the potential cyclic softening, low temperature irradiation (slip localization) and compatibility with coolants.<fnref n="96" /> The main issue affecting ODS cladding is the **high cost** associated with the multiple fabrication steps needed in a classical powder metallurgy route, which is also a bottleneck for upscaling to industrial production, 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 temperatures 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 temperature radiation embrittlement**.

The problem of **corrosion by dissolution and erosion** can be faced either by applying suitable **surface modification** (see below), or by producing **corrosion resistant steels** that produce such a protective layer by themselves, e.g. through **Al addition (FeCrAl)**.<fnref n="97" /> These steels exist but still need composition optimisation, to minimise side-effects of the addition of Al (decrease of ductility despite fully ferritic microstructure), while guaranteeing the formation 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 irradiation 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 expected to be complex.

## SiCf/SiC composites and other ceramics

Increasing the temperature of operation beyond 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-resistant 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 materials available on earth. These composites are also proposed in the commercial deployment of the SFR for the realization of the hexagonal 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 resistance,<fnref n="98" /> while being considered also for advanced accident tolerant fuel cladding for GenIII+ reactors.<fnref n="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:

* behaviour under irradiation;

* corrosion issues;

* joining;

* standardization of testing procedures.

<footnote n="96">⁹⁶ R.L. Klueh, et al, Scripta Mater., 53 (2005), p. 275</footnote>

<footnote n="97">⁹⁷ Jun Lim et al. Journal of Nuclear Materials, Volume 441, Issues 1-3, October 2013, Pages 650.</footnote>

<footnote n="98">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.</footnote>

<footnote n="99">⁹⁹ http://www.westinghousenuclear.com/Portals/0/Tchnovation%20Stuff/Accident%20Tolerant%20Fuel%20Brochure%20.pdf; http://www.ga.com/accident-tolerant-fuel.</footnote>
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, $Al_2O_3$, $TiO_2$, ZrC, ZrN, B4C, graphite or graphene.

## SiC behaviour under irradiation

Nuclear-grade $SiC_f$/SiC composites have shown to be stable to extremely high irradiation 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 component (fuel cladding tube). The **loss of thermal conductivity** of $SiC_f$/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 behaviour 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 $CO_2$, $CO$, $H_2O$, $H_2$, $CH_4$, $O_2$, as well as solid particles coming from a variety of sources throughout the reactor system.<fnref n="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 oxidation, active oxidation and bubble formation. The use of SiC composites requires extensive testing to evaluate the impact of non-passivating oxidation and corrosion in relevant environments. Barrier coating to hinder or limit corrosion may have to be envisaged.

## SiC Joining

$SiC_f$/SiC tubes are initially fabricated with one closed end; the open end must then be hermetically 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 technology 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 assessment of reliable joining technologies

## Standardization of testing procedures for $SiC_f$/SiC

The qualification and eventual codification of $SiC_f$/SiC requires a vast effort of pre-normative research, starting from the problem of the standardization of tests as a pre-requisite to characterize the behaviour of the material in environment (tightness against fission products, contact with flowing He, irradiation). **Standards on mechanical tests for nuclear grade $SiC_f$/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 $SiC_f$/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 preferably fabricated in a net-shape design, with the ability to be joined. Achieving reasonable leak tightness is one of the most important

<footnote n="100">¹⁰⁰ Estimation for ALLEGRO from HTR-10 experiments.</footnote>
issues, since He gas permeates easily through most materials. While also some metallic materials 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 using CVI), also other materials such as carbon composites, mullite, $Al_2O_3$, $TiO_2$, 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 melting 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 applications, as **metallic alternative to SiC/SiC** and also as competitors to advanced steels for in-core applications. However, these alloys exhibit **high affinity for oxygen, hydrogen, nitrogen 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 determined 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 establish 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 discussed as prominent examples, both studied to some extent in the past and recently experiencing 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 engineering-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 operation temperature limits. This involves finding the most appropriate approach (composition, 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 vulnerability to oxidizing species and proneness to embrittlement by impurities that are unavoidable 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 contamination, 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
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 joinability exists, so component construction is not a major concern. Molybdenum and its alloys have also good swelling behaviour under neutron irradiation, less than 3% up to 110 dpa.

The main issues are:

* The affinity for oxygen, carbon and nitrogen, which easily diffuse in the lattice occupying the interstitial sites, causing embrittlement;

* The susceptibility to oxidation corrosion in oxidizing environments where the low melting temperature (795°C) and the formation of volatile oxides may give rise to catastrophic oxidation,<fnref n="101" /> although Mo has good corrosion resistance in several liquid metals;

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

investigate LME issues; (d) development of welding procedures; (e) development of barrier 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 treatments. 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 corrosion (and/or wear) resistance**. This need is common to (a) steels in HLM operating above 400-450°C; (b) SiC/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 development of an optimized surface layer (SL):

1. The SL should form thermodynamically stable 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 introduction of an interdiffusion barrier or substrate where the diffusion rate of the SL species is high.

4. The values of the coefficients of thermal expansion 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

<footnote n="101">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.</footnote>
performance, in general and especially under neutron irradiation.

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:

* alumina-forming metallic layers produced by **GESA**;

* ceramic $Al_2O_3$ 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.

Prospective materials

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.<fnref n="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.<fnref n="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

<footnote n="102">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.</footnote>

<footnote n="103">¹⁰³ Liu et al. Transactions of Nonferrous Metals Society of China 25 (4) (2015) 1341.</footnote>
and functional applications. Further research
should be devoted to explore their applica-
bility to the GenIV systems.

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 M<sub>n</sub>+1AX<sub>n</sub>, 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.<sup>104</sup> 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'.<sup>105</sup> 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.<sup>106</sup> 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.<sup>107</sup> 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:

* Selection of appropriate composition followed by microstructural tailoring.

* Phase purity.

* Collection of statistically-relevant experimental data.

3.1.4. Summary of structural material issues

Because of the large number of structural
materials classes and types, it is convenient
to provide a summary of the main issues con-
connected 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.

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

<footnote n="104">¹⁰⁴ Barsoum, MAX Phases: Properties of Machinable Ternary Carbides and Nitrides, 2013 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim, Germany.</footnote>

<footnote n="105">¹⁰⁵ Lapauw et al., Inorganic Chemistry 55 (2016) 5445; Tunca et al., Inorganic Chemistry 56 (2017) 3489.</footnote>

<footnote n="106">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., Journal of the European Ceramic Society 37 (2017) 2353.</footnote>

<footnote n="107">107 Low, Thermal Decomposition of MAX phases, https://www.azom.com/article.aspx?ArticleID=6711.</footnote>
<table><caption>TABLE 6: Summary of issues relevant to the different structural material classes and types.</caption><thead><tr><th>Materials</th><th>Type of related issues</th><th>Pre-normative research</th><th>Modelling</th><th>Advanced materials' solutions</th></tr></thead><tbody><tr><td rowspan="3">Austenitic steels</td><td>316L(N) (prototype irreplaceable components)</td><td>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</td><td></td><td>Improve compatibility with coolants, apply high temperature protective barriers</td></tr><tr><td>15-15Ti (cladding)</td><td>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</td><td></td><td>Improve swelling resistance and compatibility with coolants (apply high temperature protective barriers).</td></tr><tr><td>Alumina forming austenitic (AFA) steels</td><td>Exposure needed for screening between candidates</td><td>Thermodynamic models for composition optimisation, microstructure evolution models</td><td>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</td></tr><tr><td rowspan="2">Ni-based alloys (heat exchangers, valves, coaxial pipes)</td><td>Alloy 800 (high Ni austenitic steel)</td><td>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).</td><td>To identify mitigation strategies, models of radiation-induced microstructure evolution in connection with predictions of embrittlement and swelling (He production).</td><td>Improve compatibility with alternative coolants and high temperatures; increase strength properties</td></tr><tr><td>Actual Ni-based alloys (eg Inconel 617, Haynes 230, ...)</td><td>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.</td><td>Creep and creep-fatigue engineering models in support of design correlations and rules.</td><td>Compatibility with coolant at high temperatures; manufacturing and joining</td></tr></tbody></table>
<table><thead><tr><th>Materials</th><th>Type of related issues</th><th>Pre-normative research</th><th>Modelling</th><th>Advanced materials' solutions</th></tr></thead><tbody><tr><td rowspan="3">Ferritic / Martensitic (F/M) steels (cladding and core)</td><td>9-14 %Cr</td><td>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</td><td></td><td>Need solution to minimize embrittlement, improve creep resistance (e.g. by thermomechanical treatment) and improve compatibility with coolants</td></tr><tr><td>Oxide dispersion strengthened (ODS)</td><td rowspan="2">Exposure needed for screening between candidates, suitable treatments for recrystallization to eliminate anisotropy after powder metallurgy production of bars and tubes by extrusion.</td><td>Oxide formation/stability, microstructure evolution, modes of deformation</td><td>ODS steels (tubes) have better creep resistance, but manufacturing and joining are issues (optimization needed); toughness and compatibility are also issues</td></tr><tr><td>FeCrAl alloys (also ODS)</td><td>Thermodynamic models for composition optimisation, microstructure evolution models</td><td>Addition of Al increases compatibility with coolants (protective alumina layer), but worsens mechanical behaviour: compromise searched</td></tr><tr><td rowspan="3">Refractory metallic alloys (cladding and core)</td><td>Molybdenum alloys (including ODS)</td><td rowspan="2">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</td><td></td><td>Prospective materials, mainly for cladding, studied also in the past, with problems of manufacturing, compatibility with coolant and mechanical behaviour</td></tr><tr><td>Vanadium alloys</td><td></td><td></td></tr><tr><td>High Entropy Alloys</td><td colspan="3">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.</td></tr><tr><td rowspan="2">Ceramics (cladding and coating)</td><td>SiC/SiC (also C/C) composites (cladding)</td><td>Mechanical test standardization, radiation resistance (thermal conductivity, hermeticity, swelling, ...) and corrosion resistance → define design rules</td><td>Microstructure evolution models under irradiation, finite element models for composite architectures, X-ray tomography techniques</td><td>Liners to guarantee hermeticity of cladding, or other techniques to guarantee hermeticity. Limit thermal conductivity degradation under irradiation.</td></tr><tr><td>Graphite</td><td>Irradiation effects on oxidation resistant graphite, irradiation creep, Codes & Standards development</td><td>Dependence of properties on porosity, graphite structure dynamics (stress states)</td><td>SiC/Graphite "composites"</td></tr></tbody></table>
<table><thead><tr><th>Materials</th><th>Type of related issues</th><th>Pre-normative research</th><th>Modelling</th><th>Advanced materials' solutions</th></tr></thead><tbody><tr><td rowspan="3">Ceramics (cladding and coating)</td><td>Non-metallic core support structures (ad hoc ceramics)</td><td>Screening of candidates. Test standardization (mechanical & thermophysical properties)</td><td>Microstructure evolution models under irradiation</td><td>Protection against oxidation</td></tr><tr><td>Al<sub>2</sub>O<sub>3</sub> coatings</td><td>Applied with different techniques on different substrates to protect against coolant attack and temperature: exposure for screening and qualification</td><td></td><td></td></tr><tr><td>Max phases</td><td>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.</td><td></td><td></td></tr></tbody></table>

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 composition, 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 parameters. If the phase diagram of the (U-Pu-O) system is largely known<fnref n="108" />, 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 atom transport properties are at the origin of several important phenomena taking place during irradiation, for instance the redistribution of Pu in the fuel, which strongly affects the heat and fission product distribution, and

the oxygen diffusion, which governs the local 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 created by energetic particles produced through fissions and radioactive decay, as well as intense ionization.

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

## Fission Product and Helium behaviour

The cladding constitutes the second retention barrier against radioactive fission product release.<fnref n="109" /> A design criterion is the absence or

<footnote n="108">¹⁰⁸ C. Guéneau et al. J. Nucl. Mater. 419, 145 (2011).</footnote>

<footnote n="109">¹⁰⁹ The first retention barrier is the inherent retention capability of the fuel. Attempts at improving this exist both for UO₂</footnote>
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.<fnref n="110" />, <fnref n="111" />, <fnref n="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,

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

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

<footnote n="110">¹¹⁰ J. Rest et al., J. Nucl. Mater., in press 52018), https://doi.org/10.1016/j.jnucmat.2018.08.019</footnote>

<footnote n="111">¹¹¹ M. Tonks et al., J. Nucl. Mater. 504, 300 (2018).</footnote>

<footnote n="112">¹¹² L. Luzzi et al., Nucl. Eng. Des. 330, 265 (2018).</footnote>
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 dimension changes, fission gas release (FGR), volatile and non-volatile material migration, microstructure 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 information and knowledge harvesting than in the past, with thermal conductivity, fission gas retention, 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 recommendations on properties of (U,Pu)O₂ fuel made by a group of European experts <fnref n="113" />.

The FP7 ESNII+ project<fnref n="114" /> (2013-2017) has started the update of this property catalogue by continuing 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. <fnref n="115" /> Thermal properties (thermal conductivity, specific heat, melting temperature and emissivity), structural and mechanical properties (lattice parameter, thermal expansion, elastic

constant, brittle-to-ductile transition temperature, yield and ultimate stresses, thermal creep) and atomic transport properties (diffusion and migration of oxygen, uranium, plutonium or minor actinides, fission gases, as well as of fission gases pores, fission gas bubbles) were particularly 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 unveil the missing relevant data and elementary mechanisms underpinning the fuel behaviour, and also to extend the reliability regime of traditionally deduced empirical laws governing various aspects of nuclear fuel under irradiation, which are implemented in fuel performance 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 5f electrons in actinide compounds.

<footnote n="113">¹¹³ Fast Reactor Data Manual, Technical report NT CEA DRN/DEC/SPU/LPCA 2 (1990)</footnote>

<footnote n="114">¹¹⁴ http://www.snetp.eu/esnii/</footnote>

<footnote n="115">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, International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development, 26-29 June 2017, Yekaterinburg, Russian Federation.</footnote>
<table><caption>TABLE 7: Main issues concerning fuel materials qualification and design rules.</caption><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Materials concerned</th><th>Techniques/ Methods</th></tr></thead><tbody><tr><td rowspan="2">Evolution of margin to fuel melting</td><td>Establishment of phase diagrams</td><td></td><td>Determination as a function of O/M, Pu, Am content, BU of<ul><li>Melting points</li><li>Heat capacity</li><li>Vapour pressure</li></ul>Techniques: laser heating, calorimetry, Knudsen cell mass spectrometry (KEMS)</td></tr><tr><td>Evolution of thermal properties</td><td></td><td>Measurement as a function of temperature, composition (O/M, Pu/M, BU) of<ul><li>Thermal conductivity</li><li>Thermal diffusivity</li><li>Emissivity</li></ul>Techniques: laser flash analysis</td></tr><tr><td rowspan="2">Atom transport and microstructural evolution</td><td>Atom transport</td><td>Irradiated oxide fuels and fuel pins: MOX (U,Pu)<br>O<sub>2</sub><br>UO<sub>2</sub><br>(U,Pu,Am)O<sub>2</sub><br>(U,Am)O<sub>2</sub></td><td>Measurement as a function of temperature and initial composition (O/M, Pu/M, BU) of<ul><li>Lattice parameter</li><li>Diffusion of U, Pu, Am</li><li>Diffusion of O</li><li>Oxygen potential</li></ul>Techniques: SEM-WDX, EPMA, SIMS, coulometric titration, XRD</td></tr><tr><td>Microstructural Evolution</td><td>Fresh oxide fuels as reference</td><td>Measurement as a function of temperature (radial position), initial Pu/M, O/M ratios and porosity and BU of<ul><li>Density</li><li>Beginning of life restructuring</li><li>Centre void formation</li><li>Pu, Am and O/M homogeneity and content</li><li>Diffusion/migration of pores</li><li>Grain size and their distribution</li></ul>Techniques: Optical microscopy, SEM/WDX-EDX, TEM, EPMA, SIMS, XRD</td></tr><tr><td rowspan="2">Fission Products and Helium</td><td>Gas (fission gas and He) behaviour</td><td></td><td>Measure of gas release in irradiated fuel as a function of BU and irradiation history (normal, off-normal and severe accident conditions)<br>Techniques: gas puncturing, KEMS, thermal treatment, thermal desorption spectroscopy</td></tr><tr><td>Non-gaseous FP transport</td><td></td><td>Measure of fission products transport, release and compounds formed as a function of BU and history (in normal, off-normal and SA conditions)<br>Techniques: Gamma scanning of fuel pins, KEMS, optical microscopy, SEM/WDX-EDX, EPMA, TEM, XRD, ICP-MS.</td></tr></tbody></table>
<table><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Materials concerned</th><th>Techniques/ Methods</th></tr></thead><tbody><tr><td rowspan="2">Mechanical properties and mechanical interactions between fuel and cladding</td><td>Evolution of mechanical properties and "state" of fuel pellets<br>Fragmentation, cracking</td><td>Irradiated oxide fuels and fuel pins: MOX (U,Pu) O2<br>UO2 (U,Pu,Am)O2 (U,Am)O2<br>Fresh oxide fuels as reference</td><td>Measurement as a function of temperature, Pu/M and O/M ratio, microstructure and BU of:<ul><li>Thermal expansion</li><li>Elastic constants</li><li>Ultimate stress</li><li>Yield stress</li><li>Brittle to ductile transition temperature</li><li>Hardness</li><li>Creep rates</li></ul>Techniques: XRD, dilatometry, acoustic methods, micro-nano indentation, eventually miniaturised mechanical testing methods</td></tr><tr><td>Fuel Cladding<br>Mechanical Interactions</td><td></td><td>Measurement as a function of temperature, Pu/M and O/M ratio, microstructure and BU of:<ul><li>Pellet geometry</li><li>Inner pin geometry</li><li>Swelling</li><li>Pellet density</li></ul>Determination of remaining gap width between pellet and cladding<br>Techniques: Optical inspection, XRD, dilatometry, pin profilometry, immersion method, optical microscopy, SEM, neutron radiography</td></tr><tr><td rowspan="2">Chemical interactions between fuel, cladding and coolant</td><td>Chemical interactions between fuel and cladding, internal cladding corrosion</td><td rowspan="2">Irradiated and fresh fuel pins, Breached pins, Sodium, lead and lead-bismuth coolants</td><td>Determination of<ul><li>Width of corrosion layer</li><li>Composition of corrosion layer</li></ul>Techniques: Gamma scanning, SEM/WDX-EDX, optical microscopy</td></tr><tr><td>Chemical interactions between fuel, cladding and coolant in case of severe accident</td><td>Measurement (as a function of BU if possible) of:<ul><li>Size of breach</li><li>Amount of fuel / fissile matter loss</li><li>Phases produced</li></ul>Techniques: Optical inspection, Gamma scanning, SEM/WDX-EDX, optical microscopy</td></tr></tbody></table>

This type of stud studies started to develop significantly in the year 2000s on uranium fuels, especially UO₂¹¹⁰,<fnref n="116" />, and to a lesser

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

<footnote n="116">¹¹⁶ R. Konings et al., Behaviour and Properties of Nuclear Fuels, in Experimental and Theoretical Approaches to Actinide Chemistry, First Edition, Ed: J.K. Gibson, W.A. de Jong, John Wiley & Sons Ltd '2018).</footnote>

<footnote n="117">¹¹⁷ https://cordis.europa.eu/publication/rcn/16699_en.html</footnote>
thermodynamic data were determined using new measurements in well controlled conditions (oxygen potential for fuels for instance) and samples (chemical composition, impurity 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 between the atomic and mesoscopic modelling on fuels by synthesising the data needed as input in mesoscale modelling that were calculated at the atomic scale. <fnref n="118" /> On the other hand, F-BRIDGE demonstrated a first success 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 properties for oxide, nitride and carbide fuels were implemented in the TRANSURANUS code. <fnref n="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,<fnref n="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 technological issues, pending scientific questions and corresponding basic research investigations to be carried out, were built from the interaction between F-BRIDGE participants

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)<fnref n="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 released 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 fuels 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 enhance 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).

<footnote n="118">¹¹⁸ M. Bertolus et al., J. Nucl. Mater. 462, 475 (2015).</footnote>

<footnote n="119">¹¹⁹ E.A. Kotomin, J. Nucl. Mater. 393, 292 (2009).</footnote>

<footnote n="120">¹²⁰ P. Van Uffelen et al., Nucl. Eng. Design, 43, 477 (2011).</footnote>

<footnote n="121">¹²¹ https://www.oecd-nea.org/science/wpmm/expert_groups/m2f.html</footnote>

<footnote n="122">¹²² https://www.oecd-nea.org/science/docs/2015/nsc-r2015-5.pdf</footnote>
<table><caption>TABLE 8: Main issues concerning modelling and characterization</caption><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Materials concerned</th><th>Experimental techniques/ Modelling methods</th></tr></thead><tbody><tr><td rowspan="3">Temperature: Margin to fuel melting</td><td>Phase diagrams: phases as a function of composition</td><td></td><td><ul><li>Laser heating measurement of melting temperature of virgin samples with various stoichiometries</li><li>Calorimetry on virgin samples with various non-stoichiometries</li><li>Atomic scale calculations of melting point and calorific capacity vs non-stoichiometry</li><li>Calphad modelling</li></ul></td></tr><tr><td>Evolution of melting temperature with BU</td><td></td><td><ul><li>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</li></ul></td></tr><tr><td>Evolution of thermal conductivity with BU</td><td></td><td><ul><li>Calorimetry on simfuels, self-irradiated and irradiated samples with various compositions</li><li>Atomic scale calculations of thermal conductivity of fuels vs non-stoichiometry and composition</li></ul></td></tr><tr><td rowspan="6">Irradiation defects and atom transport</td><td rowspan="5">Irradiation defects: point and extended defects</td><td>Fresh, self-, ion or neutron irradiated</td><td><ul><li>Calorimetry of virgin, ion-irradiated, self-irradiated and neutron-irradiated samples (short times, T and flux controlled) for various non-stoichiometries</li><li>Determination of type of defects created using positron annihilation and Raman spectroscopies, MAS-NMR, XAS, electrical conductivity measurement</li><li>Atomic scale modelling of point defects to determine most stable configurations as a function of non-stoichiometry</li><li>Atomic scale modelling of extended defects to determine most stable configurations and mechanisms of formation and growth</li><li>Atomic scale modelling of displacement cascades</li></ul></td></tr><tr><td>MOX (U,Pu)O<sub>2</sub></td><td></td></tr><tr><td>UO<sub>2</sub></td><td></td></tr><tr><td>(U,Pu,Am)O<sub>2</sub></td><td></td></tr><tr><td>(U,Am)O<sub>2</sub></td><td></td></tr><tr><td rowspan="5">Pu relocation and O/M variation</td><td>UC</td><td></td></tr><tr><td>(U,Pu)C</td><td></td></tr><tr><td>UN</td><td></td></tr><tr><td>(U,Pu)N</td><td></td></tr><tr><td>(Pu,Zr)N</td><td></td></tr><tr><td></td><td>SIMFUELS (fresh fuels including FP)</td><td><ul><li>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</li><li>Diffusion couple experiments, including with high temperature gradient</li><li>Atomic scale modelling of thermal, irradiation-induced and irradiation-enhanced diffusion coefficients for cations vs stoichiometry and composition</li><li>Mesoscale modelling of diffusion to determine cation concentrations in grain/pellet (KMC, rate theory, phase field...)</li></ul></td></tr><tr><td></td><td>Microstructural evolution</td><td><ul><li>Modelling of grain growth, fragmentation</li><li>Mesoscale modelling of pore evolution, central void formation</li><li>Diffusion experiments with high temperature gradient (laser heating)</li></ul></td></tr></tbody></table>
<table><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Materials concerned</th><th>Experimental techniques/ Modelling methods</th></tr></thead><tbody><tr><td rowspan="3">Fission Product and Helium behaviour</td><td>Gas behaviour</td><td></td><td><ul><li>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)</li><li>Atomic scale modelling of gas incorporation, as well as thermal, irradiation-induced and irradiation-enhanced diffusion vs stoichiometry and composition</li><li>Mesoscale modelling of gas concentrations in grain/pellet (in solution, bubbles, grain boundaries) and release</li></ul></td></tr><tr><td>Non-gaseous FP transport</td><td></td><td><ul><li>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)</li><li>Atomic scale modelling of FP incorporation, as well as thermal, irradiation-induced and irradiation-enhanced diffusion vs stoichiometry and composition</li></ul></td></tr><tr><td>FP compounds and JOG</td><td>Fresh, self-, ion or neutron irradiated<br>MOX (U,Pu)O2<br>UO2<br>(U,Pu,Am)O2<br>(U,Am)O2<br>UC<br>(U,Pu)C<br>UN<br>(U,Pu)N<br>(Pu,Zr)N<br>SIMFUELS (fresh fuels including FP)</td><td><ul><li>Electronic structure (and empirical potential?) calculations of stability and thermodynamic data on fission product compounds (Cs, I, Te, Mo)</li><li>Electronic structure (and empirical potential?) calculations of stability and thermodynamic data and compounds between FP and fuel elements (uranates, plutonates...)</li><li>Synthesis of selected compounds and measurement of thermodynamic data (melting temperature, thermal conductivity...)</li><li>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</li><li>Extension of this database to (Cs-I-Te-Mo-U-Pu-O) system</li><li>Using models developed above, thermodynamic equilibria calculations to predict the phase formation with respect to oxygen potential and temperature</li></ul></td></tr><tr><td rowspan="3">Mechanical properties</td><td>Thermal expansion</td><td></td><td><ul><li>Atomic scale calculation of thermal expansion as a function of composition and defect content</li><li>Mesoscale modelling of fuel thermal expansion</li></ul></td></tr><tr><td>Creep: thermal and under irradiation</td><td></td><td><ul><li>Atomistic modelling of high temperature and irradiation effects on the mechanical properties of fuels</li><li>Atomistic modelling of fuel deformation behaviour</li><li>High temperature creep experiments with controlled conditions oxygen partial pressure control in</li><li>Measure creep under ion and neutron to evaluate the radiation induced creep component as a function of temperature and load</li></ul></td></tr><tr><td>FCMI</td><td></td><td><ul><li>Atomistic modelling of thermal expansion in presence of defects and fission products</li><li>Measurement of thermal expansion of infused or self-irradiated fuels</li></ul></td></tr></tbody></table>
<table><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Materials concerned</th><th>Experimental techniques/ Modelling methods</th></tr></thead><tbody><tr><td rowspan="3">Compatibility between fuel, cladding and coolant</td><td>Cladding corrosion: ROG</td><td>Fresh, self-, ion or neutron irradiated<br>MOX (U,Pu)O2<br>UO2<br>(U,Pu,Am)O2<br>(U,Am)O2</td><td><ul><li>Electronic structure calculations of thermodynamic data on compounds between FP (uranates, plutonates...)</li><li>Synthesis of selected compounds between FP and fuel elements and measurement of thermodynamic data (melting temperature, thermal conductivity...)</li><li>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</li><li>Using this model, thermodynamic equilibria calculations to predict the phase formation in nuclear fuel with respect to oxygen potential and temperature</li><li>Carbides, nitrides: study of carburization and nitridation reactions</li></ul></td></tr><tr><td>Fuel/coolant issues</td><td>UC<br>(U,Pu)C<br>UN<br>(U,Pu)N<br>(Pu,Zr)N<br>SIMFUELS<br>(fresh fuels including FP)</td><td><ul><li>Electronic structure calculations of thermodynamic data on compounds between FP, fuel (U, Pu, Np, Am, Cm, C, N, O) and coolant (Na, Pb) elements</li><li>Synthesis of selected compounds and measurement of thermodynamic data (melting temperature, thermal conductivity...)</li><li>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</li></ul></td></tr><tr><td>Corium composition</td><td></td><td><ul><li>Electronic structure calculations of thermodynamic data on compounds in the (U-Pu-Fe-O), (U-Pu-Fe-C) and (U-Pu-Fe-N) systems</li><li>Synthesis of selected compounds in these systems and measurement of thermodynamic data (melting temperature, thermal conductivity, miscibility gap...)</li><li>Development of a thermodynamic database for the (U-Pu-Fe-O) (U-Pu-Fe-C) and (U-Pu-Fe-N) systems</li></ul></td></tr></tbody></table>

### 3.2.3. Development of advanced fuels

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 concepts (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.

#### Advanced oxide fuels development

Powder metallurgy will still be the choice for first MOX fabrication plants, but co-conversion routes to generate MOX powder should be developed to reduce dust issues, enable optimised remote handling with degraded Pu vectors (and concomitant $^{241}$Am pollution due to $^{241}$Pu decay) when Pu multi recycling is established, and to achieve greater Pu homogeneity, 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.
<table><caption>TABLE 9: Main issues concerning the development of advanced fuel materials.</caption><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Materials concerned</th><th>Techniques/ Methods</th></tr></thead><tbody><tr><td rowspan="2">Advanced oxide fuels development</td><td>Microstructure, process and design optimisation</td><td>MOX</td><td><ul><li>Development of advanced microstructures, increased homogeneity (e.g. chemical methods, improved milling, use of dopants)</li><li>Alternative fabrication routes for MOX pellet productions: e.g. SPS, additive manufacturing...</li><li>Design, geometry improvements.</li><li>Complete characterization (fresh fuel, irradiation, PIE)</li></ul></td></tr><tr><td>Transmutation fuel development (homeogeneous and heterogeneous)</td><td>MA bearing MOX, IMF (Cer-Cer, Cer-Met)</td><td><ul><li>Development of advanced safe fabrication processes (heavily shielded or in hot cells)</li><li>Development of recycling strategies (Pu and MA multirecycling, recycling of non-actinides, e.g. Mo).</li><li>Complete characterization (fresh fuel, irradiation, PIE)</li></ul></td></tr><tr><td rowspan="2">Carbide and nitride fuels development</td><td>Fabrication and design issues</td><td>UC, (U,Pu)C, UN, (U,Pu)N, (Pu,Zr)N</td><td><ul><li>Develop safe fabrication and recycling routes for different geometries (e.g. microspheres and/or pellets), considering radiological protection and pyrophoricity</li><li>Develop innovative methods (e.g. chemical methods for MX powder synthesis, Spark Plasma Sintering...)</li><li>Solve recycling issues (e.g. gaseous release during dissolution, 14C, Tritium, fuel dissolution, options for 15N recycling and/or reduce cost of 15N...)</li></ul></td></tr><tr><td>Assessment of safety and fundamental properties</td><td></td><td><ul><li>Assessment of thermal stability (Decomposition of nitrides, vaporisation (during processing), thermal stress cracking.</li><li>Compatibility tests between fuel, cladding and coolant</li><li>Evaluation of pyrophoricity and oxidation behaviour (thermochemical modelling of the reaction with air and moisture, microstructural studies of the oxidation mechanisms)</li><li>Complete characterisation using thermo-physical, chemical, mechanical and microstructural characterisation techniques</li></ul></td></tr><tr><td>Alternative fuels concepts</td><td>Molten Salt Reactor fuels/containment and materials</td><td>Molten salt mixtures</td><td><ul><li>Synthesis and purification of halide salts</li><li>Melting temperature determination, phase diagrams</li><li>Solubility and activity coefficients of the actinides and fission products</li><li>Thermal properties of the liquid phase (heat capacity, thermal conductivity, viscosity, density, surface tension)</li><li>Retention capacity for fission products (I, Cs, Te)</li><li>Interaction with other core materials</li><li>Tritium management</li></ul></td></tr></tbody></table>
Incremental process improvements can be made. In particular, a deviation (partial or complete) 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 predetermined Pu enrichment. The (U,Pu) solution can be co-converted into a powder ready for further manipulation and processing into product pellets. Classical hydroxide precipitation is to be avoided as it results in a dangerous material – ammonium nitrate, which is explosive when dry. Other co-precipitation routes should be considered, for example (U,Pu) oxalate precipitation. Plant simplification and automation will play an important role too. The implementation 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 powder, 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 radiological pollution of the gloveboxes and the equipment therein can be reduced.<sup>123, 124</sup>

The use of dopants to create advanced microstructures 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.<sup>125, 126</sup>

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

from increasing levels of <sup>241</sup>Pu will increase the radiation hazard necessitating improved plant designs and plant automation. Industry today 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 production of such fuels.<sup>127</sup> 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, optimising core performance in terms of breeding and increased margins to melt would necessitate the adoption of mixed uranium and plutonium carbides and nitrides (MX = MC and MN). Their fabrication is not trivial, if high purities are to be achieved, and there are question marks about their volatility at temperatures below 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

<footnote n="123"><sup>123</sup> R. Parrish et al., J. Nucl. Mater., 510, 644 (2018).</footnote>

<footnote n="124"><sup>124</sup> E. Douglas et al. J. Nucl. Mater., 393, 1 (2009).</footnote>

<footnote n="125"><sup>125</sup> V. Tyrpekl et al., Rev. Sci. Inst., 86, 023904 (2015).</footnote>

<footnote n="126"><sup>126</sup> M.A. Pouchon et al., ETH Zurich, https://doi.org/10.3929/ethz-a-010699574 (2016).</footnote>

<footnote n="127"><sup>127</sup> F. Jorion et al., Research and development for the fabrication of minor actinide-bearing fuel materials and technologies, IAEA (2015).</footnote>
intimate contact with carbon needs to be improved or replaced.<fnref n="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 generation of fine highly pyrophoric powders. Improved fabrication methods must be envisaged to ease the production and improve the product quality. Electric field assisted sintering 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.<fnref n="129" />, <fnref n="130" />

In the case of nitrides, enrichment in ¹⁵N during fabrication and its recovery from CO in the off gases needs to be dealt with.<fnref n="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 developments, revolutionary synthesis routes for carbides and nitrides are needed to shake off the undesirable attributes of the CTR procedure 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 produced 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 experimental and theoretical programmes.

## 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.<fnref n="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, elements) 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 relevant 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 retention 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 important and requires the development of irradiation facilities to study the fuel behaviour during irradiation in static and dynamic conditions (loops). Since the MSR reactor concept is generally coupled 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 development of advanced fuel materials are listed in Table 9.

<footnote n="128">¹²⁸ Hj. Matzke, Science of advanced LMFBR fuels, North-Holland (1986).</footnote>

<footnote n="129">¹²⁹ P. Malkki et al., J. Nucl. Mater. 452, 548 (2014).</footnote>

<footnote n="130">¹³⁰ D. Salvato et al., Ceram. Int. 43, 866 (2017).</footnote>

<footnote n="131">¹³¹ N. Chauvin et al., State-of-the-art Report on Innovative Fuels for Advanced Nuclear Systems. NEA - 6895 (2014).</footnote>

<footnote n="132">¹³² J. Serp, Prog. Nucl. Energy 77, 308 (2014).</footnote>
4

CROSS-
CUTTING
ISSUES
<figure><img src="image_18.png" /></figure>
# 4. CROSS-CUTTING ISSUES

## 4.1. Objectives

The qualification of materials in environments corresponding to the expected service conditions, the development of models capable of anticipating materials degradation in operation, 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 enhance 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 materials between different technologies can only be beneficial and contributes to accelerating 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 temperature 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, significant effort was deployed to dialogue with the platforms involved, NUGENIA and NC2I (exploiting the MoU signed with SNETP) and also EUROfusion,<fnref n="133" /> to identify cross-cutting issues.

Commonalities exist with energy technologies 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 concerning 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 materials 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 efficiency and cost reduction.

<footnote n="133">¹³³ The H2020 project M4F is a first example of a project jointly run by fission and fusion materials communities</footnote>
## 4.2. Commonalities with nuclear GenII/III reactor materials

The first common area between GenII/III and GenIV nuclear technologies concerns integrity (performance and ageing) of structural materials. Four topics for collaboration have been identified (for more details see Technical Annex TA.2.1<sup>71)</sup>:

* 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

* Ion irradiation as a neutron irradiation surrogate to gain better understanding of microstructural evolution under irradiation and improve identification of radiation resistant 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.2.1<sup>71)</sup>). As far as qualification and assessment of procedures for safety and integrity are concerned, the following topics are of special common interest.

* Safety of oxide fuels

* Innovative fuels and synthesis routes

* Fuel performance codes development and validation

Other commonalities are:

* Modelling, including separate effect experiments and materials characterization

* Materials and coatings considered for accident tolerant claddings

* Use of experimental facilities enabling the manufacturing and characterisation of fuels

Finally, three potentially common research areas are found also with the field of design of

innovative LWR (for more details see Technical Annex TA.2.3<sup>71)</sup>):

* Supercritical Water Cooled Reactor (SCWR) materials

* Development and application of advanced or novel materials manufacturing processes

* Development of new materials more resistant to corrosion

Table 10 summarizes materials issues of potential 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 particular 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.3<sup>71)</sup>):

* High temperature resistant materials for intermediate heat exchangers, insulating structures, control rods, and other internal structures

* Design life of 60 years with the development of relevant design rules and design curves and codification

* 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

Table 10 summarizes also materials issues of potential cross cutting interest between EERA JPNM and VHTR (NC2I) communities.
<table><caption>TABLE 10: Summary of potential cross-cutting issues on materials through nuclear fission technologies</caption><thead><tr><td>General topic</td><td>Commonalities</td></tr></thead><tbody><tr><td>Integrity of structural materials</td><td><ul><li>Development and qualification of welding procedures, including the analysis of residual stresses</li><li>Testing and qualification procedures for miniaturised specimens for both mechanical characterization and crack growth under environmental conditions</li><li>Advanced characterisation and multi-scale modelling of microstructural evolution under irradiation</li><li>Charged particle irradiation as a neutron irradiation surrogate to gain better understanding of microstructural evolution under irradiation and improve identification of radiation resistant materials</li></ul></td></tr><tr><td>Fuel and cladding materials</td><td><ul><li>Safety of oxide fuels</li><li>Innovative fuels and synthesis routes</li><li>Fuel performance codes development and validation</li><li>Development of advanced mechanistic and multiscale modelling tools and the execution of separate effect experiments and detailed materials characterization</li><li>Materials and coatings for accident tolerant claddings</li><li>Experimental facilities and relevant problems of availability</li></ul></td></tr><tr><td>Innovative Light Water Reactor Designs and Technologies</td><td><ul><li>Materials issues for SCWR</li><li>Development and application of advanced or novel materials manufacturing processes</li><li>Development of new radiation-, corrosion- and high temperature resistant materials</li></ul></td></tr><tr><td>(V)HTR materials</td><td><ul><li>Issues related with development and qualification of high temperature resistant materials in gas coolant environment</li></ul></td></tr></tbody></table>

## 4.4. Commonalities with fusion materials

Despite the significant differences that exist in terms of functioning principles and therefore 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.4<sup>71)</sup>:

* Compatibility with heavy liquid metals: corrosion and liquid metal embrittlement and a need for protective (alumina) surface layers

* Codification of F/M steels, as these are not yet included in design codes such as RCC-MRx, except at probatory level (specific issues are cyclic softening, plastic flow localisation, ...); joint materials database

* Welding procedures and characterization to detect defects and evaluate their consequences, taking into proper account the presence of residual stresses

* Small specimen testing in connection with small volumes of materials that can be exposed to irradiation

* 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 reducing their susceptibility to low temperature (<350°C) embrittlement

* 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

* Advanced modelling to continually improve the level of understanding of materials behaviour by exploiting advanced microstructural characterization and evolving physical models
Table 11 summarizes materials issues of potentially cross-cutting interest between GenIV fission and fusion.

## 4.5. Commonalities with materials for other energy technologies<fnref n="134" />

The commonalities with other energy technologies 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 materials needs and materials science approaches through energy technologies in EERA, including nuclear, led to the identification of the following **common patterns**<fnref n="135" /> (for more details, see Technical Annex TA.5.1⁷¹):

* Ageing and degradation mechanisms studied by combining advanced experimental characterization with multiscale modelling

* Characterization of energy materials and devices: contribution of large scale facilities, as well as in situ and operando techniques

* Rational design of materials supported by modelling

The second group stems from the recognition of the fact that **resistance to high temperature is a requirement for materials in a wide spectrum of energy technologies**, because

the efficiency of thermodynamic cycles operating 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 hydrogen, 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.<fnref n="136" /> More details on the commonalities concerning high temperature materials are given in the Technical Annex TA.5.2.⁷¹

Within high temperature applications, **specific commonalities identified and targeted between materials issues for GenIV fission 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, geothermal, GenIV nuclear fission) concerning high temperature and environmental degradation resistant materials.

<footnote n="134">¹³⁴ This section is largely based on the outcome of two workshops: (1) the EERA inter-JP cross-fertilization workshop on materials for energy applications and technologies, held on April 28th-29th, 2015, in Brussels (http://www.eera-jpnm.eu/filesharer/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.</footnote>

<footnote n="135">¹³⁵ http://www.eera-set.eu/wp-content/uploads/EERA-JP-workshop-Materials_for_Energy_report.pdf</footnote>

<footnote n="136">¹³⁶ https://www.eera-set.eu/wp-content/uploads/FINAL-June-2018-EERA-EUMAT-position-paper-HT-materials.pdf</footnote>
<table><caption>TABLE 11: Summary of GenIV fission/fusion potential cross-cutting issues on materials</caption><thead><tr><th>General topic</th><th>Commonalities</th></tr></thead><tbody><tr><td>F/M steels for current concept designs</td><td>F/M steels<ul><li>Design rules in RCC-MRx for F/M steels</li><li>F/M property assessment at low and high temperature: plastic flow localisation, cyclic softening, thermal creep, thermal fatigue, creep-fatigue, F/M welding</li><li>F/M compatibility with HLM: corrosion, erosion, LME, and test standardization</li></ul></td></tr><tr><td>Innovative high temperature resistant steels</td><td><ul><li>Optimisation of F/M ODS &amp; TMT steel fabrication<ul><li>Identification of best compositions and TMT</li><li>Deformation modes of ODS alloys at high temperature (creep)</li><li>Stability of the microstructure after long exposure to high temperature and long time for the ODS &amp;TMT steels</li></ul></li><li>Screening methods for prospective new materials (including charged particle irradiation and small specimen testing technology)</li></ul></td></tr><tr><td>Ceramic materials</td><td><ul><li>Ceramic coatings on steels (alumina or aluminium containing)</li></ul></td></tr><tr><td>Physical modelling and modelling-oriented experiments</td><td>F/M steels<ul><li>Microstructure evolution under irradiation</li><li>Low temperature hardening and embrittlement and plastic flow localisation</li><li>Irradiation creep and swelling</li><li>Liquid metal corrosion and liquid metal embrittlement</li><li>ODS fabrication processes, precipitation coarsening and recrystallization of oxides or carbides in advanced steels, ...</li></ul></td></tr></tbody></table>

<table><caption>TABLE 12: Summary of cross-cutting issues on high temperature and environmental degradation resistant materials through low carbon energy technologies</caption><thead><tr><th>General topic</th><th>Commonalities</th></tr></thead><tbody><tr><td>High temperature (&gt;400°C) mechanical performance assessment</td><td>Resistance to<ul><li>Thermal creep deformation (rupture beyond lifetime)</li><li>Thermal-mechanical fatigue (resistance to crack initiation/propagation)</li><li>Creep-fatigue interaction</li></ul></td></tr><tr><td>Protection from aggressive environment (liquid metals, molten salts, gases,...)</td><td><ul><li>Corrosion/oxidation/dissolution/erosion processes</li><li>Coatings of proven stability</li><li>Self-healing surface protection mechanisms</li></ul></td></tr><tr><td>Other properties to be maintained</td><td><ul><li>Thermal conductivity and limited thermal expansion</li><li>Non-permeability (to specific elements, e.g. gaseous like H or He)</li></ul></td></tr><tr><td>Steels for high temperature applications: existing and advanced</td><td><ul><li>Creep-resistant and corrosion resistant (FeCrAl) F/M steels (including ODS)</li><li>Austenitic steels</li><li>Ni-based alloys</li></ul></td></tr><tr><td>Refractory materials: metals and ceramic composites: existing and advanced</td><td><ul><li>V-, Mo-based alloys</li><li>SiC/SiC</li><li>Alumina based ceramics</li><li>Max phases</li></ul></td></tr><tr><td>Materials qualification</td><td><ul><li>Exposure facilities</li><li>Test standardization</li></ul></td></tr><tr><td>Advanced modelling and characterization</td><td><ul><li>Physical mechanisms</li></ul></td></tr></tbody></table>
5

WORKFLOW
AND
TIMELINE
<figure><img src="image_19.png" /></figure>
5. WORKFLOW AND TIMELINE

The development and qualification of materials for any technology is a continuous process:

* The return of experience from the use of already qualified and licensed materials provides further input for the design codes and triggers the search for more suitable materials and materials solutions, when design solutions are insufficient or unsatisfactory.

* Fundamental research on materials behaviour mechanisms provides tools to better assess component lifetimes and suggests routes towards materials property improvement.

* Following the return of experience from use and the continuously improved understanding of the processes that govern materials behaviour, new routes to materials 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 different representation of the workflow can be provided, in which **two parallel routes supported 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:

1. Demonstrators will be built using existing 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, sustainability.

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 emphasises 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 possible to associate indicative milestones for the qualification/development of specific material solutions with the moment when it needs to be ready for application, in order to enable
<figure><img src="image_20.png" /><figcaption>**FIGURE 8:** *The two parallel paths followed in the present SRA, one with short term and another with long term perspective.*</figcaption></figure>

the design of a specific system. However, it is impossible to establish precisely the timeline according to which the different possible systems will be designed, licensed and built, because this will depend on political and technical 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 development,<fnref n="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

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 farther 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.

<footnote n="137">¹³⁷ In particular, at the time of writing of this SRA the fate of ASTRID is undefined.</footnote>
<figure><img src="image_21.png" /><figcaption>**FIGURE 9:** Schematic timeline for structural materials. Dashed outline arrows indicate potential continuation of activities towards following milestones.</figcaption></figure>

<figure><img src="image_22.png" /><figcaption>**FIGURE 10:** Schematic timeline for fuel materials. Dashed outline arrows indicate potential continuation of activities towards following milestones.</figcaption></figure>
6

INFRASTRUC-
TURES FOR
NUCLEAR
MATERIALS
R&D
<figure><img src="image_23.png" /></figure>
# 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 infrastructures for materials qualification through exposure to conditions representative of service 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-nuclear 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 contrast, **irradiation is quite specific of nuclear materials** and **irradiation facilities are scarce**, as well as are scarce the infrastructures that allow **testing and examination of irradiated materials**. Therefore, this section focuses mainly on facilities for exposure to irradiation and handling of irradiated materials.

Notwithstanding the focus of this section, it is clear that equipment in **furnaces** to perform 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 qualification of materials for sustainable nuclear energy. This section summarises the main features and issues related with the facilities 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 necessity 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 construction should be qualified in fast neutron irradiation spectra. Unfortunately, there is currently no fast neutron power or testing reactor
operating in Europe. To redress the situation, **a fast system should be built in Europe**, but this leads to a vicious circle, because fast neutron 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 facilities**. 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 enable 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' investigations. These centres are equipped with 'hot' cells, as well. **Europe's neutron irradiation capacity is currently very limited: the opportunities to test new materials in reactors are therefore restricted to the extreme.**

<table><caption>TABLE 13: Materials Testing Reactors currently operational in Europe and their irradiation characteristics.</caption><thead><tr><td>Name</td><td>Location</td><td>Maximum fast neutron flux (&gt;0.1 MeV) [10<sup>14</sup> n /cm<sup>2</sup> s] (from *)</td><td>Dose rate [dpa<sub>Fe</sub>/fpy]</td><td>Accessible temperatures [°C]</td></tr></thead><tbody><tr><td>BR2</td><td>Mol, Belgium</td><td>7</td><td>Up to 5</td><td>50-1200</td></tr><tr><td>HFR</td><td>Petten, The Netherlands</td><td>5.1</td><td>&lt;7</td><td>80-1100</td></tr><tr><td>LVR-15</td><td>Řež, Czech Republic</td><td>3</td><td>~1</td><td>50-850</td></tr><tr><td>MARIA</td><td>Świerk-Otwock, Poland</td><td>1</td><td>~1</td><td>50-100</td></tr><tr><td>Triga Pitesti</td><td>Pitesti, Romania</td><td>1.8</td><td>2.5</td><td>80-300 (→ 500 future)</td></tr><tr><td>Triga Mark II</td><td>Ljubljana, Slovenia</td><td>0.06</td><td>&lt; 0.01</td><td>20-50 (→ 300 future)</td></tr></tbody></table>

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 unknown 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

<footnote>(*) https://nucleus.iaea.org/RRDB/RR/ReactorSearch.aspx</footnote>
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 facilities<fnref n="138" />

Materials exposed to neutron irradiation
need to be handled, tested and examined
in dedicated 'hot' cells or similarly shielded

**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 experiment. 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, under severe safety rules for the operators, and only very rarely in universities.

The main 'hot' cell facilities in Europe are **located 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 several cases the maintenance cost are expected to lead to closure without replacement. In particular, 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

<footnote n="138">¹³⁸ In addition to those associated with MTRs.</footnote>
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.

6.3. Computational facilities

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**

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

6.4.1. Costs related to crucial
infrastructures build and use

The construction and maintenance of infra-
structures and facilities such as MTRs and 'hot'
cells is beyond the scope of the EERA JPNM
activities and capabilities. It is important, however, to emphasise the importance of this point. In particular, if it is out of place to attempt estimates of infrastructure investment costs here, it is useful to remind that they have been detailed in European projects such as ADRIANA,<fnref n="139" /> and vary considerably, with material testing reactors 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, instrumentation 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 dedicated 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 industrial 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 context, **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:

* Ensure that a **fast neutron flux facility** comes available soon (e.g. ESNII demonstrator);

* 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);

* Renew and possibly **extend its 'hot' cell capacity**, taking due account of transportation issues;

* Ensure that 'hot' laboratory capabilities, in particular for fuels, **do not decrease** further;

* Ensure appropriate access to **high performance computing for the fission community**;

In addition, three recommendations for policy-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 problematic** for legal, security, safety and financial reasons, namely:

* **Legal**: Protection of know-how & expertise: there is often reluctance to give full open access as this may reveal details on protected know-how (this attitude is however often inconsistent even within the same organisation);

* **Related to security**: Access to 'hot' cells requires clearance from authorities for security reasons: this takes weeks or months and a significant administrative burden;

* **Related to safety**: Only trained & skilled operators can safely use some equipment, especially in 'hot' labs (manipulators...);

* **Financial**: Availability of specific equipment has a high cost and opening it for access to external users might limit the profitability of facilities.

Notwithstanding these difficulties, **solutions can be envisaged**, for example a scheme of mutual

<footnote n="139">¹³⁹ http://ojs.ujf.cas.cz/~wagner/transmutace/erinda/presentations/05_ADRIANA_ERINDA.pdf</footnote>
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 education and training** of not only researchers, but also operators, enabling Europe to maintain 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 context of scarcity of infrastructures and financial means, and in order to foster pan-European harmonised infrastructure planning, **joint programming 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.<fnref n="140" />

The **joint forum** should mainly agree upon **irradiation 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 infrastructures.

The **joint scheme for the management of nuclear infrastructures** should fulfil the following conditions:

1. Be a **single entry point**, i.e. a single interlocutor 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 experiments possible (also the cheapest ...), establishing if and when there are advantages to access non-EU infrastructures;

3. Benefit from the **harmonised commitment of MS to pay the costs of the infrastructures and use**, as part of the engagement to innovation, but develop flexible management 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 available 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 conditions it is possible to perform inexpensive "piggy-back" irradiation experiments, in a certified 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 programming 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 actions. It remains to be seen whether the conditions for this approach to materialise actually exist.

<footnote n="140">¹⁴⁰ This could be a European version of the American NSUF: https://nsuf.inl.gov/</footnote>
7

EDUCATION,
TRAINING AND
MOBILITY OF
RESEARCHERS
<figure><img src="image_24.png" /></figure>
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 community, 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 inherently 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 nuclear 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 projects, 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 actively manage currently dwindling key nuclear competences. In this respect, **infrastructure sharing is a key driver, to be made effective within joint programming schemes** (see section 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 managed cross-European education and training and mobility (E&T&M) programme with specific
focus on nuclear structural and fuel materials should apply outcome-based teaching methods, focusing on **interdisciplinary integration** and emphasizing both **research and industry needs**. It should **extend and complement existing university curricula**, training programmes and schools, without doubling them, while being **recognised by European universities in terms of credits (ECTS)**. To be attractive, **it should enable the development of knowledge 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 extra-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 collaboration 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 institutions and to attract qualified students to participate in special E&T national and international 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 sustainable 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 available for training purposes, as well as suggesting suitable training topics and schemes.

The harmonisation of E&T&M activities between 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 making coordinated use of the funds distributed through different Euratom-supported projects for E&T purposes**. The idea here is to reverse the current situation, where each project is delegated the task of identifying suitable E&T activities, to replace it with a cross-platform 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 support 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 potential concept to develop an E&T programme that should offer different learning packages, to give the participants the possibility to acquire specific and needed competences, as described in Table 14.
<table><caption>TABLE 14: Description of a possible E&T programme on nuclear structural and fuel materials, as envisioned by the EERA JPNM.</caption><tr><td>What</td><td><ul><li>A learning package - E&T program devoted solely to nuclear materials (structural materials and fuels)</li></ul></td></tr><tr><td>When</td><td><ul><li>Annually or bi-annually, at predetermined dates</li></ul></td></tr><tr><td>Participants</td><td><ul><li>Master's students from relevant engineering disciplines</li><li>PhD students</li><li>Employees from the nuclear industry</li></ul></td></tr><tr><td>Class size</td><td><ul><li>Class size is optimally 20 people, max. 30 people</li></ul></td></tr><tr><td>Where</td><td><ul><li>Suitable locations can be identified after sending out respective explanations and a questionnaire to all known institutions asking for their availability</li><li>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</li><li>Suitable experimental facilities must offer at least partial hands-on experience to students</li></ul></td></tr><tr><td>Length</td><td><ul><li>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)</li><li>1-2 weeks of optional, preparatory online learning in advance to bring everybody up to the same level</li><li>Participation during the entire year is necessary to meet the goal of high-level competence</li></ul></td></tr><tr><td>Teaching methods</td><td><ul><li>Hands-on experimental assignments</li><li>Hands-on computational exercises</li><li>Learning through in-class and at-home comprehensive problem solving</li><li>Lectures leading in assignments and assisting in their solving</li><li>Field trips to the industry and research facilities</li><li>E-learning whenever suitable</li></ul></td></tr><tr><td>Topics</td><td><ul><li>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</li></ul></td></tr><tr><td>Teachers</td><td><ul><li>Outstanding researchers and experts from universities and the industry who are interested in teaching and applying effective teaching methods to achieve the intended learning outcomes</li></ul></td></tr><tr><td>Teaching support</td><td><ul><li>A short course to motivate applying successful teaching methods</li><li>Guidelines for desirable lecture composition and assignment design</li></ul></td></tr><tr><td>Suggested curriculum features</td><td><ul><li>Topics per session limited to a small number so that sufficient time is given to students to listen, understand and apply new knowledge</li><li>A high problem-solving-to-lecture ratio should be applied</li><li>Teacher-assisted independent work in small teams throughout all sessions</li><li>All assignments related to topical research interests and industry interests</li></ul></td></tr><tr><td>Management aspect</td><td><ul><li>Fewer topics per week means fewer teachers per week which results into smoother time management and potentially reduced costs</li><li>Annual course on predetermined dates simplifies planning and availability both for teachers and students</li></ul></td></tr></table>
8

INDUSTRY
AND
REGULATORS
INVOLVEMENT
<figure><img src="image_25.png" /></figure>
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 support 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,<fnref n="141" /> although political decisions have also an influence there. Finally, reactor operators are those that may provide return of experience on materials in service

<footnote n="141">¹⁴¹ Nuclear electro-producers have currently drawn back from investing in technologies to close the fuel cycle because of uncertain future of nuclear energy in some countries and in Europe at large and because of renewable penetration leading to load follow mode. Incomes are reduced, while cheap fossil fuel energy sources remain convenient (especially coal). Therefore without public and political support, 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 crucial to enable the development of simulation codes, design codes etc.</footnote>
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 hand-in-hand 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 manufacturers, for example ESTEP (steel-makers)<fnref n="142" />, EMIRI<fnref n="143" />, EUMAT<fnref n="144" />, ...

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 designers, 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 procedures used for materials qualification and possibly guide them from a safety point of view, in order to eventually accelerate the licensability 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 associate 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 intellectual property right -IPR- rules and based on the signature of a non-disclosure agreement) **are invited to join the stake-holders' 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.<fnref n="145" /> The wish for the future is that also TSOs and regulators may join SHGs of EERA-JPNM projects.

<footnote n="142">¹⁴² https://www.estep.eu/estep-at-a-glance/</footnote>

<footnote n="143">¹⁴³ http://emiri.eu/about</footnote>

<footnote n="144">¹⁴⁴ http://www.eumat.eu/</footnote>

<footnote n="145">¹⁴⁵ See for example the task force report on 60 years operational life of reactors, available at http://www.eera-jpnm.eu/?q=jpnm&sq=nboard</footnote>
9

INTER-
NATIONAL
COOPERATION
<figure><img src="image_26.png" /></figure>
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:<fnref n="146" />

* Optimized/harmonized use of infrastructures, making unique facilities transnationally accessible and also, importantly, adequately planning their use to avoid duplications and redundancies, while making the best use possible of the specificities of

each particular infrastructure: this should
be the rational approach to make materials
qualification complete and affordable.

* **Harmonization of procedures and methodologies** to test and characterize materials, especially innovative materials in specific environments, including protocols to perform microstructural examination with advanced techniques and to analyse the results. In several cases, completely 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.

* **Data collection and sharing**, through suitable databases that should be eventually made available for reactor designers (industry and not only) and regulators: the more complete and extensive the databases, 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

<footnote n="146">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.</footnote>
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 moment only informal contacts with the GIF project devoted to materials for VHTR, in whose framework 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 specific activities on GenIV materials. Data sharing is the essence of the GIF, but most countries should also be interested in harmonizing testing 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-NERI<fnref n="149" /> are in

place, although they are not in practice very attractive, especially on the European side.<fnref n="150" /> Collaboration with South Korea on issues related with creep design rules for austenitic and F/M steels and compatibility of these materials with HLM is underway through Euratom projects (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 difficult 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 because of the unique opportunity they may offer to expose materials to high dose in fast spectrum 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.

<footnote n="149">¹⁴⁹ E.g. http://cordis.europa.eu/fp7/euratom-fission/docs/06-haas-ws4_en.pdf</footnote>

<footnote n="150">¹⁵⁰ While in the US this scheme leads to chances for additional 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 additional financial incentive.</footnote>
10

RESOURCES
<figure><img src="image_27.png" /></figure>
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 systems. However, based on past experience and also on estimates coming from the deployment of the research agenda of the EERA JPNM, **excluding** the cost of the construction and maintenance of the needed infrastructures and facilities, which are here taken for granted, one can assess that:

In order to adequately address the most important 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 involvement 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.**

* If all costs of irradiation qualification campaigns, transport and hot cells are included,

for proper materials qualification (within the limits allowed by the facilities that are available), 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.

* The inclusion of significant industrial involvement and extensive activities also on materials currently at low or very low TRL, extended to systems beyond those included 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.**<fnref n="151" />

**These resources clearly exceed by about an order of magnitude the possibilities of funding 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

<footnote n="151">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.</footnote>
innovative systems already exists now, in particular 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 exists, for instance, in the case of fusion energy, which is taken here as a successful example.

Fusion energy receives support under different forms, including through joint undertakings (Fusion for Energy, F4E) and international agreements (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 commitment combined with Euratom support. Of course, such a funding scheme for fusion energy is explained by the widespread support 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 widespread 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 pilots, NUGENIA in primis. There are also strong connections and overlaps, in practice, between researchers working in nuclear fission and fusion materials. It is therefore believed that **an EJP, or any equivalent co-fund instrument, could be suitable also to support the established nuclear materials community**, with substantial advantages in terms of optimal use of resources, by avoiding duplications

and redundancies, as sometimes happens between fusion and fission. However, without support for projects specifically aimed at designing 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 institutional projects.

<figure><img src="image_28.png" /><figcaption>FIGURE 11: Indicative estimated subdivision of funds between research activities, based on past experience, assuming all activities are sufficiently funded.</figcaption></figure>
11

RISK
ASSESSMENT
<figure><img src="image_29.png" /></figure>
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.
<table><caption>TABLE 15. Risk assessment table on several issues addressed in this document.</caption><thead><tr><td>No.</td><td>Category</td><td>Risk denomination</td><td>Risk description</td><td>Likelihood<sup>(1)</sup> and Level of criticality of consequences</td><td>Possible consequences</td><td>Mitigation strategy</td></tr></thead><tbody><tr><td>1</td><td>GenIV systems in Europe</td><td>Low industrial interest to close the fuel cycle</td><td>Closing the fuel cycle is the main reason driving the deployment of fast reactors. While this is considered as a necessary step towards nuclear energy sustainability, the investment required, particularly if not supported at political level and/or by public funding, is likely to lead the current nuclear industry to delay the goal of closing the fuel cycle to farther times.</td><td>60% Very critical</td><td>The development of fast reactors, and GenIV systems in general, may be perceived as a lesser priority for Europe, leading to a significant reduction of financial support.</td><td>The activities can shift towards more cross-cutting issues with GenII/III, fusion or even other energy technologies</td></tr><tr><td>2</td><td>GenIV systems in Europe</td><td>ESNII projects stopped</td><td>One or more of the ESNII projects might be stopped because of lack of financial or political support.</td><td>60% Currently very critical for ASTRID. ALLEGRO might also run into difficulties</td><td>Some materials or some specific research lines (e.g. compatibility with a specific coolant) would lose importance and maybe have to be abandoned</td><td>Focus on other systems or on cross-cutting issues</td></tr><tr><td>3</td><td>Prenormative research on structural materials</td><td>Insufficient qualification of reference materials and design rules &amp; curves for HLM compatibility</td><td>Design Rules and Codes for HLM compatibility require a comprehensive and complete long-term test programme that includes reference materials and welds with relevant engineering and mechanistic models. Data obtained within the available timeframe or with the available resources/ infrastructures may not be sufficient for a demonstration of compatibility at design level.</td><td>50% Critical</td><td>HLM compatibility is critical for licensing of HLM cooled reactors. Without appropriate demonstration at design level, license for construction may not be given. Or the system, due to excessive conservativism, might turn out not to be viable from the point of view of economy/efficiency</td><td>Drastic design changes may help to guarantee sufficient conservativism, within available data.</td></tr></tbody></table>
<table><thead><tr><td>No.</td><td>Category</td><td>Risk denomination</td><td>Risk description</td><td>Likelihood<sup>(1)</sup> and Level of criticality of consequences</td><td>Possible consequences</td><td>Mitigation strategy</td></tr></thead><tbody><tr><td>4</td><td>Prenormative research on structural materials</td><td>Insufficient qualification of reference materials and design rules &amp; curves for 60 years operation life</td><td>Long-term tests (&gt; 5 years) for representative service conditions under very stable test conditions need to be performed to qualify materials and components for 60 years design life. These tests require long-term commitments by participating organizations which may dwindle over time.<br>Reliable methods for accelerated tests need to be devised and their feasibility to 5 represent long-term degradation needs to be demonstrated by integration of modelling and experiments. The extent to which this is possible is an open issue.</td><td>70%<br>Critical in some areas</td><td>Lack of design rules and codes for 60 years operation lifetime, with reduced return of investment for nuclear reactors and hence reduced willingness to invest.</td><td>Drastic design changes may help to guarantee sufficient conservatism, within available data.</td></tr><tr><td>5</td><td>Qualification of fuels for GEN IV reactors</td><td>Insufficient qualification of MOX fuels for GEN IV reactors</td><td>Qualification of new fuels or known fuels in new conditions or with new safety requirements is a long and expensive process which climaxes in an irradiation in reactor. This requires long-term commitments by participating organizations. Also possible that the time for full qualification exceeds the time needs</td><td>40%<br>critical</td><td>The efficiency of the plant will not be optimal, not fulfilling the GenIV requirements.<br>The system, due to excessive conservativism, might turn out not to be viable from the point of view of economy/efficiency</td><td>Better known fuel, such as UO<sub>2</sub>, could be used.<br>Qualification done by non-European countries for their own reactors could be used</td></tr><tr><td>6</td><td>Qualification of fuels for GEN IV reactors</td><td>Insufficient qualification of transmutation fuels for GEN IV reactors</td><td>Qualification of new fuels is a long and expensive process which climaxes in an irradiation in reactor. This requires long-term commitments by participating organizations. Also possible that the time for full qualification exceeds the time needs.</td><td>60%<br>severe</td><td>The closed fuel cycle will not be achieved, not fulfilling one of the main GenIV objectives: sustainability.</td><td>The activities can shift towards more cross-cutting issues with GenII/III. Transmutation in LWR reactors could be envisaged</td></tr></tbody></table>
<table><thead><tr><td>No.</td><td>Category</td><td>Risk denomination</td><td>Risk description</td><td>Likelihood<sup>(1)</sup> and Level of criticality of consequences</td><td>Possible consequences</td><td>Mitigation strategy</td></tr></thead><tbody><tr><td>7</td><td>Modelling</td><td>Too high complexity of processes for available modelling tools</td><td>The development of physically-based multiscale models is not equally easy in terms of modelling tools and scale bridging for all processes. The models that can be developed might not be sufficient to address the complexity of the processes of interest.</td><td>50% Critical in some areas</td><td>Lack of suitable models may force to too heavy experimental burden for the full qualification of materials and components.</td><td>Hybrid models with a component of empiricism may in some cases prove sufficient.</td></tr><tr><td>8</td><td>Advanced material solutions</td><td>Lack of ODS industrial supplier</td><td>The fabrication of ODS fuel cladding tubes is feasible and has been successfully achieved. However, the repeatability and reproducibility of the fabrication procedure at larger scale is still under optimization, in particular the recrystallization behavior of high Cr ODS grades. The production of ODS cladding tubes will remain at lab scale, taking into account that few prototypes will be built in Europe in the next 20-30 years</td><td>90% Uncritical</td><td>Fuel burn-up will not be increased and the efficiency of the plant will not be optimal, not fulfilling the GenIV requirements.</td><td>Alternative materials may be considered: swelling resistant austenitic steels, creep-strength enhanced F/M steels, currently prospective materials, ...</td></tr><tr><td>9</td><td>Advanced material solutions</td><td>Qualification as nuclear-grade of innovative materials</td><td>Several new alloys or other low TRL material solutions are under explorative R&amp;D to verify if they are as promising as expected, e.g. FeCrAl, AFA, HEA, MAX phases. The long process to become nuclear grade materials involve their qualification in terms of good mechanical properties, oxidation and corrosion resistance, radiation tolerance, compatibility with fuel and neutronic behavior (for fuel cladding) in addition to weldability and formability characteristics. Depending on the material, the possibility that some of the above mentioned characteristics are not fulfilled is high. Also possible that the time for full qualification exceeds the time needs.</td><td>50% Critical</td><td>No alternative material found to replace those considered in the short term design. Fuel burn-up is not increased, so the efficiency of the plant will remain not optimal. Corrosion mitigation by replacing the existing materials may not be possible. The goals of GenIV systems are not fully achieved.</td><td>Design improvements may compensate for the lack of materials.</td></tr></tbody></table>
<table><thead><tr><td>No.</td><td>Category</td><td>Risk denomination</td><td>Risk description</td><td>Likelihood<sup>(1)</sup> and Level of criticality of consequences</td><td>Possible consequences</td><td>Mitigation strategy</td></tr></thead><tbody><tr><td>10</td><td>Advanced fuels development</td><td>Qualification of advanced fuels</td><td>Several new low TRL fuels solutions are under explorative R&D to verify if they are as promising as expected; carbides, nitrides, molten salts<br/>Depending on the material, the possibility that some of the necessary characteristics are not fulfilled is high. Also possible that the time for full qualification exceeds the time needs.</td><td>50% Severe</td><td>No alternative material found to replace those considered in the short term design. Fuel burn-up is not increased, so the efficiency of the plant will remain not optimal.<br/>Corrosion mitigation by replacing the existing materials may not be possible.<br/>The goals of GenIV systems are not fully achieved.</td><td>Design improvements may compensate for the lack of materials.</td></tr><tr><td>11</td><td>Infrastructures</td><td>Availability of nuclear materials exposure and testing infrastructures</td><td>Some infrastructures are crucial for the qualification of materials but are either currently not available in Europe (fast spectra), or limited and expensive (MTR, 'hot cells, loops, ...'). These infrastructures might end up becoming unsustainable.</td><td>50% Can be critical</td><td>Insufficient qualification of materials for the design of demonstrators</td><td>Set up a scheme of joint planning and management (sharing, exploitation) of nuclear infrastructures in Europe</td></tr><tr><td>12</td><td>Education and training</td><td>Loss of competence</td><td>Decrease in the number of new researchers and operators entering the nuclear community, 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</td><td>50% In some areas critical</td><td>Nuclear fission research and competence in Europe may be in danger of decline</td><td>Set up a stable joint European scheme of education and training, using materials as a way to attract young people towards nuclear through collaboration between platforms and coordinated use of funding from projects</td></tr></tbody></table>
<table><thead><tr><td>No.</td><td>Category</td><td>Risk denomination</td><td>Risk description</td><td>Likelihood<sup>(1)</sup> and Level of criticality of consequences</td><td>Possible consequences</td><td>Mitigation strategy</td></tr></thead><tbody><tr><td>13</td><td>International cooperation</td><td>Lack of exploitation of the potential of cooperation</td><td>For international cooperation to be beneficial an effort is needed on the side of international organisations, as well as Euratom, to lay down the proper legal and attractive financial instruments. This may prove to be impossible, either for lack of political willingness and/or financial support</td><td>90% Uncritical</td><td>Progress worldwide will be hindered (probably with some countries emerging over the others, depending on specific political decisions and resources conceded)</td><td>Limited agreements of collaboration, based on opportunities offered by countries or on personal contacts</td></tr></tbody></table>

(1) 10%: Rare (1-20%) - The event may occur only in exceptional circumstances

30%: Unlikely (21-40%) - More likely not to occur under normal conditions

50%: Likely (41-60%) - Given time, likely to occur

70%: Very likely (61-80%) - The event will probably occur in most circumstances

90%: Almost certain (81-99%) - The event is expected to occur in most circumstances
12

SUMMARY
AND
RECOMMENDATIONS
<figure><img src="image_30.png" /></figure>
12. SUMMARY AND
RECOMMENDATIONS

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.

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

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.

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,
importance of international cooperation, estimate of resources required, risk assessment.

Besides the obvious need of adequate financial 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 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 qualification of nuclear materials, not only irradiation facilities, 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 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 nuclear 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 available 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**.

**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 stake-holders assets.
<figure><img src="image_31.png" /></figure>

Joint Programme on Nuclear Materials of the European Energy Research Alliance
Coordinating sustainable nuclear materials research for a low carbon Europe