 
Knudsen cell mass spectrometry (KEMS) 

Measurement as a function of temperature, 
composition (O/M, Pu/M, BU) of 

•  Thermal conductivity

•  Thermal diffusivity

•  Emissivity

Techniques: laser flash analysis 

Measurement as a function of temperature 
and initial composition (O/M, Pu/M, BU) of 

•  Lattice parameter

•  Diffusion of U, Pu, Am

•  Diffusion of O 

•  Oxygen potential

Techniques: SEM-WDX, EPMA, SIMS, 
coulometric titration, XRD

Measurement as a function of temperature 
(radial position), initial Pu/M, O/M ratios and 
porosity and BU of

•  Density

•  Beginning of life restructuring

•  Centre void formation

•  Pu, Am and O/M homogeneity and content 

•  Diffusion/migration of pores

•  Grain size and their distribution

Techniques: Optical microscopy, SEM/WDX-
EDX, TEM, EPMA, SIMS, XRD 

Measure of gas release in irradiated fuel as a 
function of BU and irradiation history (normal, 
off-normal and severe accident conditions)

Techniques: gas puncturing, KEMS, thermal 
treatment, thermal desorption spectroscopy

Measure of fission products transport, release 
and compounds formed as a function of BU 
and history (in normal, off-normal and SA 
conditions)

Techniques: Gamma scanning of fuel pins, 
KEMS, optical microscopy, SEM/WDX-EDX, 
EPMA, TEM, XRD, ICP-MS. 

3. Nuclear materials research agenda82 

Main issue

Breakdown in 
sub-issues

Materials 
concerned

Techniques/ Methods

Evolution of 
mechanical 
properties and 
“state” of fuel 
pellets

Fragmentation, 
cracking

Irradiated oxide 
fuels and fuel 
pins: MOX (U,Pu)
O2

UO2

(U,Pu,Am)O2

(U,Am)O2

Fresh oxide fuels 
as reference

Fuel Cladding 
Mechanical 
Interactions

Chemical 
interactions 
between fuel 
and cladding, 
internal cladding 
corrosion

Chemical 
interactions 
between fuel, 
cladding and 
coolant in case of 
severe accident

Irradiated and 
fresh fuel pins,

Breached pins,

Sodium, lead and 
lead-bismuth 
coolants

Measurement as a function of temperature, 
Pu/M and O/M ratio, microstructure and BU 
of:

•  Thermal expansion

•  Elastic constants

•  Ultimate stress

•  Yield stress

•  Brittle to ductile transition temperature

•  Hardness 

•  Creep rates

Techniques: XRD, dilatometry, acoustic 
methods, micro-nano indentation, eventually 
miniaturised mechanical testing methods

Measurement as a function of temperature, 
Pu/M and O/M ratio, microstructure and BU 
of:

•  Pellet geometry

•  Inner pin geometry

•  Swelling

•  Pellet density

Determination of remaining gap width 
between pellet and cladding

Techniques: Optical inspection, XRD, 
dilatometry, pin profilometry, immersion 
method, optical microscopy, SEM, neutron 
radiography

Determination of 

•  Width of corrosion layer

•  Composition of corrosion layer

Techniques: Gamma scanning, SEM/WDX-
EDX, optical microscopy

Measurement (as a function of BU if possible) 
of:

•  Size of breach

•  Amount of fuel / fissile matter loss

•  Phases produced 

Techniques: Optical inspection, Gamma 
scanning, SEM/WDX-EDX, optical microscopy

Mechanical 
properties and 
mechanical
interactions 
between fuel 
and cladding 

Chemical 
interactions 
between fuel, 
cladding and 
coolant

This type of stud studies started to develop 
significantly  in  the year  2000s  on  uranium 
110,  116,  and  to  a  lesser 
fuels,  especially  UO2 

116  R. Konings et al., Behaviour and Properties of Nuclear Fu-
els, in Experimental and Theoretical Approaches to Actinide 
Chemistry,  First  Edition,  Ed: J.K.  Gibson, W.A.  de Jong, John 
Wiley & Sons Ltd ‘2018).

extent uranium carbide and uranium nitride. 
They progressed significantly in the F-BRIDGE 
FP7 project (2008-2012).117 Basic data such as 
diffusion coefficients of oxygen, cation and 
He,  melting  temperatures  and  consistent 

117   https://cordis.europa.eu/publication/rcn/16699_en.html 

Strategic Research Agenda of the EERA JPNMthermodynamic data were determined using 
new measurements in well controlled condi-
tions (oxygen potential for fuels for instance) 
and samples (chemical composition, impu-
rity rates, density,…), as well as state-of-the-art 
modelling from the atomic to the grain scale. 
Emphasis was put on improving the reliability 
of the data obtained. Results have in particular 
shown that it is now feasible to use electronic 
structure and empirical potential calculations 
to obtain precise data on fuels to feed higher 
scale models and help interpret experiments 
on fuels. 

In addition, the links between the scales and 
between modelling and experiments were 
strengthened. On the one hand, the multiscale 
modelling exercise on transport properties in 
uranium dioxide stressed the links built be-
tween the atomic and mesoscopic modelling 
on fuels by synthesising the data needed as 
input in mesoscale modelling that were cal-
culated at the atomic scale. 118 On the other 
hand,  F-BRIDGE  demonstrated  a  first  suc-
cess in updating existing fuel performance 
codes  using  advanced  material  properties 
and  models  obtained  from  basic  research 
and the multiscale modelling approach. In 
particular, recently obtained material proper-
ties for oxide, nitride and carbide fuels were 
implemented in the TRANSURANUS code.119 
Two models for Poisson’s ratio and emissivity 
of UO2, a specific porosity correction model, 
as well  as  a  formulation  for  intra-granular 
diffusion of fission gas,120 were also added 
to TRANSURANUS. The new version of the 
code  was  successfully  compared  against 
experimental results from the SUPERFACT 
experiment in the Phénix reactor.

Finally, data request lists, gathering key tech-
nological issues, pending scientific questions 
and corresponding basic research investiga-
tions to be carried out, were built from the 
interaction between F-BRIDGE participants 

83 

and industry representative members of the 
user group. These lists were at the origin of 
many investigations started amongst others 
in the Joint Programme of Nuclear Materials.

The ICME approach on nuclear fuels has also 
been reviewed by the Expert Group on the 
Multiscale Modelling on Fuels of the Working 
Party on Multi-scale Modelling of Fuels and 
Structural  Materials  for  Nuclear  Systems 
(WPMM)121 established under the auspices of 
the OECD NEA Nuclear Science Committee. A 
first state-of-the art report on the multi-scale 
modelling of nuclear fuels, synthesising the 
modelling  approaches  from  the  atomic  to 
the macroscopic scale devoted to nuclear 
fuels in support of current fuel optimisation 
programmes  and  innovative  fuel,  was  re-
leased in 2015122. This report also includes 
critical analyses of the mid- and long-term 
challenges for the future, i.e. approximations, 
methods,  scales,  key  experimental  data, 
characterisation techniques missing or to be 
strengthened.

The approach must now be extended to the 
various types of FR fuel materials: MOX and 
minor actinide bearing oxide fuels, but also 
actinide carbides and nitrides, as well as to fu-
els containing non-gaseous fission products. 
It also requires the use of a large number of 
facilities: hot labs, materials research reactors, 
large facilities accepting radioactive materials, 
such as ion accelerators or synchrotrons for 
the experimental characterization, as well as 
supercomputer centres for the modelling. The 
new data and models obtained will then be 
implemented in fuel performance codes to en-
hance their reliability in normal and off-normal 
situations. 

The  modelling  and  characterizations  to  be 
done on the issues, as well as the methods 
and techniques to be employed, are listed in 
Table 8 (next page). 

118  M. Bertolus et al., J. Nucl. Mater. 462, 475 (2015).

119  E.A. Kotomin, J. Nucl. Mater. 393, 292 (2009).

120  P. Van Uffelen et al., Nucl. Eng. Design, 43, 477 (2011).

121  https://www.oecd-nea.org/science/wpmm/expert_
groups/m2f.html 

122  https://www.oecd-nea.org/science/docs/2015/nsc-
r2015-5.pdf 

3. Nuclear materials research agenda84 

Table 8: Main issues concerning modelling and characterization

Main issue

Breakdown in 
sub-issues

Materials 
concerned

Experimental techniques/ Modelling methods

Temperature: 
Margin to fuel 
melting

Irradiation 
defects and 
atom transport

Phase 
diagrams: 
phases as a 
function of 
composition

Evolution 
of melting 
temperature 
with BU

Evolution 
of thermal 
conductivity 
with BU

Irradiation 
defects: point 
and extended 
defects

Pu relocation 

and O/M 
variation

Microstructural 
evolution

Fresh, self-, 
ion or neutron 
irradiated
MOX (U,Pu)O2
UO2
(U,Pu,Am)O2
(U,Am)O2
UC

(U,Pu)C

UN

(U,Pu)N

(Pu,Zr)N

SIMFUELS 
(fresh fuels 
including FP)

•  Laser heating measurement of melting 

temperature of virgin samples with various 
stoichiometries

•  Calorimetry on virgin samples with various non-

stoichiometries

•  Atomic scale calculations of melting point and 

calorific capacity vs non-stoichiometry

•  Calphad modelling

•  Laser heating measurement of melting 

temperature of simfuels, self-irradiated and 
neutron-irradiated samples (short times, T and flux 
controlled) with various compositions and non-
stoichiometries

•  Calorimetry on simfuels, self-irradiated and 

irradiated samples with various compositions 

•  Atomic scale calculations of thermal conductivity of 

fuels vs non-stoichiometry and composition

•  Calorimetry of virgin, ion-irradiated, self-irradiated 
and neutron-irradiated samples (short times, T and 
flux controlled) for various non-stoichiometries

•  Determination of type of defects created using 

positron annihilation and Raman spectroscopies, 
MAS-NMR, XAS, electrical conductivity 
measurement

•  Atomic scale modelling of point defects to 

determine most stable configurations as a function 
of non-stoichiometry

•  Atomic scale modelling of extended defects 
to determine most stable configurations and 
mechanisms of formation and growth

•  Atomic scale modelling of displacement cascades

•  Measurement of thermal, irradiation-induced 

and irradiation-enhanced diffusion coefficients 
for cations as a function of stoichiometry and 
composition using ion and neutron irradiated 
samples, radioactive tracers, electrical conductivity 
measurements, SIMS

•  Diffusion couple experiments, including with high 

temperature gradient

•  Atomic scale modelling of thermal, irradiation-
induced and irradiation-enhanced diffusion 
coefficients for cations vs stoichiometry and 
composition

•  Mesoscale modelling of diffusion to determine 
cation concentrations in grain/pellet (KMC, rate 
theory, phase field...)

•  Modelling of grain growth, fragmentation

•  Mesoscale modelling of pore evolution, central void 

formation

•  Diffusion experiments with high temperature 

gradient (laser heating)

Strategic Research Agenda of the EERA JPNMMain issue

Breakdown in 
sub-issues

Materials 
concerned

Experimental techniques/ Modelling methods

85 

Gas behaviour

Non-gaseous 
FP transport

Fission Product 
and Helium 
behaviour 

FP compounds 
and JOG

Thermal 
expansion

Mechanical 
properties

Creep: thermal 
and under 
irradiation

FCMI

Fresh, self-, 
ion or neutron 
irradiated

MOX (U,Pu)O2

UO2

(U,Pu,Am)O2

(U,Am)O2

UC

(U,Pu)C

UN

(U,Pu)N

(Pu,Zr)N

SIMFUELS 
(fresh fuels 
including FP)

•  Measurement of thermal, irradiation-induced and 

irradiation-enhanced diffusion coefficients for fission 
gases and helium as a function of stoichiometry and 
composition (ion implantation, short and controlled 
neutron irradiation, TDS, SIMS, mass spectrometer 
coupled to Knudsen cell)

•  Atomic scale modelling of gas incorporation, 
as well as thermal, irradiation-induced and 
irradiation-enhanced diffusion vs stoichiometry and 
composition

•  Mesoscale modelling of gas concentrations in 

grain/pellet (in solution, bubbles, grain boundaries) 
and release

•  Measurement of thermal, irradiation-induced and 
irradiation-enhanced diffusion coefficients for 
fission products as a function of stoichiometry and 
composition (SIMS, mass spectrometer coupled to 
Knudsen cell)

•  Atomic scale modelling of FP incorporation, 
as well as thermal, irradiation-induced and 
irradiation-enhanced diffusion vs stoichiometry and 
composition

•  Electronic structure (and empirical potential?) 

calculations of stability and thermodynamic data on 
fission product compounds (Cs, I, Te, Mo)

•  Electronic structure (and empirical potential?) 

calculations of stability and thermodynamic data 
and compounds between FP and fuel elements 
(uranates, plutonates…)

•  Synthesis of selected compounds and 

measurement of thermodynamic data (melting 
temperature, thermal conductivity…)

•  Development of thermodynamic database 

describing Gibbs energy functions of the phases of 
the (Cs-I-Te-Mo-O) as a function of temperature and 
composition

•  Extension of this database to (Cs-I-Te-Mo-U-Pu-O) 

system 

•  Using models developed above, thermodynamic 

equilibria calculations to predict the phase formation 
with respect to oxygen potential and temperature

•  Atomic scale calculation of thermal expansion as a 

function of composition and defect content

•  Mesoscale modelling of fuel thermal expansion

•  Atomistic modelling of high temperature and 

irradiation effects on the mechanical properties of 
fuels 

•  Atomistic modelling of fuel deformation behaviour

•  High temperature creep experiments with controlled 

conditions oxygen partial pressure control in 

•  Measure creep under ion and neutron to evaluate 

the radiation induced creep component as a 
function of temperature and load

•  Atomistic modelling of thermal expansion in 
presence of defects and fission products

•  Measurement of thermal expansion of infused or 

self-irradiated fuels

3. Nuclear materials research agenda86 

Main issue

Breakdown in 
sub-issues

Materials 
concerned

Experimental techniques/ Modelling methods

Compatibility 
between fuel, 
cladding and 
coolant 

Fresh, self-, 
ion or neutron 
irradiated

MOX (U,Pu)O2

UO2

(U,Pu,Am)O2

(U,Am)O2

UC

(U,Pu)C

UN

(U,Pu)N

(Pu,Zr)N

SIMFUELS 
(fresh fuels 
including FP)

Cladding 
corrosion: ROG

Fuel/coolant 
issues

Corium 
composition

•  Electronic structure calculations of thermodynamic 

data on compounds between FP (uranates, 
plutonates…)

•  Synthesis of selected compounds between FP and 
fuel elements and measurement of thermodynamic 
data (melting temperature, thermal conductivity…)

•  Development of thermodynamic database 

describing Gibbs energy functions of the phases 
of the (Cs-I-Te-Mo)-(U,Pu)-O system as function of 
temperature and composition 

•  Using this model, thermodynamic equilibria 

calculations to predict the phase formation in 
nuclear fuel with respect to oxygen potential and 
temperature

•  Carbides, nitrides: study of carburization and 

nitridation reactions

•  Electronic structure calculations of thermodynamic 
data on compounds between FP, fuel (U, Pu, Np, 
Am, Cm, C, N, O) and coolant (Na, Pb) elements

•  Synthesis of selected compounds and 

measurement of thermodynamic data (melting 
temperature, thermal conductivity…)

•  Development of thermodynamic database 

describing Gibbs energy functions of the phases 
of the (Cs-I-Te-Mo)-(U,Pu)-O system as function of 
temperature and composition 

•  Electronic structure calculations of thermodynamic 

data on compounds in the (U-Pu-Fe-O), (U-Pu-Fe-C) 
and (U-Pu-Fe-N) systems

•  Synthesis of selected compounds in these systems 
and measurement of thermodynamic data (melting 
temperature, thermal conductivity, miscibility gap…) 

•  Development of a thermodynamic database for the 
(U-Pu-Fe-O) (U-Pu-Fe-C) and (U-Pu-Fe-N) systems 

3.2.3.  Development of advanced fuels

Advanced oxide fuels development 

The needs for longer term R&D concerning 
fuel  and  relative  fabrication  processes  are 
identified in the following areas: (i) oxide fuels 
with optimized microstructure, including fuels 
incorporating Pu coming from multirecycling 
and transmutation fuels, (ii) advanced carbide 
and nitride fuels and (iii) alternative fuel con-
cepts (see Table 9). 

For each case dedicated irradiation testing will 
also be necessary in material testing or power 
reactors. Once removed from the reactor, PIEs 
must be performed.

Powder metallurgy will still be the choice for 
first MOX fabrication plants, but co-conver-
sion routes to generate MOX powder should 
be developed to reduce dust issues, enable 
optimised remote handling with degraded 
Pu vectors (and concomitant 241Am pollution 
due to 241Pu decay) when Pu multi recycling 
is established, and to achieve greater Pu ho-
mogeneity, enabling greater reliability and 
accuracy in properties, lower risk of hot spots 
under  power,  and  a  product  that  is  more 
readily dissolvable to close the nuclear fuel 
cycle. 

Strategic Research Agenda of the EERA JPNMTable 9: Main issues concerning the development of advanced fuel materials. 

Main issue

Breakdown in sub-
issues

Materials 
concerned

Techniques/ Methods

87 

Microstructure, 
process and design 
optimisation

MOX

Advanced 
oxide fuels 
development

Transmutation 
fuel development 
(homeogeneous and 
heterogeneous)

MA bearing 
MOX, IMF 
(Cer-Cer,  
Cer-Met)

Fabrication and 
design issues

Carbide and 
nitride fuels 
development

UC, (U,Pu)C, 
UN, (U,Pu)N, 
(Pu,Zr)N

Assessment of safety 
and fundamental 
properties

Alternative fuels 
concepts

Molten Salt Reactor 
fuels/containment 
and materials

Molten salt 
mixtures

•  Development of advanced microstructures, 

increased homogeneity (e.g. chemical methods, 
improved milling, use of dopants)

•  Alternative fabrication routes for MOX pellet 

productions: e.g. SPS, additive manufacturing…

•  Design, geometry improvements. 

•  Complete characterization (fresh fuel, 

irradiation, PIE)

•  Development of advanced safe fabrication 
processes (heavily shielded or in hot cells)

•  Development of recycling strategies (Pu and 

MA multirecycling, recycling of non-actinides, 
e.g. Mo). 

•  Complete characterization (fresh fuel, 

irradiation, PIE)

•  Develop safe fabrication and recycling routes 

for different geometries (e.g. microspheres and/
or pellets), considering radiological protection 
and pyrophoricity 

•  Develop innovative methods (e.g. chemical 
methods for MX powder synthesis, Spark 
Plasma Sintering…)

•  Solve recycling issues (e.g. gaseous release 

during dissolution, 14C, Tritium, fuel dissolution, 
options for 15N recycling and/or reduce cost of 
15N…)

•  Assessment of thermal stability (Decomposition 

of nitrides, vaporisation (during processing), 
thermal stress cracking.

•  Compatibility tests between fuel, cladding and 

coolant 

•  Evaluation of pyrophoricity and oxidation 

behaviour (thermochemical modelling of the 
reaction with air and moisture, microstructural 
studies of the oxidation mechanisms)

•  Complete characterisation using thermo-

physical, chemical, mechanical and 
microstructural characterisation techniques

•  Synthesis and purification of halide salts

•  Melting temperature determination, phase 

diagrams

•  Solubility and activity coefficients of the 

actinides and fission products

•  Thermal properties of the liquid phase (heat 
capacity, thermal conductivity, viscosity, 
density, surface tension)

•  Retention capacity for fission products (I, Cs, Te)

•  Interaction with other core materials

•  Tritium management

3. Nuclear materials research agenda88 

Incremental process improvements can be 
made. In particular, a deviation (partial or com-
plete) from traditional powder blending should 
be considered to yield a more homogeneous 
MOX fuel devoid of Pu rich regions, enabling 
a more uniform burnup and a material that is 
readily soluble under PUREX conditions. Such 
an evolutionary step can be achieved at the 
conversion stage at the recycling plant, where 
the U and Pu solutions are unified in a prede-
termined Pu enrichment. The (U,Pu) solution 
can be co-converted into a powder ready for 
further manipulation and processing into prod-
uct pellets. Classical hydroxide precipitation is 
to be avoided as it results in a dangerous ma-
terial – ammonium nitrate, which is explosive 
when dry. Other co-precipitation routes should 
be considered, for example (U,Pu) oxalate pre-
cipitation. Plant simplification and automation 
will play an important role too. The implemen-
tation of such processes can be accelerated 
if important issues pertaining to dust can be 
eliminated. Ideally the co-precipitation steps 
should result in an essentially dust-free pow-
der, and should in an optimum situation not 
contain particles with diameters less than 20 
µm  aerodynamic  diameter.  Such  particles 
should also be spherical or near spherical to 
ease their transport in the production lines. If 
dust can be eliminated, the potential for radi-
ological pollution of the gloveboxes and the 
equipment therein can be reduced.123, 124 

The use of dopants to create advanced mi-
crostructures  can  also  be  envisaged  (e.g. 
enhanced grain size and creep rates), as well 
as alternative fabrication routes for MOX pellet 
productions in the long term including, but not 
limited to, spark plasma sintering or additive 
manufacturing.125, 126

In the longer term, as described in section 2.2, 
Pu multi recycling in fast reactors must be de-
veloped. The increased presence of americium 

123  R. Parrish et al., J. Nucl. Mater., 510, 644 (2018).

124  E. Douglas et al. J. Nucl. Mater., 393, 1 (2009).

125  V. Tyrpekl et al., Rev. Sci. Inst., 86, 023904 (2015).

126  M.A. Pouchon et al., ETH Zurich, https://doi.org/10.3929/
ethz-a-010699574 (2016).

from increasing levels of 241Pu will increase the 
radiation hazard necessitating improved plant 
designs  and  plant  automation.  Industry  to-
day has vast experience in automated pellet 
fabrication in France and UK Melox and SBR 
plants. Despite their high level of automation, 
both are based on operator intervention (for 
maintenance and adjustment of equipment) 
using glovebox technology. The viability of this 
type of intervention needs to be evaluated for 
the future. 

An important challenge to be overcome for 
the production of transmutation fuel will be 
the need for increased biological protection 
because of the increased Am and Cm content 
present in these fuels. It is likely that highly 
shielded hot cells will be needed for the pro-
duction of such fuels.127 Intervention then would 
only be possible through engineered solutions 
to move equipment from the production lines 
into specially constructed facilities for repair or 
adjustment. These challenges will extend all 
the way to assembly production, heat removal 
during transport, and shielding at the reactor 
for storage before loading in the reactors.

Carbide and nitride fuels development

As indicated in section 2.2, in the long term, op-
timising core performance in terms of breeding 
and increased margins to melt would neces-
sitate  the  adoption  of  mixed  uranium  and 
plutonium carbides and nitrides (MX = MC and 
MN). Their fabrication is not trivial, if high puri-
ties are to be achieved, and there are question 
marks about their volatility at temperatures be-
low the melting point. 

Advanced driver fuels of the future (MC and 
MN) have never been manufactured on the 
large scale, and there are many needs to be 
fulfilled. The tried and trusted carbothermal 
reduction (CTR) route used widely in the past 
to convert the oxide feed stocks to carbide 
or nitride by their thermal treatment when in 

127  F. Jorion et al., Research and development for the fabrica-
tion  of  minor  actinide-bearing  fuel  materials  and  technolo-
gies, IAEA (2015).

Strategic Research Agenda of the EERA JPNMintimate contact with carbon needs to be im-
proved or replaced.128 This process is relatively 
simple, but the high temperatures result in 
powders with very low specific surface area, 
rendering them difficult to sinter. Comminution 
steps are necessary, and with them the gen-
eration  of  fine  highly  pyrophoric  powders. 
Improved fabrication methods must be envis-
aged to ease the production and improve the 
product quality. Electric field assisted sinter-
ing procedures such as spark plasma sintering 
(SPS) are being developed and offer the (as yet 
unproven) potential to yield high quality pellets 
without deleterious milling steps.129, 130 

In the case of nitrides, enrichment in 15N dur-
ing  fabrication  and  its  recovery  from  CO  in 
the off gases needs to be dealt with.131 Due 
caution must be given to maintain controlled 
atmospheres in the gloveboxes, with 10 ppm 
in oxygen and water or less being essential at 
all times. Furthermore, contingency plans (e.g. 
nitrogen or argon flooding) must be elaborated 
to combat inadvertent rupture of the glovebox 
containment. 

Beyond these evolutionary synthesis devel-
opments, revolutionary synthesis routes for 
carbides and nitrides are needed to shake off 
the undesirable attributes of the CTR proce-
dure completely. These can be as simple as 
new routes to co-precipitate the carbon oxide 
precursor to reduce the CTR temperatures, 
or radical to generate the carbide and nitride 
powders via an organometallic route. 

Moreover, dedicated samples must be pro-
duced and the intrinsic properties of these 
materials determined to the highest accuracy, 
after which, the in pile behaviour of these fuel 
forms must be investigated, leveraging past 
knowledge with further dedicated experimen-
tal and theoretical programmes. 

128  Hj. Matzke, Science of advanced LMFBR fuels, North-Hol-
land (1986).

129  P. Malkki et al., J. Nucl. Mater. 452, 548 (2014).

130  D. Salvato et al., Ceram. Int. 43, 866 (2017).

89 

Alternative fuels concepts (Molten 
Salt Reactor fuels/containment and 
materials)

In addition to the conventional pellet-in-pin 
fuel designs of the ESNII reactors, fuels for the 
Molten Salt Reactor (MSR) designs deserve 
some attention. MSR’s generally use molten 
fluorides as fuel carrier of the fissile (U, Pu) or 
fertile (Th) elements, but also chloride salts 
are  under  consideration.132  The  liquid  salts 
must have favourable physical and chemical 
properties. Then, one of the challenges for the 
development of this fuel is the optimization of 
the composition in relation to neutronic and 
clean-up conditions. Due to the liquid state, 
traditional  fuel  synthesis  (pellets,  pins,  ele-
ments) is not applicable here. The fabrication 
aspects will be dominated by the synthesis via 
fluorination of the fuel components, and their 
purification from impurities such as oxygen and 
water. The scientific and technical issues rele-
vant to the in-reactor behaviour of this type of 
nuclear fuel under irradiation are very different 
from solid fuels. Radiation effects are expected 
to play little role in the fuel behaviour, thermal 
transfer of the fission heat is strongly coupled 
to the fluid dynamics, and solubility and reten-
tion of the fissile material and fission products 
in the molten salt are of major importance for 
the safety characteristics of the reactor. Many 
fission products dissolve in the liquid fuel, and 
are strongly bonded. Noble gases and the noble 
metals are exceptions, and their removal during 
operation is an important issue. Experimental 
demonstration of these characteristics is impor-
tant and requires the development of irradiation 
facilities to study the fuel behaviour during irra-
diation in static and dynamic conditions (loops). 
Since the MSR reactor concept is generally cou-
pled to on-line clean-up of the salt to allow long 
term continuous operation, the separation of the 
fission products from the salt is integrated in the 
fuel research. 

The key research topics involved in the devel-
opment of advanced fuel materials are listed 
in Table 9.

131  N. Chauvin et al., State-of-the-art Report on Innovative Fu-
els for Advanced Nuclear Systems. NEA - 6895 (2014).

132  J. Serp, Prog. Nucl. Energy 77, 308 (2014).

3. Nuclear materials research agenda4

CROSS-
CUTTING 
ISSUES

© SCK-CEN

4.  CROSS-CUTTING 

ISSUES

4.1.  Objectives

The qualification of materials in environments 
corresponding to the expected service con-
ditions, the development of models capable 
of anticipating materials degradation in op-
eration, as well as the development of new 
material solutions for better performance and 
fitness  for  purpose,  are widely  recognized 
through all energy technologies, nuclear and 
non-nuclear, to have a huge potential to en-
hance the safety, efficiency and lifetime of 
energy production devices and to contribute 
to lower maintenance costs. The EERA JPNM 
firmly believes that cross-fertilisation on ma-
terials between different technologies can 
only be beneficial and contributes to accel-
erating progress. Moreover, joint research on 
materials topics of common interest, sharing 
methodologies and facilities, is an effective 
way to make better use of available human, 
infrastructural and financial resources.

Many common or contiguous research topics 
exist with other nuclear fission technologies, 
namely GenII/III reactors and (very) high tem-
perature reactors for co-generation, but also 
with  fusion  energy  technology.  The  EERA 
JPNM is in the privileged position of acting as 

catalyser for collaboration between these two 
nuclear technology branches. Thus, signifi-
cant effort was deployed to dialogue with the 
platforms involved, NUGENIA and NC2I (ex-
ploiting the MoU signed with SNETP) and also 
EUROfusion,133 to identify cross-cutting issues.

Commonalities exist with energy technolo-
gies outside the nuclear field, as well. While 
each energy system faces different materials’ 
challenges, cross-cutting issues with nuclear 
materials can be identified, in particular con-
cerning materials that operate under extreme 
conditions (high temperature and aggressive 
operational environments). It is believed that 
the competences developed in the nuclear 
field on materials can be of use also for other 
communities. 

As a general principle, collaboration on mate-
rials qualification and development across low 
carbon energy technologies on cross-cutting 
issues is a way to accelerate the energy system 
transformation towards the highest possible 
standards of safety, reliability, increased effi-
ciency and cost reduction.

133  The H2020 project M4F is a first example of a project joint-
ly run by fission and fusion materials communities

94 

4.2.  Commonalities with nuclear 
GenII/III reactor materials

innovative LWR (for more details see Technical 
Annex TA.2.371):

The first common area between GenII/III and 
GenIV nuclear technologies concerns integrity 
(performance and ageing) of structural mate-
rials. Four topics for collaboration have been 
identified (for more details see Technical Annex 
TA.2.171): 

 y Development and qualification of welding 
procedures, including the analysis of resid-
ual stresses 

 y Testing  and  qualification  procedures  for 
miniaturised specimens for both mechani-
cal characterization and crack growth under 
environmental conditions

 y Advanced characterisation and multi-scale 
modelling of microstructural evolution un-
der irradiation

 y Ion  irradiation  as  a  neutron  irradiation  su-
rrogate  to  gain  better  understanding  of 
microstructural evolution under irradiation 
and  improve  identification  of  radiation  re-
sistant materials. 

A second area of close collaboration between 
GenII/II and IV communities is the research 
on nuclear fuels and claddings (see Technical 
Annex TA.2.171). As far as qualification and as-
sessment of procedures for safety and integrity 
are concerned, the following topics are of spe-
cial common interest. 

 y Safety of oxide fuels
 y Innovative fuels and synthesis routes
 y Fuel performance codes development and 

validation

Other commonalities are: 

 y Modelling, including separate effect exper-
iments and materials characterization 
 y Materials and coatings considered for acci-

dent tolerant claddings 

 y Use  of  experimental  facilities  enabling  the 
manufacturing and characterisation of fuels

 y Supercritical Water Cooled Reactor (SCWR) 

materials

 y Development and application of advanced 
or novel materials manufacturing processes
 y Development of new materials more resis-

tant to corrosion

Table 10 summarizes materials issues of poten-
tial cross cutting interest between EERA JPNM 
and GenII/III (NUGENIA), as well as NC2I (see 
next chapter).

4.3.  Commonalities with 
materials for (V)HTR

The existence of subjects of common interest 
and common challenges related to materials 
between NC2I and EERA JPNM stems in par-
ticular from the strong similarities between 
GFR (ESNII system) and (V)HTR, which is the 
core  business  of  NC2I:  except  for  the  fact 
that the latter includes the use of graphite 
as moderator, many components are similar 
or are subjected to the same requirements, 
namely high temperature operation in contact 
with flowing pressurized helium. Therefore the 
materials of interest are largely the same (for 
more details see Technical Annex TA.371):

 y High temperature resistant materials for 
intermediate heat exchangers, insulating 
structures, control rods, and other internal 
structures 

 y Design life of 60 years with the develop-
ment of relevant design rules and design 
curves and codification 

 y Study of the degradation of the properties 
of  relevant  materials  in  operation,  due  to 
the synergistic effect of high temperatures, 
mechanical  stresses,  radiation  and  gas 
coolant environments

Finally, three potentially common research ar-
eas are found also with the field of design of 

Table 10 summarizes also materials issues of 
potential cross cutting interest between EERA 
JPNM and VHTR (NC2I) communities.

Strategic Research Agenda of the EERA JPNM95 

Table 10: Summary of potential cross-cutting issues on materials through nuclear fission technologies

General topic

Commonalities

Integrity of 
structural 
materials

•  Development and qualification of welding procedures, including the analysis of 

residual stresses

•  Testing and qualification procedures for miniaturised specimens for both mechanical 

characterization and crack growth under environmental conditions

•  Advanced characterisation and multi-scale modelling of microstructural evolution 

under irradiation

•  Charged particle irradiation as a neutron irradiation surrogate to gain better 

understanding of microstructural evolution under irradiation and improve identification 
of radiation resistant materials

Fuel and cladding 
materials

•  Safety of oxide fuels

•  Innovative fuels and synthesis routes

•  Fuel performance codes development and validation

•  Development of advanced mechanistic and multiscale modelling tools and the 
execution of separate effect experiments and detailed materials characterization

•  Materials and coatings for accident tolerant claddings

•  Experimental facilities and relevant problems of availability

Innovative Light 
Water Reactor 
Designs and 
Technologies

(V)HTR materials

•  Materials issues for SCWR

•  Development and application of advanced or novel materials manufacturing 

processes

•  Development of new radiation-, corrosion- and high temperature resistant materials

•  Issues related with development and qualification of high temperature resistant 

materials in gas coolant environment

4.4.  Commonalities with fusion 

materials

Despite the significant differences that exist 
in terms of functioning principles and there-
fore design between GenIV fission and fusion 
energy reactors, a number of materials issues 
of common interest have been identified, that 
concern the only possible cross-cutting class 
of  materials  between  fission  and  fusion, 
namely F/M steels (for more details see the 
Technical Annex TA.471):

 y Compatibility  with  heavy  liquid  metals: 
corrosion and liquid metal embrittlement 
and a need for protective (alumina) surface 
layers

 y Codification  of  F/M  steels,  as  these  are 
not yet included in design codes such as 
RCC-MRx, except at probatory level (spe-
cific issues are cyclic softening, plastic flow 
localisation, …); joint materials database

 y Welding procedures and characterization 
to detect defects and evaluate their con-
sequences, taking into proper account the 
presence of residual stresses

 y Small specimen testing in connection with 
small volumes of materials that can be ex-
posed to irradiation

 y Development of F/M steels with better high 
and low temperature properties, with the 
goal of improving the creep resistance of 
F/M steels to above 650°C and of reduc-
ing their susceptibility to low temperature 
(<350°C) embrittlement

 y Establishment of a methodology to screen 
among radiation-resistant materials, e.g. by 
combination of charged particle irradiation, 
coupled with suitable PIE and, in particular, 
mechanical property probing techniques on 
small volumes 

 y Advanced  modelling  to  continually  im-
prove  the  level  of  understanding  of  ma-
terials  behaviour  by  exploiting  advanced 
microstructural  characterization  and  evol-
ving physical models 

4. Cross-cutting issues96 

Table 11 summarizes materials issues of po-
tentially cross-cutting interest between GenIV 
fission and fusion.

4.5.  Commonalities with 

materials for other energy 
technologies134

The commonalities with other energy tech-
nologies can be classified in two groups, one 
generic and one specific, namely: 

1. General methodological common patterns
2. Materials for high temperature applications

Concerning the first group, an analysis of mate-
rials needs and materials science approaches 
through energy technologies in EERA, includ-
ing  nuclear,  led  to  the  identification  of  the 
following common patterns135 (for more de-
tails, see Technical Annex TA.5.171):

 y Ageing and degradation mechanisms stud-
ied by combining advanced experimental 
characterization with multiscale modelling
 y Characterization of energy materials and 
devices: contribution of large scale facilities, 
as well as in situ and operando techniques
 y Rational design of materials supported by 

modelling

The second group stems from the recognition 
of the fact that resistance to high tempera-
ture is a requirement for materials in a wide 
spectrum of energy technologies, because 

134  This section is largely based on the outcome of two work-
shops: (1) the EERA inter-JP cross-fertilization workshop on ma-
terials for energy applications and technologies, held on April 
28th-29th,  2015,  in  Brussels  [http://www.eera-jpnm.eu/files-
harer/documents/JPNM-related_Meetings/2015_04_28-29_
EERA_InterJP_Workshop_Energy_Materials]; (2) the workshop 
on cross-cutting issues in structural materials R&D for future 
energy systems, held on November 25th-26th, 2015, in Petten, 
NL  [http://www.eera-jpnm.eu/filesharer/documents/JPNM-
related_Meetings/2015_11_24-25_MatISSEXcuttingWshop_
Petten]. Moreover, an initiative on high temperature materials 
through energy technologies is ongoing, that involved several 
EERA JPs and should lead to a joint EERA position paper on 
the subject.

the efficiency of thermodynamic cycles operat-
ing between two heat reservoirs is improved by 
increasing the temperature of the hot one and 
some energy production systems inherently 
require high temperature to function, because 
of the physical-chemical processes involved. 
Thus,  low  carbon  energy  technologies  as 
different from each other as fuel cells and hy-
drogen, concentrated solar power, bioenergy, 
geothermal, GenIV nuclear fission, and fusion, 
find  commonalities  in  the  need  to  operate 
at temperatures above, and sometimes well 
above, 400°C. During the transition to a fully 
low-carbon economy, this problem affects also 
clean fossil fuel plants, which will also need to 
operate at the highest temperature possible 
to increase efficiency and minimize emissions. 

At high temperature, environmental aggression 
seriously limits the performance of materials 
and the lifetime of components, so the two 
issues  of  high  temperature  operation  and 
compatibility with aggressive environments 
cannot really be separated. The components 
and the classes of materials of interest, as well 
as the properties that need to be evaluated 
and improved, are largely the same through 
the different energy technologies, thus there 
are serious grounds for the establishment of a 
joint materials qualification and development 
research pool, that would share facilities and 
resources, exploiting also modern modelling 
techniques.136 More details on the commonali-
ties concerning high temperature materials are 
given in the Technical Annex TA.5.2.71

Within high temperature applications, spe-
cific commonalities identified and targeted 
between  materials  issues  for  GenIV  fis-
sion energy and concentrated solar power 
(CSP). Details on this point are given in the 
Technical Annex TA.5.3.71 Table 12 summarizes 
the cross-cutting issues through low carbon 
energy technologies (fuel cells and hydrogen, 
concentrated solar power, bioenergy, geother-
mal, GenIV nuclear fission) concerning high 
temperature and environmental degradation 
resistant materials.

135  http://www.eera-set.eu/wp-content/uploads/EERA-JP-
workshop-Materials_for_Energy_report.pdf

136  https://www.eera-set.eu/wp-content/uploads/FINAL-
June-2018-EERA-EUMAT-position-paper-HT-materials.pdf

Strategic Research Agenda of the EERA JPNM97 

Table 11: Summary of GenIV fission/fusion potential cross-cutting issues on materials

General topic

Commonalities

F/M steels for current 
concept designs

Innovative high 
temperature resistant 
steels

F/M steels

•  Design rules in RCC-MRx for F/M steels

•  F/M property assessment at low and high temperature: plastic flow 

localisation, cyclic softening, thermal creep, thermal fatigue, creep-fatigue, 

•  F/M welding

•  F/M compatibility with HLM: corrosion, erosion, LME, and test 

standardization

•  Optimisation of F/M ODS & TMT steel fabrication

Identification of best compositions and TMT

 ੋ
 ੋ Deformation modes of ODS alloys at high temperature (creep)
 ੋ Stability of the microstructure after long exposure to high temperature and 

long time for the ODS &TMT steels

•  Screening methods for prospective new materials (including charged 

particle irradiation and small specimen testing technology)

Ceramic materials

•  Ceramic coatings on steels (alumina or aluminium containing)

F/M steels

•  Microstructure evolution under irradiation

Physical modelling and 
modelling-oriented 
experiments

•  Low temperature hardening and embrittlement and plastic flow localisation

•  Irradiation creep and swelling

•  Liquid metal corrosion and liquid metal embrittlement

•  ODS fabrication processes, precipitation coarsening and recrystallization of 

oxides or carbides in advanced steels, …

Table 12: Summary of cross-cutting issues on high temperature and environmental degradation resistant mate-
rials through low carbon energy technologies 

General topic

Commonalities

High temperature (>400°C) 
mechanical performance 
assessment

Resistance to 
•  Thermal creep deformation (rupture beyond lifetime)
•  Thermal-mechanical fatigue (resistance to crack initiation/propagation)
•  Creep-fatigue interaction

Protection from aggressive 
environment (liquid metals, 
molten salts, gases,…)

•  Corrosion/oxidation/dissolution/erosion processes
•  Coatings of proven stability
•  Self-healing surface protection mechanisms

Other properties to be 
maintained

•  Thermal conductivity and limited thermal expansion
•  Non-permeability (to specific elements, e.g. gaseous like H or He)

Steels for high 
temperature applications: 
existing and advanced

•  Creep-resistant and corrosion resistant (FeCrAl) F/M steels (including ODS)
•  Austenitic steels
•  Ni-based alloys

Refractory materials: 
metals and ceramic 
composites: existing and 
advanced

Materials qualification

•  V-, Mo-based alloys
•  SiCf/SiC
•  Alumina based ceramics
•  Max phases

•  Exposure facilities
•  Test standardization

Advanced modelling and 
characterization

•  Physical mechanisms

4. Cross-cutting issues5

WORKFLOW 
AND 
TIMELINE

© ENEA

5.  WORKFLOW 

AND TIMELINE

The development and qualification of materials 
for any technology is a continuous process:

1. 

 y The return of experience from the use of al-
ready qualified and licensed materials pro-
vides further input for the design codes and 
triggers the search for more suitable mate-
rials and materials solutions, when design 
solutions are insufficient or unsatisfactory. 
 y Fundamental research on materials behav-
iour mechanisms provides tools to better 
assess component lifetimes and suggests 
routes towards materials property improve-
ment. 

 y Following  the  return  of  experience  from 
use  and  the  continuously  improved  un-
derstanding  of  the  processes  that  govern 
materials  behaviour,  new  routes  to  mate-
rials  fabrication  and  processing  and  new 
materials solutions are explored. 

The virtuous circle of Figure 2 represents well 
this continuous process. In this SRA, also a dif-
ferent representation of the workflow can be 
provided, in which two parallel routes sup-
ported by modelling are pursued,  one  for 
the shorter term application (demonstrators), 
the other for the longer term (FOAK and then 
commercially deployed GenIV reactors). This 
as illustrated in Figure 8, which shows that:

 Demonstrators will be built using exist-
ing materials, that need to be qualified 
for the expected service conditions and, 
in turn, determine the allowable service 
conditions due to their limitations in terms 
of performance. Thus the demonstrators 
will not offer the best efficiency and will 
be unlikely to meet all the requirements 
of GenIV systems;

2.  FOAK prototypes and then truly GenIV 
commercially  deployed  systems will 
need innovative material solutions  in 
order to achieve the expected targets of 
efficiency, optimal use of resources, waste 
reduction, and economy or, in short, sus-
tainability.

Figure 8 summarises the crucial issues for each 
ESNII system, concerning both structural and 
fuel materials, clarifies that physical modelling 
is the crucial support for both paths and em-
phasises the overarching principle of safety.

Figure 8 also highlights that, depending on 
the system, some materials and issues are 
more relevant than others. It is therefore pos-
sible to associate indicative milestones for the 
qualification/development of specific mate-
rial solutions with the moment when it needs 
to be ready for application, in order to enable 

102 

Figure 8: The two parallel paths followed in the present SRA, one with short term and another with long term 
perspective.

the design of a specific system. However, it 
is impossible to establish precisely the time-
line according to which the different possible 
systems will be designed, licensed and built, 
because this will depend on political and tech-
nical decisions that are out of the scope of the 
present SRA.

Figures  9  and  10  provide  a  simplified 
timeline, for structural and fuel materials 
respectively, defined based on an arbitrary 
“chronological” order, where the milestones 
are given by the fast reactor system devel-
opment,137 associated with the different TRL 
attributed to each of them,12 rather than on an 
actual timeline expressed in years. Each row 
refers to experimental work and supporting 

137  In particular, at the time of writing of this SRA the fate of 
ASTRID is undefined.

models aimed at a specific goal, as expressed. 
The longest arrows are those projected farther 
in the future and are more uncertain, but are 
expected to benefit from the operation of the 
demonstrators. If the goal has to be reached 
on time, all activities should start now (in most 
cases they have started already), but the far-
ther the corresponding milestone, the more 
the relevant activities should be considered 
as  underlying  technology  research,  rather 
than high priorities. In some cases, if the goal 
is not reached as closer milestone, the activity 
remains of interest for farther milestones: this 
is indicated by an empty arrow with dashed 
outline. Goals that are somehow a pre-requisite 
to the main one are indicated as lighter colour 
arrows: the length is in this case shorter but 
arbitrary, and does not refer to any milestone. 
These charts are of course only indicative and 
will require progressive updating over time.

Strategic Research Agenda of the EERA JPNM103 

Figure 9: Schematic timeline for structural materials. Dashed outline arrows indicate potential continuation of 
activities towards following milestones.

Figure 10: Schematic timeline for fuel materials. Dashed outline arrows indicate potential continuation of activi-
ties towards following milestones.

5. Workflow and timeline6

INFRASTRUC-
TURES FOR 
NUCLEAR 
MATERIALS 
R&D

© SCK-CEN

6.  INFRASTRUCTURES 

FOR NUCLEAR 
MATERIALS R&D

The research activities described in sections 
3  and  4  have  as  an  essential  prerequisite 
the availability of suitable facilities and infra-
structures for materials qualification through 
exposure to conditions representative of ser-
vice and subsequent characterization, both 
mechanical and microstructural. Modelling, 
on the other hand, implies the availability of 
suitable computational facilities.

The three service conditions of interest for 
GenIV  reactor  nuclear  materials  are high 
temperature, contact with coolants and a 
high level of irradiation. The first two can be 
of interest and application also for non-nu-
clear technologies (e.g. facilities for tensile 
and fracture toughness tests, standard creep 
and corrosion tests, slow strain rate tests in 
environment…) and may be in some cases 
even redundant. Moreover, microstructural 
studies on materials exposed to corrosive 
environment and/or high temperature can 
be performed in many laboratories. In con-
trast, irradiation is quite specific of nuclear 
materials  and  irradiation  facilities  are 
scarce, as well as are scarce the infrastruc-
tures that allow testing and examination of 
irradiated materials. Therefore, this section 
focuses mainly on facilities for exposure to ir-
radiation and handling of irradiated materials. 

Notwithstanding the focus of this section, it 
is clear that equipment in furnaces to per-
form for example very long-term creep and 
creep-fatigue testing, or loops for exposure 
to flowing coolants, are crucial facilities that 
are in fact not so common and may not be 
available in sufficient quantity for a full qual-
ification of materials for sustainable nuclear 
energy. This section summarises the main 
features and issues related with the facili-
ties for irradiation and handling of irradiated 
materials. More details can be found in the 
Technical Annex, section TA.6,71 which is an 
extended version of the present one.

6.1.  Irradiation Facilities

6.1.1.  Fast Neutron Facilities

Neutron irradiation facilities and associated 
‘hot’ cells laboratories are a central neces-
sity for performance and safety testing of all 
types of structural and fuel materials. Since 
the ESNII demonstrators are all fast neutron 
spectrum systems, materials for their con-
struction should be qualified in fast neutron 
irradiation spectra. Unfortunately, there is cur-
rently no fast neutron power or testing reactor 

108 

operating in Europe. To redress the situation, 
a fast system should be built in Europe, but 
this leads to a vicious circle, because fast neu-
tron data are needed for this purpose. This 
means that at the moment Europe is totally 
dependent  on  non-European  countries 
to have access to fast neutron flux facili-
ties. The construction of at least one of the 
ESNII demonstrators within the coming two 
decades is thus essential in order to provide 
Europe with a facility to qualify structural and 
fuel materials for commercial GenIV reactors.

6.1.2.  Material Testing Reactors and 

associated ‘hot’ cells

Materials  testing  reactors  (MTR)  are  high 
power research facilities that can be used to 
expose materials to operation-like conditions. 
Flexibility  and  ability  to  adapt  to  changing 
needs  is  a  fundamental  principle  for  such 
reactors. MTR can perform a variety of test 

irradiations  simultaneously,  the  number  of 
these depending on their design and on the 
nature of the tests. In specific cases, MTR can 
be augmented by loops emulating the coolant 
flow conditions in power reactors, although in 
practice devices of this type are very limited 
in existence and availability – and none exist 
in a fast spectrum. Almost all MTR have on-
site ancillary ‘hot’ cells to make preliminary 
or detailed examinations, as well as to ena-
ble packaging and distribution of samples to 
other laboratories. Among the approximately 
40 MTR currently operational in the world, 
Europe  hosts  seven  (see Table  13). Two  of 
these can be used for fuel testing and have 
appropriate ‘hot’ cells available. The others do 
not have currently a license to handle fuel, 
but are available for structural materials’ in-
vestigations. These centres are equipped with 
‘hot’ cells, as well. Europe’s neutron irradia-
tion capacity is currently very limited: the 
opportunities to test new materials in reac-
tors are therefore restricted to the extreme. 

Table 13: Materials Testing Reactors currently operational in Europe and their irradiation characteristics.

Name

Location

BR2

HFR

Mol, Belgium

Petten, The Netherlands

LVR-15

Řež, Czech Republic

MARIA

Świerk-Otwock, Poland

Triga Pitesti

Pitesti, Romania

Maximum fast neutron 
flux (>0.1 MeV)
[1014 n /cm2 s] (from *)

7

5.1

3

1 

1.8

Dose rate
[dpaFe/fpy]

Up to 5

<7

~1

~1

Accessible 
temperatures
[°C]

50-1200

80-1100

50-850

50-100

 2.5

80-300 (→ 500 future)

Triga Mark II

Ljubljana, Slovenia

0.06

< 0.01 

20-50 (→ 300 future)

(*) https://nucleus.iaea.org/RRDB/RR/ReactorSearch.aspx

For irreplaceable component materials, the 
doses  expected  at  the  end  of  life  can  be 
reached in existing MTR, although in order to 
have exposure times much shorter than the 
real ones the fluxes used are much higher. 
While this may be a problem especially when 
the operating temperatures are high and the 
synergy between mild irradiation and thermal 

ageing and/or thermal creep may lead to un-
known effects, MTR available in Europe are 
at  least  usable  to  reach  end-of-life  doses, 
with the only caveat of the different neutron 
spectrum. However, for fuel and replaceable 
component materials, especially cladding, the 
doses expected in service are much higher 
than those accessible in thermal MTRs. The 

Strategic Research Agenda of the EERA JPNMonly way round this problem to reach doses 
up to or in excess of 100 dpa is currently the 
use of charged particles, as described in the 
next section.

6.1.3.  Charged particle irradiation facilities

Charged particles can be electrons, protons or 
ions of different weight. Although the penetra-
tion of protons and light ions can be significant, 
all  these  particles  are  obviously  efficiently 
stopped by electrons and ions inside the ma-
terials, thereby affecting only limited volumes, 
insufficient to produce standard specimens for 
e.g. mechanical property testing. Moreover, 
there are spectral differences with respect to 
neutrons in terms of damage that is produced 
and the progressive slowing down results in 
damage production gradients, while chemical 
species initially absent in the material and/
or atoms in excess are injected in the target 
material, including sometimes unwanted im-
purities that are difficult to control, e.g. carbon. 
Despite these shortcomings and limitations, 
which certainly prevent their use for full qual-
ification purposes, charged particle irradiation 
is useful to get insight into the behaviour of 
materials under irradiation (see sections 2.3.2, 
3.1.2 and 3.2.2).

A significant number of facilities permitting 
charged particle irradiation, which can be 
exploited for modelling and screening pur-
poses, exist in Europe at nuclear research 
centres and universities, including one facility 
usable for fuel. 

6.2.  Handling of irradiated 

materials

6.2.1.  ‘Hot’ Cells and shielded facilities138

Materials  exposed  to  neutron  irradiation 
need to be handled, tested and examined 
in dedicated ‘hot’ cells or similarly shielded 

138  In addition to those associated with MTRs.

109 

facilities. In particular, capsules taken out of 
the reactors need to be safely dismantled and 
‘hot’ cells on the sites of all European MTRs 
are an integrated component of any exper-
iment.  For  specimen  fabrication  and  large 
specimen  testing,  e.g.  mechanical  testing, 
dedicated ‘hot’ cells are absolutely necessary. 
For microstructural specimen fabrication and 
characterization,  or  in  other  specific  cases 
where the quantity of material to handle is 
small,  simpler  types  of  shielding  are  often 
sufficient. But invariably these are very costly 
infrastructures, both to build and maintain, 
and are available and employed only in a few 
equipped and licensed research centres, un-
der severe safety rules for the operators, and 
only very rarely in universities. 

The main ‘hot’ cell facilities in Europe are lo-
cated next to the European MTRs (see section 
6.1.2). In addition some centres are equipped 
with ‘hot’ cells even without possessing an 
MTR. Yet the number of centres that dispose of 
facilities of this type remains limited and in sev-
eral cases the maintenance cost are expected 
to lead to closure without replacement. In par-
ticular, the availability of ‘hot’ cells suitable for 
fuel handling is very limited and endangered. 
On the other hand, the number of sites where 
new ‘hot’ cell facilities have been recently 
built or refurbished is limited. 

The limited availability of ‘hot’ cell facilities, 
especially those licensed for fuel handling, 
whose availability has declined steadily over 
the decades, limits enormously the number 
of tests, measurements and examinations 
that can be performed on irradiated mate-
rials. This determines the fact that complete 
post-irradiation examination (PIE) of samples 
from irradiation experiments may take sev-
eral years to be completed. This problem 
is exacerbated by the ever rising costs and 
difficulties of irradiated materials transport 
(see section 6.2.2) and the fact that current 
PIE include advanced microstructural exami-
nation with new techniques that did not exist 
when most ‘hot’ cell facilities were built in 
Europe. In the case of fuel, the limited ‘hot’ 
cell capacity also has important negative 
consequences  on  Europe’s  capability  of 

6. Infrastructures for nuclear materials R&D110 

fuel procurement, since these facilities are 
indispensable for the preparation and safety 
testing of new fuel.

In conclusion, despite fair availability, Europe’s 
aged ‘hot’ cell capacity is already insufficient 
today and, as they age further, their replace-
ment will be imperative. Given the time and 
budget necessary to build and commission a 
‘hot’ cell facility, new replacements should 
be  planned  right  now  at  national  and/or 
European level.

6.2.2. Transportation of samples for 

dedicated analysis

Suitable, flexible transport flasks and contain-
ers for irradiated test specimens are essential 
to make an effective use of Europe’s currently 
geographically separated test reactors, irra-
diation facilities and ‘hot’ cells, as well as of 
the specificities of the various facilities. The 
capability to ship irradiated materials to and 
between  appropriate  facilities  is  an  essen-
tial  component  of  European  infrastructure. 
Currently, in addition to the inherently high 
costs of this type of shipments, the existence 
of differences in the laws and rules applied 
to  radioactive  material  transport  through 
European  countries  often  causes  this  type 
of transports to suffer from unduly very large 
delays, which effectively increase the overall 
costs, reduce the effectiveness of the research 
and prevent the timely completion of projects. 
Improved coordination and standardisation 
of regulations and transport containers are 
here essential to reduce transport times and 
costs, and to break the strangle hold of ever 
increasing costs for fewer transports. A paral-
lel effort can be done in miniaturising the size 
of the samples to be transported for specific 
analysis, for example by employing FIB.

described in section 2.3.2, as well as the sim-
ulations using design and fuel performance 
codes, are computationally intensive and call 
for access to world class high performance 
computing (HPC) systems. 

More generally, this access is essential for 
international competiveness in science and 
engineering. The importance of developing 
HPC capacity has been recognized by coun-
tries such as USA, Japan or France since the 
1990’s. Europe has recently acknowledged 
that HPC was a strategic resource for Europe’s 
future  and  the  need  for  a  European-level 
policy to optimise national and European in-
vestments and adopted its HPC Strategy in 
2012. Industry and SMEs are increasingly rely-
ing on the power of supercomputers to come 
up with innovative solutions, reduce cost and 
decrease time to market for products and ser-
vices. This is also true in the development of 
nuclear systems. 

Therefore, in addition to the development of 
the  global  European  HPC  capacity,  efforts 
must be made in the nuclear community to 
have  the  investigations  on  nuclear  mate-
rials recognized as a top priority subject in 
Europe and guaranteed access to significant 
computational  resources.  Currently, while 
fusion research can rely, as part of the EURO-
fusion consortium, on a funded HCP facility, no 
equivalent exists for fission research. This is a 
foremost condition for advanced modelling to 
bring the appropriate support to the develop-
ment and qualification of the materials needed 
for, especially, GenIV systems. 

6.4.  General considerations  

on nuclear infrastructures  
in Europe

6.3.  Computational facilities

6.4.1.  Costs related to crucial 

infrastructures build and use

Even if the need for computing facilities is not 
specific to the investigation of nuclear materi-
als, it is worth stressing that the ICME approach 

The construction and maintenance of infra-
structures and facilities such as MTRs and ‘hot’ 
cells is beyond the scope of the EERA JPNM 

Strategic Research Agenda of the EERA JPNMactivities and capabilities. It is important, howe-
ver, to emphasise the importance of this point. 
In particular, if it is out of place to attempt esti-
mates of infrastructure investment costs here, it 
is useful to remind that they have been detailed 
in European projects such as ADRIANA,139 and 
vary considerably, with material testing reac-
tors lying in excess of 1000 M€, and ’hot’ cells, 
depending on scope, in the 100-300 M€ range. 
Typical single fuel irradiation experiments cost 
1-2M€, depending on complexity, instrumen-
tation required, and PIE effort. The cost of the 
transport of irradiated fuel can reach 150 k€, 
although in a well-planned oriented approach 
can drop to as little as 15 k€; active structural 
material specimens can be transported with 
costs varying between 5-30 k€. The advent of 
micro sampling techniques (e.g. FIB) for dedi-
cated examination on ultra-small samples can 
reduce these costs even further.

In addition, Tier-0 HPC resources, which do not 
serve the whole European scientific and indus-
trial community, have to be renewed every 2-3 
years, with construction costs between 200 
and 400 M€ and annual running costs around 
100 M€/yr.

6.4.2. Renewal of nuclear infrastructures

A research agenda can only be fulfilled if it is 
matched by appropriate and timely available 
infrastructures. Therefore, in the present con-
text, in order to maintain its competences 
and progress further in the field of nuclear 
materials, Europe should, through its research 
organisations located in different MS, and thus 
with the full commitment of the latter:

 y Ensure  that  a  fast  neutron  flux  facility 
comes available soon (e.g. ESNII demon-
strator);

 y Construct at least one new MTR in addition 
to JHR (i.e. PALLAS, MYRRHA, the latter with 
the added value for GenIV applications to 
offer a fast neutron spectrum);

111 

 y Renew and possibly extend its ‘hot’ cell 
capacity, taking due account of transpor-
tation issues;

 y Ensure that ‘hot’ laboratory capabilities, in 
particular for fuels, do not decrease further;
 y Ensure appropriate access to high perfor-
mance computing for the fission commu-
nity;

In addition, three recommendations for poli-
cy-makers, that are further developed in the 
next section, are in order:

1. 

 Plan major infrastructures judiciously at a 
pan-European level in a harmonised way;
2.  Implement consequent open access and 

infrastructure sharing initiatives;

3.  Foster joint programming (fusion energy 

provides an example to follow).

6.4.3. Sharing and joint programming  
of nuclear infrastructures

Access to and sharing of facilities for nuclear 
materials exposure, testing and examination, 
especially controlled zones where radioactive 
materials are manipulated, can be problem-
atic  for  legal,  security,  safety  and  financial 
reasons, namely:

 y Legal: Protection of know-how & expertise: 
there is often reluctance to give full open 
access as this may reveal details on pro-
tected know-how (this attitude is however 
often inconsistent even within the same or-
ganisation);

 y Related to security: Access to ‘hot’ cells re-
quires clearance from authorities for secu-
rity reasons: this takes weeks or months and 
a significant administrative burden;

 y Related  to  safety:  Only  trained  &  skilled 
operators can safely use some equipment, 
especially in ‘hot’ labs (manipulators…);
 y Financial: Availability of specific equipment 
has a high cost and opening it for access to 
external users might limit the profitability of 
facilities. 

139  http://ojs.ujf.cas.cz/~wagner/transmutace/erinda/pre-
sentations/05_ADRIANA_ERINDA.pdf

Notwithstanding these difficulties, solutions can 
be envisaged, for example a scheme of mutual 

6. Infrastructures for nuclear materials R&D112 

compensation between organisations within a 
“virtual research centre”, in-kind or in-cash. An 
in-kind type of compensation is for example by 
seconding employees from A to B in a stable 
way. This could happen under specific bilateral, 
as well as multilateral, agreements concerning 
non-disclosure of know-how, use of manpower 
for host purposes, etc. 

It should be stressed that mobility scheme of 
this type provide also a key motor to drive ed-
ucation and training of not only researchers, 
but also operators, enabling Europe to main-
tain and actively manage key competences 
currently dwindling in the nuclear field, for the 
benefit of all nuclear research centres (see 
section 7).

To facilitate both sharing and mobility, in a con-
text of scarcity of infrastructures and financial 
means, and in order to foster pan-European 
harmonised infrastructure planning , joint pro-
gramming is key. In particular, it is believed that 
the creation of a permanent joint European 
planning forum and management scheme of 
nuclear infrastructures is a timely need.140 

The joint forum should mainly agree upon ir-
radiation future needs, in order to plan and 
propose harmonised European investments. 
It  may  also  express  views  on  the  rules  for 
financial coverage for the use of nuclear infra-
structures.

The joint scheme for the management of nu-
clear infrastructures should fulfil the following 
conditions:

1.  Be a single entry point, i.e. a single inter-
locutor whom users should address for the 
use of available infrastructures;

2.   Coordinate the best use of infrastructures 
for a given experiment, by pooling facilities 
and experts in Europe, to design the best

140  This could be a European version of the American NSUF: 
https://nsuf.inl.gov/

  experiments possible (also the cheapest ...), 
establishing if and when there are advan-
tages to access non-EU infrastructures;

3.  Benefit from the harmonised commitment 
of MS to pay the costs of the infrastruc-
tures and use, as part of the engagement 
to innovation, but develop flexible manage-
ment attracting collaborative projects and 
also industrial users;

4 

 Distinguish between R&D users (low cost 
open access, selection based on merit of 
proposal, open access data and results, ... 
also open to non-MS) and industrial users 
(charged services, data protection, …).

Within  such  a  scheme,  tenders  could  be 
launched and/or differences between avail-
able  facilities  could  be  optimally  used  to 
explore specific effects, including, whenever 
suitable, fundamental studies and the use of 
ion irradiation. In particular, such a scheme 
would allow the best use of available space 
in reactors to be made, through the design of 
joint campaigns. For instance, while it is true 
that MTR operation is expensive and cannot 
be offered for free, irrespective of the open 
access nature of the facility, under some con-
ditions it is possible to perform inexpensive 
“piggy-back” irradiation experiments, in a con-
certed framework, especially if experiments 
funded by industrial users can be partially 
used for this purpose at no detriment of the 
customer.

Of course, for this scheme to see the light 
the willingness of sharing and joint pro-
gramming must exist at all levels, including 
high management of research centres as 
well as MS policy-makers. This should also 
be properly fostered by Europe-driven ac-
tions.  It  remains  to  be  seen whether  the 
conditions for this approach to materialise 
actually exist.

Strategic Research Agenda of the EERA JPNM7

EDUCATION, 
TRAINING AND  
MOBILITY OF  
RESEARCHERS

© SCK-CEN

7.  EDUCATION,  

TRAINING AND  
MOBILITY OF  
RESEARCHERS

Nuclear fission research and competence in 
Europe may be in danger of decline due to the 
decrease in the number of new researchers 
and operators entering the nuclear commu-
nity, while they are required to replace previous 
generations. This endangers the possibility of 
preserving the existing knowledge and skills. 
Several factors can be identified as origin of 
this problem, mainly related to the uncertain 
professional perspectives that the nuclear field 
currently offers in Europe, as well as in part to 
the lack of sufficiently funded challenging and 
attractive projects, such as those offered in 
contrast by fusion energy, renewable energy 
or energy integration. A suitable and inher-
ently attractive education and training (E&T) 
programme is therefore required to reduce 
the risk of a future shortage of nuclear skills 
and ensure the maintenance of the acquired 
knowledge and expertise. 

In this respect, nuclear materials can be a 
way to attract young researchers to the nu-
clear field, thanks to the inherent cross-cutting 
nature of materials science through several 
technologies. Nuclear materials courses with 
a specific focus on nuclear structural and fuel 
materials  with  interdisciplinary  integration 
and emphasising both research and industry 
needs, may act as catalysers to attract more 

young researchers to nuclear energy pro-
jects, integrating expertise from other fields 
into specific nuclear applications. GenIV plays 
here an important role, by involving interesting 
scientific challenges.

Young researchers and operators should be 
trained in particular by using both common and 
specialized facilities that, regardless of size, 
provide a key engine to drive education and 
training, enabling Europe to maintain and ac-
tively manage currently dwindling key nuclear 
competences. In this respect, infrastructure 
sharing is a key driver, to be made effective 
within joint programming schemes (see sec-
tion 6.4.3). A homogenous and coordinated 
network of nuclear facilities and infrastructures, 
together with a stable scheme of mobility of 
young researchers and operators, is expected 
to enable wider competences to be built on 
various aspects of nuclear technology and, by 
exploiting its inner cross-cutting aspects, also 
towards other technologies. Some exchange 
schemes have already been conducted or are 
planned in the framework of FP7 and H2020 
projects (e.g. GENTLE and ENEN+).

To obtain maximum benefit, an actively man-
aged cross-European education and training 
and mobility (E&T&M) programme with specific 

116 

focus on nuclear structural and fuel materials 
should apply outcome-based teaching meth-
ods, focusing on interdisciplinary integration 
and emphasizing both research and industry 
needs.  It  should  extend  and  complement 
existing  university  curricula,  training  pro-
grammes and schools, without doubling them, 
while being recognised by European universi-
ties in terms of credits (ECTS). To be attractive, 
it should enable the development of knowl-
edge and skills in materials science that can 
also be useful outside the nuclear field, as 
well as the possibility of hands-on training via 
access to available experimental facilities. 
The ultimate goal here is that an attractive ex-
tra-curricular course on materials for advanced 
nuclear systems of the type described should 
be offered periodically to all students and 
young researchers. 

In order to reach this goal, the EERA JPNM 
considers  that  a  coordination  in  terms  of 
E&T&M initiatives through nuclear platforms, 
i.e. SNETP and its pillars, in close collabora-
tion with the European Nuclear Education 
Network (ENEN), is the right way to go. ENEN 
has indeed the aim to encourage E&T related 
co-operation and integration between insti-
tutions  and  to  attract  qualified  students  to 
participate in special E&T national and interna-
tional programmes. In collaboration with ENEN 
and in coordination with SNETP and its pillars, 
the  EERA  JPNM  can  then  act  as  the  focal 
starting point to assure up-to-date and sus-
tainable continuity in the E&T programmes 

for nuclear structural and fuel materials, by 
identifying and providing the high-level experts 
and research infrastructures which are availa-
ble for training purposes, as well as suggesting 
suitable training topics and schemes.

The  harmonisation  of  E&T&M  activities  be-
tween EERA-JPNM, SNETP and its pillars, with 
the support of ENEN, should also move in the 
direction of identifying a series of periodic 
short training courses, summer schools or 
workshops that address topics considered 
important  for  the  whole  nuclear  energy 
community, and organising them by mak-
ing coordinated use of the funds distributed 
through different Euratom-supported pro-
jects for E&T purposes. The idea here is to 
reverse the current situation, where each pro-
ject is delegated the task of identifying suitable 
E&T activities, to replace it with a cross-plat-
form  joint  medium  term  planning  of  E&T 
activities, to be funded through the sources 
that case-by-case are available from projects, 
topped up with the financial and logistic sup-
port that the different organisations involved 
may offer. Similarly, mobility schemes should 
be set up, as well.

Depending on the existence of suitable funding 
schemes, the EERA-JPNM proposes here a po-
tential concept to develop an E&T programme 
that should offer different learning packages, 
to give the participants the possibility to ac-
quire specific and needed competences, as 
described in Table 14.

Strategic Research Agenda of the EERA JPNM117 

Table 14: Description of a possible E&T programme on nuclear structural and fuel materials, as envisioned by 
the EERA JPNM.

What

•  A learning package - E&T program devoted solely to nuclear materials (structural materials 

and fuels)

When

•  Annually or bi-annually, at predetermined dates

•  Master’s students from relevant engineering disciplines

Participants

•  PhD students

•  Employees from the nuclear industry

Class size

•  Class size is optimally 20 people, max. 30 people

•  Suitable locations can be identified after sending out respective explanations and a 

questionnaire to all known institutions asking for their availability 

Where

•  Sessions focusing on computational exercises, lectures and in-class problem solving 
are not particularly location-dependent and can be held alongside other workshops, 
conferences etc. of the nuclear material community

•  Suitable experimental facilities must offer at least partial hands-on experience to students

•  4 weeks in total in suitable institutions, consisting of 2-4 individual sessions (2+1+1 weeks or 

2+2 weeks or 1+1+1+1 weeks)

Length

•  1-2 weeks of optional, preparatory online learning in advance to bring everybody up to the 

same level

•  Participation during the entire year is necessary to meet the goal of high-level competence

•  Hands-on experimental assignments

•  Hands-on computational exercises

Teaching 
methods

•  Learning through in-class and at-home comprehensive problem solving

•  Lectures leading in assignments and assisting in their solving

•  Field trips to the industry and research facilities

•  E-learning whenever suitable

Topics

•  Suitable topics for the 4-week period should be defined by the trends for future research 

needs, the needs of students and the competence of teachers willing to participate

•  Outstanding researchers and experts from universities and the industry who are interested 

Teachers

in teaching and applying effective teaching methods to achieve the intended learning 
outcomes

Teaching 
support

•  A short course to motivate applying successful teaching methods

•  Guidelines for desirable lecture composition and assignment design 

•  Topics per session limited to a small number so that sufficient time is given to students to 

Suggested 
curriculum 
features

listen, understand and apply new knowledge

•  A high problem-solving-to-lecture ratio should be applied

•  Teacher-assisted independent work in small teams throughout all sessions

•  All assignments related to topical research interests and industry interests 

Management 
aspect

•  Fewer topics per week means fewer teachers per week which results into smoother time 

management and potentially reduced costs

•  Annual course on predetermined dates simplifies planning and availability both for teachers 

and students

7. Education, training and mobility of researchers8

INDUSTRY 
AND  
REGULATORS 
INVOLVEMENT

© PF.Grosjean/CEA

8.  INDUSTRY AND 
REGULATORS 
INVOLVEMENT

The EERA-JPNM is a public research organisa-
tion platform that does not deal with the actual 
design, licensing, construction and operation of 
GenIV prototypes in which, ultimately, materials 
are going to be actually used and therefore fully 
qualified, based on return of experience, in the 
proper operational environment. It is however 
essential to have a close link to the industrial 
application, because in terms of technology 
readiness level (TRL6) the EERA-JPNM mainly 
works between TRL 2 and 5, i.e. just up to the 
level where the material becomes industrially 
usable, by having been developed and quali-
fied to the pre-normative level.

Three classes of stake-holders are the indus­
trial counterpart of the EERA JPNM:

A.  Reactor designers and constructors: The  
EERA  JPNM  has  to  know  very  clearly 
the goals and the needs of the reactor 
designers and constructors, in order to 
orientate the research towards the sup-
port  of  the  processes  of  licensing  and 
construction of advanced nuclear systems, 
for which return of experience is limited 
or  even  non-existent.  In  addition,  the 
results of the research of the EERA JPNM 
need to be appropriately and efficiently 
transferred  to  the  reactor  designers 

and  constructors,  in  the  form  of  data/
knowledge, through suitably updated and 
managed databases (see section 2.4). Fast 
reactor  designers  and  constructors  are 
almost all involved in ESNII, which explains 
the strong link with ESNII.

B.  Reactor  operators  (nuclear  electro-
producers):  The  reactor  operators  are 
the  customers  of  the  reactor  designers 
and constructors, they essentially decide 
whether or not innovative nuclear systems, 
and  which  ones,  are  attractive  for  their 
business. In particular, they decide about 
the economic relevance of closing the fuel 
cycle,141 although political decisions have 
also  an  influence  there.  Finally,  reactor 
operators  are  those  that  may  provide 
return of experience on materials in service 

141  Nuclear  electro-producers  have  currently  drawn  back 
from  investing  in  technologies  to  close  the  fuel  cycle  be-
cause  of  uncertain  future  of  nuclear  energy  in  some  coun-
tries and in Europe at large and because of renewable pen-
etration leading to load follow mode. Incomes are reduced, 
while  cheap  fossil  fuel  energy  sources  remain  convenient 
(especially coal). Therefore without public and political sup-
port, also financial, towards sustainability of nuclear, GenIV 
technologies will have difficulties to emerge on the nuclear 
arena. Nonetheless, even if industrial interest is shrinking at 
the moment, the research on materials at TRL<5 remains cru-
cial to enable the development of simulation codes, design 
codes etc.

122 

behaviour  (whether  relevant  or  not  for 
innovative systems). Although designers 
and constructors are those mostly interested 
in interacting with reactor operators, as a 
research platform the EERA JPNM has to 
be aware of the industrial experience and 
needs concerning materials in service, as 
well as to know towards which innovative 
solutions they consider moving. Operators 
are mainly involved in NUGENIA, hence 
the  importance  of  establishing  a  close 
relationship with this platform.

C.  Materials manufacturers:  the process of 
development of new materials, especially 
when innovative fabrication routes are 
explored, eventually requires upscaling 
to industrial production. This upscaling 
is not necessarily simple, because it is not 
guaranteed that a certain type of material 
can be efficiently and affordably produced 
at  larger  production  scale,  outside  the 
laboratory where it has been developed. For 
example, currently no large scale industrial 
producer of ODS steels exists in Europe, 
despite the fact that several laboratories can 
do that, in small or even very small batches. 
Industrial production may require adaptation 
or even complete changes in the fabrication, 
processing and treatment of the materials. 
Moreover, innovative fabrication processes 
(for instance additive manufacturing) are 
currently emerging, the industrial use of 
which should be enabled. Thus the EERA 
JPNM should strive to work hand­in­hand 
with the materials manufacturing industry, 
as early as possible along the materials 
development route, in order to take into 
account  industrial  production  upscaling 
as  a  criterion.  Several  platforms  include 
materials manufacurers, for example ESTEP 
(steel-makers)142, EMIRI143, EUMAT144, …

In addition to industries, safety authorities and 
regulators, or at least technical and scientific 
support organizations (TSO), would ideally be 
an important counterpart for the EERA JPNM. 
Although the interaction with regulators and 
TSOs is mainly the duty of the system design-
ers, because of the need to refer to specific 
designs and service conditions in connection 
with the licensing, it is considered beneficial 
for regulators and TSOs to follow the proce­
dures used for materials qualification and 
possibly guide them from a safety point of 
view, in order to eventually accelerate the li­
censability of nuclear components. However, 
contacts with TSOs and regulators are quite 
difficult to establish.

The industrial involvement in the EERA JPNM 
activities is currently pursued in two ways:

1. Industries may join the EERA JPNM as as­
sociate members. Currently a few industries 
(reactor designers/constructors, operators, 
and materials manufacturers/steel-makers), 
participate directly in the EERA JPNM projects, 
also with in-kind contributions, as partners.

2. Industrial representatives that prefer not 
to commit themselves to active participation 
in the EERA JPNM projects, but are ready to 
provide guidelines and feedback to the EERA 
JPNM activities and to have partial access to 
the results obtained (following case by case in-
tellectual property right –IPR- rules and based 
on the signature of a non-disclosure agree-
ment) are invited to join the stake­holders’ 
group (SHG) of projects under the umbrella 
of the EERA JPNM.

Stake-holders are expected to be involved in 
task forces aimed at drawing specific research 
plans.145 The wish for the future is that also 
TSOs and regulators may join SHGs of EERA-
JPNM projects.

142  https://www.estep.eu/estep-at-a-glance/

143  http://emiri.eu/about

144  http://www.eumat.eu/

145  See  for  example  the  task  force  report  on  60 years  ope-
rational  life  of  reactors,  available  at  http://www.eera-jpnm.
eu/?q=jpnm&sq=nboard

Strategic Research Agenda of the EERA JPNM9

INTER­
NATIONAL 
COOPERATION

© SCK-CEN

9.  INTERNATIONAL 
COOPERATION

International cooperation can provide a real 
boost to progress towards innovation in the 
nuclear energy field, which is much needed 
along the path leading to the deployment of 
GenIV systems. 

However, for international cooperation to be 
effective, appropriate instruments and in-
centives need to be put in place in Europe 
towards other countries, given that those that 
exist right now are insufficient, or not sufficiently 
efficient, or in the best case not sufficiently at-
tractive (and known) to researchers. 

Areas where effective international coopera-
tion could determine an important boost in the 
development and qualification of materials for 
advanced nuclear systems, provided the right 
instruments are offered, are for example:146

 y Optimized/harmonized  use  of 

infras-
tructures,  making  unique  facilities  trans-
nationally accessible and also, importantly, 
adequately planning their use to avoid du-
plications and redundancies, while making 
the best use possible of the specificities of 

146  Clearly,  these  are  first  and  foremost  areas  where  closer 
cooperation within Europe would already make a difference. 
Here they are put in a more general international framework.

each  particular  infrastructure:  this  should 
be the rational approach to make materials 
qualification complete and affordable.

 y Harmonization  of  procedures 

and 
methodologies  to  test  and  characterize 
materials,  especially  innovative  materials 
in  specific  environments,  including  pro-
tocols  to  perform  microstructural  exami-
nation  with  advanced  techniques  and  to 
analyse the results. In several cases, com-
pletely new tests need to be designed and 
standardized.  Moreover  the  design  codes 
RCC-MRx and ASME BVP Section III share 
the  same  goals,  so  also  in  this  area  there 
is  mutual  interest  to  share  and  compare 
methodologies for design rules and design 
curves.

 y Data collection and sharing, through sui-
table databases that should be eventually 
made  available  for  reactor  designers  (in-
dustry  and  not  only)  and  regulators:  the 
more  complete  and  extensive  the  data-
bases, the safer the corresponding design 
rules and the more conducive the action of 
the regulators. Importantly, data collection 
and  sharing  makes  sense  provided  that 
tests  are  homogeneous  as  to  procedures 
used,  so  this  point  links  strongly  with  the 

126 

above one. Ideally, the database produced 
in the framework of EERA JPNM could be-
come  an  official  database  for  the  design 
and construction of GenIV reactors.

data:  this  is  always  the  bottleneck  in  the 
case  of  databases,  thus  grants  should  be 
accorded  to  support  financially  data  see-
kers and collectors.

 y Synergy on modelling: modern modelling 
approaches  at  different  scales  inherently 
require  inter-disciplinarity,  computer  re-
sources,  theoretical  developments  and, 
crucially,  extensive  experimental  effort  for 
the characterization of materials at all sca-
les after exposure to a variety of conditions. 
Structured  international  coordination  and 
exchange can only be beneficial. The three 
points  above  (optimal  use  of  infrastructu-
res,  harmonization  of  test  and  characte-
rization  procedures,  and  data  collection 
and sharing) clearly enter here, as well. In 
addition,  there  is  a  need  to  make  mode-
lling  approaches  more  compatible  and 
complementary  with  each  other,  in  order 
to  better  focus  the  development  towards 
platforms of linkable codes and models.

Enablers can be identified in terms of inter-
national cooperation schemes. For example 
expert committees could be enabled to work 
on (some of) the issues listed here below. 

 y Harmonisation  of  test  and  characterisa-
tion  methodologies  and  procedures,  as 
pre-normative  step  to  standardization  by 
bespoke bodies (ASME, CEN, ISO, …): sup-
port  to  round-robin  exercises  addressing 
non-standardized  tests;  best  practices 
for  microstructural  characterization  tech-
nique  application  and  data  analysis,  with 
exercises of inter-comparison between la-
boratories for specific microstructural cha-
racterization techniques (TEM, APT, SANS, 
PAS, …).

 y Identification  of  unified  international  da-
tabases  of  reference  for  the  collection  of 
materials  testing  data;  establishment  and 
distribution of relevant data templates com-
patible with selected reference database(s); 
establishment  of  rules  concerning  protec-
tion  and  disclosure  of  data  collected  in 
databases (e.g. 10 years embargo on pro-
prietary  data);  encouragement  to  upload 

 y Overview  the  use  of  major  research  fa-
cilities worldwide (mainly MTRs,  but also 
other  materials  exposure  facilities:  high 
temperature and coolants), based on exis-
ting  lists  and  maps;  optimised  and  com-
plementary  use  of  these  facilities,  driving 
similar  facilities  to  non-overlapping  uses; 
design  of  large  joint  experimental  pro-
grammes  of  cross-cutting  interest,  using 
available  large  facilities  in  a  coordinated 
way (this may specifically apply in the case 
of modelling-oriented experiments).

 y Harmonization  in  the  development  of 
computer  simulation  materials  models 
for  better  mutual  complementarity  and 
com patibility;  identification  of  gaps;  mo-
delling data collections (according to cri-
teria similar to the above ones applying to 
testing and characterisation).

The role of international organisations, e.g. 
OECD-NEA or IAEA, is pivotal in this respect, 
to facilitate cooperation on nuclear energy 
at global level. For this reason, EERA signed a 
MoU with OECD-NEA, that is intended exactly 
to bring the JPNM to a higher level of interna-
tional visibility, by actively participating in, and 
providing expertise to, NEA initiatives such as 
working parties, expert groups, task forces, and 
so on. Specifically, an opportunity for effective 
international cooperation in which the EERA 
JPNM is fully involved is currently offered by 
the NEA Nuclear Innovation 2050 Initiative.147

In concrete terms, a more effective interaction 
with the GenIV International Forum (GIF148) is 
advocated. GIF is a cooperative international 
endeavour based on a Charter that was signed 
first in 2001 by a number countries, that has now 
grown to 14. While some countries are “dormant” 
members, especially active are United States, 

147  https://www.oecd-nea.org/ndd/ni2050/

148  https://www.gen-4.org/gif/jcms/c_9260/public

Strategic Research Agenda of the EERA JPNM127 

South Korea, France, Switzerland, and Australia 
that recently joined, as well as, to a lesser extent, 
Japan and also China. The Russian Federation is 
also member of the GIF. The goal is here to share 
data of relevance for GenIV systems. Euratom 
is a signatory of the GIF Charter, although only 
some of the EU countries are involved. Thus 
the connection with GIF should occur mainly 
through Euratom, with the pivotal role of JRC 
as interface. The EERA JPNM has for the mo-
ment only informal contacts with the GIF project 
devoted to materials for VHTR, in whose frame-
work some activities of cross-cutting interest 
through several GenIV systems are included. 
Stronger connections can only be established 
within a legal support provided by Euratom.

Benefit is quite obviously expected through 
collaboration with non EU countries involved in 
the GIF, so USA, South Korea, Australia, Japan, 
China and Russian Federation, perhaps also 
Canada, because all these countries have spe-
cific activities on GenIV materials. Data sharing 
is the essence of the GIF, but most countries 
should also be interested in harmonizing test-
ing and characterization procedures, without 
which data sharing becomes less meaningful. 
Collaboration with the US on several fronts, 
from materials qualification to modelling and 
development of new materials, is relatively 
easy and instruments such as I-NERI149 are in 

place, although they are not in practice very 
attractive, especially on the European side.150 
Collaboration with South Korea on issues re-
lated with creep design rules for austenitic and 
F/M steels and compatibility of these materials 
with HLM is underway through Euratom pro-
jects (FP7/MATTER, H2020/GEMMA), in which 
Korean institutions participate as partners. In 
principle, a similar type of collaboration could 
be extended to US partners directly involved in 
the development of the ASME code. More dif-
ficult or less exploited is the collaboration with 
all other countries. Modelling could be a good 
ground for collaboration with Australia. Russian 
Federation, China and potentially also Japan 
are important partners to be considered be-
cause of the unique opportunity they may offer 
to expose materials to high dose in fast spec-
trum reactors, such as BOR60 in Russia, where 
for example experiments were performed in 
the framework of the FP7/GETMAT project. 
However, efficiency can only be guaranteed 
if (whenever needed) suitable agreements 
are  signed  (e.g.  between  Euratom  and  EU 
members  states  and  other  countries)  and 
appropriate funding is provided, not only to 
cover meeting and travel expenses of experts, 
but also for the work performed in preparation 
of them. Ideally, funded schemes for mobility 
of researchers should be set up in support of 
the four enablers listed above.

149  E.g. http://cordis.europa.eu/fp7/euratom-fission/docs/06-
haas-ws4_en.pdf

150  While in the US this scheme leads to chances for addition­
al funding, this is not the case in Europe. While collaboration 
with international key partners is always useful, there is often 
reluctance to disclose costly data, especially without any ad­
ditional financial incentive.

9. International cooperation10

RESOURCES

© S.Le Couster/CEA

10.  RESOURCES

It is challenging to foresee correctly the costs 
associated with research that is not meant to 
lead directly to the development of a defined 
technology and its release on the market, or 
to the construction of a system, but rather to 
provide the suitable and continuous R&D&I 
background to enable the development and 
release of such technology, as is the case for 
nuclear materials for advanced nuclear sys-
tems.  However,  based  on  past  experience 
and also on estimates coming from the de-
ployment of the research agenda of the EERA 
JPNM, excluding the cost of the construction 
and maintenance of the needed infrastruc-
tures and facilities, which are here taken for 
granted, one can assess that:

In order to adequately address the most im-
portant issues, strictly related with the needs 
of the ESNII demonstrators and their follow up 
(so, fast GenIV reactors, see Figure 9

Figure 10), with very limited industrial involve-
ment and also limited activities devoted to 
perspective innovative materials, excluding 
the cost of irradiation campaigns, subsequent 
transport and use of hot cells, a cost of about 
12-15 M€ per year should be budgeted.

 y If  all  costs  of  irradiation  qualification  cam-
paigns, transport and hot cells are included, 

for proper materials qualification (within the 
limits allowed by the facilities that are avai-
lable),  then  the  resources  should  range 
between 20-25 M€ per year. This amount 
should  also  enable  adequate  support  for 
education,  training  and  mobility  of  young 
researchers.

 y The  inclusion  of  significant  industrial  in-
volvement and extensive activities also on 
materials currently at low or very low TRL, 
extended to systems beyond those inclu-
ded  in  ESNII  (future  power  reactors,  but 
also  other  systems  such  as  VHTR,  MSR, 
SCWR,…),  would  about  double  the  costs, 
leading  to  the  upper  bound  estimate 
of  50  M€  per  year,  as  expressed  in  the 
Integrated Roadmap of the SET-plan.151

These resources clearly exceed by about an 
order of magnitude the possibilities of fund-
ing offered by Euratom. Therefore, not only 
the involvement, but also the commitment 
of the Member States is crucial. MS funding 
for research activities on nuclear materials for 

151  Heading 5 of Part II: “Supporting Safe Operation of Nuclear 
Systems  and  Development  of  Sustainable  Solutions  for  the 
Management of Radioactive Waste”. Action: “Qualify nuclear 
materials for operation under Gen IV conditions and develop 
innovative  materials  to  improve  plant  safety  and  efficiency”. 
See also Section 1.6.2.

132 

innovative systems already exists now, in par-
ticular stable salaries for researchers and the 
possibility of opening new positions in the field 
are guaranteed through MS provisions and 
support to research centres and universities, 
meaning that there is already substantial MS 
involvement. However, there is currently no 
coordinated commitment to earmarked MS 
funding through Europe. In contrast, this ex-
ists, for instance, in the case of fusion energy, 
which is taken here as a successful example. 

Fusion energy receives support under differ-
ent forms, including through joint undertakings 
(Fusion for Energy, F4E) and international agree-
ments (ITER, Broader Approach). Another form 
of support, in the EU, is through the co-fund 
instrument called European Joint Programme 
(EJP),  namely  EUROfusion. This  instrument 
appears to be especially suitable to enable 
the joint implementation of a roadmap, using, 
among other tools, internal calls for projects, 
schemes of E&T&M, etc. with clear MS com-
mitment  combined  with  Euratom  support. 
Of course, such a funding scheme for fusion 
energy is explained by the widespread sup-
port to this research through essentially all EU 
Members States.

It is here suggested that nuclear materials at 
large are a subject that can also find wide-
spread support for an EJP through several 
MS, by virtue of its cross-cutting nature (see 
section 4). The present SRA is a good starting 
point for this, as it shows the existence of an 
already  established  and  thriving  European 
nuclear materials research community, partly 
under the umbrella of the EERA JPNM, and 
partly also under the umbrella of SNETP pil-
lars, NUGENIA in primis. There are also strong 
connections  and  overlaps,  in  practice,  be-
tween researchers working in nuclear fission 
and fusion materials. It is therefore believed 
that an EJP, or any equivalent co-fund instru-
ment, could be suitable also to support the 
established nuclear materials community, 
with substantial advantages in terms of opti-
mal use of resources, by avoiding duplications 

and  redundancies,  as  sometimes  happens 
between fusion and fission. However, without 
support for projects specifically aimed at de-
signing and constructing GenIV demonstrators, 
such as those considered in ESNII, the driving 
force to sustain activities on nuclear materials 
for GenIV will be scarce, beyond fundamental 
research motivations.

Figure 11 shows the approximate subdivision 
of resources through EERA JPNM activities, 
based on the estimated costs of EERA JPNM 
pilot projects from the recent past, as well as 
from the actual cost of research activities in the 
framework of European and national or institu-
tional projects.

Figure 11: Indicative estimated subdivision of funds 
between research activities, based on past experi-

ence, assuming all activities are sufficiently funded.

Strategic Research Agenda of the EERA JPNM11

RISK 
ASSESSMENT

134 

© CVR

Strategic Research Agenda of the EERA JPNM11. RISK ASSESSMENT

This SRA suggests a path to be followed to 
ensure that suitable materials are made avail-
able for the design, licensing and construction 
of GenIV demonstrators, with a view also to 
the longer term construction of FOAK proto-
types and then deployment of commercial 

reactor. However, all activities described in 
this SRA are exposed to sometimes very crit-
ical risks. Table 15 (next page) provides a list of 
those that are perceived as more prominent 
in a number of categories addressed in this 
document.

136 

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Strategic Research Agenda of the EERA JPNM 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
12

SUMMARY 
AND  
RECOMMEN-
DATIONS

142 

© S.Le Couster/CEA

Strategic Research Agenda of the EERA JPNM12. SUMMARY AND
RECOMMEN-
DATIONS

This SRA identifies the research lines to be pur-
sued in order to ensure that suitable structural 
and fuel materials are available for the design, 
licensing, construction and safe long-term op-
eration of GenIV demonstrators, prototypes 
and later commercial reactors, with empha-
sis on ESNII fast neutron spectrum systems, 
namely SFR, HLM-cooled systems (ADS and 
LFR), and GFR. Fast reactors create more fuel 
than they burn and operate at high tempera-
ture using passive safety systems, so increasing 
enormously the sustainable use of resources 
and guaranteeing safe energy production for 
centuries, also reducing waste production.

and  fuel  materials,  of  a  well-established 
materials  science  approach  based  on  the 
combination of three classes of activities: (1) 
materials testing for full qualification and defi-
nition of design rules in a pre-normative spirit; 
(2) development of mechanistic and physical 
models in support of materials behaviour cor-
relations used to define design rules and of 
the improvement of materials properties; (3) 
development of innovative material solutions 
through experimental screening, assisted by 
models rooted in the understanding of the 
physical processes that govern materials be-
haviour.

The materials considered here cover as priority 
the needs of the ESNII demonstrators, but at-
tention is given also to materials solutions that 
are intended for FOAK and commercial GenIV 
systems, in which higher energy efficiency and 
longer burnups than in the demonstrators are 
targeted, seeking also connections with other 
GenIV systems, not included in ESNII, i.e. VHTR, 
SCWR and MSR. The content of this SRA is con-
sistent with other relevant strategic documents 
and roadmaps compiled by other platforms, in 
other frameworks.

The research activities are organized in blocks 
that result from the application, for structural 

Effort  has  been  devoted  to  identify  issues 
addressed  in  a  GenIV  materials  research 
framework that are of interest for other nuclear 
(current LWR, fusion) and also non-nuclear (e.g. 
concentrated solar, geothermal, bioenergy, 
fuel cells and hydrogen, …) energy technol-
ogies, with  a view  to  optimising  the  use  of 
resources, whenever possible, by joining forces 
with other research communities.

This document also addresses issues that are 
necessary corollaries to the proposed research 
activities,  namely:  infrastructures  needed, 
need for education and training and mobility 
schemes, industry and regulators involvement, 

144 

importance of international cooperation, esti-
mate of resources required, risk assessment.

Besides  the  obvious  need  of  adequate  fi-
nancial  resources  in  order  to  address  the 
research problems outlined in this document, 
a  few  recommendations  emerge  that  this 
document intends to bring to the attention of 
stake-holders, particularly research managers 
and decision-makers:

R1:  Data from materials property measure-
ments after exposure to relevant conditions 
are the essential ingredient for robust design 
curves and rules. Plenty of data were produced 
in the past that are now de facto unusable; this 
is either because they are covered by confi-
dentiality or because they were not properly 
archived. Correct data management to guar-
antee availability for future re-assessment is 
therefore essential and should be encouraged 
and fostered. In particular, financially sup-
ported policies to foster data sharing and 
encourage  old  data  disclosure  should  be 
implemented.

R2: Some infrastructures are absolutely es-
sential to enable the correct qualification of 
nuclear materials, not only irradiation facili-
ties, but also suitable ‘hot’ cells where active 
materials can be safely handled and tested, 
nuclearized characterization techniques, loops 
and pools for compatibility experiments, etc. 
They are also crucial for education and train-
ing of young researchers and operators. These 
infrastructures are costly to build and maintain. 
Other research facilities are, on the other hand, 
more common and sometimes redundant. A 
rational and harmonised, pan-European man-
agement of infrastructures, based on joint 
programming, including trans-national in-
frastructure renewal planning and a scheme 
for facility sharing and exploitation, would 
be highly desirable and, at the end of the day, 
beneficial for all. 

R3:  International  cooperation with  non-EU 
countries where research on nuclear materi-
als is pursued can be very valuable for Europe. 
Quite clearly, the goals of this cooperation are 
in the end the same as in the case of internal 

European cooperation, namely coordination 
of activities, sharing of data, and access to 
infrastructures. Currently, however, the instru-
ments available in Europe for international 
cooperation are not sufficiently attractive to 
motivate significant cooperation with non-EU 
researchers. Efforts should be made to im-
prove their attractiveness and ease of access. 
International organisations  such  as  OECD.
NEA, IAEA, but also Euratom and JRC for the 
connection with GIF, have here a crucial role.

R4: The  nuclear  materials  research  com-
munity  in  Europe  is  currently  strongly 
integrated and engaged in thriving collabo-
ration, in a bottom-up sense. This is in contrast 
with the inadequacy of the top-down instru-
ments offered to make this integration efficient 
and functional. This SRA is largely the result 
of matching bottom-up research proposals 
with top-down strategies. The appropriate 
instrument to allow this community to deliver 
according to the SRA goals should provide 
the conditions to implement the agreed re-
search agenda and to set up suitable E&T&M 
schemes  that  allow  knowledge,  data,  and 
facility sharing. Since the financial support of 
Euratom will never be sufficient, earmarked 
funding from the MS dedicated to support 
integrated research on nuclear materials is 
crucial. In this sense, a co-fund instrument, 
such as a European Joint Programme, seems 
to be most suitable.

These recommendations are clearly based 
on the willingness to pursue a policy of in-
creased integration rather than of isolation 
at  all  levels:  research  organisations,  EU 
Member States, and European Commission. 
Given the amount of resources that can be 
reasonably  allocated  to  cover  a  need  that 
has been estimated to range -depending on 
the ambition of the goals- between 15 and 
50 M€/yr, this requires the finding of a diffi-
cult equilibrium between the need to make 
the best use possible of the limited resources 
available, in a framework of nuclear energy 
where support is politically difficult to obtain, 
and the legitimate ambition, in a context of 
healthy competition, to preserve each stake-
holders assets.

Strategic Research Agenda of the EERA JPNMJoint Programme on Nuclear Materials of the European Energy Research Alliance

Coordinating sustainable nuclear materials research for a low carbon Europe

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Joint Programme on Nuclear Materials of the European Energy Research Alliance

Coordinating sustainable nuclear materials research for a low carbon Europe

          doi: 10.2760/59099 
ISBN 978-92-76-03318-9