# Towards a single European strategic research and innovation agenda on materials for all reactor generations through dedicated projects

Lorenzo Malerba, Pietro Agostini, Massimo Angiolini, Marjorie Bertolus

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Lorenzo Malerba, Pietro Agostini, Massimo Angiolini, Marjorie Bertolus. Towards a single European strategic research and innovation agenda on materials for all reactor generations through dedicated projects. EPJ N - Nuclear Sciences & Technologies, 2022, 8, pp.36. 10.1051/epjn/2022042. hal-03899616

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# Towards a single European strategic research and innovation agenda on materials for all reactor generations through dedicated projects

Lorenzo Malerba<fnref n="1" />, Pietro Agostini<fnref n="2" />, Massimo Angiolini<fnref n="2" />, and Marjorie Bertolus<fnref n="3" />,\*

Received: 21 September 2022 / Received in final form: 2 August 2022 / Accepted: 21 October 2022

**Abstract.** The goal of the ORIENT-NM action is to produce a single European strategic vision on research and innovation concerning nuclear materials in the EU, serving all reactor generations and nuclear systems. The key in this endeavour is to focus on advanced materials science practices that, combined with digital techniques, will enable acceleration in materials development, manufacturing, supply, qualification, and monitoring, in support of nuclear energy safety, efficiency, economy and sustainability. The research agenda will be rooted in existing virtuous examples of nuclear materials science projects. Here the results of three of them are summarised, thereby covering different reactor applications and families of materials, as well as a range of advanced material research approaches. GEMMA addressed a number of key areas concerning the development and qualification of metallic structural materials for GenIV reactor conditions, focusing on austenitic steels and their compatibility with several non-aqueous coolants, their welds and the modelling of their stability under irradiation. INSPYRE was an integrated project applying a basic science approach to (U,Pu)O₂ fuels, to develop physics-based models for the behaviour of nuclear fuels under irradiation and improve fuel performance codes. Modelling was also the focus of the M4F project, which brought together the fission and fusion materials communities to study the effects of localised deformation under irradiation in ferritic/martensitic steels and to develop good practices to use ion irradiation as a tool to evaluate radiation effects on materials.

# 1 Introduction

The ORIENT-NM project is elaborating a single European strategic research and innovation agenda (SRIA) that should set the path for future activities on nuclear materials in the EU until 2040, serving all reactor generations. The key in this endeavour is to focus on advanced materials science practices that, combined with modern digital techniques, will enable acceleration in materials development, manufacturing, supply, qualification, and monitoring, in support of nuclear energy safety, efficiency, economy and sustainability. This research agenda, which pursues five transversal research lines, is rooted in existing virtuous examples of materials science projects that target nuclear energy innovation. The objectives and key results of three of them, which cover different reactor applications and families of materials, are presented in this paper, illustrating the advances brought by the various research lines considered in ORIENT-NM. GEMMA addresses a number of

key areas concerning the development and qualification of metallic structural materials for Gen IV reactor conditions: (1) corrosion-resistance of austenitic steels for application in heavy-liquid metal-cooled systems; (2) production of welds of available austenitic steels and their characterization in terms of internal stresses; (3) testing of these materials (baseline, welds and advanced) under representative conditions in contact with heavy liquid metals and helium; (4) development of physical models for the prediction of the behaviour of austenitic alloys under long term irradiation. The INSPYRE project applied a basic science approach to (U,Pu)O₂ fuels to get further insight into the mechanisms underlying their changes under irradiation, and accordingly develop physics-based models that describe the behaviour of nuclear fuels under irradiation and improve European fuel performance codes. Finally, the M4F project created a bridge between fission and fusion materials communities by applying physical modelling techniques to target two objectives: (1) understand and predict the origin and effects of localised deformation under irradiation in ferritic/martensitic steels,

<footnote n="1">¹ CIEMAT, Division MXE, Avda. Complutense, 40, 28040 Madrid, Spain</footnote>

<footnote n="2">² ENEA-FSN-ING Division, C.R. Brasimone, 40032 Camugnano, Italy</footnote>

<footnote n="3">³ CEA, DEs, IRESNE, DEC/SESC, 13108 Saint-Paul Paul-Lez-Durance, France</footnote>

<footnote>\* e-mail: marjorie.bertolus@cea.fr</footnote>
<figure><img src="image_2.png" /><figcaption>Fig. 1. Classes of nuclear materials and their variety.</figcaption></figure>

which affect the mechanical behaviour of components for future fission and fusion reactors, to enable their design based on robust standards; (2) to develop good practices to use ion irradiation as a tool for the evaluation of radiation effects on materials, also applied to ferritic-martensitic alloys. The objectives and structures of these projects are described in [1]. All the projects described in this article are part of the activities of the Joint Programme on Nuclear Materials of the European Energy Research Alliance [2].

## 2 ORIENT-NM: a single vision on nuclear materials research for all reactor generations until 2040

ORIENT-NM (Organisation of the European Research Community on Nuclear Materials) is a Coordination and Support Action (CSA) partially funded by the Euratom research and training work programme 2019–2020. It was launched to explore the opportunity of setting up a European partnership on nuclear materials, establishing the relevant Strategic Research Agenda (SRA) and governing structure, while defining its interactions with external stakeholders. The starting point for this was the analysis of the national energy and climate plans (NECP) [3], as well as other documents that are available online, such as the IAEA and WNA country nuclear power profiles [4,5]. This enabled the different national priorities to be identified, in connection with nuclear energy and materials in the EU member states. This analysis [6] revealed, among other things, that a significant number of EU member states intend to maintain or even expand their nuclear-installed power through long-term operation

(LTO), power uprates and new builds, from now to 2040. In addition, game-changers such as small modular reactors and advanced designs of interest throughout the continent may lead nuclear energy to be even more widespread in 2040 than currently foreseeable. Recent dramatic geopolitical events might also have an impact on this.

In this framework, materials research plays a crucial role to enhance the safety, efficiency, economy, and overall sustainability of current reactors and enabling the commissioning and deployment of next-generation reactors, as well as fusion. The nuclear materials science community in Europe is therefore called to provide the tools, knowledge and skills to enable each European country to maintain the wished and needed nuclear capacity and/or, depending on national policies and interests, to develop advanced nuclear systems, within the time horizon of 2040. Specifically, it should contribute to enabling EU member states to:

* ensure safe and affordable LTO of current generation reactors;

* design, license and construct Gen III+ new builds over the next two decades;

* deploy light water SMRs within the next decade;

* facilitate and reduce the time and costs for design, licensing, and construction of competitive next-generation nuclear reactors, including advanced SMRs, within the 2040 horizon.

The materials involved a priori range from concrete and metallic structural alloys, to polymers, fuel and compounds for neutron control, as is shown in Figure 1. In order to enable progress towards enhanced safety, efficiency and economy, the development, manufacturing and qualification of innovative materials must be accelerated,
<figure><img src="image_3.png" /><figcaption>Fig. 2. The five transversal research lines and Grand Goals proposed in ORIENT-NM.</figcaption></figure>

thus reducing their time to market. Accurate health monitoring while operating is also needed. This implies a shift from the traditional “observe and qualify” to the modern “design and control” materials science approach. Advanced digital techniques and suitable models are the enablers of this shift.

Five materials science practices and relevant research lines constitute accordingly the Grand Goals to be pursued within the next decade:

(a) establishment of an integrated European system for the efficient application of advanced and suitably standardized experimental procedures and methodologies for nuclear materials characterization, testing and qualification, be they destructive, non-destructive or microstructural; i.e., nuclear test-beds;

(b) development of methodologies for accelerated, targeted and systematic nuclear materials improvement, or even discovery, including the whole range of variables of relevance; i.e., nuclear materials acceleration platforms (MAPs), which promise to be key to reducing the time to market of innovation;

(c) development of combined physical and data-driven models and predictive methodologies of direct application for industrial needs;

(d) development of advanced methods for materials and component health monitoring through non-destructive examination and testing, coupled with diagnostics and simulation tools, to enable the implementation of digital twins;

(e) establishment and use of efficient platforms and procedures for data collection, storage and management (European nuclear materials FAIR<fnref n="1" /> database).

Nuclear test beds and MAPs inherently require coordinated use of (present and future) European facilities and infrastructures and exploitation of available schemes and roadmaps for access to and use of major infrastructures

(chiefly materials testing reactors and ion irradiation facilities, but also facilities for the exposure to fluids and the subsequent characterization). The transversal nature of the above five research lines and Grand Goals is illustrated in Figure 2. The first corresponding proposal of the Strategic Research Agenda has been published in [7].

## 3 GEMMA: qualification of austenitic steels for application in GenIV prototypes and demonstrators

The GEMMA project (GenIV Materials Maturity), which ended in November 2021, was partially funded by the Euratom research and training work programme 2014–2018. It mainly addressed the following areas: (1) production and characterization of advanced alumina forming corrosion-resistant austenitic steels for application in heavy-liquid metal-cooled systems; (2) characterization of welds on available austenitic steels in terms of internal stresses; (3) testing under representative conditions in contact with heavy liquid metals and helium; (4) development of physical models for the prediction of the behaviour of austenitic alloys under long term irradiation. The results on these topics are summarised below.

### 3.1 Advanced alumina forming corrosion-resistant austenitic steels for application in heavy-liquid metal-cooled systems

Alumina-forming austenitic stainless steels (AFA) and Al-based diffusion coatings are candidates for application in heavy liquid metals at high temperatures, to replace conventional stainless steels, which exhibit poor performance in these environments [8]. In the GEMMA Project, more than 20 AFA model alloys were produced by the industrial partner SANDVIK and compared.

<footnote n="1">¹ Findable, Accessible, Interoperable, Reusable.</footnote>
<figure><img src="image_4.png" /><figcaption>Fig. 3. Diffractometric measurements of internal stresses, before and after PWHT.</figcaption></figure>

They were all exposed to liquid Pb under various condi-
tions. Selected AFA alloys were also exposed to steam at
1200°C to explore their potential use in a steam environ-
ment. Their microstructure was then analysed to inves-
tigate their corrosion resistance. The AFA model alloys
with the composition formulas Fe-(15.2-16.6)Cr-(3.8-
4.3)Al-(22.9-28.5)Ni (wt.%) and Fe-16Cr-(23-25.5)Ni-
(3-4.5)Al-C(N)-(Y) have all shown corrosion resistance
to oxygen-containing molten Pb at 600°C. By increas-
ing the exposure temperature to 650°C, only the alloy
of the first family, with the lower Ni content (~23 wt.%),
and with minor additions of Nb and Y, forms a protective
oxide scale. Higher Ni contents increase the susceptibility
to corrosion attack at 650°C in Pb. The passivating scale
was based on (AlCr)₂O₃ at both temperatures and formed
a homogeneously alloyed coating with the expected com-
position. To evaluate the microstructural stability of AFA
alloys at elevated temperatures, thermal ageing experi-
ments were performed at 600°C for 1000 and 2000 h and
650°C for 3550 h. The austenitic phase was preserved, but
all the alloys exhibited secondary phases and precipita-
tions [9].

## 3.2 Characterization of welds on available austenitic steels in terms of internal stresses

High-resolution neutron diffraction residual stress mea-
surements on a welded coupon after Post-Weld Heat
Treatment (PWHT) were carried out at the spalla-
tion neutron source SINQ in Villigen (CH). An identi-
cal coupon had been studied by high-resolution neutron
diffraction stress measurements in the as-welded state at
FRM-II in Garching. The PWHT consisted of a dwelling
temperature of 700°C for 24 h with limited heating and
cooling rates: these relatively high temperatures and
long dwelling time were chosen to ensure tangible relax-
ation of residual stresses, but bore no connection with
the heat treatment process prescribed by the RCC-MRx
code. Diffractometric measurements were carried out on
both coupons at the POLDI instrument (Pulsed Over-

Lap Diffractometer). Figure 3 shows that the PWHT was
effective to reduce the stresses inside the weld and the
heat-affected zone, as well as to reduce the width of the
stress gradient, particularly in the longitudinal direction.
The maximum stress values are more than halved and
stress lower than 100 MPa are found already at a 10 mm
distance from the weld centre.

3.3 Testing under representative conditions in contact
with heavy liquid metals and helium

An extensive experimental campaign was carried out
to obtain data on the compatibility of base materials
(austenitic alloys 316L and 15-15Ti), their welds, AFA,
and coated specimens (alumina forming diffusion coatings,
ceramic coatings by pulsed laser deposition and by deto-
nation gun spray) with heavy liquid metals (HLM) and He
gas. The goal was to clarify the degradation mechanisms
and to establish correlations for the design of MYRRHA
[10], ALFRED [11] and ALLEGRO [12]. For the first time,
a large amount of data has been collected on the com-
patibility of steels and their welds in contact with these
fluids. Thermodynamic simulations with the CALPHAD
method and with atomistic modelling have supported the
experimental work, to gather information on the relevant
part of the phase diagrams of the complex oxide systems
and the mass transport properties of the compounds that
form.

Contrary to common perception, the corrosion in
molten lead and lead-bismuth eutectic (LBE) differed
both for the severity of the attack and the mechanisms, the
dissolutive attack in LBE being more severe. Furthermore,
while the lead dissolving process mainly concerns nickel,
in LBE both chromium and nickel are dissolved. In partic-
ular, the corrosion campaign in flowing Pb in the testing
loop MATLOO at CV-Rez showed that, after 16 000 to
18 000 h at 480 °C, with oxygen concentration 10⁻⁷ wt.%
and 1.6 m/s flow velocity, corrosion was negligible in all the
exposed materials [13]. The ENEA corrosion campaign,
carried out in the BID1 facility, in liquid Pb at 550 °C
with low oxygen concentration (10⁻⁸–10⁻⁷% wt.%) up to
8000 h, led to the interesting observation that the mag-
nitude of the corrosion of welds in terms of the depth of
penetration was low compared to the corrosion observed in
base materials under the same conditions. The microstruc-
tural investigations have not unambiguously clarified the
relevant mechanisms, but the results suggest precipita-
tion of the δ ferrite at grain boundaries and triple joints,
which limits or prevents the rapid diffusion of nickel there.
Moreover, pulsed-laser deposited alumina coatings con-
firmed their superior protective action on 15-15Ti, while
both selected AFA alloys showed low yearly corrosion
[14,15].

Opposite results were obtained in tests in LBE, with
the poorer performance of welded joints versus the base
material. The unacceptable corrosive attack on welded
joints calls for additional measures for their protection
in LBE. Investigations aimed at studying the corrosion
mechanisms highlighted that the initiating process of
the attack in HLM is the dissolution of iron from the
<figure><img src="image_5.png" /><figcaption>Fig. 4. (a) Gibbs free energy of mixing of the Fe–Ni solid solution of the pair interaction model (full lines) and CALPHAD (dotted lines), at different temperatures. For the sake of clarity, each curve (except at 700 K) is shifted downwards by the value given in the figure. (b) the Fe–Ni FCC phase diagram: comparison between the pair interaction model and experiments.</figcaption></figure>

native iron-chromium oxide that protects the stainless steel in many environments. This dissolution leads then to the formation of a porous structure with interconnected channels along the grain boundaries, which is permeable to oxygen and the components of the steel. The inner Fe-Cr oxide provides channels for oxygen entering and iron leaching. Investigations have also shown that the subsequently formed layer of magnetite in turn loses its protective capability, due to the onset of wetting processes of the grain boundaries by the liquid metal, above temperatures around 450 °C. These results set limits on the applicability of oxygen control for corrosion control of HLM; on the other hand, they set a limit below which steels can be used safely in lead [14,16].

Finally, the mechanical testing campaign in HLM confirmed the absence of liquid metal embrittlement in the austenitic steels 316L and 15-15Ti. Yet, the intensity of the corrosion processes strongly depends on the microstructure of the steel. The presence of extended defects with negligible effects on the mechanical properties in the air could lead to deep corrosive attacks in HLM, with mechanical degradation that is solely linked to the reduction of the resistant section [17].

## 3.4 Development of physical models for the prediction of the behaviour of austenitic alloys under long-term irradiation

In the temperature range of interest for operation (300–700 °C), irradiation produces in austenitic alloys new microstructural features, such as new families of precipitates, solute segregations at structural defects, point defect clusters and voids, leading to dimensional changes (swelling) and worsening the mechanical properties (loss of ductility, hardening). These microstructural features induce adverse effects and limit the lifetime of the materials. Predictive models for microstructure evolution under irradiation are therefore useful tools in support of assessing the component lifetime. With this aim in mind, as

the first step kinetic models were developed to describe the phase diagrams of Fe–Ni–Cr model alloys, their diffusion properties and the kinetics of segregation and ordering. The composition range of the ternary model alloy has been chosen to encompass those of 316L(N) and AIM1 steels, targeting temperatures around 500 °C. Interdiffusion experiments in Fe–Ni–Cr multi-layers during isothermal annealing and under ion irradiation have been performed to assess the validity of the atomistic kinetic models. Two thermodynamic descriptions were developed, one based on an implicit treatment of magnetism and a second one explicitly treating the magnetic interactions between Fe and Ni atoms. Relying on the former, a rigid lattice pair interaction diffusion model was developed and used for Atomic Kinetic Monte Carlo (AKMC) simulations in binary Fe–Ni alloys. The pair interactions were fitted to 0 K ab initio calculations of formation enthalpies of ordered and disordered structures and also systematically fitted on the Gibbs free energy of the γ Fe–Ni solid solution, as described in a CALPHAD (CALculation of PHAse Diagrams) study [18]. The model reproduces well the data for the solid solution (Fig. 4a) and the FeNi₃–L₁₂ ordered phase in a large domain of composition and temperature, using first and second neighbour pair interactions which depend on temperature and local Ni concentration. The FCC phase diagram of the Fe–Ni system was determined using Monte Carlo simulations in the semi-grand canonical ensemble (Fig. 4) and compared with experimental studies and other models. In addition to the γ solid solution and the FeNi₃–L₁₂ phase, the FeNi–L₁₀ phase is found to be stable below 684 K. The model also predicts a phase separation between ferromagnetic (γf) and paramagnetic (γp) solid solutions.

The pair interaction model was then extended to include vacancy diffusion. Vacancy-pair and saddle-point interactions were fitted on electronic structure calculations of vacancy formation and migration energy and to experimental tracer diffusion coefficients in dilute alloys. AKMC simulations were performed to calculate tracer and
interdiffusion coefficients in concentrated solid solutions and compared with available data. Finally, a BCC and FCC Fe-Ni phase diagram below the magnetic-transition temperatures based only on electronic structure calculation results was constructed. The free energy of the various phases was determined by including the vibrational and the ideal configurational entropies. The resulting phase diagram is in very good agreement with the experimental data and suggests possible improvements for upper-scale semi-empirical approaches (such as CALPHAD). In order to go further in the thermodynamic prediction of the Fe-Ni alloys at higher temperatures, when magnetic excitations and transitions occur, an effective interaction model for Fe-Ni, which describes chemical and magnetic interactions explicitly and the effect of vibrational entropy, was explored. This model is currently extended to describe the properties of defects (vacancies and self-interstitial atoms) in Fe-Ni for various chemical and magnetic states [19].

## 4 INSPYRE: basic research on U-Pu mixed oxide fuels to improve fuel performance codes

The INSPYRE project (Investigations Supporting MOX Fuel Licensing in ESNII Prototype Reactors), which ended in February 2022, was partially funded by the Euratom research and training work programme 2014–2018. INSPYRE aimed at:

(1) using a basic research approach that combines out-of-pile separate effect experiments with physical modelling and simulation to get further insight into the underlying phenomena that govern the behaviour of (U, Pu)O₂ mixed oxide (MOX) fuels under irradiation.

(2) Performing additional post-irradiation examinations (PIE) on selected samples to complete the results available in the literature.

(3) Extending the reliability regime of empirical laws that describe nuclear fuel under irradiation using the results obtained in the project, as well as those of previous integral neutron irradiation tests.

(4) Using these results to improve the reliability of European operational fuel performance codes in normal and off-normal situations to facilitate nuclear fuel licensing and improve safety.

(5) Training the next generation of researchers on nuclear fuels.

These objectives are presented in the project's video [20]. The results of the project are synthesised in [21] and summarised below.

### 4.1 Advances in the understanding and description of the behaviour of nuclear fuels

INSPYRE brought significant advances in the understanding of mechanisms that are crucial for the safety assessment and qualification of MOX fuels for future nuclear reactors, as well as in the simulation of the fuel behaviour in the reactor. The new experimental and modelling

<figure><img src="image_6.png" /><figcaption>Fig. 5. Heat capacity as a function of temperature and Pu content yielded by atomic-scale calculations and analytical law (solid lines) fitted on these results that were implemented in the GERMINAL fuel performance code.</figcaption></figure>

results significantly enhance the knowledge and understanding of the physical, thermal, chemical and mechanical properties of MOX fuels. In addition, several setups were developed to enable the detailed characterization of UO₂ and plutonium- and americium-bearing fuels in the hot laboratories of the partners.

Laser heating experiments and atomic scale calculations enabled the determination of high temperature and melting properties of (U,Pu)O₂, (Pu,Am)O₂, and O₂. Together with literature data and data from the FP7 ESNII+ project, these results improved the thermodynamic modelling of the (U-Pu-Am-O) system [22]. The specific heat of (U, Pu)O₂ was also evaluated using atomic scale calculations (see Fig. 5), yielding information at very high temperatures and over the full composition range, which is very difficult to measure experimentally [23].

Atomic scale calculations provided further insight into the elementary processes of formation and migration of defects in (U,Pu)O₂ and a new diffusion model for plutonium in MOX fuel and an associated mobility database were developed [24]. In addition, diffusion couples were designed to enable refined experimental measurements of diffusion coefficients.

Diffusion coefficients of inert gases Xe, Kr, and He in fresh UO₂ and MOX were determined using a combination of experimental techniques and modelling methods, from the atomic to the grain scale [25]. Finally, MOX fuel samples from past irradiation campaigns with various Pu contents and at various locations of the pellet, i.e., subjected to different burnups and irradiation temperatures, were characterised using inter alia electron microscopy, down to the nanometer scale.

Complementary techniques and methods were combined to a precise control of materials and test conditions to determine mechanical properties and get further insight into the elementary processes governing their evolution. Elastic properties, ductile-fragile transition and ultimate strength in MOX single crystals as functions of
<figure><img src="image_7.png" /><figcaption>**Fig. 6.** Apparatus for online creep test under ion irradiation and swelling as a function of damage dose on UO₂ polycrystalline sample.</figcaption></figure>

Pu content, temperature and irradiation dose were studied using atomic scale simulations [26]. Then, thermal creep in UO₂ under controlled conditions (in particular the oxygen partial pressure [27]) was measured in a dedicated setup, while new micro and nano-indenters were used to determine the room temperature elastic properties and micro-hardness at the grain scale of well-characterized fresh MOX fuel samples.

Irradiation-induced swelling in UO₂ was studied in situ using ion irradiation at the CNRS cyclotron in Orléans and MOX in the HFR reactor in Petten. As shown in Figure 6, macroscopic swelling of UO₂ was proved to increase with damage dose, in agreement with the trends shown by simulations and post-irradiation examinations [28]. In addition, deformation was recorded in situ in MOX fuels irradiated under very small temperature gradients in the HFR reactor for 6 months.

INSPYRE also improved significantly the knowledge of the “Joint Oxyde-Gaine” (JOG) layer, which is formed by the migration of volatile fission products, in particular caesium, tellurium, iodine and molybdenum, from the centre to the periphery of the fuel pellet. The incorporation and transport properties of Cs and I in MOX fuel were evaluated using electronic structure calculations. Thermodynamic data on fission product compounds in the chemical system (Cs-I-Te-Mo)-(U-Pu-O) were measured and calculated at the atomic scale [29]. The thermodynamic models on the Te-U, O-Te-U, Mo-O-Te, Cs-Mo-O, Cs-Te-O and Cs-Mo-Te-O sub-systems were then improved using these results.

Finally, properties such as melting temperatures of phases containing U, Pu, Fe and O (PuO₂-Fe₃O₄ system), which form from the interaction between MOX and cladding in the case of a severe accident, were measured for the first time using an advanced laser heating setup. In addition, a novel experimental setup enabled new data to be gathered on the liquid miscibility gap of the Fe-U-O system, in which two immiscible liquid oxide and metallic phases coexist. Thermodynamic models of the U-Fe-O and Pu-Fe-O systems were then derived using these experimental data.

## 4.2 Advances in the simulation of nuclear fuels

Fuel performance codes used to simulate the behaviour of Gen IV fuels under irradiation were improved using the results presented above. Physics-based correlations giving melting temperature and thermal conductivity versus fuel temperature, Pu and minor actinide content, deviation from stoichiometry, porosity and burn-up were developed for MOX and minor actinide-bearing MOX fuels under GEN IV reactor relevant conditions [30]. Several physics-based models were implemented in the SCIENTIX grain-scale module to improve the description of fission gas and helium behaviour in MOX fuels [31]: (1) a burn-up model to evaluate helium production in MOX; (2) a reduced order model of fission gas diffusion in columnar grains; (3) a model for the intra-granular helium behaviour that accounts for helium diffusivity, solubility and interaction with other fission gases; (4) an improved description of the formation of the High Burn-up Structure and the corresponding porosity evolution at the periphery of fuel pellets. Then, improved correlations were obtained for the MOX thermal expansion and Young's modulus as functions of fuel temperature, Pu content, porosity and deviation from stoichiometry. A 3D micro-mechanical model based on representative volume elements, which includes a description of the fuel rupture behaviour, was also developed for both normal operation and off-normal conditions.

The models and correlations developed in the project were used to improve three major European fuel performance codes: GERMINAL, developed by CEA, MACROS developed by SCKCEN, and TRANSURANUS, developed by JRC with the support of academic organisations. All of them were coupled with the SCIENTIX grain-scale module for inert gas behaviour [32]. The improved versions of these codes were then assessed by simulating the three past fast-neutron irradiation experiments SUPERFACT-1 (see Fig. 7), RAPSODIE-I and NESTOR-3. In addition, data obtained in the second half of the project are available to further improve models and codes.

The code validation against local and integral experimental data reveals harmonized predictions of fuel
<figure><img src="image_8.png" /><figcaption>**Fig. 7.** (a) Relative linear rate and fission gas release at peak power node yielded by improved FPC. (b) Integral fission gas release yielded by pre and post-INSPYRE versions of FPC and comparison with experimental data.</figcaption></figure>

restructuring (as fuel inner void formation). In particular, fuel central temperature predictions are now more consistent between the various codes [33].

## 4.3 Simulation of ESNII reactor cores

The new versions of the codes were used to simulate normal operation conditions in the ASTRID sodium fast reactor prototype [34], as well as normal and transient conditions in the LBE-cooled reactor of MYRRHA [35]. This enabled the evaluation of the safety margins of the fuel designs (e.g., margin to fuel melting, cladding plasticity) under relevant conditions.

# 5 M4F: bringing together fusion and fission materials communities to predict the behaviour of ferritic-martensitic steels under irradiation

The M4F project (Multiscale Modelling for Fusion and Fission Materials), which ended in December 2021, was partially funded by the Euratom research and training work programme 2014–2018. It brought together fusion and fission materials research communities to work on the prediction of irradiation-induced microstructural damage and deformation mechanisms of irradiated ferritic/martensitic (F/M) steels, applying a multidisciplinary approach that integrates modelling and experiments at different scales. The main objectives were:

* to develop physical understanding and predictive models of the origin of localised deformation under irradiation in F/M steels and its consequences on the

<figure><img src="image_9.png" /><figcaption>**Fig. 8.** Simulation of tensile test on irradiated EUROFER97 (0.65 dpa@300°C): comparison of simulated and experimental true stress vs. true strain curves.</figcaption></figure>

mechanical behaviour of components. This is meant to set the basis for the future elaboration of physically-motivated, non-over-conservative design rules for this class of steels, of use for both fission and fusion applications.

* to identify good practices for the use of ion irradiation as a tool to evaluate radiation effects on materials for modelling and screening purposes, minimising artefacts with respect to neutron irradiation experiments and allowing the evaluation of not only microstructural changes but also mechanical property changes.

The rationale of these objectives, reviewing past work and summarizing research pathways and first results of the M4F project, is the subject of reference [36], while the overall results are summarized in its Final Report [37]. Here we, therefore, provide only a brief highlight of the main results of the project with respect to the above objectives, emphasizing result exploitation and impact aspects.

The project impact concerns two main items:

(1) cross-fertilization between fission and fusion;

(2) overcoming bottlenecks that are limiting developments in fission and fusion.

In connection with cross-fertilization, a continuous effort was made to identify and disseminate cross-cutting research activities of equal interest for fission and fusion applications. Thanks to this effort, the spectrum of these commonalities has progressively widened over time. The pioneering role of the project is proven by the fact that synergies between fission and fusion are currently being sought for also by the IAEA [38]. The outcome of the M4F project in terms of the identification of commonalities between the two materials communities will certainly impact the outcome of this IAEA effort.

Concerning the goal of overcoming the bottlenecks that are limiting developments in fission and fusion, it has
<figure><img src="image_10.png" /><figcaption>**Fig. 9.** Ion irradiation damage for two different beam energy values (5 and 8 MeV) as a function of depth: transmission electron microscopy micrographs; dpa and implanted ion profiles calculated.</figcaption></figure>

to be evaluated in terms of advancing towards the two above-stated goals.

Regarding the first one, the problem of plastic flow localization has been addressed covering all scales, from the study of elementary processes concerning dislocation motion in the matrix, through obstacles [39] and across grain boundaries [40,41], using atomistic simulation techniques, to dislocation dynamics [42] and finite element approaches [43–47]. In particular, the visco-plastic modelling approach developed in [45–47] allows the simulation of flow localization in F/M steel structures under neutron irradiation, taking into account finite strains and ductile damage. The model is a powerful tool for the assessment and development of new criteria for immediate plastic flow localization and can also be used to assess other failure modes caused by ductile damage. In case of irradiation causing loss of ductility, it predicts immediate local fracture due to exhaustion of ductility and its dependence on triaxiality. As an example of application, Figure 8 shows how the model can reproduce to a good approximation the experimental true-stress/true-strain curve of irradiated EUROFER97.

Advances were made also in the direction of identifying good practices for ion irradiation experiments [48], with the support of advanced microstructure evolution models [49–51]. The importance of choosing a high ion beam energy value, to mimic as much as possible neutron irradiation, has been clearly evidenced. A characteristic pattern of defect bands linked to the dpa and to the implanted ion profile has been identified, as shown in Figure 9.

One of the bands, between the surface and the damage peak, “safely” resembles the microstructure that is observed under neutron irradiation, provided that the Fe-ion beam energy is ≥5 MeV [48]. 8 MeV also give neutron-closer results than 5 MeV in terms of solute-rich

cluster formation, but differences are detected as compared to irradiations performed with 5 MeV. It is the first time that the influence of ion energy on the microstructural features has been observed [52]. For the interpretation of the experiments, three microstructure evolution models were developed which, combined, can provide a complete description of the radiation-induced microstructure under neutron and also ion irradiation, including all the specific features of the latter [49–51].

The project included also an important activity on nanoindentation as a tool to evaluate mechanical property changes due to irradiation using ion-implanted specimens [53–56], which are only affected within a depth of a few µm from the surface. In connection with this, the publication of the CEN NATEDA (NAnindentation TEsting DAta) Workshop Agreement [57] represents an important precedent for the standardization of nanoindentation testing, as a tool for the characterization of the mechanical behaviour of ion-irradiated specimens.

# 6 Summary and conclusions

GEMMA, INSPYRE and M4F represent examples of multidisciplinary projects with similar goals and approaches, but applied to different materials and conditions, showing the cross-cutting character of these approaches. GEMMA was mainly dedicated to the qualification of structural materials in corrosive environments, stress relief and irradiation effects. INSPYRE and M4F had a heavier modelling component, with a view to improving engineering tools, such as fuel performance codes or component design codes. In the three projects, advanced modelling tools were combined with microstructure examination and experimental measurements in order to improve our
level of understanding of the mechanisms and rationalise collected data, enriching experimental data with data obtained from models. These and other cross-cutting materials science approaches, organized into five research lines, constitute the warp, while materials classes constitute the weft, on which the SRIA of ORIENT-NM is being built. In addition, all projects had an intensive and partly common education and training programme not described here, which corresponds to another important cross-cutting activity. It is expected that, by centralising these activities within the same initiative, such as a co-funded partnership, they can maintain their continuity, while benefitting on the one hand from cross-fertilisation and on the other hand from a single management and coordination structure.

## Conflict of interests

The authors declare that they have no competing interests to report.

## Acknowledgements

GEMMA, INSPYRE, and M4F received funding from the Euratom research and training programme 2014–2018, under Grant Agreements No. 754329 (INSPYRE), 755039 (M4F) and 755269 (GEMMA). ORIENT-NM is receiving funding from the Euratom research and training programme 2019–2020, under Grant Agreement No. 899997. All these projects contribute to the activities of the Joint Programme on Nuclear Materials of the European Energy Research Alliance (EERA-JPNM). The authors wish to thank the work package and task leaders, as well as all the participants, to whom goes the merit of the project's success.

## Funding

INSPYRE Project – Horizon 2020 Framework Programme – Grant Agreement No. 754329; GEMMA Project – Horizon 2020 Framework Programme – Grant Agreement No. 755269; M4F Project – Horizon 2020 Framework Programme – Grant Agreement No. 755039; ORIENT-NM Project – Horizon 2020 Framework Programme – Grant Agreement No. 899997.

## Data availability statement

This article has no associated data generated and/or analyzed/ Data associated with this article cannot be disclosed due to legal/ethical/other reasons.

## Author contribution statement

The four authors wrote the manuscript and coordinated the reported projects. The content of the work is the merit of the long list of project participants.

## References

1. L. Malerba, P. Agostini, M. Bertolus, F. Delage, A. Gallais-During, Ch. Grisolia, K. Liger, P.-F. Giroux, Advances on GenIV structural and fuel materials and cross-cutting activities between fission and fusion, EPJ Nuclear Sci. Technol. **6**, 32 (2020)

2. https://www.eera-jpnm.eu

3. European Commission, National Energy and Climate Plans (NECPs), https://ec.europa.eu/energy/topics/energy-strategy/national-energy-climate-plans_en

4. International Atomic Energy Agency, Country Nuclear Power Profiles, https://www.iaea.org/publications/13448/country-nuclear-power-profiles

5. World Nuclear Association, Country Profiles, https://world-nuclear.org/information-library/country-profiles.aspx

6. L. Malerba, Summary of national programmes on nuclear materials, ORIENT-NM Deliverable D1.3 (2021), http://www.eera-jpnm.eu/orient-nm/filesharer/documents/Deliverables_and_Milestones/Public%20deliverables

7. L. Malerba, A. Al Mazouzi, M. Bertolus, M. Cologna, P. Efsing, A. Jianu, P. Kinnunen, K.-F. Nilsson, M. Rabung, M. Tarantino, Materials for sustainable nuclear energy: a European strategic research and innovation agenda for all reactor generations, Energies **15**, 1845 (2022)

8. M. Angiolini, P. Agostini, S. Bassini, F. Fabbri, M. Tarantino, F. Di Fonzo, Challenges for coolants in fast neutron spectrum systems, in *IAEA-TECDOC-1912*, International Atomic Energy Agency TECDOC Series (2020), pp. 195–203, https://www.iaea.org/publications/13657/challenges-for-coolants-in-fast-neutron-spectrum-systems

9. I. Proriol Serre, I. Ponsot, J.-B. Vogt, Alumina-Forming Austenitic (AFA) steels and aluminium-based coating on 15-15 Ti steel to limit mechanical damage in presence of liquid lead-bismuth eutectic and liquid lead, in *MATEC Web of Conference* (EDP Sciences, 2021), Vol. 349, p. 02007 (2021)

10. H. Aït Abderrahim, P. Baeten, D. de Bruyn, R. Fernandez, MYRRHA – a multi-purpose fast spectrum research reactor, Energy Convers. Manag. **63**, 4 (2012)

11. M. Tarantino, M. Angiolini, S. Bassini, S. Cataldo, C. Ciantelli, C. Cristalli, A. Del Nevo, I. Di Piazza, D. Diamanti, M. Eboli, et al., Overview on lead-cooled fast reactor design and related technologies development in ENEA, Energies **14**, 5157 (2021)

12. B. Kvizda, G. Mayer, P. Vácha, J. Malesa, A. Siwiec, A. Vasile, S. Bebjak, B. Hatala, ALLEGRO Gas-cooled Fast Reactor (GFR) demonstrator thermal hydraulic benchmark, Nucl. Eng. Des. **345**, 47 (2019)

13. L. Rozumová, L. Košek, J. Vít, A. Hojná, P. Halodová, Comparison of corrosion behavior of the austenitic stainless steel 316 L in static and flowing liquid lead, ASME J. Nucl. Rad. Sci. **7** 021605 (2021)

14. S. Bassini, M. Angiolini, Assessment of liquid metal corrosion for candidate materials and welds of LFR reactor, GEMMA D4.2 Report, 2021

15. S. Gavrilov, Assessment of liquid metal corrosion for candidate materials and welds of MYRRHA primary circuit, GEMMA D4.1 Report, 2021

16. M. Angiolini, A. Antonelli, P. Agostini, S. Bassini, F. Fabbri, M. Falconieri, F. Mura, Oxidation of P91 steel in oxygen saturated lead, to be submitted to J. Nucl. Mater. (2022)

17. K.-F. Nilsson, Effect of liquid lead on mechanical properties of austenitic steels, their welds and implications for design, GEMMA D4.7 Report, 2021

18. G. Cacciamani, A. Dinsdale, M. Palumbo, A. Pasturel, The FeeNi system: thermodynamic modelling assisted by atomistic calculations, Intermetallics **18**, 1148 (2010)
19. K. Li, C.C. Fu, Ground-state properties and lattice-vibration effects of disordered Fe-Ni systems for phase stability predictions, Phys. Rev. Mater. 4, 023606 (2020)

20. www.eera-jpmn/inspyre

21. M. Bertolus et al., Synthesis of the INSPYRE results on MOX fuel behaviour, INSPYRE Deliverable D7.6 (2022), http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

22. P. Fouquet-Métivier, Ph.D. thesis, Université Paris-Saclay, France, 2022

23. D. Bathellier, M. Lainet, M. Freyss, P. Olsson, E. Bourasseau, A new law of heat capacity for UO₂, PuO₂ and (U, Pu)O₂ derived from molecular dynamics simulations and useable in fuel performance codes, J. Nucl. Mater. 549, 152877 (2021)

24. P. Chakraborty, C. Guéneau, A. Chartier, Modelling of plutonium diffusion in (U, Pu)O₂ₓ mixed oxide, Solid State Ion 357, 115503 (2020)

25. M. Gérardin, E. Gilabert, D. Horlait, M.-F. Barthe, G. Carlot, Experimental study of the diffusion of Xe and Kr implanted at low concentrations in UO₂ and determination of their trapping mechanisms, J. Nucl. Mater. 556, 153174 (2021)

26. H. Balboa Lopez, Ph.D. thesis, Université Paris-Saclay, France, 2018

27. P. Garcia, A. Miard, T. Helfer, J.-B. Parise, X. Iltis, G. Antou, The effect of oxygen partial pressure on dislocation creep in polycrystalline uranium dioxide, J. Eur. Ceram. Soc. 41, 2124 (2021)

28. J. Chen, J.-B. Parise, P. Garcia, A. Miard, P. Desgardin, P. Sigot, M.F. Barthe, Evolution of mechanical properties under ion irradiation, INSPYRE Deliverable D3.4 (2022), http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

29. A.L. Smith, E. Epifano, A. Quaini, C. Guéneau, M. Bertolus, K. Samuelsson, P. Olsson, Thermodynamic data of fission product phases for the JOG modelling, INSPYRE Deliverable D4.1 (2022), http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

30. A. Magni, L. Luzzi, D. Pizzocri, A. Schubert, P. Van Uffelen, A. Del Nevo, Modelling of thermal conductivity and melting behaviour of minor actinide-MOX fuels and assessment against experimental and molecular dynamics data, J. Nucl. Mater. 557, 153312 (2021)

31. D. Pizzocri, T. Barani, L. Cognini, L. Luzzi, A. Magni, A. Schubert, P. Van Uffelen, T. Wiss, Synthesis of the inert gas behaviour models developed in INSPYRE, INSPYRE Deliverable D6.4 (2020), http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

32. P. Van Uffelen, A. Schubert, L. Luzzi, T. Barani, A. Magni, D. Pizzocri, M. Lainet, V. Marelle, B. Michel, B. Boer, S. Lemehov, A. Del Nevo, Incorporation and verification of models and properties in fuel performance codes, INSPYRE Deliverable D7.2 (2020), http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

33. L. Luzzi, T. Barani, B. Boer, A. Del Nevo, M. Lainet, S. Lemehov, A. Magni, V. Marelle, B. Michel, D. Pizzocri, A. Schubert, P. Van Uffelen, M. Bertolus, Assessment of INSPYRE-extended fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment, submitted to Nucl. Eng. Technol. (2022), https://doi.org/10.1016/j.net.2022.10.038

34. B. Michel, M. Lainet, A. Magni, L. Luzzi, D. Pizzocri, Results of the applicative benchmark between TRANSURANUS and GERMINAL on the ASTRID case study, INSPYRE Deliverable D7.4 (2021) http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

35. A. Magni, M. Bertolus, M. Lainet, V. Marelle, B. Michel, A. Schubert, P. Van Uffelen, L. Luzzi, D. Pizzocri, B. Boer, S. Lemehov, A. Del Nevo, Fuel performance simulations of ESNII prototypes: results on MYRRHA normal and transient conditions, INSPYRE Deliverable D7.5 (2022), http://www.eera-jpmn.eu/inspyre/filesharer/documents/Deliverables%20&%20Milestones/Public%20deliverables

36. L. Malerba, et al., Multiscale modelling for fusion and fission materials: the M4F project, Nucl. Mater. Energy 29, 101051 (2021)

37. L. Malerba, M4F project – Final Report, Deliverable D8.6/D38, http://www.h2020-m4f.eu/filesharer/documents/Deliverables%20&%20Milestones/Deliverables/Public%20Deliverables

38. International Atomic Energy Agency, Technical Meeting on Synergies between Nuclear Fusion Technology Developments and Advanced Nuclear Fission Technologies, 6–10 Jun 2022, Vienna, Austria (hybrid event), https://www.iaea.org/events/evt2103079

39. G. Bonny, A. Bakaev, D. Terentyev, Combined effect of carbon and chromium enrichment on 〈100〉 loop absorption in iron, Comput. Mater. Sci. 211, 111533 (2022)

40. N. Kvashin, A. Ostapovets, N. Anento, A. Serra, On the migration of {332} (110) tilt grain boundary in bcc metals and further nucleation of {112} twin, Comput. Mater. Sci. 196, 110509 (2021)

41. N. Kvashin, P.L. García-Müller, N. Anento, A. Serra, Atomic processes of shear-coupled migration in {112} twins and vicinal grain boundaries in bcc-Fe, Phys. Rev. Mater. 4, 73604 (2020)

42. L. Dupuy, C. Robertson, I. Simonovski, T. Yalçinkaya, Towards simulations of dislocation channeling using large-scale dislocation dynamics simulations, M4F Deliverable D4.4 (2021), http://www.h2020-m4f.eu/filesharer/documents/Deliverables%20&%20Milestones/Deliverables/Public%20Deliverables

43. M. Sauzay, L. Dupuy, C. Robertson, I. Simonovski, T. Yalçinkaya, Continuum scale modelling of slip localization, M4F Deliverable D4.5 (2021), http://www.h2020-m4f.eu/filesharer/documents/Deliverables%20&%20Milestones/Deliverables/Public%20Deliverables

44. D. Gonçalves, I. Simonovski, M. Sauzay, T. Yalçinkaya, Polycrystal mean-field and full-field homogenization predictions of tensile behaviour, M4F Deliverable D5.3 (2021), http://www.h2020-m4f.eu/filesharer/documents/Deliverables%20&%20Milestones/Deliverables/Public%20Deliverables

45. R. Rajakrishnan, E. Gaganidze, J. Aktaa, Physically-based constitutive equations for describing deformation damage behaviour of F/M steels, M4F Deliverable D5.1 (2019), http://www.h2020-m4f.eu/filesharer/documents/
Deliverables%20&%20Milestones/Deliverables/
Public%20Deliverables

46. R. Rajakrishnan, E. Gaganidze, J. Aktaa, Simulation of post yield and post necking behavior by using the UMAT implementation of the developed-constitutive equations Impact of rational use of post yield post necking behavior of F/M steels on the development of the advanced design rules, M4F Deliverable D5.6 (2021), http://www.h2020-m4f.eu/filesharer/documents/Deliverables%20&%20Milestones/Deliverables/Public%20Deliverables

47. R. Rajakrishnan, E. Gaganidze, D. Terentyev, J. Aktaa, Macro-scale modeling of finite strain viscoplasticity in irradiated F/M steels: a continuum thermodynamic framework, submitted to the J. Mech. Phys. Solids (2022)

48. K. Vogel, P. Chekhonin, C. Kaden, M. Hernández-Mayoral, S. Akhmadaliev, F. Bergner, Depth distribution of irradiation-induced dislocation loops in an Fe-9Cr model alloy irradiated with Fe ions: the effect of ion energy, Nucl. Mater. Energy **27**, 101007 (2021)

49. N. Castin, et al., The dominant mechanisms for the formation of solute-rich clusters in low-Cu steels under irradiation, Mater. Today Energy **17**, 100472 (2020)

50. J.P. Balbuena, L. Malerba, N. Castin, G. Bonny, M.J. Caturla, An object kinetic Monte Carlo method to model precipitation and segregation in alloys under irradiation, J. Nucl. Mater. **557**, 153236 (2021)

51. L. Malerba, et al., Physical mechanisms and parameters for models of microstructure evolution under irradiation

in Fe alloys – Part I: pure Fe, Nucl. Mater. Energy **29**, 101069 (2021)

52. P.-M. Gueye, Ph.D. thesis, Université de Rouen Normandie, 2022

53. L. Veleva, P. Hähner, A. Dubinko, T. Khvan, D. Terentyev, A. Ruiz-Moreno, Depth-sensing hardness measurements to probe hardening behaviour and dynamic strain ageing effects of iron during tensile pre-deformation, Nanomaterials **11**, 71 (2021)

54. A. Ruiz Moreno, P. Hähner, L. Kurpaska, J. Jagielski, P. Spatig, M. Trebala, S. Hannula, S. Merino, G. De Diego, H. Namburi, O. Libera, D. Terentyev, T. Khvan, C. Heintze, N. Jennett, Round Robin into best practices for the determination of indentation size effects, Nanomaterials **10**, 130 (2020)

55. K. Vogel, C. Heintze, P. Chekhonin, S. Akhmadaliev, E. Altstadt, F. Bergner, Relationships between depth-resolved primary radiation damage, irradiation-induced nanostructure and nanoindentation response of ion-irradiated Fe-Cr and ODS Fe-Cr alloys, Nucl. Mater. Energy **24**, 100759 (2020)

56. A. Ruiz-Moreno, P. Hähner, F. Fumagalli, V. Haiblikova, M. Conte, N. Randall, Stress-strain curves and derived mechanical parameters of P91 steel from spherical nanoindentation at a range of temperatures, Mater. Des. **194**, 108950 (2020)

57. https://www.cencenelec.eu/media/CEN-CENELEC/News/Publications/2022/cen-cenelec_work_programme2022.pdf

Cite this article as: Lorenzo Malerba, Pietro Agostini, Massimo Angiolini, and Marjorie Bertolus. Towards a single European strategic research and innovation agenda on materials for all reactor generations through dedicated projects, EPJ Nuclear Sci. Technol. **8**, 36 (2022)