MoD-PMI 2025

26 May 2025, 08:40 → 28 May 2025, 13:20 Europe/Vienna

Press Room (IAEA)

MoD-PMI 2025

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Introduction

The 2025 International Workshop on Models and Data for Plasma-Material Interactions in Fusion Devices (MoD-PMI) is organized in cooperation with the International Atomic Energy Agency (IAEA) for the seventh time. The event will discuss the underlying effects and related data on interactions between the fusion plasma fuel and reactor component materials. It brings together researchers and scientists from the areas of fusion energy and materials science to review advances in modelling of processes relevant to plasma-wall interactions (PWIs) and plasma-material interactions (PMIs) in fusion devices. 

The Mod-PMI workshop aims to provide a bridge between fundamental computations and interpretation of experimental PMI data; it will address a variety of processes spanning a wide range of scales including erosion, transport and trapping of fusion fuel species in the wall, changes and evolution in the material microstructure, composition and morphology. 

The MoD-PMI 2025 workshop will continue to act as a forum between the fusion PMI data users and the data providers with an aim to advance the fruitful communication between these two communities for finding gaps and needs in PMI data and to review the recent activities and to recommend new ones.  

As the private fusion companies and enterprises have entered the fusion sector in the recent few years, the MoD-PMI 2025 welcomes participation of the private sector and promotes for further private-public partnerships in fusion knowledge exchange. 

MoD-PMI 2025 will take place at IAEA Headquarters (Vienna, Austria) and is a satellite workshop for the 20th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC-20).

 

Participants are invited to propose a presentation of their research, either as an oral contribution or as a poster. 

 

Audience

Research scientists related to nuclear fusion, at any stage of their career, who are involved in the quantitative and qualitative study, measurement, analysis, modelling, simulation or prediction of processes in plasma-material interactions in fusion devices, using or producing fundamental data on these processes. Experimental and theoretical approaches are welcomed. 

The IAEA is committed to the promotion of gender equality in all its activities and women scientists are particularly encouraged to participate in the meeting.

 

Participation and Registration

All persons wishing to participate the event must be designated by an IAEA Member State or should be a member of an organization that has been invited to attend. 

There is no registration fee.

IAEA does not cover participation costs of the workshop attendees. However, limited amount of grants may be awarded for eligible participants upon application. The grant application must be submitted together with the registration. 

Please note: deadline for grant applications was due on 7 March 2025. Grant applications submitted beyond this deadline will not be processed.

UPDATE: new deadline for registrations (without grant application) has been extended until 28 March 2025.

Registration through InTouch+ system. Please follow information provided in Participation and Registration.

Quick guide can be found here.

 

Presenters

We welcome contributions in the form of orals and posters. 

Participants wishing to give an oral presentation or a poster presentation, are requested to submit an abstract through Call for Abstracts. 

Abstract submission deadline: 7 March 2025  New deadline 28 March 2025.

Guidelines for preparing and submitting presentation slides and posters are given in Presentation Guidelines and Poster Guidelines. 

 

About the Organizer 

Atomic and Molecular Data Unit, IAEA

The IAEA's Atomic and Molecular Data (AMD) Unit has facilitated collaborative international research for the production, evaluation and recommendation of data for global fusion R&D [1]. The AMD Unit maintains both numerical and bibliographical databases for fusion [2] and has organized several international coordinated research programmes, meetings, workshops and schools to review atomic and molecular (AM) and plasma-surface interaction (PSI) as well as plasma-material interaction (PMI) research for fusion [3-5]. 

 

 

References

[1] https://amdis.iaea.org/

[2] https://amdis.iaea.org/databases/

[3] https://amdis.iaea.org/CRP/

[4] https://amdis.iaea.org/meetings/

[5] https://amdis.iaea.org/workshops/

 

 

MoD-PMI 2025 flyer and poster MoD-PMI 2025 flyer.pdf MoD-PMI 2025 poster.pdf

MoD-PMI 2025 timetable timetable_250515.pdf

Templates Template_-_Poster.pptx

Monday 26 May

Welcome

invited

1 Alternative plasma facing material for nuclear fusion reactors

The development of plasma facing materials (PFMs) able to withstand the harsh conditions (large thermal loads and radiation-induced damage) in reactors is one of the key parameters for nuclear fusion to upscale to commercial power plants both, in inertial confinement fusion (ICF) and in the magnetic confinement fusion (MCF) approaches.
The radiation environment that PFMs will face in nuclear fusion reactors operating in the two previously mentioned pathways will be distinct. These disparities will lead to different radiation-induced damage, mainly: fuzz formation in MCF and exfoliation and cracking in ICF.
In this contribution, we will discuss the radiation environment that PFMs will withstand under MCF and ICF, both in the indirect (ID) and direct (DD) drive approaches, especially focusing on the differences between them [1,2]. Subsequently, we will illustrate the peculiarities of the radiation-induced damage associated to the irradiation conditions. Then, we will discuss the need of developing superior radiation-resistant PFMs, alternative to coarse-grained W (CGW), paying special attention to the capabilities of nanostructured W. We will also talk about the need for validating and improving computer codes to interpret experimental results and to predict material behavior under those extreme conditions not achievable in experiments. Finally, we will show that, in comparison to CGW, 3D isolated W nanocolumns exhibit retardation in fuzz formation [3], a lower sputtering yield under Ar [4,5] and D [6] irradiation, as well as a flattening in their angular dependence [4].

References
[1] J. Alvarez, R. Gonzalez-Arrabal, A. Rivera, E. Del Rio, D. Garoz, E.R. Hodgson, F. Tabares, R. Vila, M. Perlado, Potential common radiation problems for components and diagnostics in future magnetic and inertial confinement fusion devices, Fusion Engineering and Design 86 (2011) 1762–1765. https://doi.org/10.1016/j.fusengdes.2011.01.080.
[2] D. Garoz, A.R. Páramo, A. Rivera, J.M. Perlado, R. González-Arrabal, Modelling the thermomechanical behaviour of the tungsten first wall in HiPER laser fusion scenarios, Nucl. Fusion 56 (2016) 126014. https://doi.org/10.1088/0029-5515/56/12/126014.
[3] W. Qin, F. Ren, R.P. Doerner, G. Wei, Y. Lv, S. Chang, M. Tang, H. Deng, C. Jiang, Y. Wang, Nanochannel structures in W enhance radiation tolerance, Acta Materialia 153 (2018) 147–155. https://doi.org/10.1016/j.actamat.2018.04.048.
[4] A. Lopez-Cazalilla, C. Cupak, M. Fellinger, F. Granberg, P.S. Szabo, A. Mutzke, K. Nordlund, F. Aumayr, R. González-Arrabal, Comparative study regarding the sputtering yield of nanocolumnar tungsten surfaces under ${\mathrm{Ar}}^{+}$ irradiation, Phys. Rev. Mater. 6 (2022) 075402. https://doi.org/10.1103/PhysRevMaterials.6.075402.
[5] C. Cupak, A. Lopez-Cazalilla, H. Biber, J. Brötzner, M. Fellinger, F. Brandstätter, P.S. Szabo, A. Mutzke, F. Granberg, K. Nordlund, R. González-Arrabal, F. Aumayr, Sputter yield reduction and fluence stability of numerically optimized nanocolumnar tungsten surfaces, Phys. Rev. Mater. 7 (2023) 065406. https://doi.org/10.1103/PhysRevMaterials.7.065406.
[6] J. Brötzner, C. Cupak, M. Fellinger, H. Biber, A. Lopez-Cazalilla, F. Granberg, F. Kporha, A. Mutzke, R. González-Arrabal, F. Aumayr, Sputtering yield reduction for nano-columnar W surfaces under D ion irradiation, Nuclear Materials and Energy 37 (2023) 101507. https://doi.org/10.1016/j.nme.2023.101507.

• Corresponding author. raquel.gonzalez.arrabal@upm.es

Speaker: Raquel Gonzalez-Arrabal (Universidad Politécnica de Madrid)

2 Experiments and modeling of hydrogen isotope inventory in damaged tungsten

In a fusion reactor, tungsten will be exposed to high heat flux, neutrons, helium ash, and tritium-containing fuel plasma. Neutron irradiation generates defects in tungsten, and the migration of these defects under irradiation leads to their clustering and annihilation. These irradiation-induced defects serve as strong trapping sites for hydrogen isotopes. Therefore, predicting the accumulation of irradiation-induced defects during reactor operation is crucial for evaluating the tritium inventory in the vacuum vessel and assessing the hazard of a loss-of-vacuum accident.
In this study, the accumulation of irradiation-induced defects and the trapping of hydrogen isotopes in tungsten were experimentally evaluated. These phenomena were also modeled using rate equations, which were combined to predict the tritium inventory in the first wall of the fusion reactor.
In this presentation, we will compare experimentally obtained results on hydrogen isotope retention and release behaviors in damaged tungsten with predictions from the model. Additionally, recent advancements in experiments and modeling related to this study will be discussed.

Speaker: Makoto Kobayashi (National Institute for Fusion Science)

contributed

3 A kinetic model based on DFT for H atoms transport at the W/Cu interface

This study explores the diffusion of hydrogen (H) across the tungsten (bcc) - copper (hcp) W(001)/Cu(11-20) interface [1-3]. It combines DFT electronic structure calculations and kinetic modeling based on diffusion coefficients and macroscopic rate equations (MRE). The copper lattice reconstructs significantly near the interface, inducing complex energetics for hydrogen atoms inserted at the interface. The many solution energies of hydrogen absorbed in these different interstitial sites have been calculated, as well as the complex hydrogen diffusion energy profile at the interface. Two diffusion paths across the interface were considered, connected by a third path in the plane of the W/Cu interface.

The activation barriers within the copper network displayed significant variations among these paths due to the reconstruction of the Cu fcc bulk into a hcp structure in the vicinity of the interface. Additionally, diffusion properties are established in perfect tungsten and copper as a reference. They are compared to diffusivity across and parallel to the W/Cu interface. Notably, diffusion parallel to the interface is shown to be lower than that within the tungsten and copper bulks across the temperature range from 260K to 1000K. Subsequently, the hydrogen diffusion perpendicular to the interface plane was modeled and analyzed according to a kinetic model we built based on 0-dimension Macroscopic Rate Equations. The complex energy pattern of the diffusion path across the interface behaves like a two steps model. Consequently, a model is proposed to model the kinetic behavior of H transport across the W/Cu interface. This reduced model reproduces the results of the full kinetic model with great accuracy. It also proves robust when the boundary solutions applied to solve the system of differential equations are modified.

[1] Y. Silva-Solis, J. Denis, E. A. Hodille, Y. Ferro, J. Phys: Condense Mater 36 (2024) 465001
[2] Y. Silva-Solis, J. Denis, E. A. Hodille, Y. Ferro, J. Phys: Condense Mater 37 (2025) 129501 - Corrigendum
[3] Y. Silva-Solis, J. Denis, E. A. Hodille, Y. Ferro, Nuclear Materials and Energy Fusion, 37 (2023) 101516

Speaker: Dr Yves FERRO (Aix-Marseille University)

10:55 break

invited

4 A foundation model for atomistic materials chemistry

The general problem of first principles force fields is to create surrogate models for quantum mechanics that yield the energy of a configuration of atoms in 3D space, as we would find them in reality for materials or molecules. Over the last decade significant advances were made in the attainable accuracy, and today we can model materials and molecules with a per-atom energy accuracy of up to 1 part in 10,000 with a speedup of over a million or more compared to the explicit quantum mechanical calculation, enabling molecular dynamics on large length and time scales. The most surprising aspect of the best model is its extreme generalisation: fitted only on small periodic crystals, it shows stable trajectories on arbitrary chemical systems, from water to nanoparticles and proteins. I will show some of the technical details behind the success of our models: equivariant many-body graph polynomials with very few and weak nonlinearities. The relationship between the architectural elements and the extreme generalisation is still a mystery. The locality of the graph structure is key to its success, as well as high body order and message passing.

Speaker: Gabor Csanyi (University of Cambridge)

5 From Grains to Gigabytes: Generating Massive Virtual Specimens for Irradiation Damage Study

In nuclear fusion environments, structural materials must withstand severe thermal loads and high-energy particle bombardment. Accurately simulating these conditions requires large-scale atomistic models capable of capturing complex grain and interface arrangements. PolyPal provides a big solution by generating massive polycrystalline specimens—scaling to tens of billions of atoms—in only a few minutes.

Existing atomic structure generators typically operate in serial and thus can handle just a few million atoms, limiting their capacity to represent realistic microstructural features. In contrast, PolyPal runs in parallel and supports flexible memory configurations, offering virtually unlimited scalability. The code also features a fully parallelized file I/O scheme for seamless integration with large-scale molecular dynamics (MD) simulations. Its domain-preserving file format removes the need for time-consuming atom sorting during MD initialization, an operation that can otherwise take hours for extremely large virtual specimens.

In tungsten and other fusion-relevant alloys, grain boundaries and solute distributions significantly influence defect migration, damage clustering, and embrittlement phenomena. PolyPal accommodates polycrystalline structures at the micrometer scale, allowing precise control over grain size distribution, crystallographic orientation, and compositional variations. This capability enables systematic exploration of how different microstructures may either enhance or diminish material performance under fusion conditions. Notably, such versatility enables more realistic multi-PKA simulations, transient thermal shock analyses, and other advanced MD studies relevant to reactor component design.

In this presentation, we will describe the computational framework underlying PolyPal, highlighting its parallel I/O architecture, performance benchmarks, and customizable microstructure settings. We will also explore its potential application to cutting-edge fusion-material research.

Speaker: Dr Younggak Shin (Yonsei University)

6 Point-Defect-Induced Metastable Phase Diagrams

Excess lattice point defects can drive phase transformations, alter phase selection, or even lead to the formation of new phases. The removal of these defects serves both as a mechanism for accommodating precipitate eigenstrain and as a driving force for semi-coherent precipitation [1].

We present a thermodynamic framework for point-defect-induced precipitations, considering both precipitate eigenstrain and the Gibbs free energy associated with point defect formation. Additionally, we introduce the concept of a volume-constrained metastable phase diagram to rationalize phase selection in model alloys subjected to irradiation.

To compute these metastable phase diagrams, we rely on lattice parameters and CALPHAD databases, which provide the Gibbs free energy of equilibrium phases. For calculating the Gibbs free energies of point defects and phase formation in cases where reliable thermodynamic data are lacking, we develop statistical physics methods. These methods extract short-range order (SRO) contributions from ab initio random energy sampling [2,3]. This approach allows us to account for local atomic arrangements, leading to more accurate predictions of defect energetics, phase stability and metastability.

References
[1] Maylise Nastar, Lisa Belkacemi, Estelle Meslin, Marie Loyer-Prost, Commun. Materials 2 (2021), 32.
[2] T. Schuler, M. Nastar, K. Li, C.C. Fu, Towards Accurate Thermodynamics from Random Energy Sampling, Acta Mat 281 (2024) 120305
[3] K. Li, T. Schuler, C. C. Fu, M. Nastar, Vacancy Formation Free Energy in Concentrated Alloys: Equilibrium vs. Random Sampling, Acta Mat. 1 (2024) 120274

Speaker: Maylise Nastar (CEA)

contributed

7 Microstructure impact on tritium retention and permeation in tungsten/oxide interface from first-principles based phase field modelling

D. Nguyen-Manh a)*, K. Starkey b), M. Christensen b),
E. Wimmer b), C. Geller b) and M..R. Gilbert a)

a) Materials Division, United Kingdom Atomic Energy Authority, Culham Campus, Abingdon, OX14 3DB, United Kingdom
b) Materials Design, 42 Avenue Verdier, 92120 Montrouge, France

Tungsten is a promising plasma-facing material for future fusion power plants owing to several favourable properties, including exceptionally high melting temperature, and excellent strength at high temperatures. However, lattice defects or “traps” created under fusion conditions in structural materials by 14.1 MeV neutrons from deuterium-tritium (D-T) fusion can increase the total HI retention in W by several orders of magnitude compared with unirradiated W. Moreover, due to its affinity for oxygen, W readily forms a natural oxide film on its surface at ambient temperature. This oxide layer, which may thicken under fusion device operating conditions on the "dark" side, or during maintenance, will critically influence HI uptake, T retention in, and release of T. Minimizing the loss of tritium through the exterior sides of a fusion device makes it imperative to have a better understanding of tritium solubility and diffusion in the sequence of defected WOx phases on W tile surfaces. A fundamental understanding of tritium permeation and retention behaviour in fusion materials components is essential for efficient tritium recovery, waste classification and performance of in-vessel materials under neutron irradiation. There is little literature investigating the detritiation of tungsten under these fusion-specific conditions.

In this work, the current results of an ongoing multiphysics investigation of the properties and microstructure of tungsten and its oxides relevant to tritium retention and transport, which is being undertaken for the UKAEA. A multi-scale materials modelling approach using first principles calculations combined with the machine learning potential developments and phase field simulation tools have been brought to bear on W oxidation and detritiation behaviour to provide mechanistic insights into key trends and input values for microstructural simulations. Diffusion of tritium in WO3 has been found to be like that in bulk W while diffusion of tritium is significantly slower in WO2 [1]. W oxide morphology, metal/oxide interface topology, and oxidation kinetics were qualitatively predicted successfully. Our work demonstrated that these characteristics of W oxidation are principally dictated by the interaction of phase nucleation, grain boundary diffusion and the emerging stress field. Tritium transport and prospects for efficient tritium recovery were shown potentially to depend on crack nucleation and growth processes in the W oxidation film, and on the dynamic metastability of a persistent WO2 -rich layer within the film. The results demonstrated a promising computational basis for informing the relative efficacy of various possible post-irradiation W tile handling scenarios bearing on tritium recovery.

[1M. Christensen et al., Materials and Energy, 38 (2024) 101611

*Corresponding author: duc.nguyen@ukaea.uk (D. Nguyen-Manh)

Speaker: Duc Nguyen-Manh (United Kingdom Atomic Energy Authority)

13:15 lunch

invited

8 Detection of defects in displacement-damaged tungsten and iron

Due to their advantageous properties, tungsten (W) and steels are the main candidates for plasma-facing and structural materials for future fusion reactors. However, exposure to 14 MeV neutrons from D-T fusion reaction in nuclear environment will introduce significant displacement damage, altering the crystal structure of the materials and affecting their physical properties. To examine displacement damage created in the W and Fe lattice (the primary constituent of steels), we employed Rutherford backscattering spectrometry in channeling configuration (RBS-C), a well-established technique for characterization of lattice disorder and defect evolution induced by irradiation. Disorder was quantified by analysing the change in the ion yield of light ions backscattered along a specific crystallographic direction [1].

MeV heavy ion irradiation was used as surrogate for the displacement damage neutrons will cause. W and Fe single crystals (SC) were irradiated at two different doses (0.02 and 0.2 dpa), at two temperatures (300 and 800 K) for W and at 300 K for Fe. Our goal was to create samples containing different amounts of defects with different nature. In W samples open volume defects (vacancies and vacancy clusters) were analysed by positron annihilation spectroscopy while interstitial-type defects were characterized by transmission electron microscopy (TEM) analysis. Dislocation lines and loops of different sizes were observed in W (111) SC, depending on the irradiation dose and temperature.

Multi-energy RBS-C spectra analysis was used to study the disorder in the materials and to obtain complementary information to the TEM analysis. For the first time for W, we employed molecular dynamics (MD) simulations of overlapping cascades as input for the RBSADEC code [2], to simulate the RBS-C spectra. These simulations showed remarkable agreement with the experiment for the lower dose sample, while discrepancies at higher doses are attributed to the formation of large dislocation structures observed by TEM, which cannot be formed in finite-size MD cells [3]. Presentation of ongoing work will include new simulation results with a larger MD cell and different MD potential for the high dose and high-temperature samples. Preliminary first results on irradiated Fe SCs will be presented and discussed.

References:
[1] Feldman et al., Academic Press, San Diego, (1982), pp. 88–116
[2] Zhang et al. Phys. Rev. E 94, 043319 (2016).
[3] Markelj et al, Acta Materialia 263 (2024) 119499

Speaker: Sabina Markelj (Jozef Stefan Institute)

9 Positron Annihilation Spectroscopy for vacancy defects studies in irradiated tungsten: Combination of modelling and experiments for vacancy size distribution and impurities interaction determination

Tungsten has been selected as the divertor material for ITER and is also considered for the first wall in fusion reactors due to its exceptional properties, including a high melting point, excellent thermal conductivity, low thermal expansion, high strength at elevated temperatures, and a high sputtering threshold energy. In these environments, tungsten will be exposed to neutron irradiation, intense Helium and Hydrogen fluxes, and extreme thermal loads, with temperatures reaching up to 1780 K. These harsh conditions could drastically affect its macroscopic properties, potentially causing embrittlement and swelling as a result of microstructural evolution.

Multi-scale modeling plays a crucial role in understanding and predicting tungsten microstructure evolution in future fusion reactors. To validate these models, experimental data are essential, particularly regarding the behavior of radiation-induced defects. Additionally, one of the key open questions is the interaction between these defects and impurities present in both the material and the plasma. Theoretical studies suggest strong interactions that could significantly influence microstructure evolution under irradiation.

This work combines modeling and experimental approaches to investigate the nature and properties of irradiation-induced defects in tungsten. On the experimental side, well-controlled irradiations are followed by advanced characterizations using Positron Annihilation Spectroscopy (PAS), including lifetime and Doppler Broadening (PAS-DB), along with Transmission Electron Microscopy (TEM) to analyze defects across different scales, from single vacancies to nanoscale cavities. On the modeling side, calculations using Two-Component Density Functional Theory (TC-DFT) provide positron annihilation characteristics—lifetimes and Doppler broadening spectra—essential for defect identification. Additionally, defect evolution is simulated as a function of irradiation conditions and impurity concentration using Molecular Dynamics, Cluster Dynamics, and Object Kinetic Monte Carlo methods.

A novel approach is introduced to estimate vacancy-type defect concentrations from experimental PAS data [1]. This method utilizes a quadratic solver calibrated with a positron trapping model [1], incorporating TC-DFT-calculated annihilation characteristics for various vacancy defects [2]. The analysis reveals small vacancy clusters that are undetectable by TEM, with concentrations significantly exceeding those of TEM-visible defects (10²⁴ m⁻³) in self-irradiated tungsten at 0.02 dpa and 500°C [1]. Moreover, the effect of the tungsten purity on the vacancy type defects distribution is shown [2,3]. Accounting for positron trapping in oxygen-vacancy complexes enables an accurate reproduction of high-temperature irradiation experimental data [1].

References
[1] Z. Hu, J. Wu, Q. Yang, F. Jomard, F. Granberg, M-F Barthe, New insight into quantifying vacancy distribution in self-ion irradiated tungsten: a combined experimental and computational study, submitted to Nano Letters
[2] Q. Yang, Z. Hu, I. Makkonen, P. Desgardin, W. Egger, M.-F. Barthe, P. Olsson, A combined experimental and theoretical study of small and large vacancy clusters in tungsten, Journal of Nuclear Materials 571 (2022) 154019. https://doi.org/10.1016/j.jnucmat.2022.154019.
[3] Z. Hu, P. Desgardin, C. Genevois, J. Joseph, B. Décamps, R. Schäublin, M.-F. Barthe, Effect of purity on the vacancy defects induced in self–irradiated tungsten: A combination of PAS and TEM, Journal of Nuclear Materials 556 (2021) 153175. https://doi.org/10.1016/j.jnucmat.2021.153175.
[4] Z. Hu, Q. Yang, F. Jomard, P. Desgardin, C. Genevois, J. Joseph, P. Olsson, T. Jourdan, M.-F. Barthe, Revealing the role of oxygen on the defect evolution of electron-irradiated tungsten: A combined experimental and simulation study, Journal of Nuclear Materials 602 (2024) 155353. https://doi.org/10.1016/j.jnucmat.2024.155353.

Speaker: Dr Marie-France Barthe (CNRS)

contributed

10 Impact of the trapping model on Tritium retention and permeation in DEMO Tungsten/Eurofer First Wall

In future fusion devices, the tritium retention and permeation in plasma facing components (PFCs) are important safety concerns. In order to predict the tritium retention and permeation during operation of a reactor, numerical modelling of tritium transport and trapping at defects can be done. This allows to know how much tritium atoms are retained in the wall and how much permeate to the cooling system of the PFCs.
In this work, we want to focus on the Water Cooled Lithium Lead design (WCLL), a specific part of the wall of DEMO which is the First Wall (FW) facing the plasma on one side and the breeding material on the other side. One element of the first wall (125x27x255 mm) is composed of a 2 mm thick tungsten (W) on top of a few cm of Eurofer actively cooled by 4 square channels. We want to investigate the impact of the different available trapping models for these 2 materials on the tritium transport and permeation. For W, we use 2 models: the first one includes only the native traps at low detrapping energies [1]. The second adds high detrapping energies that simulates the formation of vacancy clusters due to neutron impact [2]. For Eurofer, we also use 2 models: the first one contains only one low detrapping energy trapping site, as observed in several permeations studies [3, 4] and the second one adds trapping sites with high detrapping energies observed with thermal desorption spectrometry [4]. The calculation are done with FESTIM [5] in a 2D geometry (125x27 mm).
On one hand, the addition of trapping sites with high detrapping energies, whether in W or Eurofer is very similar. It makes increase the breakthrough time for the tritium permeation flux to increase by orders of magnitude going from few minutes to days (or even hundreds of days). On the other hand, it drastically increases the tritium retention in a single FW element: the retention goes from below 1 µg/m-1 to few mg/m-1.

[1] E. A. Hodille et al, J. Nucl. Mater. 467 (2015) 424-431
[2] J. Dark et al, Nucl. Fusion 64 (2024) 086026
[3] G. Esteban et al, J. Nucl. Materials 367-370 (2007) 473-477
[4] F. Montupet-Leblond et al, Nucl. Mater. Energ. 29 (2021) 101062
[5] R. Delaporte-Mathurin et al, Inter. J. Hydrogen Energ. (2024) 786-802

Speaker: Etienne Hodille (CEA-IRFM, F-13108 Saint Paul Lez Durance, France)

15:50 break

invited

11 Adsorption and in-diffusion of Hydrogen at metal and metal oxide surfaces

Hydrogen adsorption and absorption at solid surfaces is a prototype surface reaction, where fundamental concepts of gas-surface interaction are embodied. As such, hydrogen interaction with solid surfaces has been extensively studied in the surface science community [1]. To investigate the behavior of hydrogen at surfaces, our group has developed nuclear reaction analysis (NRA), which allows us to quantitatively measure the H depth distribution and H location in materials [2], and resonance-enhanced multiphoton ionization combined with atomic/molecular hydrogen beams, which allows us to detect the internal quantum states of hydrogen including the spin and ro-vibrational motion [3]. With these techniques, we have studied molecular physisorption, dissociative adsorption, and diffusion into the bulk of hydrogen on various metals such as Pd, V, Fe and Ti and metal oxides such as TiO2, CeO2 and perovskite oxides.

Palladium and Vanadium are typical H-absorbing metals. We demonstrate the hydrogen diffusion across the Pd surface can be controlled by molecular cap and surface alloy formation [4]. By applying low-energy hydrogen ions, furthermore, metastable hydride states were formed in the near-surface region of Pd and V films. The resistance of the films was found to change as the films relax from the metastable states to stable states. By measuring the time evolution of the film resistance, the diffusion of H in the films were analyzed, which revealed a crossover from a classical thermal regime to a quantum regime [5]. Even if the substrate does not absorb hydrogen exothermically, hydrogen ions penetrate the surface and migrate into the near-surface region causing heavy H doping. The depth distribution measured by NRA was consistent with SRIM simulations. As observed by photoemission spectroscopy and transport measurements, it is shown that hydrogen acts as an electron donor for most metal oxides including TiO2 and SrTiO3 [6]. From the depth profile of H, the diffusion coefficient in TiO2 is analyzed.

1. K. Fukutani et al., Chem. Rec. 17, 233 (2017).
2. M. Wilde and K. Fukutani, Surf. Sci. Rep. 69 (2014) 196; T. Ozawa et al., Nat. Commun. 15, 9558 (2024).
3. H. Ueta et al., Front. Chem. 11, 1258035 (2023); Y. Nagaya et al., J. Chem. Phys. 155, 194201 (2021).
4. S. Ogura et al., J. Phys. Chem. C 117 (2013) 9366; K. Namba et al., PNAS 115, 7896 (2018).
5. T. Ozawa et al., J. Phys. Chem. Solids 185, 111741 (2024).
6. Y. Ohashi et al., J. Phys. Chem. C 123, 10319 (2019); N. Nagatsuka et al., J. Chem. Phys. 152, 074708 (2020); G.C. Lim et al., J. Am. Chem. Soc. 146, 32013 (2024).

Speaker: Katsuyuki Fukutani (University of Tokyo)

12 Interface-induced enhanced deuterium plasma-driven permeation in chemical vapor deposition tungsten-copper composite

With tungsten (W)-copper (Cu) bonding potentially used in the plasma-facing components (PFCs) in fusion devices, hydrogen isotope (HI) transport through the W/Cu interface is a key concern for tritium self-sustainment and operation safety.
To investigate HI permeation through W/Cu interface, a series of low-energy deuterium (D) plasma-driven permeation (PDP) experiments were performed on chemical vapor deposition tungsten (CVD-W)/Cu composite, bare CVD-W, and bare Cu, across a temperature range of ~600 K-800 K. The effective D diffusion coefficient of CVD-W was found to be higher than that of rolled W and forged W, likely due to the grain boundary as a high diffusivity path of D diffusion. Under the identical experimental conditions, an unexpected result is found that the steady state permeation flux in CVD-W/Cu was higher than that in bare Cu, with 3.1E18 m-2 s-1 in CVD-W/Cu and 4.1E17 m-2 s-1 in Cu at 741 K. And the time required in CVD-W/Cu to reach steady state exceeded the sum of time required for CVD-W and Cu individually. Rate equation simulations suggest that a high D concentration segment with a low HI solution energy of 0.65 eV is necessary to replicate the high permeation flux in CVD-W/Cu. The thickness of the segment is in line with the experimental observation of the W-Cu mixture region near the CVD-W/Cu interface. And density functional theory calculation confirms that Cu in W could reduce the HI solution energy in W. Furthermore, an analytical solution for the steady state permeation flux in a generalized three-layer composite is derived from a modified analytical equation for the fast evaluation of permeation flux.
This work provides experimental data evaluating HI transportation in W/Cu composite which have been rarely reported before. The measured permeation flux and the implication of the high HI concentration in W/Cu mixture could provide a better understanding of the tritium transportation through PFC and the bonding strength between W and Cu during tokamak operation.

Speaker: Long Cheng (Beihang University)

Tuesday 27 May

invited

13 Quantum-accurate large-scale atomistic simulation of fusion materials with LAMMPS and FitSNAP

Molecular dynamics (MD) is a powerful materials simulation approach whose accuracy is limited by the interatomic potential (IAP). The quest for improved accuracy has resulted in a decades-long growth in the complexity of IAPs, many of which are implemented in the LAMMPS MD code.[1] Traditional physics-based IAPs are now being rapidly supplanted by machine-learning potentials (MLIAPs). The SNAP (Spectral Neighbor Analysis Potential) [2] approach is an early example of this, but new improved MLIAP approaches continue to emerge each year. The FitSNAP software[3], tightly integrated with LAMMPS, provides an automated methodology for generating accurate and robust application-specific MLIAPs, including support for Atomic Cluster Expansion (ACE) descriptors, and the PyTorch and JAX neural network libraries. Each MLIAP is trained on a large set of quantum electronic structure calculations of energy, force, and stress for many small configurations of atoms. The resultant potentials enable high-fidelity large-scale MD simulations of diverse materials, yielding insight into their behavior on lengthscales and timescales unreachable by other methods. The relatively large computational cost of using MLIAPs is offset by combining LAMMPS' spatial parallel algorithms with Kokkos-based hierarchical multithreading, enabling the efficient use of Exa-scale CPU and GPU platforms, allowing large-scale production simulations at speeds approaching 30 ns/day with millions to billions of atoms. These capabilities have been used by myself and collaborators to study diverse materials systems including shock compression of diamond, free expansion of molten aluminum, the magnetic/structural phase transition in shocked iron, and plasma-exposed materials for fusion energy.
[1] Thompson et al., Comp. Phys. Comm., 271:108171, 2022. DOI 10.1016/j.cpc.2021.108171
[2] Thompson et al., J. Comp. Phys., 285:316, 2015. DOI 10.1016/j.jcp.2014.12.018
[3] Rohskopf et al., Journal of Open Source Software, 8: 5118, 2023. DOI 10.21105/joss.05118

Speaker: Aidan Thompson (Sandia National Laboratories)

14 Multiscale modeling of diffusion phenomena in nuclear fuels: from the atomic scale to phase-field polycristalline microstructure evolution simulations

Nuclear materials are subjected to extreme environments involving high-flux irradiation, strong temperatures gradients, and severe mechanical stresses. These conditions induce complex macroscopic behavior and property evolution, originating from microstructural phenomena and below, with diffusion playing a crucial role. Understanding and predicting this evolution requires multiscale modeling approaches that link atomic-level processes to continuum-scale behavior.

We focus in this work on fission gas behavior in polycrystalline uranium dioxide, integrating thermomechanical evolution to improve our understanding on fission gas release, swelling, and high-burnup restructuring. A multiphase, multifield phase-field approach simulates grain boundary and gas bubble evolution under temperature and stress gradients, with irradiation-induced defect and gas source terms. The phase-field model is parameterized using atomic-scale calculations of free energy and diffusion data.

We obtain atomic-scale diffusion coefficients through direct simulations of diffusion trajectories or via mean-field approaches, and explore diffusion in multicomponent solutions, such as mixed actinide oxides or high-entropy alloys, using a generative machine-learning approach to optimize computational time [1]. This multiscale methodology can be relevant for fusion materials, where irradiation-induced microstructural heterogeneities impact material properties and behavior. Understanding and predicting these phenomena by integrating relevant physical data and behavioral laws into higher-scale models can improve the physical reliability and predictive power of macro-scale simulations, contributing to the development of robust and safe nuclear materials.

References:
[1] M. Karcz et al., “Targeting the partition function of chemically disordered materials with a generative approach based on inverse variational autoencoders”, arXiv:2408.14928 (2024).

Speaker: Luca Messina (CEA, DES, IRESNE, DEC, France)

contributed

15 Multi-scale modelling of hydrogen isotopes retention and diffusion in highly irradiated materials

A fundamental description of gas transport and retention in plasma-facing materials is crucial for tritium inventory modelling. During future reactor operations, the material microstructure is expected to evolve and reveal defects which act as “traps” for diffusing gas atoms, thereby compromising material performance. The classical formalism for gas diffusion and trapping is given by the McNabb-Foster equation [1] and Oriani equilibrium condition [2] to describe the reduction in diffusivity in the presence of traps. This “effective” diffusivity was initially proposed for single-occupancy traps under a dynamic equilibrium, where trapping and detrapping rates are equal.

In this work, we have developed a multi-gas, multi-trap, multi-occupancy and multi-dynamic transport code ‘Palioxis’ to bridge the gap between microscopic first principles and macroscopic partial differential equation approaches for modelling the diffusivity under irradiated conditions. We calculate the effective diffusivities of hydrogen isotopes in vanadium, tungsten and iron, as a function of homogenous vacancy concentration and gas concentration. In tungsten, the diffusivity monotonically increases with increasing hydrogen/gas content. However, a small drop in diffusivity is identified in vanadium and iron when gas and vacancy concentrations are comparable. The difference in behaviour between the elements is traced back to the difference in incremental binding energy of the nth hydrogen to an (n-1) vacancy complex. We validate the results from Palioxis by directly comparing effective diffusivities to those from molecular dynamics simulations performed with classical and machine-learning interatomic potentials.

[1] A. McNabb, P. Foster Trans. Metall. Soc. AIME 227 (1963)
[2] R. Oriani Acta Metall. 18 (1970)

Speaker: Sanjeet Kaur (United Kingdom Atomic Energy Authority)

10:20 break

invited

16 Machine Learned Interatomic Potentials for Extreme Environments

Classical molecular dynamics (MD) is in principle an ideal tool to investigate the long-time evolution of materials in extreme environments, as ab initio-based MD simulations remain limited to very short time. While modern machine learning MD potentials report errors on the order 1 meV/atom, these errors are only typical of configurations that are similar to those found in the training set used to fit the potential, and transferability to genuinely new configurations remain limited. This poses a challenge to the accuracy of long-time MD simulations for two reasons: i) transition rates are exponentially sensitive to energy barriers, and ii) saddle configurations form a very small subset of the whole configuration space and are unlikely to appear in traditional hand-crafted datasets, or even as part of conventional active-learning approaches. We propose a large-scale automated workflow to develop and validate transferable machine learned interatomic potentials (MLIP) for long-time simulations in extreme environments. The workflow is implemented in the pyiron workflow framework, developed in our group, which accelerates the rapid prototyping and up-scaling of computational materials science workflows up to the latest generation of Exascale Computers. Finally, we apply the resulting MLIPs to calculate bulk phase stabilities and the stability of defect phases.

Speaker: Jan Janssen (Max Planck Institute for Sustainable Materials)

17 Towards Advanced Wall Modeling Using Machine Learning and Its Integration with Neutral Transport Simulations

In fusion reactors, the control of plasma parameters in the edge plasma plays a crucial role in maintaining high core plasma performance and achieving efficient burning conditions. The edge plasma is in direct contact with the reactor wall, where complex interactions occur due to plasma irradiation. This interaction leads to the release of neutral hydrogen atoms and molecules from the wall material. These released particles, in turn, undergo further reactions within the edge plasma, influencing its density, temperature, and overall behavior.

Understanding and accurately modeling the dynamic plasma–wall interactions is thus essential for predicting and controlling edge plasma behavior, especially under reactor-relevant conditions. However, the interaction mechanisms involve multi-scale and multi-physics processes, ranging from atomic-scale surface reactions to macroscopic plasma transport, making it a highly challenging task for conventional modeling approaches.

To address this challenge, we are developing a novel numerical simulation scheme that incorporates machine learning-based wall models into plasma transport simulations. Our approach utilizes molecular dynamics simulations to generate fundamental interaction data, which are then used to train machine learning models capable of predicting particle desorption. By integrating these trained models into a plasma fluid simulation framework, we aim to achieve a self-consistent simulation of plasma–wall interactions that evolve dynamically with changing plasma conditions.

This paper presents the development of this machine learning–integrated simulation scheme, with a particular focus on its application to the edge plasma environment of the JA-DEMO reactor. Simulation results demonstrate that incorporating realistic wall response models alters the edge plasma profiles such as rotational state distribution of hydrogen molecules. These changes affect the overall heat and particle balance in the edge region.

Speaker: Seiki Saito (Yamagata University)

contributed

18 Applying machine learning potential models to the study of hydrogen in metals: accurate property calculations, complex dynamics simulations, and challenges

The transport and retention of hydrogen isotopes, deuterium and tritium, in materials affect the economics and sustainability of the fusion fuel cycle, as well as the integrity of materials due to the detrimental effects of hydrogen and helium produced by the alpha decay of tritium. Therefore, accurately predicting the behavior and effects of hydrogen in materials used as reactor components has been an important research topic in fusion engineering. However, the prediction is not a trivial task, mainly due to the complexity and diversity of hydrogen dynamics caused by lattice imperfections such as impurities, point defects, dislocations and grain boundaries, which limit the accuracy of experimental measurements. At the same time, the applicability of atomistic simulations has also been limited by a severe trade-off between computational accuracy and efficiency: to be specific, classical molecular dynamics (MD) is not accurate enough and density functional theory (DFT) calculations are not efficient enough.

Recently, the accuracy-efficiency tradeoff of atomistic simulations has been greatly improved by the advent and advanced implementation of machine learning potential models (MLIPs), which can achieve accuracy comparable to DFT calculations at much lower computational cost. In this presentation, focusing on hydrogen in metals, we will present our recent research progress and experience in the application of MLIPs, including (i) accurate calculations of fundamental properties, such as diffusivity, solubility, and permeability of hydrogen in metals, and (ii) calculations of kinetic parameters for complex surface processes to be used for rate models to simulate hydrogen behavior at the engineering scale. In addition, we will share our experiences in the construction and use of MLIPs, highlighting the challenges we have faced and what should be overcome to better utilize MLIPs for the study of hydrogen in metals.

Speaker: Takuji Oda (Seoul National University)

12:20 lunch

invited

19 GRACE universal interatomic potential for materials discovery and design

Universal interatomic potentials parameterize the interactions between all chemical elements in the periodic table simultaneously. In my talk I will introduce the Graph Atomic Cluster Expansion (GRACE). GRACE builds on a complete set of graph basis functions and can be viewed to generalize equivariant message passing neural networks and other machine learning interatomic potentials. I will then discuss the parameterization of a GRACE foundation model across the periodic table and compare the performance of GRACE foundation model to element-specific potentials.

The ability to simulate thousands or millions of atoms with complex chemistries for extended time scales opens completely new routes for materials discovery and design. I will demonstrate usage scenarios for widely different materials. Finally, I will discuss limitations of current foundation models and suggest steps to overcome these.

Speaker: Ralf Drautz (Ruhr Universitaet Bochum)

20 Sensitivity analysis and optimization of multi-scale models for microstructural evolution in metal materials under neutron irradiation

Developing materials resistant to neutron irradiation is one of the key challenges in fusion energy applications. Due to the lack of fusion neutron sources for irradiation experiments, computational simulations provide valuable information and references for addressing such challenges. Multiscale simulation methods through a hierarchical, information-passing paradigm are often employed to study microstructural evolution of irradiated materials. Large-scale methods usually take the computational results of small-scale methods (such as Molecular Dynamics, MD) as input parameters. The uncertainties in small-scale methods can then propagate in large-scale methods, leading to inevitable uncertainties in the model predictions. In order to improve the accuracy of model predictions, it is necessary to perform parameter sensitivity analysis for multi-scale simulation methods.

This study conducted a sensitivity analysis on the Object Kinetic Monte Carlo (OKMC) simulations of tungsten (W) microstructural evolution under neutron irradiation. Initially, a total of 14 input parameters including capture radius, defect migration energies and pre-exponential factors were selected for analysis. The Latin hypercube sampling was used to construct the input space. The average size and number density of defect clusters were taken as quantity of interest (QoI), and the Spearman correlation coefficient was used as the sensitivity index. The results show that single-vacancy migration energy, self-interstitial rotation energy and capture radius are the most sensitive. These sensitive parameters were then characterized in a more accurate way to reduce model uncertainty and improve the accuracy of model predictions. For example, we performed a more accurate calculation of the capture radius of defects through molecular dynamics simulations. We used the optimized capture radius as the input parameter for OKMC simulations and found that the resulting average sizes and number densities of the defect clusters were in better agreement with the experimental results, confirming that the optimization of sensitive input parameters can improve the accuracy of OKMC model.

Speaker: Zhangcan Yang (Huazhong University of Science and Technology)

contributed

21 Trap-diffusion modelling of diffusion in restricted geometries

Trap-diffusion modelling is of fundamental importance for the analysis of plasma-material interaction experiments (e.g. to extract information about hydrogen isotope trapping energies using thermal effusion spectroscopy) and is also indispensable to estimate tritium retention/permeation in future fusion devices.

Thus in the recent years a number of simulation codes have been developed (e.g. TESSIM-X, MHIMS, TMAP8, RAVETIME, FESTIM), all implicitly relying on the mathematical assumption of an underlying 3-dimensional random walk based transport process. However, in many cases of practical interest this assumption may not hold. For example, hydrogen permeation experiments on tungsten foils, evaluated using hydrogenography have shown the importance of hydrogen transport along grain boundaries. The geometric properties (i.e. the aspect ratio) of typical grain boundaries (e.g. in tungsten) suggest that a 2-d random walk is better suited than a 3-d random walk model to describe the hydrogen transport in grain boundaries.

The consequences for the transport properties and how the macroscopic trap-diffusion equations have to be modified to account for diffusion in restricted geometries are presented.

Speaker: Udo von Toussaint (Max-Planck-Institute for Plasmaphysics)

15:10 break

invited

22 MD simulations of high-dose irradiation in Tungsten: The Role of Defect Boundary & Morphology

The molecular dynamics (MD) simulations of successive collision cascades (SCC) within a single simulation domain have recently been employed to predict radiation damage at varying dpa levels [1,2]. We carry out SCC simulations with different primary knock-on atoms (PKAs) and interatomic potentials (IPs)—namely, traditional EAM and machine learning potentials (MLIPs). Since dpa serves as the primary metric for quantifying irradiation-induced changes in key material properties irrespective of incident energy, we analyze defect properties as a function of dpa while noting their sensitivity to the PKA energy, choice of IP, and agreement with experiments. We compare defect characteristics—such as overall defect density, boundary defect density (accounting for all point defects, except that for dislocation loops only peripheral defects are counted), swelling, defect morphology and size distribution—with experimental data from Transient Grating Spectroscopy (TGS) [3] and transmission electron microscopy (TEM) at various dpa levels. Notably, the boundary defect density remains consistent across different energies and IPs and aligns well with TGS measurements, whereas the total defect count varies significantly. Differences in defect morphologies observed across potentials are discussed in the context of formation energies for various self-interstitial atom (SIA) configurations, as well as the influence of training data and parameters in MLIP development [4,5]. We also discuss the computational methods and challenges for a scalable and detailed analysis of SCC performed by extending the Csaransh [6] and SaVi [7] algorithms. Our work thus provides a comprehensive assessment of the sensitivity of defect properties to both incident energy and potential choice, showing that the boundary defect density is an important factor that exhibits energy-independent behavior and experimental agreement similar to dpa. The results highlight the immense predictive capabilities—and current limitations—of SCC in fusion materials research.

REFERENCES:

1. F. Granberg, J. Byggmästar, K. Nordlund, Molecular dynamics
   simulations of high-dose damage production and defect evolution in
   tungsten, Journal of Nuclear Materials 556 (2021) 153158.
2. U. Bhardwaj, M. Warrier, Molecular dynamics simulations of the
   defect evolution in tungsten on successive collision cascades,
   arXiv:2405.03344.
3. A. Reza, H. Yu, K. Mizohata, F. Hofmann, Thermal diffusivity
   degradation and point defect density in self-ion implanted tungsten,
   Acta Materialia 193 (2020) 270–279.
4. J. Byggmästar, F. Granberg, K. Nordlund, Effects of the short-range
   repulsive potential on cascade damage in iron, Journal of Nuclear
   Materials 508 (2018) 530–539.
5. U.Bhardwaj, V.Mishra, S.Mondal, M. Warrier: A Robust Machine Learned
   Interatomic Potential for Nb: Collision Cascade Simulations with
   accurate Defect Configurations, arxiv: 2502.03126.
6. U. Bhardwaj, H. Hemani, M. Warrier, N. Semwal et. al., Csaransh:
   Software suite to study MD simulations of collision cascades, JOSS
   (Sep2019). doi: 10.21105/joss.01461.
7. U. Bhradwaj, A. Sand, M. Warrier, Graph theory based approach to
   characterize self interstitial defect morphology, Computational
   Materials Science, 195 (2021) 110474.

Speaker: Utkarsh Bhardwaj (Bhabha Atomic Research Centre)

contributed

23 Transport and deuterium interaction with defects in W, EUROFER and W-Cr-Y alloys damaged by heavy ions

To simulate neutron-induced defects, materials have been irradiated with Fe ions with an energy of 5.6 MeV at a temperature range of 250-500C and dose range of 3-50 dpa. The radiation defects have been investigated by transmission electron microscopy (TEM), energy-dispersive X-ray spectrometry (EDXS), atom probe tomography (APT) and positron annihilation spectroscopy (PALS). To decorate radiation-induced defects, D was implanted in damaged materials with an energy of 670 eV to a fluence of 1022 D/m2. The D retention was studied by in-situ thermal desorption spectroscopy (TDS) in order to avoid changes in the surface conditions as a result of the sample’s contact with air after irradiation.
It was found that (i) radiation-induced defects are mainly vacancy-type defects in W and (ii) the radiation-induced Cr clusters suppresses the formation of both vacancy-type and dislocation-type defects in the W-Cr-Y alloy. In contrast, the formation of radiation-induced Cr-Mn clusters in Eurofer does not mitigate the dislocation loop development. It was shown that TEM method leads to underestimation of small particles and clusters: smaller particle sizes and vacancy clusters with densities an one order of magnitude higher than those measured by TEM were measured using APT and PALS methods, respectively. Therefore, it is necessary to use several experimental methods to determine the micro- and nanostructure of materials, since each of them has its own limitations. Validation of models against only TEM data (which is currently often used) may be incorrect and can lead to misunderstanding of the underlying physics.
The accumulation of D in radiation defects in W and Eurofer is an one order of magnitude higher than in undamaged materials, but is either does not change or only by a factor of 1.4 higher in damaged W-Cr-Y alloy compared to undamaged W-Cr-Y alloy. The accumulation of D in radiation defects correlates with a formation of (i) vacancy clusters in W, (ii) Cr clusters in W-Cr-Y alloy and (iii) Cr-Mn clusters, dislocation loops and vacancy clusters in Eurofer. The modelling of migration and trapping of D in each type of intrinsic and radiation-induced defects has been performed, and the binding energies of D with defects were determined.

Speaker: Dr Olga Ogorodnikova (National Research Nuclear University “MEPHI” (Moscow Engineering Physics Institute))

24 Vacancy dynamics and hydrogen retention in beryllium

This contribution summarizes experimental and modeling studies of hydrogen retention in ion-implanted beryllium, focusing on the role of vacancy and self-interstitial atoms (SIA) dynamics during the implantation process. As the main modelling tool macroscopic rate equations (MRE) implemented in the CRDS code [1] and supported by density functional theory (DFT) calculations are used to simulate the dynamic behavior of irradiation induced defects along with hydrogen transport and trapping processes. The DFT data provide information on the energy landscape for hydrogen trapping and indicate that annihilation of irradiation induced vacancies and hydrogen-vacancy complexes with self-interstitial beryllium atoms occurs without an additional energy barrier, irrespective of the number of hydrogen atoms trapped in a vacancy. It also follows from DFT calculations that mobility of hydrogen-vacancy complexes can be neglected in MRE simulations. Effects of these assumptions are investigated in application to experimental conditions reported in [2], with a focus on the fluence dependence of the retained amount of hydrogen isotopes, including the experimentally observed saturation of retention at fluences above ~1022 m 2. Finally, an attempt is made to link the observed low temperature desorption stage to hydrogen release from cavities [3] formed by vacancy clustering.

[1] D. Matveev et al, Nucl. Instrum. Methods Phys. Res. B 430 (2018) 23
[2] M. Eichler et al, Nucl. Mater. Energy 19 (2019) 440
[3] M. Zibrov and K. Schmid, Nucl. Mater. Energy 30 (2022) 101121

Speaker: Dr Dmitry Matveev (Forschungszentrum Juelich)

18:30 Meeting of the Scientific Committee (closed event)

Wednesday 28 May

invited

25 Deuterium trapping and release from high-temperature ion irradiated tungsten: experiments and reaction-diffusion simulations

Tungsten (W) is considered as a promising plasma-facing material for future fusion reactors. W components will be subjected to an intense flux of 14 MeV neutrons. This will result in the production of displacement damage and material transmutation. W components will operate at elevated temperatures (673-1300 K), which will favor the formation of irradiation-induced voids. These voids will act as trapping sites for the hydrogen (H) isotopes. So far, little is known about their H trapping characteristics.

In order to simulate the neutron-induced displacement damage, recrystallized W samples were irradiated by 20 MeV self-ions at 1350 K to peak damage doses ranging from 0.001 to 2.3 dpa. To decorate the introduced defects with deuterium (D), the samples were exposed to a low-temperature D plasma at 370 K. Concentration profiles of trapped D in the samples were measured using D(3He,p)4He nuclear reaction analysis.

In contrast to the irradiations near room temperature, where the trapped D concentration reaches saturation at damage doses above 0.1 dpa, no clear trend towards saturation is visible at 1350 K. The D concentrations in the samples irradiated at 1350 K to 0.1-2.3 dpa are considerably higher compared with the extrapolation of the existing data on ion-irradiated tungsten at different temperatures (up to 1243 K).

Thermal desorption spectra (TDS) from the samples irradiated at 1350 K differed significantly from the spectra of the samples irradiated near room temperature, demonstrating the change of the D trapping mechanism. Transmission electron microscopy (TEM) investigations of the samples irradiated to 0.1, 0.5, and 2.3 dpa revealed the presence of nm-sized irradiation-induced voids. The damage dose dependence of the void swelling correlated with the damage dose dependence of the trapped D concentration in these samples.

Reaction-diffusion simulations assuming that all trapped D is located in the voids as D2 gas in the volume and as D atoms at the surface were performed. Void number densities and average sizes derived from the TEM observations were used in the simulations. The model could reasonably reproduce the kinetics of D uptake in the samples during the D plasma exposure and the TDS spectra.

Speaker: Mikhail Zibrov (Max Planck Institute for Plasma Physics)

26 Deuterium Retention Behavior in Tungsten After Plasma Exposure at Varying Temperatures

Temperature plays a critical role in the behavior of deuterium (D) in tungsten (W), a key material for plasma-facing components in fusion reactors. This study investigates the effects of temperature variation on D retention, surface blistering, and defect evolution in W. Two sets of temperature rising (TR) and temperature declining (TD) irradiation experiments were conducted in the temperature ranges of 420 K–590 K and 320 K–720 K.
The deuterium-induced surface blistering behavior was markedly different between the two temperature ranges. In the 420 K–590 K range, the TD exposure resulted in more severe surface blistering, while the TR exposure reduced the formation of intragranular blisters. For the 320 K–720 K irradiation, both TR and TD exposures led to dense blister formation on the tungsten surface, including both intragranular and intergranular blisters, though with different blister densities and average sizes. Statistical analysis of the surface blisters showed that the TD exposure favored the formation of smaller intergranular and intragranular blisters, whereas TR exposure promoted the growth of larger blisters.
In the 320 K–720 K irradiation experiment, both pristine and pre-damaged tungsten were compared to evaluate the influence of temperature variation on the pre-damage effect. The results showed that pre-damage suppressed intragranular blister formation, accelerated intergranular blister growth, and enhanced D retention in both TR and TD exposures, in agreement with previous studies. However, the exacerbating effect on intergranular blisters was more pronounced in the TR exposure, while the suppressive effect on intragranular blisters was reduced under the TD exposure.
Thermal desorption spectra revealed that D retention was significantly higher in the TD exposures compared to the TR exposures. The analysis of deuterium desorption peaks suggests that intrinsic defects in tungsten could evolve and cluster into larger ones in the TR exposure, which acted as higher-energy D traps. The evolution of pre-damaged defects in TR exposure was even more pronounced.
These studies highlight the significant impact of temperature variation during plasma exposure on D behavior and defect evolution in tungsten. Understanding these effects is crucial for optimizing the performance of plasma-facing materials in fusion reactors, where temperature fluctuations during exposure are inevitable.
The above experimental results have been published in the Journal of Nuclear Materials [J Nucl Mater 537 (2020), J Nucl Mater 606 (2025) 155607].

Speaker: Xiu-Li Zhu (North China Electric Power University)

27 Deuterium effect on defect evolution in tungsten, from bulk to surfaces

Tungsten will be used as the plasma facing material in fusion reactors, and will be subjected to irradiation, which will affect its properties. Fusion fuels, such as hydrogen isotopes, will be implanted into the material, which will affect the material properties and be a problem as radioactive tritium is retained. Experimentally it has been seen that having deuterium present during irradiation will substantially affect the defect evolution of tungsten, compared to pure tungsten [1]. For instance, about double the amount more deuterium was trapped in the dual beam experiments compared to the case where tungsten first was irradiated and then as a second step subjected to deuterium. This indicates that the defect structure and defect production mechanisms are different. The tungsten surfaces have also been seen to be decorated by deuterium, which might affect the sputtering of the tungsten surfaces, when they are not pure tungsten anymore.

To remedy this, we carried out atomistic simulations with Molecular Dynamics to understand these phenomena better. In the bulk case we could reproduce the experimental trends by deuterium, as found the underlying mechanisms [2]. Among these were increased defect production and significantly less recombination if deuterium is present. We systematically studied the effect of different deuterium concentrations and vacancy cluster sizes. To shed some light on how deuterium affects the sputtering of tungsten surfaces, different deuterium decoration levels of the tungsten were studied. Sputtering simulations by various fusion relevant ions were studied under different conditions. We found that deuterium decoration will affect the tungsten sputtering, in addition to being sputtered itself. The next steps involve multiscale modelling to achieve longer timescales and Machine Learned interatomic potential development for these systems, for more accurate representation.

References:
[1] S. Markelj, T. Schwarz-Selinger, M. Pecovnik et al. Nucl. Fusion 59 (8) (2019) 086050
[2] V. Lindblad, D. R. Mason, F. Granberg, J. Nucl. Mater. 603 (2025) 155422

Speaker: Fredric Granberg (University of Helsinki)

10:30 break

invited

28 Investigating the Influence of Helium on Screw Dislocation Mobility in Tungsten

Tungsten, a candidate for nuclear fusion reactor armour, is susceptible to helium contamination. Such impurities are believed to interfere with the movement of screw dislocations, which typically propagate by nucleation and migration of kink pairs. Modelling this requires large simulations cells that are well beyond the limits of density functional theory, with a few heroic exceptions [1]. Therefore, we have built upon an existing machine learning interatomic potential for tungsten [2], with the aim of accurately modelling extended defects such as dislocation kinks and dislocation-helium interactions.
As has been previously reported in QM/MM studies [3], we observe a reconstruction to the split-core local to helium in the dilute regime. We then investigate the influence of helium on the kink pair mechanism, finding that the nucleation energy is reduced from 1.6eV (in the pure metal) to 0.5eV, caused by the preference for helium atoms to bind to the vacancy-like kink [4]. This preference is so large, that when bound to He₂/He₃ clusters, the kinked dislocation line becomes more stable than the straight one.
Using our potential, we have also run large-scale molecular dynamics simulations to obtain dislocation velocities. To accelerate these simulations, we are testing the new LAMMPS plugin: ML-MIX [5], where potentials of different complexities (and costs) can be spatially mixed, in the hope that our complex potential is only required at the dislocation core.

[1] L. Ventelon, D. Caillard, B. Lüthi, E. Clouet, D. Rodney, and F. Willaime, Acta Materialia 247, 118716 (2023).
[2] W. J. Szlachta, A. P. Bartók, and G. Csányi, Physical Review B 90, (2014).
[3] P. Grigorev, A. M. Goryaeva, M.-C. Marinica, J. R. Kermode, and T. D. Swinburne, Acta Materialia 247, 118734 (2023).
[4] M. Nutter, J. R. Kermode, and A. P. Bartók, https://arxiv.org/abs/2406.08368, (2024).
[5] F. Birks, T. D. Swinburne, & J. R. Kermode, https://arxiv.org/abs/2502.19081, (2025).

Speaker: Matt Nutter (University of Warwick)

contributed

29 Surface Characterization of Advanced Tungsten and Ceramic Plasma-Facing Materials

This study focuses on using a combination of surface characterization techniques to assess the response of advanced tungsten alloys and ultra-high temperature ceramics (UHTCs) to high-flux plasmas. While these materials are at an early stage of development, they potentially offer superior thermomechanical properties and microstructural stability relative to other existing candidate materials. Our ongoing work focuses on assessing their performance from an erosion, defect nucleation, and hydrogen isotope retention standpoint.

One important diagnostic requirement is the ability to directly detect and quantify how hydrogen isotopes interact with other species present at the exposed surface. Most electron spectroscopy techniques provide only limited sensitivity to hydrogen isotopes. To overcome this challenge, our group employs low energy ion beam techniques that provide direct sensitivity to chemisorbed deuterium through analysis of scattered and recoiled particles. Our basic approach relies on an ultrahigh vacuum system equipped with an electrostatic analyzer, allowing us to perform low energy ion scattering (LEIS), direct recoil spectroscopy (DRS), and x-ray photoelectron spectroscopy (XPS), and thermal desorption spectroscopy (TDS) in one measurement chamber. LEIS and DRS measure the surface chemical composition, including hydrogen isotopes, with sub-monolayer sensitivity. While XPS is not strongly sensitive to D, it provides complementary information on the chemical bonding between species and their electronic structures. Finally, TDS is used to assess surface-to-bulk transport of hydrogen isotopes in these materials.

In this report, we will summarize recent results from applying these techniques to different dispersoid-strengthened W-TiO$_{2}$ alloys and K-doped W materials. Incorporation of additive species was found to have a modest effect on hydrogen isotope chemisorption behavior on sputter-cleaned surfaces, since the surface is dominated by W which readily dissociates molecular hydrogen. For most materials, the surface hydrogen concentration diminishes during heating to 500 °C, following an Arrhenius trend when dosing with a partial pressure of $10^{-6}$ Torr D$_{2}$(g). The presence of O and C impurities, however, has a much more dramatic effect on reducing the chemisorbed H concentration. The effect of the alloy components was found to be more pronounced from a hydrogen retention and sputtering standpoint. Hydrogen retention in W-TiO$_{2}$ dispersoid-strengthened material has been observed to be a factor of 2-3 higher than in pure polycrystalline W, and enhanced sputtering of the dispersoid material was also observed relative to the surrounding tungsten matrix. In addition to these measurements, we will also discuss preliminary results for newer ZrC UHTC’s and nanostructured W materials as part of a new project focusing on the development of advanced plasma-facing materials.

Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC (NTESS), a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration (DOE/NNSA) under contract DE-NA0003525.

Speaker: Dr Robert Kolasinski (Sandia National Laboratories)

30 Nanoindentation and Defect Behavior in Irradiated FCC NiFe Alloys: Experimental Insights and Atomistic Modeling

Fcc NixFe1-x single-crystal alloys are key model systems for studying defect evolution under self-ion irradiation at room temperature, with fluences ranging from 4 × 10¹³ to 2 × 10¹⁵ ions/cm² [1,2]. This study investigates the effects of irradiation-induced defects on the nanomechanical response of NiFe alloys through a combination of experimentally guided nanoindentation and atomistic simulations. Defects in the irradiated materials were characterized using transmission electron microscopy (TEM), providing essential input for molecular dynamics simulations of overlapping collision cascades to prepare irradiated samples at various doses [3]. The simulations revealed mechanisms of defect formation and evolution, including the emergence of an A15 Frank-Kasper phase within the {111} plane in pure Ni, Ni₀.₈₈Fe₀.₁₂, and Ni₀.₇₇Fe₀.₂₃ at fluences up to 2 × 10¹⁴ ions/cm², serving as a precursor to Frank loop nucleation. The findings also highlight the critical role of Fe atoms in influencing dislocation nucleation and evolution. In Ni₀.₇₇Fe₀.₂₃, compact 3D precipitates heavily decorated with Fe atoms were observed. However, these precipitates remained too small to evolve into Frank loops, even at elevated fluences. Nanoindentation experiments and simulations demonstrated a notable increase in hardness in irradiated alloys compared to their pristine counterparts [4,5]. From the atomistic modeling, we provide detailed analyses of the irradiated samples revealed surface morphologies, dislocation densities, and strain mappings to discuss the mechanisms of dislocation pinning during mechanical loading due to irradiation-induced defects. These defects were found to obstruct dislocation motion, contributing to the enhanced hardness of the material. Overall, this study provides critical insights into the interplay between composition, irradiation-induced defect structures, and their mechanical consequences in NixFe1-x alloys. These findings enhance our understanding of the mechanisms underpinning radiation resistance in these materials, supporting their potential applications in radiation-intensive environments. [1] E. Wyszkowska et al. Nanoscale 15, 4870 (2023). [2] A. Ustrzycka et al. Int. J. Plasticity 182, 104118 (2024) [3] K. Mulewska et al. J. Nucl. Materials 586, 154690 (2023) [4] L. Kurpaska et al. Materials & Design 217, 110639 (2022). [5] F. Dominguez-Gutierrez et al. J. Appl. Phys. 135, 185101 (2024)

Speaker: Francisco Javier Dominguez Gutierrez (National Centre for Nuclear Research)

Discussion

Closing