# Materials for sustainable nuclear energy.

Lorenzo Malerba, M. Bertolus, Karl-Fredrik Nilsson

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Lorenzo Malerba, M. Bertolus, Karl-Fredrik Nilsson. Materials for sustainable nuclear energy .. EERA-JPNM. 2019. cea-05504065

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MATERIALS FOR
SUSTAINABLE
NUCLEAR ENERGY

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

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L. Malerba (CIEMAT)

M. Bertolus (CEA, DEN)

K.F. Nilsson (JRC)

<figure><img src="image_5.png" /></figure>

Joint Programme on Nuclear Materials of the European Energy Research Alliance

Coordinating sustainable nuclear materials research for a low carbon Europe
MATERIALS FOR
SUSTAINABLE
NUCLEAR
ENERGY

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

Chair:

L. Malerba (CIEMAT)*

Main authors:

L. Malerba (CIEMAT)*

M. Bertolus (CEA, DEN)

K.F. Nilsson (JRC)

Contributors:

M. Angiolini (ENEA), M. Cologna (JRC), A. Jianu (KIT), J. Kalivodová (CVR)

C. Pareige (CNRS), M. Serrano (CIEMAT), J. Somers (JRC)

Editing:

J. Marrow (U. Oxford)
*Cover Images:* © KIT (left); © CEA (middle and right)

*Book design and layout:* MarianaEguaras.com
<table>
  <tbody>
    <tr>
      <td>4. CROSS-CUTTING ISSUES</td>
      <td>93</td>
    </tr>
    <tr>
      <td>4.1. Objectives</td>
      <td>93</td>
    </tr>
    <tr>
      <td>4.2. Commonalities with nuclear GenII/III reactor materials</td>
      <td>94</td>
    </tr>
    <tr>
      <td>4.3. Commonalities with materials for (V)HTR</td>
      <td>94</td>
    </tr>
    <tr>
      <td>4.4. Commonalities with fusion materials</td>
      <td>95</td>
    </tr>
    <tr>
      <td>4.5. Commonalities with materials for other energy technologies</td>
      <td>96</td>
    </tr>
    <tr>
      <td>5. WORKFLOW AND TIMELINE</td>
      <td>101</td>
    </tr>
    <tr>
      <td>6. INFRASTRUCTURES FOR NUCLEAR MATERIALS R&amp;D</td>
      <td>107</td>
    </tr>
    <tr>
      <td>6.1. Irradiation Facilities</td>
      <td>107</td>
    </tr>
    <tr>
      <td>6.1.1. Fast Neutron Facilities</td>
      <td>107</td>
    </tr>
    <tr>
      <td>6.1.2. Material Testing Reactors and associated 'hot' cells</td>
      <td>108</td>
    </tr>
    <tr>
      <td>6.1.3. Charged particle irradiation facilities</td>
      <td>109</td>
    </tr>
    <tr>
      <td>6.2. Handling of irradiated materials</td>
      <td>109</td>
    </tr>
    <tr>
      <td>6.2.1. 'Hot' Cells and shielded facilities</td>
      <td>109</td>
    </tr>
    <tr>
      <td>6.2.2. Transportation of samples for dedicated analysis</td>
      <td>110</td>
    </tr>
    <tr>
      <td>6.3. Computational facilities</td>
      <td>110</td>
    </tr>
    <tr>
      <td>6.4. General considerations on nuclear infrastructures in Europe</td>
      <td>110</td>
    </tr>
    <tr>
      <td>6.4.1. Costs related to crucial infrastructures build and use</td>
      <td>110</td>
    </tr>
    <tr>
      <td>6.4.2. Renewal of nuclear infrastructures</td>
      <td>111</td>
    </tr>
    <tr>
      <td>6.4.3. Sharing and joint programming of nuclear infrastructures</td>
      <td>111</td>
    </tr>
    <tr>
      <td>7. EDUCATION, TRAINING AND MOBILITY OF RESEARCHERS</td>
      <td>115</td>
    </tr>
    <tr>
      <td>8. INDUSTRY AND REGULATORS INVOLVEMENT</td>
      <td>121</td>
    </tr>
    <tr>
      <td>9. INTERNATIONAL COOPERATION</td>
      <td>125</td>
    </tr>
    <tr>
      <td>10. RESOURCES</td>
      <td>131</td>
    </tr>
    <tr>
      <td>11. RISK ASSESSMENT</td>
      <td>135</td>
    </tr>
    <tr>
      <td>12. SUMMARY AND RECOMMENDATIONS</td>
      <td>143</td>
    </tr>
  </tbody>
</table>
List of abbreviations

<table><tr><td>ADS</td><td>Accelerator-Driven System</td></tr><tr><td>AFA</td><td>Alumina Forming Austenitic (steels)</td></tr><tr><td>AI/ML</td><td>Artificial Intelligence / Machine Learning</td></tr><tr><td>AMPEA</td><td>Advanced Materials and Processes for Energy Applications (EERA JP)</td></tr><tr><td>ATF</td><td>Accident Tolerant Fuel</td></tr><tr><td>APT</td><td>Atom Probe Tomography</td></tr><tr><td>BU</td><td>Burn-Up</td></tr><tr><td>C/M</td><td>Carbide/Metal</td></tr><tr><td>CSE</td><td>Creep-Strength Enhanced</td></tr><tr><td>CSP</td><td>Concentrated Solar Power</td></tr><tr><td>CTR</td><td>Carbothermal Reduction</td></tr><tr><td>CVD</td><td>Chemical Vapour Deposition</td></tr><tr><td>CVI</td><td>Chemical Vapour Infiltration</td></tr><tr><td>DBTT</td><td>Ductile-Brittle Transition Temperature</td></tr><tr><td>DoW</td><td>Description of Work</td></tr><tr><td>dpa</td><td>Displacements per atom</td></tr><tr><td>DS</td><td>Deployment Strategy</td></tr><tr><td>EC</td><td>European Commission</td></tr><tr><td>EELS</td><td>Electron Energy Loss Spectroscopy</td></tr><tr><td>EERA</td><td>European Energy Research Alliance</td></tr><tr><td>EII</td><td>European Industrial Initiative</td></tr><tr><td>ENEN</td><td>European Nuclear Education Network</td></tr><tr><td>EOL</td><td>End Of Life</td></tr><tr><td>ESNII</td><td>European Sustainable Nuclear Industrial Initiative</td></tr><tr><td>ESTEP</td><td>European Steel Technology Platform</td></tr><tr><td>EMIRI</td><td>Energy Materials Industrial Research Initiative</td></tr><tr><td>EUMAT</td><td>European Technology Platform for Advanced Engineering Materials and Technologies</td></tr><tr><td>ETIP</td><td>European Technology and Innovation Platform</td></tr><tr><td>ETP</td><td>European Technology Platform</td></tr><tr><td>E&T(&M)</td><td>Education and Training (and Mobility)</td></tr><tr><td>EU</td><td>European Union</td></tr><tr><td>EURATOM</td><td>European Atomic Energy Community</td></tr><tr><td>FCCI</td><td>Fuel-Clad Chemical Interaction</td></tr><tr><td>FCMI</td><td>Fuel-Clad Mechanical Interaction</td></tr><tr><td>FeCrAl</td><td>Alumina forming F/M steels (from the main composing elements)</td></tr></table>
FEM Finite Element Models

FIB Focused Ion Beam

FGR Fission Gas Release

FOAK First Of A Kind

FPC Fuel Performance Codes

F/M Ferritic/Martensitic

GenIV Generation IV

GFR Gas-cooled Fast Reactor

HPC High Performance Computing

HEA High Entropy Alloys

HIP High Isostatic Pressing

HLM Heavy Liquid Metal

HRTEM High Resolution Transmission Electron Microscopy

HT High Temperature

HTR High Temperature Reactor

ICME Integrated Computational Materials Engineering

ICT Information and Communication Technology

IAEA International Atomic Energy Agency

IEA International Energy Agency

IPR Intellectual Property Rights

IR Integrated Roadmap (SET-plan)

JPNM Joint Programme on Nuclear Materials

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

KEMS Knudsen Effusion Mass Spectrometry

LBE Lead-bismuth Eutectic

LFR Lead-cooled Fast Reactor

LME Liquid Metal Embrittlement

LT Low Temperature

LTO Long-Term Operation

LWR Light Water Reactors

MA Minor Actinides

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

MOX Mixed uranium-plutonium OXide

MoU Memorandum of Understanding

MR Energy Materials Roadmap (SET-plan)

MSR Molten Salt Reactor

MS Member State

MTR Material Testing Reactor

M2F Expert Group on Multiscale Modelling of Fuels of the WPMM (OECD/NEA)

NC2I Nuclear Cogeneration Industrial Initiative
NDE Non-Destructive Examination

NEA Nuclear Energy Agency

N/M Nitride/Metal

NPP Nuclear Power Plants

NUGENIA Nuclear GENII/III Association

ODS Oxide Dispersion Strengthened (or Strengthening)

OECD Organisation for Economic Co-operation and Development

O/M Oxide/Metal

PAS Positron Annihilation Spectroscopy

PIE Post Irradiation Examination

PLD Pulsed Laser Deposition

PyC Pyrolitic Carbon

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

R&I Research and Innovation

R&D Research and Development

SANS Small-Angle Neutron Scattering

SCW SuperCritical Water

SCWR SuperCritical Water Reactor

SEM Scanning Electron Microscopy

SETIS SET-plan Information System

SET-plan Strategic Energy Technology plan

SFR Sodium-cooled Fast Reactor

SL Surface Layer

SNETP Sustainable Nuclear Energy Technology Platform

SPS Spark plasma sintering

SRA Strategic Research Agenda

SRIA Strategic Research and Innovation Agenda

TEM Transmission Electron Microscopy

TMT ThermoMechanical Treatment

TRL Technology Readiness Level

TSO Technical and Scientific support Organizations

(V)HTR (Very) High Temperature Reactor

WPMM Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems (OECD/NEA)

XRD X Ray Diffraction

2DS 2 Degree Scenario
# PREFACE

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

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

Nuclear energy in particular requires innovation in a broader sense. While the level of safety of nuclear installations in Europe is probably one of the highest in the world, it can be further improved with better materials. In addition, there is a need for better flexibility and adaptability in a wider spectrum of applications, and to pursue the ultimate goal of sustainability by improving efficiency, reducing waste volumes and hazard and, through
fuel breeding and recycling, making available resources sufficient to fuel
power plants for the next centuries and, why not, millennia.

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

Hervé BERNARD

<figure><img src="image_6.png" /><figcaption>CEA, former EERA Chair</figcaption></figure>
EXECUTIVE
SUMMARY

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

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

This **Strategic Research Agenda (SRA)** has been prepared by the EERA-JPNM, based on a **wide consultation with the scientific and industrial community** involved, to identify the **research lines** to be pursued in the EU to ensure that **suitable structural and fuel materials are available for the design, licensing, construction and safe long-term operation of GenIV nuclear systems**. Emphasis has been put on the fast neutron spectrum systems considered in the European Sustainable Nuclear Industrial Initiative (ESNII), namely sodium-cooled fast reactor (SFR), heavy liquid metal (HLM) cooled systems (accelerator driven systems -ADS- and
lead-cooled fast reactor -LFR-), and gas-cooled fast reactor (GFR). However, links with other GenIV systems, namely (very) high temperature reactor ((V) HTR), supercritical water reactor (SCWR), and molten salt reactor (MSR) have been established. Effort has been devoted to identify **GenIV materials' research issues that are of interest for other nuclear and also non-nuclear energy technologies**, as well. This was done with a view to optimise the use of available facilities, knowledge and human/financial resources, whenever possible, by joining forces with other research communities. Importantly, the content of **this SRA is consistent with relevant strategic documents and roadmaps that have been compiled by linked organisations and platforms** (SETPlan integrated roadmap, OECD/NEA technology roadmap, SNETP SRIA, ...), of which it is an **updated extension for what concerns materials for sustainable nuclear**.

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

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

The research activities are organized in blocks that result from the application, for structural and fuel materials, of a **well-established materials science approach**, which is based on the combination of three classes of activities: (1) **materials testing** for full **qualification** in environment and definition of design rules in a pre-normative spirit; (2) development of **mechanistic and physical models** in support of materials behaviour correlations used in design rules and improvement of materials properties; and (3) **development of advanced materials solutions** through experimental screening of solutions, assisted by models that are rooted in understanding of the physical processes that govern materials behaviour. Considerations of the
circular economy and advanced manufacturing methods based on au-
tomation of procedures, as well as **materials monitoring**, also enter this
third class of activities. In all cases, **data management** is a crucial issue to
ensure data collection, storage, interoperability, and re-usability.

The **materials of interest** for all **demonstrators**, on which the qualification and pre-normative research effort is **mainly focused**, are **austenitic stainless steels** for structural functions (Ni-based alloys for out-of-core applications; in some cases ferritic/martensitic -F/M- steels) and **MOX** as fuel, with **ceramics** (mainly SiC/SiC) and $UO_2$ as fuel for the high temperature operation of e.g. the GFR demonstrator. **Surface protection** may be necessary already for demonstrators to provide sufficient compatibility with coolants. Improvements of safety, performance and economy in future prototypes and then commercial reactors advises the exploration of **improved austenitic steels**, **advanced F/M steels**, **refractory alloys** (V and Mo), **oxide dispersion strengthening**, **advanced surface protection methods** and, for the farther future, prospective materials such as **high entropy alloys** or **MAX phases**. Reduction of waste and increased safety requires the development of **minor actinide bearing fuel** from multirecycling, as well as advanced fuel forms such as **uranium nitrides or carbides**.

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

Concerning **fuel materials**, the properties and processes that govern its
behaviour in pile, on which research effort is focused, are: **margin to melt-**
**ing** (establishment of phase diagrams and evolution of thermal properties),
atomic **transport properties** and ensuing microstructural evolution, **fis-**
**sion product** (non-gaseous) and **helium** (gas) behaviour and transport,
**mechanical properties** (their evolution, subsequent fragmentation and
cracking, fuel-cladding mechanical interaction), and **compatibility with**
**cladding and coolant** (internal cladding corrosion, chemical interactions
especially in case of severe accident). These are all addressed from both an experimental and a modelling perspective.

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

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

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

Several classes of facilities and infrastructures are necessary for the qualification of nuclear materials in general, and specifically for GenIV systems. The specificity of nuclear energy makes **irradiation facilities essential for the qualification of nuclear materials**. Ideally GenIV materials should be qualified by exposure to fast neutrons up to high dose, however at the moment **Europe is totally dependent on non-European countries to have access to fast neutron facilities**. Working **materials testing reactors (MTR)** with dominantly thermal spectrum, equipped with relevant hot cell facilities, exist **in Europe**, but there are **not many** of them. Moreover, **hot cells and shielded facilities** where active materials can be handled and tested are costly infrastructures to maintain and, for this reason, are also **limited in number** and often becoming **obsolescent**, despite the construction of new facilities in some countries: this limits severely the number of tests, measurements and examinations that can be performed on neutron irradiated materials. **Transport** of active materials, especially fuel, is also becoming problematic, largely because of the **lack of harmonisation of regulations** throughout Europe. For all these reasons, **charged particle irradiation** facilities become crucial to provide data that, although unsuitable for qualification, can be exploited for **screening and modelling purposes**. For the latter, access to suitable **computational resources** becomes also important, though there
is currently **no dedicated resource for fission** material studies. Overall, a number of **actions are needed** in order to make sure that sufficient facilities and infrastructures are available in Europe for nuclear materials research. Amongst them, a better **coordinated use and development** of infrastructures throughout Europe is desirable, based on a principle of complementarity, rather than competition, including **common rules for the access to facilities and infrastructures**.

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

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

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

Besides the obvious need of adequate financial resources in order to address the research problems outlined in this SRA, as well as the necessary corollaries, **four recommendations** emerge that this document is intended to bring to the attention of stake-holders, particularly 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 stakeholders assets.
1

INTRODUCTION
<figure><img src="image_7.png" /></figure>
# 1. INTRODUCTION

## 1.1. Societal, economic and technical challenges

With a yearly production in excess of 850 TWh<sub>e</sub>, nuclear energy is the single largest source of low-carbon electricity in the European Union.<fnref n="1" /> This corresponds to more than ¼ of the electricity in Europe, and contributes to guaranteeing a secure and reliable base-load supply. Consistent with this, in the 2 degree scenario (2DS) of the International Energy Agency (IEA), it is considered that worldwide nuclear energy production should increase from 11% in 2015 to 16%/15% in 2050/2060.<fnref n="2" /> Thus, nuclear energy plays an important societal role, together with renewables, in the energy transition. However, three main open issues remain: (1) sustainable and responsible use of resources; (2) accident risk; and (3) long-lived nuclear waste.

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

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

<footnote n="1">¹ In 2014, 876 TWh were produced by nuclear. This, combined with 406 TWh of hydroelectricity covering especially peaks, corresponds to about 40% of the total electricity and represents by far the largest portion of low-carbon electricity in Europe. Wind comes third, with 253 TWh (source: Eurostat - http://ec.europa.eu/eurostat/statistics-explained/index.php/Electricity_and_heat_statistics). However, wind keeps growing and produced, in 2016, almost 300 TWh (source: Wind-europe - https://windEurope.org/about-wind/statistics/european/wind-in-power-2016/).</footnote>

<footnote n="2">² Source: Energy Technology Perspectives, IEA, 2016/2017.</footnote>

<footnote n="3">³ Technology Roadmap of the Generation IV International Forum: https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/gif-tru2014.pdf</footnote>
electricity produced for a given thermal power, but also less waste heat, less environmental impact and less need for cooling. Thermal efficiency depends on the temperature of the reactor core and on the performance of the conversion system. Present LWRs have thermal efficiencies around 33%, while modern coal-fired plants reach approximately 39% and combined-cycle gas plants are even 50 to 60% efficient. The use of non-aqueous coolants, mandatory in fast neutron reactors, will allow operation at temperatures well-above those of LWRs, thereby pushing the efficiency close to (or higher than) coal plants and, in the case of very high temperature, combined-cycle systems. With liquid metals as coolants this can be achieved close to atmospheric pressure, with a design that is based on an increased use of passive safety systems. This provides designers with an important tool to make fast breeder reactors both safe and economically attractive.

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

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

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

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

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

<footnote n="4">⁴ J. Magill, V. Berthou, D. Haas, J. Galy, R. Schenkel, H.-W. Wiese, G. Heusener, J. Tommasi, G. Youinou, "Impact limits of partitioning and transmutation scenarios on the radiotoxicity of actinides in radioactive waste", Nuclear Energy 42(2003)263</footnote>
**of nuclear materials is an essential point to make GenIV reactors a reality.** Furthermore, the development of materials with superior resistance to high temperature and exposure to aggressive coolants can also be beneficial for other energy technologies, such as concentrated solar power (CSP), geothermal energy, bioenergy and fuel cells and hydrogen.

## 1.2. The EERA JPNM in the European platform landscape

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

The European Sustainable Nuclear Energy Industrial Initiative (ESNII<fnref n="6" />), which has the task of developing GenIV fast neutron reactors in Europe using different technologies, was launched under the umbrella of the sustainable

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

nuclear energy technology platform (SNETP<fnref n="7" />) at the same time as the EERA JPNM. **The EERA JPNM provides the R&D for the structural and fuel materials needed for the development and implementation of fast reactors in Europe,** as defined by ESNII. Currently, this is the main reason of existence of the JPNM in EERA. Figure 1 illustrates the connections between energy platforms, in particular nuclear energy platforms, in Europe.

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

* Material behaviour for structural components and fuel

* Structural integrity of systems and components

* Manufacturing & assembly technology"

<footnote n="5">⁵ https://ec.europa.eu/research/participants/data/ref/h2020/wp/2014_2015/annexes/h2020-wp1415-annex-g-trl_en.pdf</footnote>

<footnote n="6">⁶ http://www.snetp.eu/esnii/</footnote>

<footnote n="7">⁷ http://www.snetp.eu/</footnote>

<footnote n="8">⁸ http://www.snetp.eu/wp-content/uploads/2016/01/SNETP-DEPLOYMENT-STRATEGY-2015-WEB.pdf</footnote>
<figure><img src="image_8.png" /><figcaption>FIGURE 1: Illustration of the connection between (nuclear) energy platforms in Europe.</figcaption></figure>

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

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

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

<footnote n="9">⁹ http://www.nugenia.org/</footnote>

<footnote n="10">¹⁰ http://www.snetp.eu/nc2i/</footnote>
**collaboration with other joint programmes,**
targeting high temperature operation in envi-
ronmentally harsh environments, for example
AMPEA (Advanced Materials and Processes for
Energy Application), CSP, geothermal energy,
bioenergy and fuel cells and hydrogen.

1.3. Generation IV systems
in Europe

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

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

* Parallel deployment of GenIV fission reactors and systems, guaranteeing more sustainable and safe nuclear energy, with the potential for nuclear heat generation.<fnref n="11" />

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

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

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

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

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

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

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

<footnote n="11">¹¹ In the long run, the gradual insertion of fusion systems in the energy production market, in cohabitation with fission systems, may also occur.</footnote>

<footnote n="12">¹² For an analysis of the prospects for GenIV systems, in-
cluding technical descriptions and economics considera-
tions, see Locatelli, Mancini and Todeschini, Energy Policy
61 (2013) 1503.</footnote>

<footnote n="13">¹³ At the time of preparation of the current document the AS-
TRID project is experiencing profound revision and may be
cancelled as known now, however it keeps being used as
reference SFR demonstrator throughout this document.</footnote>
a nuclear reactor and a non-nuclear plant, in a framework of CO₂ emission reduction in heat production.

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

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

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

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

## 1.4. Objectives and Grand Challenges of the EERA JPNM

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

1. **Better knowledge of materials behaviour** under service conditions, seeking predictive capability, to select the most suited materials and define for them safe design rules, especially considering radiation and temperature effects, and taking care of their compatibility with coolants.

2. **Development of innovative material solutions** to improve resistance to high temperature, irradiation and aggressive environments, through suitable processing/protective methods applied to existing materials or developing new types of materials

The EERA JPNM **vision paper**<fnref n="15" /> expresses a view on nuclear energy materials strongly linked with the ESNII systems and identifies three Grand Challenges. These are based on the three pillars of the EERA JPNM's research approach and strategy (see Figure 2):

* **Materials characterization:** Assessment of candidate structural and fuel materials' behaviour in operational conditions: screening, selection and qualification with a view to developing design rules;

* **Materials modeling:** Development of advanced models to rationalise materials' behaviour, underpin design correlations and provide a basis for the improvement of materials properties by providing predictive capability;

* **Innovative materials:** Development of innovative structural and fuel material solutions.

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

<footnote n="15">¹⁵ https://www.eera-set.eu/wp-content/uploads/Vision-Paper-EERA-JP-Nuclear-Materials-February-2015.pdf</footnote>
tions and advanced fabrication processes for industrial use, with superior capabilities in terms of resistance to irradiation, high-temperatures and aggressive environment.

<figure><img src="image_9.png" /><figcaption>FIGURE 2: The research approach and strategy pillars of the EERA JPNM.</figcaption></figure>

The **three grand challenges** correspondingly identified are:

* **Grand Challenge 1:** Elaboration of design correlations, assessment and test procedures for the structural and fuel materials that have been selected for the demonstrators, under the service conditions expected. This involves deployment of infrastructures for exposure to ageing environments, testing of materials and production of data and knowledge.

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

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

* **Grand Challenge 3:** Development of innovative structural and fuel material solutions and fabrication processes, in partnership with industry, to achieve superior thermomechanical properties, better compatibility with coolants and improved radiation-resistance, so as to increase safety and improve efficiency and economy.

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

## 1.5. Purpose, target and structure of this Strategic Research Agenda

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

* **Research on materials** that impact structural integrity, component ageing and advanced solutions for components in nuclear systems, **is a continuous process** without a deadline: it constitutes the research 'humus' on which innovation, and consequently better safety and efficiency for nuclear systems, can grow;

* In order to allow the **licensing and construction** of GenIV demonstrators, taking into account also recent developments, the research on materials needs to provide sufficient data for qualification and possibly codification of design rules, as well as computer simulation, depending on system and issue, in a horizon that has a span of **10 to 20 years**.<fnref n="16" />

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

<footnote n="16">¹⁶ It is not possible to establish a priori when the data about a given material can be considered sufficient, because it depends enormously on the system, the component and the targeted operating conditions. This is why a continuous interaction between materials scientists and designers is needed.</footnote>
This horizon, taking into account the specificities of nuclear energy, is relatively short. To be met, if the goal of sustainability is to be reached, the deployment is required now of significant resources that are devoted to materials research and development.

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

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

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

<figure><img src="image_10.png" /><figcaption>FIGURE 3: Documents describing the research programme of the EERA JPNM with increasing level of detail from top to bottom.</figcaption></figure>

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

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

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

* The **strategic research agenda (SRA)**, which is revised periodically,<fnref n="17" /> describes for stake-holders the R&D route to be followed (roadmap) to face the three Grand Challenges of the EERA JPNM identified in the vision paper, extending also to the other strategic activities that accompany research.

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

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

<footnote n="17">¹⁷ This SRA may need to be updated if: (a) materials considered have become obsolete; (b) new materials solutions appear; (c) the reference system conditions change substantially.</footnote>
<figure><img src="image_11.png" /><figcaption>FIGURE 4: Matrix structure of the SRA: the crossing of infratechnologies and materials identifies blocks which form the sections of the part 3 of this document.</figcaption></figure>

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

This SRA adopts a matrix structure (C. Featherstone and E. O'Sullivan<fnref n="19" />), as schemat-

ically described in Figure 4, which is a visual representation of its “fabric”: the horizontal ‘warp’ is given by the cross-cutting research strategies, or infratechnologies, while the materials addressed, structural and fuel, provide the vertical ‘weft’<fnref n="20" />. Within each matrix element, or block, different issues and relevant goals are identified.

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

<footnote n="18">¹⁸ http://ec.europa.eu/research/industrial_technologies/pdf/research-road-mapping-in-materials_en.pdf</footnote>

<footnote n="19">¹⁹ Featherstone and O'Sullivan, A review of international public sector strategies and roadmaps: a case study in advanced materials, Centre for Science Technology and Inno-</footnote>

<footnote>vation, Institute for Manufacturing, University of Cambridge, March 2014.</footnote>

<footnote n="20">²⁰ Research Road-Mapping in Materials, A. F. de Baas (Ed), European Commission, DG Research, EUR 24210 EN (2010), doi: 10.2777/87000.</footnote>
while physical models and simulation tools describing the behaviour of materials support and have the potential to accelerate both pre-normative research and development of new material solutions.

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

* Actual roadmap in terms of description of key issues and relevant activities;

* Identification of cross-cutting research with other nuclear and non-nuclear energy technologies;

* Needs for specific facilities and infrastructures and viable access schemes;

* Interaction with industry;

* Benefits of international cooperation;

* Some aspects related with education and training;

* Risk assessment.

## 1.6. Sources and links with other roadmaps and SRAs

To elaborate the present SRA, the priorities of ESNII concepts and the objectives of SNETP were taken into account, while identifying commonalities with other nuclear and also non-nuclear energy technologies. Towards this aim the EERA-JPNM launched, in 2015, a wide consultation involving ESNII representatives to report on the most recent design of the prototypes, and materials scientists active in the research performed within the six sub-programmes of the JPNM. Subsequently, contacts were established with the SNETP pillars<fnref n="21" /> and the fusion community, as well as with other joint programmes in EERA<fnref n="22" />. A MoU was signed between EERA and SNETP in December 2016, with the purpose of agreeing officially on common research topics. As a first action for the

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

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

This roadmap was issued after a wide consultation among stake-holders that was launched by the EC in December 2011.<fnref n="23" /> It highlighted the steps to be followed in the field of materials for advanced low-carbon energy technologies, defining a 10-year European Research and Innovation (R&I) agenda. For nuclear fission it proposed a Research and Development (R&D) programme on commercially available materials for the prototypes and demonstrators, and on advanced materials for industrial scale systems. This programme targeted materials for cladding, coatings for enhanced corrosion and erosion/wear resistance and novel advanced materials. Specifically: (i) manufacturing and out of pile testing of F/M 9%Cr steels for heat exchanger, (ii) manufacturing and out of pile (and possibly in-pile) testing of oxide dispersion strengthened (ODS) claddings, (iii) manufacturing and out of pile (and possibly in-pile) testing of SiCf/SiC composites cladding and (iv) manufacturing and testing of coatings. While not all of these targets are still relevant, **several technical issues mentioned in the MR are also in the focus of this SRA, which updates them to the current perception of needs.**

### 1.6.2. The SET-plan Integrated Roadmap (IR)

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

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

<footnote n="22">²² https://www.eera-set.eu/wp-content/uploads/EERA-JP-workshop-Materials_for_Energy_report.pdf</footnote>

<footnote n="23">²³ https://setis.ec.europa.eu/system/files/Materials_Roadmap_EN.pdf</footnote>
Nuclear energy is
recognized by the
SET-plan as a low-carbon
energy source. Materials
have clearly an important
role to play to reach the
SET-plan target of nuclear
safety and its related
Implementation Plan.

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

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

1.6.3. OECD/NEA Technology Roadmap for Nuclear Energy and other OECD/NEA initiatives

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

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

1.6.4. SNETP Strategic Research and
Innovation Agenda and Deployment
Strategy

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

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

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

<footnote n="24">²⁴ https://www.oecd-nea.org/pub/techroadmap/</footnote>

<footnote n="25">²⁵ http://www.snetp.eu/wp-content/uploads/2014/05/
sria2013_web.pdf</footnote>
Materials behaviour for
structural components and
fuel and more generally
the structural integrity of
systems and components
are common trunks through
GenII/III and IV, cogeneration,
and also fusion.

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

* Promote the use of nuclear energy beyond electricity generation, namely in cogeneration of heat or hydrogen production or water desalination.

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

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

1.6.5. GIF Technology Roadmap³ and R&D Outlook

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

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

1.6.6. Other roadmaps

Other documents connected with the present
SRA in terms of issues, materials and goals,
that were consulted during its preparation, are:
(i) the SRA of EuMaT, the European Technology
Platform for Advanced Engineering Materials
and Technologies<fnref n="27" />; (ii) the roadmap of Materials
Modelling of the European Materials Modelling
Council<fnref n="28" />; the Roadmap of the Metallurgy
Europe – EUREKA cluster<fnref n="29" />; the Roadmap to
Fusion Electricity (2012 version)<fnref n="30" />.

<footnote n="26">²⁶ https://www.gen-4.org/gif/jcms/c_42188/publications</footnote>

<footnote n="27">²⁷ http://eumat.eu/filehandler.ashx?file=11580</footnote>

<footnote n="28">²⁸ https://emmc.info/wp-content/uploads/2015/04/EMMC_Roadmap_V3.0.2.pdf</footnote>

<footnote n="29">29 http://metallurgy-europe.eu/</footnote>

<footnote n="30">³⁰ https://www.euro-fusion.org/fileadmin/user_upload/EUROfusion/Documents/Roadmap.pdf</footnote>
2

NUCLEAR
MATERIALS
FOR GEN IV
SYSTEMS AND
RESEARCH
APPROACHES
<figure><img src="image_12.png" /></figure>
## 2. NUCLEAR MATERIALS FOR GEN IV SYSTEMS AND RESEARCH APPROACHES

### 2.1. Material degradation processes and relevant properties

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

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

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

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

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

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

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

## 2.2. Nuclear materials, components and systems

### 2.2.1. Structural materials

Structural materials are those used to fabricate a component that bears load or stress, whatever its origin (mechanical, thermal, vibrations...).

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

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

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

* Non-replaceable components constitute the main structure of the reactor. Major examples of this type are the vessel or, in pool reactors, the upper cover of the containment. These components are characterized by the fact that their replacement, though theoretically not impossible, essentially corresponds to building a new unit, i.e. it is so costly and complex that it is
economically not affordable.<fnref n="31" /> These components need to be designed for the full lifetime of the reactor or, correspondingly, their lifetime defines the lifetime of the reactor itself.

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

The choice of the most suitable materials for a given component and application obviously comes, to a large extent, from previous experience. **The current design of GenIV demonstrators planned in Europe envisages the use of austenitic steels as the main class of in-core structural materials,** specifically 316L(N) for most components, including the vessel, and 15-15 Ti for the cladding and other fuel element parts<fnref n="32" />. Depending on the specific demonstrator, other materials may also enter, e.g. **ferritic/martensitic (F/M) steels** in the SFR core, while for the GFR core **cercamic materials** similar to those used in HTR, or perhaps **refractory metallic alloys,** must be considered. Other materials foreseen for use in **out-of-core** components are **Ni-based**

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

**F/M steels** (e.g. T91, EM10, HT9, reduced activation F/M like EUROFER or F82H for fusion,...) **potentially offer a number of desirable properties as cladding and core materials** that are

<footnote n="31">³¹ In the history of nuclear energy, however, some components initially meant to be irreplaceable have been eventually replaced, because the high cost was counterbalanced by the expected gain from the exploitation of the plant.</footnote>

<footnote n="32">³² These are AISI and customarily used names. In Europe the formally correct way to denote them is X2CrNiMo 17 12 2 controlled nitrogen content (for 316L(N)) and X10CrNiMoTiB 15 15 (for 15-15 Ti); the latter, when of German production, is also denoted as 1.4970.</footnote>
superior to those of austenitic steels, namely better thermal conductivity, lower expansion coefficient and better resistance to radiation-induced void swelling. The third point is key: while austenitic steel cladding will not sustain more than 100 dpa of irradiation dose,<fnref n="33" /> F/M steels are **expected to reach 200 dpa without experiencing any swelling**, thereby offering a promising solution to significantly increase the burnup. However, F/M steels can currently operate only below 550°C, suffer from radiation-induced embrittlement below 350-400°C and are especially prone to liquid metal embrittlement (LME), when in intimate contact (wetting) with, in particular, heavy liquid metals (HLM)<fnref n="34" />. Reliable **correlations and models for the identification of design rules** allowing for all these phenomena are **needed to make F/M steels** usable in GenIV systems; alternatively, the resistance to high temperature and embrittlement of F/M steels, as well as the compatibility with HLM, need to be improved.

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

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

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

Moreover, the design of the high temperature gas cooled reactor concept (GFR, V/HTR), 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

<footnote n="33">³³ Currently, 15-15 Ti is supposed to reach 90 dpa, better versions of it (AIM2) might allow this limit to increase up to 110 dpa.</footnote>

<footnote n="34">³⁴ Liquid lead and its alloys, such as Lead-Bismuth Eutectic (LBE) or Pb-Li, are collectively denoted as heavy liquid metals (HLM). Recently it has been observed that also austenitic steels in liquid sodium may experience LME, although the conditions for this to happen are not easy to comply with.</footnote>
<table><caption>TABLE 1: Main structural materials expected to be used in the different ESNII GenIV systems, distinguishing different phases that define the timeline, from demonstrator to prototype (FOAK), and finally commercially deployed plants.</caption><thead><tr><th rowspan="2">SYSTEMS ↓</th><th colspan="2">PHASES →</th><th colspan="2">ESNII demonstrator</th></tr><tr><th>As licensed (phase I)</th><th>Evolving (phase II)</th><th>FOAK (prototype)</th><th>Commercial deployment</th></tr></thead><tbody><tr><td rowspan="2">SFR (ASTRID)</td><td>Periodically Replaced Components</td><td>AuSS: 15-15Ti - AIM1 (cladding) F/M: EM10 (wrapper)</td><td>AIM2 or F/M ODS (cladding)</td><td>TMT F/M or F/M ODS</td><td>TMT F/M, F/M ODS, perhaps SiC/SiC</td></tr><tr><td>Permanent Structural Components</td><td>AuSS: AISI316L(N); 800SPH</td><td></td><td>AuSS: AISI316L(N); TMT F/M</td><td></td></tr><tr><td rowspan="2">ADS (MYRRHA)</td><td>Periodically Replaced Components</td><td>Cladding: 1.4790; structures: 316L(N)</td><td>Coated 15-15Ti (FeAl, FeCrSi, FeTa, MAX phases, ...) or AFA</td><td rowspan="2">N/A</td><td rowspan="2">AFA or FeCrAl ODS, or (coated) Mo-ODS, or SiC/SiC</td></tr><tr><td>Permanent Structural Components</td><td>316L(N)</td><td></td></tr><tr><td rowspan="2">LFR (ALFRED)</td><td>Periodically Replaced Components</td><td>Cladding and structures: (Al<sub>2</sub>O<sub>3</sub> coated) 15-15Ti (AIM1)</td><td>Cladding and structures: Al<sub>2</sub>O<sub>3</sub> Coated 15-15Ti or AFA</td><td>Cladding: AFA or FeCrAl ODS Structures: AFA</td><td></td></tr><tr><td>Permanent Structural Components</td><td>316L(N)</td><td></td><td>AFA or ferritic steel lined with AFA</td><td></td></tr><tr><td rowspan="2">GFR (ALLEGRO) / (V)HTR</td><td>Periodically Replaced Components</td><td>GFR: T&lt;550°C: 15-15Ti (cladding) - AFA? HTR: TRISO (SiC)</td><td>GFR: T&gt; 850°C: SiC/SiC (cladding) / HTR TRISO (SiC)</td><td>SiC/SiC, perhaps Mo-ODS/ HTR TRISO (SiC)</td><td></td></tr><tr><td>Permanent Structural Components</td><td>GFR: T&lt;550°C: 316L(N) - AFA? HTR: graphite</td><td>GFR: 550&lt;T&lt;850°C: AFA, FeCrAl? HTR: graphite</td><td>GFR: AFA or FeCrAl, perhaps Mo or V alloys HTR: graphite</td><td></td></tr></tbody></table>

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

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

are: ODS-Mo, high entropy alloys (HEA), and MAX phases.<fnref n="35" /> For all these materials, including SiC/SiC, there are issues concerning fabrication processes, optimal choice of composition/architecture of the component, joining, standardization of testing, and the usual problems of radiation resistance in contact with coolants.

<footnote n="35">³⁵ Most likely suitable as a corrosion protective coating at not-too-high temperatures.</footnote>
Based on this excursus on structural materials for GenIV systems, Table 1 attempts a classification and a prediction of the main structural materials that are expected to be used in the different GenIV systems that are currently part of ESNII, distinguishing different time phases from demonstrator (phase I and II) to prototype first-of-a-kind (FOAK) and, finally, commercially deployed plants. **This table is important because it allows the different materials and therefore the research and development devoted to them to be set in the correct timeframe and in connection with a specific system.**

The actual timing for each system is not the same and will crucially depend on the priorities that are politically associated with one system or another and the readiness with which funding is allocated: none of this can be easily predicted. However, from a conceptual viewpoint the phases schema-tized in the table remain valid and therefore, **through this table, it is possible to establish priorities and perspectives.** Table 1 is not expected to be fully comprehensive, but simply highlights which classes of materials the EERA JPNM research focuses on and where they fit, in order to immediately understand the application and context. This table is based on current knowledge and perceptions about the future and will require updates. Phases I/II of the prototypes are not necessarily defined in the ESNII plans and can be of very different nature, but essentially correspond to switching to different materials, particularly cladding and fuel element materials, for better performance. For instance, the GFR considers a first demonstrator working at low temperature (~500°C) to be later upgraded for high temperature operation (~850°C), when the qualification of the necessary materials is sufficiently advanced; a similar strategy is envisaged for the LFR demonstrator (380→520°C). Finally, concerning FOAK and commercially deployed reactors, the materials mentioned in the table are obviously an educated guess. The message is that materials that are not currently envisaged for the demonstrators remain a target of the EERA JPNM research, because of their long term potential.

### 2.2.2. Fuel materials

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

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

* Provide the power expected during their whole stay in reactor;

* Use the fissile elements as best as possible to reduce the cost of energy production;

* Confine the fission products inside the fuel elements in all operating and accidental conditions.

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

**MOX will be the fuel for the first cores of ALL ESNII fast neutron reactor prototypes**, although $U_2$ with <20% U-235 is planned to be used in the first core of ALLEGRO. Variations of MOX were used in the previous European fast reactor programmes: Phénix, Superphénix, and
Dounreay fast reactors. MOX fuel is Europe's only major knowledge and competence base for fast reactors, even if all European fast reactors are now closed. It is thus natural that this fuel is the first choice for the future.

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

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

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

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

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

Two types of concepts are envisaged for transmutation.

The first concept, known as **homogeneous mode**, involves diluting minor actinides in standard fast-reactor fuel. The advantage is the strong similarity between the structures of the various actinide oxides (fluorite-type cubic structure) and their mutual solubility. To minimize the impact of the introduction of minor actinides on reactor safety parameters, fuel minor actinide content is kept relatively low (a few % of heavy atoms). In such conditions, only a slight evolution is anticipated in fuel behaviour and performance, the properties remaining quite close to those of standard fuel, which facilitates the qualification of these fuels.

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

Because of the high neutron emission, thermal power and toxicity of the minor actinides, the **fabrication of MA-bearing fuels requires**
<table><caption>TABLE 2: Main fuel materials expected to be used in the different ESNII GenIV systems, distinguishing different phases that define the timeline, from prototype/demonstrator to FOAK and finally commercially deployed plants.</caption><thead><tr><th rowspan="2">SYSTEMS ↓</th><th colspan="2">ESNII demonstrator</th><th rowspan="2">FOAK (prototype)</th><th rowspan="2">Commercial deployment</th></tr><tr><th>As licensed (phase I)</th><th>Evolving (phase II)</th></tr></thead><tbody><tr><td>SFR (ASTRID)</td><td>MOX 20-25% Pu from UOx recycling</td><td>MOX 20-25% Pu from MOX recycling, MA bearing fuels?</td><td>MOX 20-25% Pu from MOX multirecycling, MA bearing oxide fuels</td><td></td></tr><tr><td>ADS (MYRRHA)</td><td>MOX with up to 30-35% Pu</td><td>MOX, MA bearing fuels</td><td>N/A</td><td></td></tr><tr><td>LFR (ALFRED)</td><td>MOX with up to 30% Pu</td><td>MOX, advanced MX</td><td>Advanced MX, MA bearing fuels</td><td></td></tr><tr><td>GFR (ALLEGRO) / (V) HTR</td><td>UO2 with &lt;20% U-235</td><td>MOX, advanced MX</td><td>Advanced MX, MA bearing fuels</td><td></td></tr></tbody></table>

**heavy shielding.** A down selection of optimal fabrication routes must be achieved and remote handling for fuel fabrication and assembly production engineered and qualified. Then, the assessment of the irradiation performance of these fuel types, performed in the frame of several European projects since the 90's, needs to be completed.

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

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

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

## 2.3. Materials research approaches

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

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

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

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

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

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

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

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

<footnote n="37">³⁷ In design codes, e.g. RCC-MRx, operating conditions can be defined as: (1) conditions to which component may be subjected in the course of normal operation, including normal operating incidents, start-up and shutdown; (2) emergency conditions corresponding to very low probability of occurrence, but which must nonetheless be considered, and which imply shut down and appropriate inspection of the component or of the plant; (3) highly improbable operating conditions, whose consequences on components are studied among others for safety reasons. Severe accident conditions are not necessarily part of design codes, although In the aftermath of Fukushima this is being discussed ("Design Extended Conditions").</footnote>

<footnote n="38">³⁸ HLM compatibility is not included in any code today and it is quite urgent to develop codes and standards. The two key HLM related degradation mechanisms are corrosion (by oxidation and dissolution) and liquid-metal embrittlement.</footnote>
Such accident scenarios generally translate into load excursions with high temperatures and load rates. The temperature range for mechanical properties assessment needs therefore to be extended and strain rate properties are necessary in the relevant range. When such data do not exist, specific research programmes are required.

In this context **materials qualification** means generation and maintenance of evidence to ensure that material or equipment will operate on demand, under specified service conditions, meeting system performance requirements.<fnref n="39" />

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

**Materials qualification requires dedicated experiments** both out of pile and in pile,<fnref n="40" /> to collect comprehensive and reliable data of the material properties relevant for the operational conditions.

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

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

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

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

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

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

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

<footnote n="40">⁴⁰ Pile is the core of a reactor, where the materials are exposed to neutron irradiation.</footnote>

<footnote n="41">⁴¹ No neutron irradiation facility currently exists anywhere in the world that allows doses on the order of 100 dpa to be reached within reasonable times and at affordable costs. Such facilities did exist in the past, though.</footnote>
developed to be conservative and simple to apply at the design stage. An important concept is "allowable stresses"; these depend on the material and the temperature and should ensure that deformations are in the elastic regime. Eventually, **design rules are collected in design codes**,<fnref n="42" /> together with materials specifications & design data for different materials and components, by making reference to test standards and qualification procedures.

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

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

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

In the case of the **fuel element**, which includes the cladding and other parts, the main design tools are **fuel performance codes** (FPC) that enable the simulation of the thermal and mechanical behaviour and evolution of the fuel element in reactor as a function of the irradiation and thermal parameters in normal operational, incidental and accidental conditions.<fnref n="44" /> To this aim, these codes solve coupled partial differential equations governing heat transfer, stresses and strains in the fuel element, the evolution of isotopes and the behaviour of various fission products in the fuel rod with boundary conditions defined by the reactor operational conditions. These equations involve material properties for the fuel and the cladding, which evolve with the residence time in reactor. Models are therefore needed to describe the very complex relationships between the evolution of these properties and the relevant parameters, especially the temperature, composition and microstructure.

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

<footnote n="42">⁴² The two main design codes are RCC-MRx developed by AFCEN and ASME Boiler and Pressure Vessel Code 2015 Edition. The R5 for high-temperature and R6 for defect assessment developed in the UK are the most used assessment procedure codes in Europe.</footnote>

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

<footnote>* Tome 1 addresses design and construction. Subsection Z with 20 technical appendices contains for instance basic design material properties, welded joints, elasto-visco plastic analysis and defect assessment.</footnote>

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

<footnote>* Tome 6 contains a collection of probationary phase rules which do not yet have sufficient feedback in the standard code. This corresponds to ASME's "Code Cases".</footnote>

<footnote n="44">⁴⁴ Fuel assemblies generally need full qualification in-pile and are not included in design codes. There is an AFCEN code for fuel assemblies, RCC-C, "Design and Construction rules for Fuel Assemblies of PWR Nuclear Power Plants", but it is restricted to PWR and it is essentially a synthesis of best practices of the French industry for fuel fabrication, with, however, very few quantitative criteria.</footnote>
fuels and/or in specific operating conditions and to demonstrate the satisfactory behaviour of fuels in all operational situations to support safety reports.

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

The main European codes for GenIV reactors are TRANSURANUS<fnref n="45" /> developed by JRC in collaboration with various institutions through Europe and the GERMINAL code<fnref n="46" />, which is part of the PLEIADES platform<fnref n="47" />, co-developed by CEA and EDF.

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

### 2.3.2. Advanced materials modelling and characterisation

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

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

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

<footnote n="45">⁴⁵ Lassmann, TRANSURANUS: a fuel rod analysis code ready for use, J. Nucl. Mater. 188 (1992) 295; Di Marcello et al., Extension of the TRANSURANUS plutonium redistribution model for fast reactor performance analysis, Nucl. Eng. Design, 248 (2012) 149.</footnote>

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

<footnote n="47">⁴⁷ Marelle, Validation of PLEIADES/ALCYONE 2.0 fuel performance code, Water Reactor Fuel Performance Meeting, Jeju, Korea, 2017.</footnote>

<footnote n="48">⁴⁸ 1/3 is considered sufficient for extrapolation, but this can also be a challenging duration/dose.</footnote>

<footnote n="49">⁴⁹ ICME-based microstructure-based description of the deformation of metals: theory and application, in: Helm, Butz, Raabe and Gumbsch, Microstructure-Based Description of the Deformation of Metals: Theory and Application. JOM 63 (2011) 26. http://www.dierk-raabe.com/icme/</footnote>
<figure><img src="image_13.png" /><figcaption>**FIGURE 5:** Schematic representation of the main elements that enter the process of generation of component design codes and fuel performance codes.</figcaption></figure>

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

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

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

<footnote n="50">⁵⁰ As an example, here is an idealised chain of models to predict radiation effects at the microscopic scale:</footnote>

<footnote>* Models to predict damage production in collision cascades: techniques of application here are electronic structure and interatomic potential calculations using molecular statics and dynamics.</footnote>

<footnote>* Atomistic models that allow mobility and stability to be studied for all radiation defects involved in the microstructural and microchemical evolution. Techniques of application are as above, but extend to atomistic kinetic Monte Carlo models.</footnote>

<footnote>* Parameters and mechanisms determined as above are the base of microstructural/microchemical models that describe how radiation damage evolves versus dose. Techniques of application here are object kinetic Monte Carlo, rate theory-based cluster dynamics, phase field, ...</footnote>
properties to creep) are generally needed.<fnref n="51" /> The modelling tools are generally costly-to-run and often use parallelized computer codes, so the use of **high performance computing** (HPC) is expected to be a crucial asset in many cases.

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

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

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

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

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

<footnote n="51">⁵¹ For example the following combination may be used:</footnote>

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

<footnote>* Transfer the above understanding in the form of local rules, to dislocation dynamics models, applicable at single-crystal level.</footnote>

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

<footnote>* Identify suitable homogenization schemes to apply continuum plasticity approaches and evaluate the mechanical behaviour at component level.</footnote>

<footnote n="52">⁵² E.g. high resolution TEM (HRTEM), or scanning TEM (STEM), coupled to energy dispersive X-ray (EDX) or electron energy loss spectroscopy (EELS), to identify and analyse defects and also measure the local chemical composition.</footnote>
<figure><img src="image_14.png" /><figcaption>**FIGURE 6:** Modelling and experimental characterisation techniques at various time and length scales involved in the ICME approach, used to elucidate the basic mechanisms of nuclear material behaviour and ageing, and their coupling.</figcaption></figure>

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

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

### 2.3.3. Development of advanced material solutions

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

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

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

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

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

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

<footnote n="53">⁵³ A positive consequence of the strictness of the qualification procedure that needs to be undergone for nuclear application, combined with the especially harsh conditions that materials face in nuclear reactors, especially GenIV systems, is that the materials solutions developed in this framework, where safety is an imperative, are likely to be of use also for other technologies where harsh conditions are expected.</footnote>
The screening is currently performed essentially in the same way as for existing materials that need to be qualified for the demonstrators, i.e. by exposure and testing (the "observe and qualify" paradigm). The only difference is that the first goal here is not to fully define the correlations and rules that will allow the design, licensing and construction of a reactor, but only to give a **first assessment of the behaviour of the candidate materials** to identify the most suitable one(s). It would be very desirable for the screening process to be done at an **accelerated pace**. Accelerating the screening of materials is very much connected with the shift towards a "design and control" paradigm. This implies addressing two aspects: (i) identifying experimental methods to **accelerate the exposure and subsequent testing** and characterisation of materials; and (ii) making use of **advanced modelling and characterization techniques to guide the development** of new materials, by using a targeted quantitative methodology rather than a trial-and-error approach that is based on the experience or intuition of the researchers involved.

In the case of nuclear materials, **charged particle irradiation** can be a suitable method to accelerate exposure, also in combination with other parameters (temperature and environmental attacks), provided that: (i) suitable characterization techniques, including for mechanical characterization, are identified, developed and, importantly, standardized or at least guided by accepted best practices and protocols; and (ii) models are developed that allow the assessment of the effect of irradiation by particles that are different from neutrons (accounting for effects of spectrum, dose-rate, limited penetration, injection of species, ...etc.). This links very much with the support offered by the modelling infratechnology (section 2.3.2), as well as with the accelerated testing envisaged for pre-normative research (section 2.3.1). Accelerated characterization would be best achieved by the creation and availability of suitable **integrated test-beds**, possibly including exposure and characterization in the same package.

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

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

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

<footnote n="54">⁵⁴ E.g. Chen, *High-Throughput Computing for Accelerated Materials Discovery*, in: *Computational Materials System Design*, Shin and Saal Editors, Springer (2018).</footnote>
<figure><img src="image_15.png" /><figcaption>FIGURE 7: Illustration of the process of development of new materials solutions and its desirable acceleration.</figcaption></figure>
materials”, which can be concisely referred to as **advanced materials**. Figure 7 illustrates the procedure for materials development and its desirable acceleration.

## 2.4. Data Management

Trends in the research sector, such as the adoption of data management policies by funding agencies<fnref n="55" />,<fnref n="56" />, publishing houses<fnref n="57" /> and partner organizations<fnref n="58" />, the emergence of data journals<fnref n="59" />,<fnref n="60" /> and frameworks for data citation<fnref n="61" />, are indicative that data management is becoming an integral component of the mainstream research process. While the specific reasons for this circumstance are varied, the underlying motivations are improved science and greater opportunities for innovation. Self-evidently, membership organizations that rely on data to achieve their objectives stand to benefit from improved data management practices. This is certainly the case for the EERA JPNM, where large volumes of materials' testing, characterisation and modelling data are generated that are of inherently high intellectual and/or commercial value.

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

needed to trace empirical curves or to validate/ calibrate models. Historically, large amounts of test and measurement data have been generated through national and international research programmes, but these data are often not available or, if available, they are not complete (for instance data for time to creep-rupture, but not the creep-curve itself). While appropriate web-enabled databases have been developed, e.g. the materials database MATDB of Online Data & Information Network of the European Commission Joint Research Centre<fnref n="62" /> or the International Fuel Performance Experiments database held by the OECD,<fnref n="63" /> often data are not made available because they are protected by confidentiality and therefore cannot be shared, or in other cases because they were not properly stored. Furthermore, in ongoing and future research programmes, a very large share of the effort, and the cost, will be dedicated to material testing and property measurement, including novel materials. Since experimental data provide the basis for design curves and rules, as well as for the development and validation/calibration of models, their availability must be guaranteed with a view to their re-assessment in the context of new models, different operating conditions, different regulatory requirements, etc. Given that such data are also expensive to generate, it should become standard practice for the data to be collected, preserved, and made readily accessible, with respect to intellectual property rights, of course. The latter issue may require agreements, e.g. in terms of embargo periods, before the data become disclosed.

With improved data management depending on so many (sometimes competing) factors, best practices are evolving that are designed to support improved data management; these include data management plans<fnref n="64" />,<fnref n="65" /> and the

<footnote n="55">⁵⁵ Data management - H2020 Online Manual. Retrieved 29 May 2018 from http://ec.europa.eu/research/participants/docs/h2020-funding-guide/cross-cutting-issues/open-access-data-management/data-management_en.htm</footnote>

<footnote n="56">⁵⁶ Funders' data plan requirements. Retrieved 29 May 2018 from http://www.dcc.ac.uk/resources/data-management-plans/funders-requirements</footnote>

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

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

<footnote n="59">⁵⁹ Data in Brief - Journal - Elsevier. Retrieved 29 May 2018 from https://www.journals.elsevier.com/data-in-brief.</footnote>

<footnote n="60">⁶⁰ Scientific Data. Retrieved 29 May 2018 from https://www.nature.com/sdata</footnote>

<footnote n="61">⁶¹ Welcome to DataCite. Retrieved 29 May 2018 from https://datacite.org</footnote>

<footnote n="62">⁶² https://odin.jrc.ec.europa.eu/odin/index.jsp</footnote>

<footnote n="63">⁶³ https://www.oecd-nea.org/science/wprs/fuel/ifpelst.html</footnote>

<footnote n="64">⁶⁴ DMPonline | Digital Curation Centre. Retrieved 29 May 2018 from http://www.dcc.ac.uk/dmponline</footnote>

<footnote n="65">⁶⁵ EUROPEAN COMMISSION Directorate-General for Research & Innovation, Guidelines on FAIR Data Management in Horizon 2020 (2016). Retrieved 29 May 2018 from http://ec.europa.eu/research/participants/data/ref/h2020/grants_manual/hi/oa_pilot/h2020-hi-oa-data-mgt_en.pdf</footnote>
FAIR (Findable, Accessible Interoperable and
Re-usable) Guiding Principles for scientific
data management and stewardship<fnref n="66" />. The FAIR
guiding principles are concerned with data
becoming findable, accessible, interoperable
and reusable, so probably the most pressing
need in the engineering materials sector is for
**data interoperability**. In this context, standards
for engineering materials data are needed to
ensure consistency of the data coming from
different laboratories and to facilitate transfer
between different information systems. **There is
therefore a urgent need to develop standard**

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

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

NUCLEAR
MATERIALS
RESEARCH
AGENDA
<figure><img src="image_16.png" /></figure>

200 nm

<figure><img src="image_17.png" /></figure>
3. NUCLEAR
MATERIALS
RESEARCH
AGENDA

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

**3.1. Structural materials**

3.1.1. Pre-normative research: qualification,
test procedures, design rules

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

design, licensing, and construction of the ESNII
demonstrators, with a view also towards the
FOAK prototypes and commercial reactors to
be deployed later.<fnref n="67" />

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

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

<footnote n="68">⁶⁸ Technology Roadmap Update for Generation IV Nuclear Energy Systems, January 2014, https://www.gen-4.org/pdf/upload/docs/application/pdf/2014-03/gif-tru2014.pdf</footnote>
codes and standards. For instance, for creep an extrapolation by a time factor 3 is consid- ered feasible, but this would require tests of 20 years duration for reliable 60 years design data. In a future low-carbon energy mix, the nuclear contribution will also need to operate in a load-following mode, whereby the reac- tor components will accumulate more load cycles, which should also be taken into ac- count in a long-term operation perspective. Although the 60-years operational life is a re- quirement for commercial deployment, and is thus a long-term need, there is a short-term R&D need to start in due time, i.e. now, in view of long-term test programmes<fnref n="69" />.

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

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

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

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

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

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

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

<footnote n="70">⁷⁰ See Technical Area 1b of the MoU between EERA and SNETP: *Development of Design Rules, qualification and mitigation strategies for reactor components in heavy liquid metal coolants*.</footnote>

<footnote n="71">⁷¹ http://www.eera-jpnm.eu/?q=jpnm&sq=nboard</footnote>
materials are austenitic steels, on which most research effort is focused, in particular 316L and 316L(N), that are planned to be used for the core of all prototypes. In the case of the GFR, Ni-base alloys are also reference materials, in particular the high-Ni alloy 800H for out-of-core components. Alloy 800H is qualified up to 750°C for the steam generator; but for the other high temperature components – turbine blades, hot gas header, intermediate heat exchangers – Ni-base alloys with higher creep rupture strength are required, e.g. Inconel 617, Haynes 230, or Hastalloy-XR.<fnref n="72" /> Some emphasis remains on gCr F/M steels which, although not used for MYRRHA and ALFRED, and only to a limited extent considered in ASTRID, remain candidate materials for the phase II of some demonstrators. For cladding the reference material is the 15-15Ti austenitic steel.

## High temperature behaviour and degradation of metals

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

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

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

* creep and fatigue damage and crack propagation: develop models that incorporate the initiation of damage and crack propagation until final failure and integrate creep and fatigue with proper damage criteria;

* thermal ageing: include thermal ageing effects in design codes and assessment procedures predicting the microstructural evolution with thermodynamic codes;

* interaction creep, creep-fatigue and thermal ageing and upscaling: develop models that combine effect of creep, fatigue and ageing for total life assessments of components, upscaling from mechanistic to polycrystal continuum models through homogenization schemes;

* representative data collection and production for 60 years design life: (i) compilation and assessment of 'historical' data, (ii) mechanical testing and microstructural analysis of materials from reactor components operated in the past, (iii) new test programmes.

## Environmental compatibility between coolant and structural materials

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

<footnote n="72">⁷² There has also been effort to develop special ODS alloys with high temperature strength and stability (MA6000).</footnote>

<footnote n="73">⁷³ For a general overview of high-temperature assessment approaches see Comprehensive Structural Integrity, Fracture of Materials from Nano to Macro Vol. 4 Cyclic Loading and Fatigue and Vol. 5 Creep and High-Temperature Failure, 2007. Elsevier-Pergamon. For nuclear applications see Structural Materials for Generation IV Nuclear Reactors, Edited by Pascal Yvon, Woodhead Publishing Series in Energy, Elsevier, 2017.</footnote>
<table><caption>TABLE 3: Main issues concerning pre-normative R&D: materials qualification, design rules, assessment & test procedures.</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="3">High temperature behaviour and degradation of metals</td><td>Creep, relaxation and cyclic deformation</td><td rowspan="3">Austenitic steels (316L, alloy 800), F/M steels (Grade91), Ni-base alloys (Inconel 617, Haynes 230, Hastalloy-XR)</td><td><strong>Experiments:</strong><br>For long-term operation: Mechanical tests of in-service material, long-term tests, accelerated tests.<br>Basic tests for model calibration: creep, low-cycle fatigue, crack propagation tests. Special emphasis on long hold times.<br>Microstructural analysis to link mechanism-based models to experiments are needed.<br><br><strong>Models:</strong><br>For creep, relaxation and cyclic deformation, emphasis on unified visco-plastic continuum models, mechanistic models for different creep mechanism and damage crack propagation fracture mechanics<br>Models need to be translated into Design Rules or Assessment Procedures.</td></tr><tr><td>Creep and creep-fatigue damage and crack propagation</td><td></td></tr><tr><td>Thermal ageing</td><td></td></tr><tr><td rowspan="4">Environmental compatibility between coolant and structural materials</td><td>Liquid Metal Corrosion and erosion (LMC)</td><td rowspan="2">Austenitic steels:316L, 15-15Ti and 15-15Ti with alumina surface protection</td><td><strong>Mechanical tests:</strong> slow-strain rate tensile; fracture, fatigue, and creep-fatigue in flowing and stagnant conditions;<br><br><strong>Corrosion tests:</strong> Erosion and corrosion (oxidation and solution tests) in flowing conditions<br><br><strong>Qualification tests</strong> (mechanical and corrosion) for 316SS and welded components.<br><br><strong>Accelerated tests</strong> to map bounding conditions.<br><br>Emphasis on long-term tests: A very careful documented control and monitoring of the test conditions (in particular oxygen control) is required for all tests.</td></tr><tr><td>Liquid Metal Embrittlement in HLM (LME)</td></tr><tr><td>Compatibility with HT He</td><td>Austenitic steels 316L and alloy 800, Ni-based alloys (Inconel 617, Haynes 230, Hastalloy-XR), as well as SiC/SiC and AFA steels.</td><td>Tests to be complemented by detailed microstructural analysis (e.g. SEM, EBSD, XRD, TEM);<br><br><strong>Engineering related approaches need to be developed</strong>. Coupling with models developed in 3.1.2</td></tr><tr><td rowspan="3">Radiation effects</td><td>Low temperature embrittlement & plastic flow localisation</td><td rowspan="2">Austenitic and F/M steels (irreplaceable & structural components)</td><td>Exposure to irradiation, also including coolant environment.<br><br>Standard mechanical test in hot cells of neutron irradiated materials (irradiated in test reactors or in-service exposed) and complementary ion/proton irradiation.</td></tr><tr><td>Long-term/low dose irradiation in environment</td><td>These tests need to be supported by irradiation models from 3.1.2</td></tr><tr><td>High dose irradiation swelling and creep</td><td>Fuel cladding materials: 15-15-Ti austenitic steel</td><td>Transfer the data into reduction factors for material properties.</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="3">Assessment of complex loadings</td><td>Non-isothermal thermo-mechanical loads</td><td rowspan="3">All</td><td>Experiments:<br>Component or specimen tests that simulate thermo-mechanical loads (e.g. thermal shocks), variable amplitude tests for fatigue; tensile high strain rate tests, multi-axial tests, e.g. cruciform specimen for biaxial loading.</td></tr><tr><td>Complex stress distributions</td><td>Modelling:<br>Finite element models of complex load cases and simplification to translate into design rule load cases.</td></tr><tr><td>Load transients and beyond design conditions</td><td>Modelling:<br>Finite element models of complex load cases and simplification to translate into design rule load cases.</td></tr><tr><td rowspan="4">Integrity and qualification of weldments and welded components</td><td>Residual stresses</td><td rowspan="4">All welds:<br>austenitic, F/M; Ni-based alloys and dissimilar metal welds</td><td>Experiments:<br>Residual stress measurements by neutron diffraction and simpler but less accurate methods such as X-ray diffraction, contour method.<br>Standard test for welded specimens (tensile, fracture toughness tests, fatigue)</td></tr><tr><td>Weld procedures</td><td>Characterization of the different regions of a weld (small punch, nano and micro indentation)<br>Mock-up test of welded component for validation.</td></tr><tr><td>Degradation modes and defect assessment</td><td>Microstructural analysis (SEM, TEM, XRD, EDX)<br>Modelling:<br>Simulation of weld process and post-weld heat treatment for residual stresses;<br>Structural integrity assessment (defect assessment crack propagation, damage) of welded specimens and components by FEM.</td></tr><tr><td>Compatibility with HLM</td><td>Translation of structural analysis assessment into Design Rules.</td></tr><tr><td rowspan="4">Sub-size and miniature specimen test standardization</td><td>Sub-sized/ miniaturised specimens for mechanical property</td><td rowspan="2">Fuel cladding material, 15-15Ti</td><td>Experiments:<br>Various fuel cladding tests (internal pressure, ring-compression, small-punch, cone mandrel) with emphasis on hot-cell tests;<br>Small-punch test for tensile and creep properties;<br>Nano-indentation, micro-pillar tests for tensile properties;</td></tr><tr><td>Thin-walled cladding tubes</td><td>Miniature specimen fracture and fatigue tests.</td></tr><tr><td>Small Punch test</td><td>All</td><td>Modelling:<br>Test to be complemented with finite element analyses, and meso-scale models (dislocation dynamics and crystal plasticity)</td></tr><tr><td>Micro-pillar tests and nano-indentation</td><td>All</td><td>Modelling:<br>Test to be complemented with finite element analyses, and meso-scale models (dislocation dynamics and crystal plasticity)</td></tr><tr><td>Component and material health monitoring</td><td>---</td><td>All</td><td>Patterns of response of material to NDE techniques as part of codification;<br>Exploration of possibility of lifetime estimation based on NDE in view of online monitoring</td></tr></tbody></table>
**characterization.** LME is a serious issue in, especially, F/M steels, that lacks a complete understanding: engineering models exist but need to account for the microstructure.

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

## Radiation effects

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

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

## Assessment of complex loadings

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

## Integrity and qualification of weldments and welded components

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

## Sub-size and miniature specimen test standardization

Tests methods and associated procedures and standards for very small volumes of material using small and miniature specimens<fnref n="74" /> become essential for:

* Neutron irradiated materials that are available in limited quantity, or in order to minimize the radioactivity;

<footnote n="74">⁷⁴ Hyde et al., Requirement for and use of miniature test specimens to provide mechanical and creep properties of materials: a review, *Int. Materials Reviews* 52 (2007) 213.</footnote>
* Very thin material layers affected by charged particle irradiation (typically a few µm);

* Service-exposed materials, i.e. tested in service during inspections;

* Thin-walled cladding tubes;

* Evaluation of local properties, for instance heat affected zone of a weld (mm scale), coatings, single crystals (µm scale), or graded functional materials.

**There is a need to review in depth the test methods in terms of strength, weaknesses and restrictions to provide material data.**

This should be followed by the development of **common procedures and standardization** of these tests. The standardization work must go through the involvement of national and international standard committees with participation of main testing laboratories and designers. These activities should concern:

* *miniaturised specimens* for standard mechanical property assessment: these are smaller versions of already used and standardized specimens;

* *thin-walled* cladding tube tests: these tests require specific equipment and standardization;

* *micro- and nano-mechanical testing methods*: these are unconventional and/or truly miniaturised specimens/techniques, e.g. nano-indentation, compression/tension/bending tests on micropillars, etc.

*Component and material health monitoring*

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

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

### 3.1.2. Advanced structural materials modelling and characterization

The issues to be addressed with an ICME approach (section 2.3.2) are necessarily strongly linked to those addressed in pre-normative research (previous section), as they correspond to the main degradation processes that affect structural materials for GenIV systems. Since the early 2000s, several European projects have addressed, from a modelling point of view, material degradation processes connected to irradiation, such as radiation hardening and embrittlement (in bainitic, austenitic and ferritic/martensitic steels), as well as, to a certain extent, plastic flow localisation and irradiation assisted stress corrosion cracking in water, namely FP6/PERFECT,<fnref n="75" /> FP7/GETMAT,<fnref n="76" /> FP7/PERFORM60,<fnref n="77" /> FP7/MatISSE.<fnref n="78" /> Some of the relevant activities currently continue in the H2020/SOTERIA<fnref n="79" /> and the H2020/M4F<fnref n="80" /> projects. Issues connected with the modelling of the high temperature behaviour of F/M steels have also been partially addressed in FP7/MATTER<fnref n="81" /> and FP7/MatISSE,<fnref n="77" /> while the

<footnote n="75">⁷⁵ FP6 IP PERFECT Project: Prediction of Irradiation Damage Effects in Reactor Components, Journal of Nuclear Materials - Special Issue, vol. 406, issue 1 (2010).</footnote>

<footnote n="76">⁷⁶ http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/46/040/46040904.pdf?r=1</footnote>

<footnote n="77">⁷⁷ http://www.sciencedirect.com/science/article/pii/S0029549311001804?via%3Dihub; http://proceedings.asmedigitalcollection.asme.org/proceeding.aspx?articleId=1628620; https://www.openaire.eu/search/project?projectId=corda_-----:190daa7e4bf2ff682991a1c9e23ccfd2</footnote>

<footnote n="78">⁷⁸ www.fp7-matisse.eu; http://www.fp7-matisse.eu/wp-content/uploads/2015/12/MatISSE-2015-GEN-IV-materials-Holmstrom.pdf</footnote>

<footnote n="79">⁷⁹ www.soteria-project.eu</footnote>

<footnote n="80">⁸⁰ http://www.h2020-m4f.eu/</footnote>

<footnote n="81">⁸¹ http://www.eera-jpnme.eu/?q=jpnme&sq=nboard; http://proceedings.asmedigitalcollection.asme.org/proceeding.aspx?articleId=1627574</footnote>
modelling of oxidation and dissolution, as well
as prolonged irradiation, in austenitic steels is
being partly addressed in H2020/GEMMA⁸².
Several modelling activities are also included
in EERA JPNM pilot projects. All these projects
provide solid bases on which to progress fur-
ther, both in terms of modelling techniques
that have been developed and results ob-
tained, as well as in terms of experimental work,
performed with a view to understanding mech-
anisms and for model validation. However, the
development of models necessarily proceeds
from simpler to more complex materials and
issues, therefore there is not only still a lot of
room for the improvement of existing models
and there are still several open issues to be
addressed, but there are in fact cases where no
defined computer-simulation based physical
modelling approach has been identified yet,
beyond classical continuum approaches.

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

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

Microstructural/microchemical evolution

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

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

**Machine learning schemes based on artificial intelligence**, e.g. artificial neural networks, could be of valuable help to address the modelling problems related with high chemical complexity, as well as to bridge between scales.<fnref n="84" />

Mechanical behaviour after and under
irradiation

**Microstructural and microchemical evolution models** and the corresponding understanding should be used as **input to move to larger scales** and address, for example, the **prediction of the subsequent plastic behaviour**, particularly in tensile tests, as a function of materials type and composition, dose received, irradiation and test temperatures, and also applied strain-rate, as summarily described in section 2.3.2. Issues to be addressed are **radiation hardening and embrittlement, plastic flow localisation** and deformation at constant load due to **irradiation creep**. Model developments are necessary to address these issues, as described in section 2.3.2, essentially aiming at describing in a **single framework** the kinetic evolution of the irradiation-induced and the dislocation microstructures, to deduce **constitutive laws for crystal plasticity**<fnref n="85" /> and devise

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

<footnote n="84">⁸⁴ Castin et al. Comp. Mater. Sci. 148 (2018) 116.</footnote>

<footnote n="85">⁸⁵ Volegov et al., Physical Mesomechanics 20 (2017) 174 https://link.springer.com/article/10.1134/S1029959917020072.</footnote>
**homogenization schemes**<sup>86</sup> for continuum mechanics approaches applied to components. The challenge remains, in particular, to introduce **chemical effects** (e.g. decorated dislocation loops or heterogeneously nucleated precipitates). In the case of (irradiation) creep, the coupling between microstructural and dislocation evolution needs to be effective because both microstructures evolve simultaneously over time.

*Mechanical behaviour at high temperature*

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

*Fracture mechanics*

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

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

*Coolant compatibility models*

The processes that govern the **interaction between solids and coolants** are crucial in the context of GenIV reactors and need to be appropriately modelled beyond current capabilities.<fnref n="87" />,<fnref n="88" />

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

<footnote n="86">⁸⁶ Allen, Comp. Sci. & Technol. 61 (2001) 2223.</footnote>

<footnote n="87">⁸⁷ Zhang et al., Review - Models of liquid metal corrosion, J. Nucl. Mater. 404 (2010) 82.</footnote>

<footnote n="88">⁸⁸ Zhang, A review of steel corrosion by liquid lead and leadbismuth, Corrosion Science 51 (2009) 1207.</footnote>
<table><caption>TABLE 4: Main issues concerning advanced structural materials modelling and characterization.</caption><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Main materials concerned</th><th>Techniques/Methods</th></tr></thead><tbody><tr><td rowspan="7">Microstructural and microchemical evolution</td><td>Formation of radiation hardening and embrittling microstructural features (low temperature)</td><td rowspan="7">F/M (possibly also ODS) and austenitic steels, ceramics</td><td rowspan="7">Molecular statics and dynamics using either electronic structure techniques or interatomic potentials for energy calculation, as well as atomistic kinetic Monte Carlo methods, as atomistic models to study stability and mobility of radiation defects, or phase stability. Object kinetic Monte Carlo, or rate theory based cluster dynamics, or to some extent phase field to develop microstructural/microchemical evolution models. Machine learning / artificial intelligence. Electron microscopy, atom probe, scattering techniques, positron annihilation, etc used in a combined way to obtain complete description. Specific computational techniques to simulate the response of experimental techniques.</td></tr><tr><td>Formation of voids and onset of swelling (high temperature/dose)</td></tr><tr><td>Radiation-induced segregation/precipitation at extended defects</td></tr><tr><td>Microstructural evolution under load in relation with irradiation creep</td></tr><tr><td>Thermal ageing and subsequent precipitation/ segregation</td></tr><tr><td>Oxide formation/ dissolution in ODS during fabrication, thermal ageing and irradiation</td><td>ODS alloys</td></tr><tr><td>Correct interpretation of microstructural examination technique signals and raw data</td><td>Any</td></tr><tr><td rowspan="3">Mechanical behaviour after and under irradiation (low temperature)</td><td>Radiation hardening: flow behaviour</td><td rowspan="3">F/M and austenitic steels</td><td rowspan="3">Dislocation dynamics models for single crystal behaviour, informed to atomistic models, to deduce constitutive laws. (Strain gradient) crystal plasticity models for aggregates. Homogenisation techniques and continuum mechanics for the component scale. Wide range of mechanical testing, from micropillars to tensile and impact, for model validation/calibration.</td></tr><tr><td>Plastic flow localisation</td></tr><tr><td>Irradiation creep</td></tr><tr><td rowspan="3">Mechanical behaviour at high temperature:</td><td>Cyclic plasticity (softening and hardening)</td><td rowspan="3">Austenitic and F/M steels</td><td rowspan="3">Different models at different length-scales are needed: from dislocation based to continuum mechanics. The techniques and methods are as in the previous case, but their use is more challenging because of the number and complexity of possible mechanisms, involving not dislocations and grain boundaries, and complex thermo-mechanical loading conditions. The up-scaling includes crystal plasticity and visco-plastic models.</td></tr><tr><td>Thermal creep</td></tr><tr><td>Creep-fatigue interaction</td></tr><tr><td rowspan="2">Fracture mechanics</td><td>Crack initiation</td><td rowspan="2">Austenitic and F/M steels</td><td rowspan="2">(Visco)-plastic models including damage criteria; classical fracture mechanics governed by crack tip parameters; local approaches where the fracture processes are explicitly modelled</td></tr><tr><td>Crack propagation</td></tr></tbody></table>
<table><thead><tr><th>Main issue</th><th>Breakdown in sub-issues</th><th>Main materials concerned</th><th>Techniques/Methods</th></tr></thead><tbody><tr><td rowspan="3">Compatibility with coolants & coolant chemistry</td><td>High temperature oxidation/corrosion</td><td>F/M and austenitic steels</td><td rowspan="3">Atomistic and thermodynamic models for corrosion and LME; phase field models; extensive microstructural characterization, especially electron microscopy</td></tr><tr><td>HLM dissolution corrosion of steels</td><td></td></tr><tr><td>Liquid metal embrittlement</td><td>F/M steels</td></tr><tr><td rowspan="2">Properties of composite/ ceramic materials depending on microstructure/ architecture</td><td>Tomography of composite and correlation with their macroscopic properties (mechanical, thermal, ...)</td><td rowspan="2">SiC/SiC, Max phases, other</td><td rowspan="2">In addition to known atomistic and microstructural/microchemical evolution models as for steels, larger scale models that take into account the architecture of these materials need to be identified and developed</td></tr><tr><td>Development of models for thermal and mechanical composite behaviour</td></tr><tr><td rowspan="2">Development of methodology to perform ion/electron irradiation experiments</td><td>Design of ion irradiation experiments and interpretation of microstructural data</td><td>Model materials, to extend to all</td><td>Essentially the same techniques as for microstructural/microchemical evolution, but specifically targeted to charged particle irradiation issues</td></tr><tr><td>Use and interpretation of micromechanical techniques (e.g. nanoindentation)</td><td>Model materials, to extend to all</td><td>Specific simulation tools combining atomistic to continuum descriptions to be developed for the simulation of micromechanical techniques</td></tr><tr><td>Other issues</td><td>Residual stresses after welding<br>Modes of deformation in steels</td><td>Austenitic and F/M steels<br>ODS steels</td><td>In principle a full suite of codes and methods, through all scales, should be deployed to simulate what happens in the welding process or under deformation in ODS steels. Neither of them exists to date.</td></tr></tbody></table>

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

### Models in support of the use of charged particle irradiation

Neutron irradiations are long and expensive experiments and the access to the relevant facilities is generally restricted. Charged particle irradiation (ions, protons, electrons, ...) is a very valuable and affordable tool for the purpose of modelling-orientated experiments and materials screening. However, transferability issues exist between charged particle and neutron irradiation environment (see also section 6.1.3).<fnref n="89" /> It is thus **essential that suitable models address the specificities of these modes of irradiation**. Difficulties need also to be overcome

concerning the post-irradiation examination, particularly the possibility of obtaining a meaningful assessment of mechanical properties after charged particle irradiation. Overall, the **goal** is to develop an established and possibly **standardized methodology**, based on **models**, to perform **charged particle irradiation and relevant PIE**, including **mechanical characterization**.

### Development of standard methodology for model validation

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

<footnote n="89">⁸⁹ Was, J. Mater. Res. 30 (2015) 1158.</footnote>
apply **microstructural characterization** techniques and analyse their outcome. Methods to **simulate** the response of materials to experimental techniques are needed. **Protocols** for the **comparison** between **simulation and experiment** need to be established, as well.

*Note on materials of interest and different levels of model development*

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

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

However:

* **The modelling of austenitic steels needs to be pushed forward,** "catching up" with the level of maturity that characterises models for F/M steels on the low-scale side.

**Ceramic materials** are often chemically and crystallographically complex systems (e.g. MAX phases), that require specific developments. In terms of atomic-level modelling, SiC and $Al_2O_3$ are probably the most studied, although the activities are globally very limited and scattered. Importantly, these materials are generally used in the form of composites, which opens different types of problems for modelling, one of them being the monitoring of the behaviour of the architecture, such as via 3D tomography, and the assessment of the role of the interfaces between e.g. fibres and matrix.

### 3.1.3. Development of advanced structural materials

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

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

* Improved austenitic steels;

* Ferritic/martensitic steels;

* SiC/SiC composites;

* Refractory metallic alloys;

* Modified surface layers for protection against corrosion;

* Other perspective materials.

For all these materials there are general issues of **fabrication processes to be developed/optimized** and/or **made less expensive**. For some of them **additive manufacturing** may be a way forward that is worth pursuing. There are also general issues of **protection against coolant attack** for those solutions that do not explicitly address this problem. Finally, for all of them different solutions need to be considered and **screening procedures** applied, which are especially costly when irradiation is involved: for some, but not for all, ion irradiation can be a way to go. In what follows, possible development paths for each class of materials are briefly listed. At the end, Table 5 (page 68) summarizes the main issues related with the different classes of advanced structural materials. More details can be found in section TA.1.1.3
of the Technical Annex, which is an extended version of the present one (next page).⁷¹

## Improved austenitic steels

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

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

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

## Ferritic/martensitic steels

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

* Loss of strength at high temperature (T>550°C), including softening under cyclic operation;<fnref n="92" />

* Neutron irradiation embrittlement at low temperature (T<350°C), including plastic flow localisation with subsequent drastic reduction of uniform elongation;<fnref n="93" />,<fnref n="92" />

* Susceptibility to liquid metal embrittlement in contact with HLM.<fnref n="94" />

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

* Improvement of high temperature mechanical behaviour;

* Improvements of ductility after low temperature irradiation, including slip localization with loss of uniform elongation;

* Improvement of the compatibility with HLM, by investigation the mechanisms responsible for corrosion and, even more importantly, liquid metal embrittlement.

<footnote n="90">⁹⁰ C. Cawthorne, E.J. Fulton Nature, 216 (1967), pp. 575-576.</footnote>

<footnote n="91">⁹¹ K. Lambrinou, et al Journal of Nuclear Materials, Volume 490, 2017, pp. 9-27.</footnote>

<footnote n="92">⁹² Raj & Vijayalakshmi, Ferritic Steels and Advanced Ferritic-Martensitic Steels, in: Comprehensive Nuclear Materials 4.03 (2012) 97.</footnote>

<footnote n="93">⁹³ Farrell, Byun and Hashimoto, J. Nucl. Mater. 335 (2004) 471.</footnote>

<footnote n="94">⁹⁴ Ersoy et al., J. Nucl. Mater. 472 (2016) 171.</footnote>
<table><caption>TABLE 5: Main issues concerning classes of advanced structural 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">Improved austenitic steels</td><td>Reduction of susceptibility to swelling</td><td>Multi-stabilized steels</td><td>Addition of elements (Ti, Si, P, Ni ...) that (a) in solid solution enhance point-defect recombination (b) create nanoprecipitates that stabilize the dislocation network</td></tr><tr><td>Improvement of compatibility with HLM</td><td>AFA steels</td><td>Addition of aluminium with double benefit of alumina-SL formation and NiAl particle strengthening</td></tr><tr><td rowspan="4">Ferritic/ martensitic steels</td><td>Improvement of high temperature behaviour ODS and CSE F/M steels</td><td>Dispersion of oxide particles by powder metallurgy (target > 700°C) or tuning of composition and TMT</td><td>ODS steel fabrication requires optimization/standardization and industrial upscale of fabrication process, and possibly development of alternative fabrication procedures.</td></tr><tr><td>Welding and relevant qualification</td><td>ODS steels</td><td>Development, optimization, standardization of welding procedures for ODS steels</td></tr><tr><td>Improvement of compatibility with HLM</td><td>FeCrAl or modified surface layers</td><td>Current F/M as well as ODS or CSE F/M can be coated or their composition modified with addition of Al (FeCrAl, incl. ODS)</td></tr><tr><td>Mitigation of low temperature irradiation embrittlement</td><td>All F/M starting from baseline steel</td><td>(Strategy may be identified via modelling, see section 3.1.2)</td></tr><tr><td rowspan="5">Development and qualification of SiC/SiC cladding</td><td>Degradation under neutron irradiation, especially thermal conductivity</td><td rowspan="5">SiC<sub>t</sub>/SiC tube with metallic liner ("sandwich" tubes) or ceramic layer</td><td>Neutron irradiations & PIE by SEM & TEM (especially PyC layer), thermal conductivity and hermeticity measurements (thermal characterisation methods by steady state or transient measurements)</td></tr><tr><td>Corrosion by oxidation, especially in moist oxygen deficient, high temperature conditions (in He, but also HLM Na)</td><td>Corrosion testing in service environment under careful control of the chemistry and varying the concentration of the contaminating species expected.<br>Post-test microstructural characterization<br>Thermodynamic modelling of the SiC environment/coolant system.<br>Testing after application of protective SL</td></tr><tr><td>Joining and relevant qualification</td><td>Screening of materials/processes/designs assisted by thermodynamic modelling<br>Realization of sample joints and prototypes<br>Mechanical testing and performance assessment in the operative conditions<br>Post-test examination by SEM, TEM, XRD</td></tr><tr><td>Standardization of mechanical tests and analysis for tubular structures of continuous fiber ceramic composites</td><td>Standardization of testing procedures<br>Development of standards & design rules</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="3">Refractory alloys</td><td>Development of processing routes</td><td rowspan="3">ODS Mo, V-Cr-Ti alloys</td><td>Alloy design & synthesis, regain processing skills. Development of welding procedures.</td></tr><tr><td>Environmental issues (oxygen corrosion and interstitial impurity embrittlement)</td><td>Protection with coatings and relevant qualification.</td></tr><tr><td>Characterise and improve behaviour towards low temperature radiation embrittlement</td><td>Improve DBTT shift under irradiation through composition and processing changes. Verification through neutron irradiation at relevant DPA doses, post irradiation characterization by TEM, SEM, mechanical testing.</td></tr><tr><td rowspan="5">Surface layers for protection against coolant attack</td><td>Optimisation of SL composition and deposition method</td><td>GESA surface modification on steels</td><td>Assessment of the long term behaviour of the system by thermal aging tests & thermodynamic modelling of the coating substrate system, with a view to progressive optimization.</td></tr><tr><td>Evaluation of effectiveness of protection against prolonges exposure to HLM, especially high speed erosion</td><td rowspan="4">GESA and PLD alumina protected austenitic steels, F/M steels, refractory alloys</td><td>Corrosion testing in service and off-normal conditions. Post-test microstructural characterization (XRD, SEM, TEM, Raman, etc.) to identify the compounds formed.</td></tr><tr><td>Evaluation of performance under irradiation</td><td>Ion and neutron irradiations & post-irradiation microstructural examination.</td></tr><tr><td>Mechanical compatibility with the substrate & thermal aging.</td><td>Mechanical testing under relevant conditions: fracture toughness, cyclic loads (thermo-mechanical fatigue, high cycle fatigue, creep and creep rupture)</td></tr><tr><td>Welding of steels with modified SL</td><td>Analysis of welded samples under above conditions.</td></tr><tr><td rowspan="2">Prospective materials</td><td>High entropy alloys</td><td></td><td>Explore different combinations of elements, test them from the high-temperature mechanical, corrosion resistance and irradiation behaviour viewpoint.</td></tr><tr><td>MAX phases</td><td></td><td>Explore different compositions (solid solutions) and microstructures to optimize the material for the specific application. Perform relevant qualification.</td></tr></tbody></table>

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

radiation resistance,<fnref n="95" /> and (ii) **creep-strength enhancement (CSE)**, through composition and thermomechanical treatment optimization, leading to optimised carbide distributions and microstructures, including the assessment of

<footnote n="95">⁹⁵ P. Dubuisson, Journal of Nuclear Materials, Volume 428, Issues 1-3, September 2012, pp. 6-12.</footnote>