# Focus on plasma-facing materials in nuclear fusion reactors

Dwaipayan Dasgupta, Elodie Bernard, Haishan Zhou, Robert D Kolasinski,

Brian D Wirth, Dimitrios Maroudas

► To cite this version:

Dwaipayan Dasgupta, Elodie Bernard, Haishan Zhou, Robert D Kolasinski, Brian D Wirth, et al. Focus on plasma-facing materials in nuclear fusion reactors. Materials Research Express, 2024, 11 (4), pp.040201. 10.1088/2053-1591/ad36b5. hal-04593951

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# Focus on plasma-facing materials in nuclear fusion reactors

Dwaipayan Dasgupta<fnref n="1" />,* , Elodie Bernard<fnref n="2" /> ID, Haishan Zhou<fnref n="3" /> ID, Robert D Kolasinski<fnref n="4" /> ID, Brian D Wirth<fnref n="1" />,<fnref n="5" /> ID and Dimitrios Maroudas<fnref n="6" /> ID

E-mail: ddasgupta@utk.edu, elodie.bernard@cea.fr, haishanzhou@ipp.ac.cn, rkolasi@sandia.gov, bdwirth@utk.edu and maroudas@ecs.umass.edu

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16 April 2024

Original content from this work may be used under the terms of the Creative Commons Attribution 4.0 licence.

Any further distribution of this work must maintain attribution to the author(s) and the title of the work, journal citation and DOI.

<figure><img src="image_2.png" /></figure>

## Abstract

Fusion energy is a promising, safe, and reliable green energy solution to the increasing energy demand. However, there are several materials challenges that need to be overcome to increase the technical readiness to a level that enables a fusion pilot plant on the grid. This focus issue aims to identify and address a set of such key impediments for realizing deuterium-tritium (D–T) fusion power in a tokamak reactor and highlight the most recent progress on those research frontiers. The main emphasis of this collection is on materials development challenges resulting from helium irradiation, neutron-induced degradation, thermomechanical loading, and the corrosive environment faced by the divertor and first-wall materials, commonly known as plasma-facing components, and blanket systems for tokamak fusion reactors.

Fusion energy is a promising, safe, and reliable base-load solution to two of the major conflicting concerns of the 21st century: increasing energy availability to meet increased projected demand and reducing the carbon footprint consistent with environmental considerations. However, challenges in integrating materials science and plasma physics into fusion technologies have long delayed the cherished goal of building a reactor to harness the immense potential of nuclear fusion. In the past two decades, significant scientific [1] and technological progress has been made to overcome some of the impediments to realizing the promise of nuclear fusion; because of such progress, according to the recent IAEA World Fusion Outlook 2023 [2], over 130 experimental fusion devices are operating, while several demonstration fusion power plants are currently in their design stage. Most operational fusion facilities are magnetic confinement devices in a tokamak design using deuterium–deuterium (D–D) fuel; going forward, considering energy efficiency and operational challenges, deuterium–tritium (D–T) fuel will be the most common choice for fusion power plants. The development of commercially viable, controlled fusion power still faces a host of engineering challenges, which include long-lasting confinement of the hot plasma, with a plasma core temperature exceeding 100 million degrees Celsius, inside the reactor chamber; creating and securing an adequate tritium supply to fuel a D–T fusion reactor and closing the fuel cycle to sustain it; the design and demonstration of materials for the reactor plasma-facing components (PFCs) which are able to withstand the extreme reactor operating conditions and the development of strategies to minimize detrimental effects on reactor materials by fusion reaction by-products; and extracting and harnessing the energy released by the fusion reactions. Considering the future potential of fusion, along with the challenges to realizing its implementation, the United States National Academy of Engineering has ranked the quest for fusion as one of the major grand challenges for engineering in the 21st century [3]. This *Materials Research Express* (MRX) article collection focuses on the aforementioned critical issues, primarily related to materials science and engineering.

<footnote n="1">¹ Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996, United States of America</footnote>

<footnote n="2">² Institute for Magnetic Fusion Research (IRFM), CEA, 13108 St Paul lez Durance, France</footnote>

<footnote n="3">³ Institute of Plasma Physics, HFIPS, Chinese Academy of Sciences, Hefei, 230031, People's Republic of China</footnote>

<footnote n="4">⁴ Sandia National Laboratories, Plasma & Reacting Flow Science Department, Livermore, CA 94550, United States of America</footnote>

<footnote n="5">⁵ Fusion Energy Division, Oak Ridge National Laboratory, Oak Ridge, TN 37830, United States of America</footnote>

<footnote n="6">⁶ Department of Chemical Engineering, University of Massachusetts, Amherst, MA 01003, United States of America</footnote>

<footnote>* Author to whom any correspondence should be addressed.</footnote>
<figure><img src="image_2.png" /></figure>

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16 April 2024

Original content from this work may be used under the terms of the Creative Commons Attribution 4.0 licence.

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EDITORIAL

# Focus on plasma-facing materials in nuclear fusion reactors

Dwaipayan Dasgupta<fnref n="1" />,* , Elodie Bernard<fnref n="2" /> ID, Haishan Zhou<fnref n="3" /> ID, Robert D Kolasinski<fnref n="4" /> ID, Brian D Wirth<fnref n="1" />,<fnref n="5" /> ID and Dimitrios Maroudas<fnref n="6" /> ID

E-mail: ddasgupta@utk.edu, elodie.bernard@cea.fr, haishanzhou@ipp.ac.cn, rkolasi@sandia.gov, bdwirth@utk.edu and maroudas@ecs.umass.edu

## Abstract

Fusion energy is a promising, safe, and reliable green energy solution to the increasing energy demand. However, there are several materials challenges that need to be overcome to increase the technical readiness to a level that enables a fusion pilot plant on the grid. This focus issue aims to identify and address a set of such key impediments for realizing deuterium-tritium (D–T) fusion power in a tokamak reactor and highlight the most recent progress on those research frontiers. The main emphasis of this collection is on materials development challenges resulting from helium irradiation, neutron-induced degradation, thermomechanical loading, and the corrosive environment faced by the divertor and first-wall materials, commonly known as plasma-facing components, and blanket systems for tokamak fusion reactors.

Fusion energy is a promising, safe, and reliable base-load solution to two of the major conflicting concerns of the 21st century: increasing energy availability to meet increased projected demand and reducing the carbon footprint consistent with environmental considerations. However, challenges in integrating materials science and plasma physics into fusion technologies have long delayed the cherished goal of building a reactor to harness the immense potential of nuclear fusion. In the past two decades, significant scientific [1] and technological progress has been made to overcome some of the impediments to realizing the promise of nuclear fusion; because of such progress, according to the recent IAEA World Fusion Outlook 2023 [2], over 130 experimental fusion devices are operating, while several demonstration fusion power plants are currently in their design stage. Most operational fusion facilities are magnetic confinement devices in a tokamak design using deuterium–deuterium (D–D) fuel; going forward, considering energy efficiency and operational challenges, deuterium–tritium (D–T) fuel will be the most common choice for fusion power plants. The development of commercially viable, controlled fusion power still faces a host of engineering challenges, which include long-lasting confinement of the hot plasma, with a plasma core temperature exceeding 100 million degrees Celsius, inside the reactor chamber; creating and securing an adequate tritium supply to fuel a D–T fusion reactor and closing the fuel cycle to sustain it; the design and demonstration of materials for the reactor plasma-facing components (PFCs) which are able to withstand the extreme reactor operating conditions and the development of strategies to minimize detrimental effects on reactor materials by fusion reaction by-products; and extracting and harnessing the energy released by the fusion reactions. Considering the future potential of fusion, along with the challenges to realizing its implementation, the United States National Academy of Engineering has ranked the quest for fusion as one of the major grand challenges for engineering in the 21st century [3]. This *Materials Research Express* (MRX) article collection focuses on the aforementioned critical issues, primarily related to materials science and engineering.

<footnote n="1">¹ Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996, United States of America</footnote>

<footnote n="2">² Institute for Magnetic Fusion Research (IRFM), CEA, 13108 St Paul lez Durance, France</footnote>

<footnote n="3">³ Institute of Plasma Physics, HFIPS, Chinese Academy of Sciences, Hefei, 230031, People's Republic of China</footnote>

<footnote n="4">⁴ Sandia National Laboratories, Plasma & Reacting Flow Science Department, Livermore, CA 94550, United States of America</footnote>

<footnote n="5">⁵ Fusion Energy Division, Oak Ridge National Laboratory, Oak Ridge, TN 37830, United States of America</footnote>

<footnote n="6">⁶ Department of Chemical Engineering, University of Massachusetts, Amherst, MA 01003, United States of America</footnote>

<footnote>* Author to whom any correspondence should be addressed.</footnote>
In the D–T fusion reaction, deuterium and tritium within the plasma undergo nuclear fusion which results in the production of a high-energy helium atom (alpha particle) and a 14.1 MeV neutron per reaction. In a magnetic confinement device, where the plasma is confined using strong magnetic fields, the high-energy neutron escapes from the plasma core and directly penetrates the PFCs and blanket materials surrounding the plasma core, causing significant degradation of the structural materials. The divertor and first-wall armor materials, commonly known as plasma-facing components (PFCs), act as the interface between the plasma boundary and the remaining reactor components. The ‘breeder blanket’ is the layer just beyond the first wall surrounding the reactor core. The function of this blanket is multi-fold, primarily for tritium production from neutron reactions with lithium, which is a crucial step since tritium is scarce [4]. Additionally, the blanket must function to translate the kinetic energy of the neutrons to heat, which can be extracted and converted to electricity, and the blanket can also assist in shielding the magnets from the 14 MeV-peaked neutron flux. On the other hand, to maintain a constant fusion power during operation, helium ash generated as a nucleus (alpha) at extremely high kinetic energies (3.5 MeV per alpha particle) needs to be constantly removed from the reactor core and replaced with additional D–T fuel. Typically, the helium nucleus is swept up by the magnetic field and undergoes energy loss before extraction as a neutral gas particle at near room temperature in a region outside of the fusion plasma, called divertor, which acts as the exhaust system of the fusion reactor. Tungsten is the plasma-facing material of choice for ITER (International Thermonuclear Experimental Reactor), currently under construction in southern France [5]. Due to its low hydrogen solubility, low sputtering yield, and high melting point, tungsten is considered a suitable candidate for withstanding the extreme fusion environment. However, under low-energy (10–100 eV) but high-flux helium implantation tungsten undergoes major surface modifications with the observed formation of dislocation loops, bubbles, or tungsten nanotendrils (so-called ‘fuzz’) in the crystal [6]. If the same inert gas (i.e., helium which also is insoluble in metals) at intermediate (> 1 keV) or high (~1 MeV) energy impinges into the inner wall, then it can produce helium bubbles and cavities or cause blistering and spallation, which have detrimental effects on the operational efficiency of the reactor and the properties of the plasma-facing materials.

In addition to ejection of high-Z tungsten into the plasma core, which causes rapid cooling of the plasma, generation of brittle ‘fuzz’ under helium-plasma exposure poses other significant challenges including tritium inventory management. Helium bubbles that form close to the plasma-exposed surface trap hydrogenic species. Retained helium and hydrogenic species, in any form (blisters, in ‘fuzz’, and/or bubbles) lead to a significant decrease in the thermal conductivity and mechanical strength of tungsten, which is further accelerated under neutron irradiation. In a reactor environment, the feedback effect from materials into the plasma such as fuel recycling would decrease plasma confinement and increase energy loss; an electric sheath is formed above the material surface due to the neutralization of plasma particles, which further accelerates ions toward the material surface and increases materials damage. Furthermore, under burning plasma conditions, PFCs are expected to experience a steady thermal load on the order of 10 MW m⁻² which can further increase to ~1 GW m⁻² during transient plasma conditions such as type I Edge-Localized Modes (ELMs). Plasma-material interactions under the harsh conditions in a fusion reactor pose arguably the single greatest materials engineering challenge, as well as a challenge for designing in-plasma sensors and diagnostics to monitor the physical conditions in the plasma.

This *MRX* focus issue consists of a collection of 3 topical review articles and 12 contributed original research articles, and provides an excellent representative snapshot of the ongoing effort in overcoming a set of such key impediments for realizing deuterium-tritium (D–T) fusion power in a tokamak reactor. This focus issue highlights the most recent progress on those research frontiers, by placing emphasis on materials development challenges resulting from helium irradiation, neutron-induced degradation, thermomechanical loading, and the corrosive environment faced by PFCs and blanket systems for tokamak fusion reactors. Additional emphasis is placed on examining the characterization techniques for plasma-surface interactions toward advancing both *in situ* and *in operando* diagnostic tools.

The central thrust of materials research on PFCs is to understand fusion environment-induced damage mechanisms over a long operating period and use that knowledge to develop improved materials and design strategies that mitigate deleterious radiation effects, providing improvements in safety and lifetime, as well as increasing efficiency and performance of materials and components. Due to a lack of experimental testing facilities that replicate the extreme operating environment of the fusion reactor, computational materials modeling is the mainstay of this research field. Research advances on plasma-facing materials in nuclear fusion reactors based on both computer simulation findings and experimental studies are covered in this focus collection. While the materials science community has a long history of conducting fundamental and applied research on plasma-materials interactions [6–8], this collection focuses on the same type of physics with respect to a nuclear fusion environment. The topics covered are unique to *'Plasma-Facing Materials in Nuclear Fusion Reactors'* such as: deuterium and tritium retention in PFCs [9–13]; fundamental processes at the plasma-surface interface [10, 14–20]; evolution of structure and properties under fusion-reactor-relevant heat loads [21]; material degradation under ion exposure [15, 16, 19]; material degradation under neutron irradiation [9, 21];
In the D–T fusion reaction, deuterium and tritium within the plasma undergo nuclear fusion which results in the production of a high-energy helium atom (alpha particle) and a 14.1 MeV neutron per reaction. In a magnetic confinement device, where the plasma is confined using strong magnetic fields, the high-energy neutron escapes from the plasma core and directly penetrates the PFCs and blanket materials surrounding the plasma core, causing significant degradation of the structural materials. The divertor and first-wall armor materials, commonly known as plasma-facing components (PFCs), act as the interface between the plasma boundary and the remaining reactor components. The ‘breeder blanket’ is the layer just beyond the first wall surrounding the reactor core. The function of this blanket is multi-fold, primarily for tritium production from neutron reactions with lithium, which is a crucial step since tritium is scarce [4]. Additionally, the blanket must function to translate the kinetic energy of the neutrons to heat, which can be extracted and converted to electricity, and the blanket can also assist in shielding the magnets from the 14 MeV-peaked neutron flux. On the other hand, to maintain a constant fusion power during operation, helium ash generated as a nucleus (alpha) at extremely high kinetic energies (3.5 MeV per alpha particle) needs to be constantly removed from the reactor core and replaced with additional D–T fuel. Typically, the helium nucleus is swept up by the magnetic field and undergoes energy loss before extraction as a neutral gas particle at near room temperature in a region outside of the fusion plasma, called divertor, which acts as the exhaust system of the fusion reactor. Tungsten is the plasma-facing material of choice for ITER (International Thermonuclear Experimental Reactor), currently under construction in southern France [5]. Due to its low hydrogen solubility, low sputtering yield, and high melting point, tungsten is considered a suitable candidate for withstanding the extreme fusion environment. However, under low-energy (10–100 eV) but high-flux helium implantation tungsten undergoes major surface modifications with the observed formation of dislocation loops, bubbles, or tungsten nanotendrils (so-called ‘fuzz’) in the crystal [6]. If the same inert gas (i.e., helium which also is insoluble in metals) at intermediate (> 1 keV) or high (~1 MeV) energy impinges into the inner wall, then it can produce helium bubbles and cavities or cause blistering and spallation, which have detrimental effects on the operational efficiency of the reactor and the properties of the plasma-facing materials.

In addition to ejection of high-Z tungsten into the plasma core, which causes rapid cooling of the plasma, generation of brittle ‘fuzz’ under helium-plasma exposure poses other significant challenges including tritium inventory management. Helium bubbles that form close to the plasma-exposed surface trap hydrogenic species. Retained helium and hydrogenic species, in any form (blisters, in ‘fuzz’, and/or bubbles) lead to a significant decrease in the thermal conductivity and mechanical strength of tungsten, which is further accelerated under neutron irradiation. In a reactor environment, the feedback effect from materials into the plasma such as fuel recycling would decrease plasma confinement and increase energy loss; an electric sheath is formed above the material surface due to the neutralization of plasma particles, which further accelerates ions toward the material surface and increases materials damage. Furthermore, under burning plasma conditions, PFCs are expected to experience a steady thermal load on the order of 10 MW m⁻² which can further increase to ~1 GW m⁻² during transient plasma conditions such as type I Edge-Localized Modes (ELMs). Plasma-material interactions under the harsh conditions in a fusion reactor pose arguably the single greatest materials engineering challenge, as well as a challenge for designing in-plasma sensors and diagnostics to monitor the physical conditions in the plasma.

This *MRX* focus issue consists of a collection of 3 topical review articles and 12 contributed original research articles, and provides an excellent representative snapshot of the ongoing effort in overcoming a set of such key impediments for realizing deuterium-tritium (D–T) fusion power in a tokamak reactor. This focus issue highlights the most recent progress on those research frontiers, by placing emphasis on materials development challenges resulting from helium irradiation, neutron-induced degradation, thermomechanical loading, and the corrosive environment faced by PFCs and blanket systems for tokamak fusion reactors. Additional emphasis is placed on examining the characterization techniques for plasma-surface interactions toward advancing both *in situ* and *in operando* diagnostic tools.

The central thrust of materials research on PFCs is to understand fusion environment-induced damage mechanisms over a long operating period and use that knowledge to develop improved materials and design strategies that mitigate deleterious radiation effects, providing improvements in safety and lifetime, as well as increasing efficiency and performance of materials and components. Due to a lack of experimental testing facilities that replicate the extreme operating environment of the fusion reactor, computational materials modeling is the mainstay of this research field. Research advances on plasma-facing materials in nuclear fusion reactors based on both computer simulation findings and experimental studies are covered in this focus collection. While the materials science community has a long history of conducting fundamental and applied research on plasma-materials interactions [6–8], this collection focuses on the same type of physics with respect to a nuclear fusion environment. The topics covered are unique to *'Plasma-Facing Materials in Nuclear Fusion Reactors'* such as: deuterium and tritium retention in PFCs [9–13]; fundamental processes at the plasma-surface interface [10, 14–20]; evolution of structure and properties under fusion-reactor-relevant heat loads [21]; material degradation under ion exposure [15, 16, 19]; material degradation under neutron irradiation [9, 21];
material erosion, migration, and deposition [14, 15, 18, 20, 22]; plasma fueling [12]; and diagnostics for plasma-materials interactions [23]. Although the details of the underlying mechanisms that govern the above phenomena remain largely unresolved, the results presented here will drive the emergence of engineering solutions to the amelioration of plasma-facing materials degradation. It is our hope and goal that this focus collection will encourage the broader materials research community to join the effort in mitigating materials challenges to accelerate the promethean effort of bringing fusion energy to the electrical power grid.

## Data availability statement

No new data were created or analysed in this study.

## ORCID iDs

Dwaipayan Dasgupta https://orcid.org/0000-0001-9405-922X

Elodie Bernard https://orcid.org/0000-0002-9291-7654

Haishan Zhou https://orcid.org/0000-0001-9773-8121

Robert D Kolasinski https://orcid.org/0000-0002-9633-1845

Brian D Wirth https://orcid.org/0000-0002-0395-0285

Dimitrios Maroudas https://orcid.org/0000-0001-9297-8839

## References

[1] Gilbert M R et al 2021 Perspectives on multiscale modelling and experiments to accelerate materials development for fusion *J. Nucl. Mater.* 554 153113

[2] IAEA World Fusion Outlook 2023 *Outlooks* (IAEA) 2023

[3] National Academy of Engineering 2008 NAE Grand *Challenges for Engineering* (The National Academies Press)

[4] Kovari M, Coleman M, Cristescu I and Smith R 2018 Tritium resources available for fusion reactors *Nucl. Fusion* 58 026010

[5] https://iter.org/newsline/-/3935 Retrieved on 14 March 2024

[6] Allain J P and Shetty A 2017 Unraveling atomic-level self-organization at the plasma-material interface *J. Phys. D: Appl. Phys.* 50 283002

[7] Eriguchi K 2017 Modeling of defect generation during plasma etching and its impact on electronic device performance—plasma-induced damage *J. Phys. D: Appl. Phys.* 50 333001

[8] Profijt H B, Potts S E, van de Sanden M C M and Kessels W M M 2011 Plasma-assisted atomic layer deposition: basics, opportunities, and challenges *J. Vac. Sci. Technol.* 29 050801

[9] Schwarz-Selinger T 2023 A critical review of experiments on deuterium retention in displacement-damaged tungsten as function of damaging dose *Mater. Res. Express* 10 102002

[10] Cusentino M A, Sikorski E L, McCarthy M J, Thompson A P and Wood M A 2023 Dynamic formation of preferentially lattice oriented, self-trapped hydrogen clusters *Mater. Res. Express* 10 106513

[11] Sun Y-W, Cheng L, Zhou H-S, Yuan Y and Lu G-H 2023 Modelling of deuterium diffusion and thermal desorption in tungsten exposed to high-flux deuterium plasma *Mater. Res. Express* 10 106522

[12] Zuo G Z, Wang Z, Sun Z, Xu W, Zhou Z T, Guan Y H, Huang M, Maingi R and Hu J S 2023 Deuterium recycling and wall retention characteristics during boron powder injection in EAST *Mater. Res. Express* 10 126402

[13] Aissou T, Ghiorghiu F, Minissale M, Angot T, De Temmerman G and Bisson R 2024 Deuterium retention and ammonia production from D-implanted 316 L stainless steel: insights for future fusion reactors *Mater. Res. Express* 11 026510

[14] Lasa A, Dasgupta D, Baldwin M J, Cusentino M A, Hatton P, Perez D, Uberuaga B P, Yang L and Wirth B D 2024 Assessment of the literature about Be-W mixed material layer formation in the fusion reactor environment *Mater. Res. Express* 11 032002

[15] Zhang R, Kajita S, Hwangbo D, Tanaka H, Feng S and Ohno N 2023 Field emission properties of nano-tendril bundles formed via helium plasma exposure with various additional impurity gases *Mater. Res. Express* 10 054002

[16] Patel O N, Dasgupta D, Wirth B D and Maroudas D 2023 Helium bubble size effects on the surface morphological response of plasma-facing tungsten *Mater. Res. Express* 10 076513

[17] Xu B-C, Li X-C, Wang J, Li Y-W, Pan X-D, Lyu Y-M, Zhou H-S and Luo G-N 2023 Interaction of 1/2⟨111⟩ interstitial dislocation loop with hydrogen and helium in tungsten: molecular dynamics simulation *Mater. Res. Express* 10 086509

[18] Abrams T et al 2023 Recent DIII-D progress toward validating models of tungsten erosion, re-deposition, and migration for application to next-step fusion devices *Mater. Res. Express* 10 126503

[19] Ito A M, Takayama A and Nakamura H 2023 Tungsten nanostructure growth by sputtering and redeposition in BCA-MD-KMC hybrid simulation *Mater. Res. Express* 10 125002

[20] Shams-Latifi J, Pitthan E, Tran T T, Kaur R and Primetzhofer D 2024 Sputter-deposition of ultra-thin film stacks from EUROFER97 and tungsten: characterisation and interaction with low-energy D and He ions *Mater. Res. Express* 11 016518

[21] Neu R et al 2023 Material and component developments for the DEMO divertor using fibre reinforcement and additive manufacturing *Mater. Res. Express* 10 116516

[22] Drobny J T and Curreli D 2023 Enabling attractive-repulsive potentials in binary-collision-approximation monte-Carlo codes for ion-surface interactions *Mater. Res. Express* 10 126513

[23] Pardanaud C et al 2023 Raman microscopy to characterize plasma-wall interaction materials: from carbon era to metallic walls *Mater. Res. Express* 10 102003
material erosion, migration, and deposition [14, 15, 18, 20, 22]; plasma fueling [12]; and diagnostics for plasma-materials interactions [23]. Although the details of the underlying mechanisms that govern the above phenomena remain largely unresolved, the results presented here will drive the emergence of engineering solutions to the amelioration of plasma-facing materials degradation. It is our hope and goal that this focus collection will encourage the broader materials research community to join the effort in mitigating materials challenges to accelerate the promethean effort of bringing fusion energy to the electrical power grid.

## Data availability statement

No new data were created or analysed in this study.

## ORCID iDs

Dwaipayan Dasgupta https://orcid.org/0000-0001-9405-922X

Elodie Bernard https://orcid.org/0000-0002-9291-7654

Haishan Zhou https://orcid.org/0000-0001-9773-8121

Robert D Kolasinski https://orcid.org/0000-0002-9633-1845

Brian D Wirth https://orcid.org/0000-0002-0395-0285

Dimitrios Maroudas https://orcid.org/0000-0001-9297-8839

## References

[1] Gilbert M R et al 2021 Perspectives on multiscale modelling and experiments to accelerate materials development for fusion *J. Nucl. Mater.* 554 153113

[2] IAEA World Fusion Outlook 2023 *Outlooks* (IAEA) 2023

[3] National Academy of Engineering 2008 NAE Grand *Challenges for Engineering* (The National Academies Press)

[4] Kovari M, Coleman M, Cristescu I and Smith R 2018 Tritium resources available for fusion reactors *Nucl. Fusion* 58 026010

[5] https://iter.org/newsline/-/3935 Retrieved on 14 March 2024

[6] Allain J P and Shetty A 2017 Unraveling atomic-level self-organization at the plasma-material interface *J. Phys. D: Appl. Phys.* 50 283002

[7] Eriguchi K 2017 Modeling of defect generation during plasma etching and its impact on electronic device performance—plasma-induced damage *J. Phys. D: Appl. Phys.* 50 333001

[8] Profijt H B, Potts S E, van de Sanden M C M and Kessels W M M 2011 Plasma-assisted atomic layer deposition: basics, opportunities, and challenges *J. Vac. Sci. Technol.* 29 050801

[9] Schwarz-Selinger T 2023 A critical review of experiments on deuterium retention in displacement-damaged tungsten as function of damaging dose *Mater. Res. Express* 10 102002

[10] Cusentino M A, Sikorski E L, McCarthy M J, Thompson A P and Wood M A 2023 Dynamic formation of preferentially lattice oriented, self-trapped hydrogen clusters *Mater. Res. Express* 10 106513

[11] Sun Y-W, Cheng L, Zhou H-S, Yuan Y and Lu G-H 2023 Modelling of deuterium diffusion and thermal desorption in tungsten exposed to high-flux deuterium plasma *Mater. Res. Express* 10 106522

[12] Zuo G Z, Wang Z, Sun Z, Xu W, Zhou Z T, Guan Y H, Huang M, Maingi R and Hu J S 2023 Deuterium recycling and wall retention characteristics during boron powder injection in EAST *Mater. Res. Express* 10 126402

[13] Aissou T, Ghiorghiu F, Minissale M, Angot T, De Temmerman G and Bisson R 2024 Deuterium retention and ammonia production from D-implanted 316 L stainless steel: insights for future fusion reactors *Mater. Res. Express* 11 026510

[14] Lasa A, Dasgupta D, Baldwin M J, Cusentino M A, Hatton P, Perez D, Uberuaga B P, Yang L and Wirth B D 2024 Assessment of the literature about Be-W mixed material layer formation in the fusion reactor environment *Mater. Res. Express* 11 032002

[15] Zhang R, Kajita S, Hwangbo D, Tanaka H, Feng S and Ohno N 2023 Field emission properties of nano-tendril bundles formed via helium plasma exposure with various additional impurity gases *Mater. Res. Express* 10 054002

[16] Patel O N, Dasgupta D, Wirth B D and Maroudas D 2023 Helium bubble size effects on the surface morphological response of plasma-facing tungsten *Mater. Res. Express* 10 076513

[17] Xu B-C, Li X-C, Wang J, Li Y-W, Pan X-D, Lyu Y-M, Zhou H-S and Luo G-N 2023 Interaction of 1/2⟨111⟩ interstitial dislocation loop with hydrogen and helium in tungsten: molecular dynamics simulation *Mater. Res. Express* 10 086509

[18] Abrams T et al 2023 Recent DIII-D progress toward validating models of tungsten erosion, re-deposition, and migration for application to next-step fusion devices *Mater. Res. Express* 10 126503

[19] Ito A M, Takayama A and Nakamura H 2023 Tungsten nanostructure growth by sputtering and redeposition in BCA-MD-KMC hybrid simulation *Mater. Res. Express* 10 125002

[20] Shams-Latifi J, Pitthan E, Tran T T, Kaur R and Primetzhofer D 2024 Sputter-deposition of ultra-thin film stacks from EUROFER97 and tungsten: characterisation and interaction with low-energy D and He ions *Mater. Res. Express* 11 016518

[21] Neu R et al 2023 Material and component developments for the DEMO divertor using fibre reinforcement and additive manufacturing *Mater. Res. Express* 10 116516

[22] Drobny J T and Curreli D 2023 Enabling attractive-repulsive potentials in binary-collision-approximation monte-Carlo codes for ion-surface interactions *Mater. Res. Express* 10 126513

[23] Pardanaud C et al 2023 Raman microscopy to characterize plasma-wall interaction materials: from carbon era to metallic walls *Mater. Res. Express* 10 102003