Christian Linsmeier

Forschungszentrum Jülich · Institut für Energie- und Klimaforschung - Plasmaphysik (IEK-4)

Successful upscaling of tungsten fiber‐reinforced tungsten composites (W f /W) on industrial level could represent an important milestone for future nuclear fusion reactors. The primary objective of these materials is to enhance the durability and operational lifespans of critical components. However, developing mature manufacturing approaches rema...

Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development for a fusion power plant. SMART exhibits similar sputtering resistance as that of pure tungsten during regular plasma operation. Under accident conditions, SMART demonstrates the suppression of oxidation-a remarkable advantage over pure tungsten. The viability...

Our talk at the International Symposium on Fusion Nuclear Technology (ISFNT 15) in Las Palmas, Gran Canaria back in September 2023. Fantastic scientific forum in fantastic location. Primary focus on industrial scale-up of our SMART materials for fusion and other new energy sources. Stay tuned :)

In order to estimate the fuel loss in ITER and further future fusion devices, the deuterium permeation through different wall and structural materials are studied. In order to determine the effective permeability, gas-driven deuterium permeation measurements are performed on Cu and ITER grade CuCrZr. The obtained permeabilites for Cu and ITER grade...

In the present study, the influence of powder characteristics on the densification behavior and microstructure evolution of the W-Cr-Zr alloy during consolidation was investigated. The W-Cr-Zr alloy powders were prepared through mechanical alloying in different modes, i.e. continuous mechanical alloying (CMA) and intermittently mechanical alloying...

Deuterium (D) retention in tungsten-chromium-yttrium (W–Cr–Y) alloy depending on the irradiation fluence of 10¹⁹–5×10²¹ D/m² at different sample temperatures in the range of 300–900 K was investigated using in-situ thermal desorption spectroscopy (TDS). The irradiation was carried out using 2 keV D3⁺ ions (670eV/D). An increased D retention compare...

16 MeV protons have been used to irradiate 300 μm thick macroscopic W samples in a pilot experiment to 0.006 dpa damage dose under low and high temperature scenarios of ∼373 K and ∼1223 K, respectively. The linear pre-Bragg region has been used for damage where the electronic loss (heat) in the sample amounts to 1.5 MW · m⁻². Post high-temperature...

Material related limitations are one of the main challenges for the design of future fusion reactors. Tungsten (W) as the primary material choice is considered resilient against erosion, has the highest melting point of any metal and shows low activation after neutron irradiation. However, W is intrinsically brittle and faces operational embrittlem...

Low energy ion scattering is a technique to detect the energy of ions which are scattered from a surface. For noble gas ions, it is predominantly sensitive to the topmost surface layer due to strong neutralisation processes. Depending on the combination of projectile ion and target material, the scattering spectra can exhibit contributions resultin...

Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development for the primary application as plasma-facing materials of the first wall in a fusion DEMOnstration power plant (DEMO). SMART materials must combine the suppressed oxidation in case of an accident and an acceptable plasma performance during the regular operatio...

Tungsten (W) has the unique combination of excellent thermal properties, low sputter yield, low hydrogen retention, and acceptable activation. Therefore, W is presently the main candidate for the first wall and armor material for future fusion devices. However, its intrinsic brittleness and its embrittlement during operation bears the risk of a sud...

For the ITER core CXRS diagnostic, a shutter is mandatory to protect the first mirror when no measurement takes place. With the continuing design development of the diagnostic and the knowledge obtained from previous shutter prototypes, the frictionless fast shutter did undergo some design changes and manufacturing improvements without degrading th...

Tungsten (W) has a unique combination of excellent thermal properties, low sputter yield, low hydrogen retention, and acceptable activation. Therefore, W is presently the main candidate for the first wall material in future fusion devices. However, its intrinsic brittleness and its further embrittlement during operation bears the risk of a sudden a...

The article is devoted to materials used in controlled fusion facilities. Material solutions for magnetic confinement fusion devices and inertial confinement fusion facilities are described. Fusion materials have to preserve their performance often under extremely harsh conditions characterized by intensive particle fluxes, high levels of neutron i...

The research on plasma facing components in view of plasma wall interactions is advancing towards a new level with the availability of long pulse plasma machines. With Wendelstein 7-X (W7-X) the first stellarator provides steady-state discharges with up to 30 min in the next operation phase (OP2). To support the investigation of PWI processes with...

An overview presentation on the progress in R&D on self-passivating SMART Alloys for a future fusion power plant. This is an invited (online) talk at the THERMEC 2021, Vienna, Austria

The tungsten (W) foil laminate is an advanced material concept developed as a solution for the low temperature brittleness of W. However, the deformed W foils inevitably undergo microstructure deterioration (crystallization) during the joining process at a high temperature. In this work, joining of the W foil laminate was carried out in a field-ass...

The tungsten (W) foil laminate is an advanced material concept developed as a solution for the low temperature brittleness of W. However, the deformed W foils inevitably undergo microstructure deterioration (crystallization) during the joining process at a high temperature. In this work, joining of the W foil laminate was carried out in a field-ass...

The newly developed ICRH antenna system for the stellarator W7-X will be available for Operational Phase 2.1. The design and application of various components in the ICRH system was validated using a prototype of the antenna box with straps. Electromagnetic characteristics were measured and optimized while reconsidering thermal and mechanical aspec...

To overcome the brittleness of tungsten, tungsten fiber-reinforced tungsten composites (Wf/W) have been developed using an extrinsic toughening mechanism. In this work, a novel type of Wf/W with porous matrix produced by field assisted sintering technology (FAST) is studied. The material is optimized regarding mechanical behavior, standing on the a...

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m2. Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat lo...

Field-assisted sintering technology (FAST), as a fast densification method with low process temperature, was used to manufacture self-passivating tungsten alloys (SPTAs) of W-Cr-Zr in this work. To clarify the behaviors of grain growth and Cr-rich phase precipitation under the action of electric current during the densification process, interrupted...

Field assisted sintering technology (FAST) has been widely employed for powder materials consolidation. The assisted-current and applied pressure in FAST facility strongly affect the microstructures of the consolidated powder materials. In this work, the influence of the applied pressure on the W-11.4Cr-0.6Y-0.4Zr alloy during FAST consolidation at...

The roughness of metallic surfaces has a vital impact on the erosion of plasma-facing materials. Roughness determines the effective sputtering yield Yeff of the facing material. the angular/energy distribution of sputtered particles, and the spatial erosion and deposition distribution. The model for simulation the effect of the surface roughness wa...

Tungsten test is currently the baseline first-wall armor material for a future DEMOnstration power plant. Smart alloys, containing tungsten (W), 11.4 weight (wt) % chromium (Cr), and 0.6 wt% yttrium (Y), aim at achieving passive safety in case of air ingress into the vacuum vessel and a loss-of-coolant accident causing a temperature rise above 1200...

The Toroidal Magnetized System device has been significantly upgraded to enable development of various wall conditioning techniques, including methods based on ion and electron cyclotron (IC/EC) range of frequency plasmas, and to complement plasma–wall interaction research in tokamaks and stellarators. The toroidal magnetic field generated by 16 co...

In fusion power plants a tritium permeation barrier is required in order to prevent the loss of the fuel. Moreover, the tritium permeation barrier is necessary to avoid that the radioactive tritium accumulates in the first wall, the cooling system, and other parts of the power plant. Oxide thin films, e.g. Al2O3, Er2O3 and Y2O3, are promising candi...

Self-passivating, so-called smart alloys are under development for a future fusion power plant. These alloys containing tungsten, chromium and yttrium must possess an acceptable plasma performance during a regular plasma operation of a power plant and demonstrate the suppression of non-desirable oxidation of tungsten in case of an accident. The up-...

The present study addresses the uncertainties that affect the recently performed predictions of beryllium (Be) erosion and migration in ITER using the Monte-Carlo code ERO2.0. The focus of the study is a D-T baseline discharge with fusion power gain Q=10, scrape-off layer (SOL) input power PSOL=100MW, toroidal plasma current Ip=15MA, and central to...

In order to investigate the effect of neon seeding on deuterium retention and surface modification of ITER-like forged tungsten with grains elongated perpendicular to the surface, pure and neon-seeded deuterium plasma exposures were performed in the linear plasma device PSI-2. The ion percentage of neon in the mixed plasma was around 10%. The sampl...

An improved quartz crystal microbalance measurement method is described, which allows us to determine erosion, implantation, and release rates of thin films, during changing temperatures and up to 700 K. A quasi-simultaneous excitation of two eigenmodes of the quartz resonator is able to compensate for frequency drifts due to temperature changes. T...

Fuel permeation and retention in fusion reactor wall materials are important issues for plasma operation and safety reasons in ITER. The austenitic stainless steel 316L(N)-IG will be used as structural material in the first wall components in ITER. The impact of deuterium plasma exposure on the deuterium permeation and retention was studied. Polish...

Intra-ELM tungsten sources, which dominate the total W source, are quantified in the inner and outer divertor of JET-ILW. The amount of the sputtered W atoms for individual ELMs demonstrates a clear dependence on the ELM frequency. It decreases when the pedestal temperature is lower and, correspondingly, the ELM frequency is higher. Nevertheless, t...

Poster on optimization of FAST technology for production of self-passivating SMART alloys

Tungsten (W) is a prime candidate as first wall armor material of future fusion power plants as W withstands extreme particle, heat, and radiation loads without forming long-lived radioactive waste. The release of radioactive material from the reactor to the environment should be suppressed in case of an accident such as a loss of coolant (LOCA) wi...

The dynamic behaviour of thermally driven segregation of Cr to the surface of WCrY smart alloy is studied with low energy ion scattering (LEIS). Sputtering the WCrY sample with 500 eV D2+ ions at room temperature results in preferential removal of the lighter alloy constituents and causes an almost pure W surface layer. At elevated temperatures abo...

Tungsten samples have been irradiated with 3 MeV protons with dose rates of 1×10-04 to 5×10-05 dpa/s to doses of 0.01 - 0.67 dpa at 360 K in a pilot experiment. Micro- and macro-indentation were used to measure irradiation hardening in the samples. An initial irradiation hardening of 1.23±0.09 GPa and 1.88±0.83 GPa was measured by micro and macro i...

Micro structured tungsten is a new approach to address one of the main issues of tungsten as high heat flux (HHF) plasma facing material (PFM), which is its brittleness and its propensity to crack formation under pulsed, ELM like, heat loads [2], [3]. With power densities between 100 MW/m² and 1 GW/m², progressive thermal fatigue induced damages li...

During an accident with loss-of-coolant and air ingress in DEMO, the temperature of tungsten first wall cladding may exceed 1000 °C and remain for months leading to tungsten oxidation. The radioactive tungsten oxide can be mobilized to the environment at rates of 10–150 kg per hour. Smart tungsten-based alloys are under development to address this...

For the in situ application of LID (Laser-Induced Desorption) as a space-resolved tritium retention diagnostic in ITER, the desorption behaviour of co-deposited deuterium (D) from beryllium (Be) layers is studied. In particular, the desorption efficiency dependence on laser pulse parameters is investigated for pulse durations of 1–20 ms and absorbe...

Plasma-wall interaction (PWI) research is an active field of study in long-pulse operation in current magnetic confinement fusion devices, such as the Experimental Advanced Superconducting Tokamak (EAST). It is an urgent requirement to be able to investigate several key PWI issues, such as fuel retention, by in situ diagnostic methods. In this work...

The development and application of robust tritium permeation barrier coatings is crucial for a safe and economic fusion reactor operation. Three different tungsten and tungsten nitride layers on Eurofer97 substrates were investigated by deuterium permeation measurements and compared. The microstructure and crystal structure was characterized before...

In this work, W-Cr-Zr as a self-passivating tungsten alloy is studied. Spark plasma sintering (SPS) is used to prepare the sample. The influence of the heating rate on the densification process and microstructure evolution during SPS was investigated. The increasing the heating rate enhance the homogeneity of the microstructure, however, did not sh...

Fuel retention and hydrogen permeation in the first wall of future fusion devices are crucial factors. Due to safety issues and in order to guarantee an economical reactor operation, tritium accumulation into reactor walls and permeation through walls have to be estimated and prevented. Therefore, studies of permeation in the fusion materials are p...

Material issues pose a significant challenge for the design of future fusion reactors. Recently progress has been made towards fully dense multi short-fibre powder metallurgical production of tungsten-fibre reinforced tungsten (Wf/W) as well as optimizing the process understanding for the routes using chemical vapour deposition (CVD). For CVD-Wf/W...

Deuterium release from Be/D layers co-deposited using high power impulse magnetron sputtering (HiPIMS) is modelled with rate equations using the CRDS code under conditions of temperature programmed (TPD) and laser induced (LID) Desorption experiments. TPD results are simulated to fit D trapping parameters that are in turn applied to simulate the LI...

On JET with fully metallic first wall, highly radiative conditions with N2, Ne and Ar as well as their mixture as radiators are approached in high density H-mode plasmas. The confinement increases from H98(y,2) = 0.65 in unseeded pulses with γrad ∼ 30% to a value of H98(y,2) = 0.75 at γrad ∼ 50% with N2 injection. A degradation of the pedestal prof...

The surface morphology of plasma-facing components (PFCs) and its evolution during plasma irradiation has been shown to have a significant effect on the erosion and subsequent transport of sputtered particles in plasma. This in turn can influence the resulting lifetime of PFCs. A model for treatment of the effect of surface roughness on the erosion...

The dynamic behaviour of thermally driven segregation of Cr to the surface of WCrY smart alloy is studied with low energy ion scattering (LEIS). Sputtering the WCrY sample with 500 eV D$_2^+$ ions at room temperature results in preferential removal of the lighter alloy constituents and causes an almost pure W surface layer. At elevated temperatures...

For future fusion reactors, tungsten (W) is currently the main candidate for the application as plasma‐facing material due to its several advanced properties. To overcome the brittleness of W, randomly distributed short W fiber‐reinforced W (Wf/W) composites have been developed using field‐assisted sintering technology (FAST). Herein, Wf/W material...

Tungsten is the main candidate for the plasma-facing material in future fusion reactors. To overcome the brittleness of tungsten, tungsten fiber-reinforced tungsten (Wf/W) composites have been developed using a powder metallurgy process. In this study, a novel type of Wf/W with a porous matrix has been developed using field-assisted sintering techn...

In future fusion reactors, the blanket is foreseen to remove heat from the reactor and enclose material to breed tritium. While the blanket structure is made of steel, for a successful operation its first wall (FW) needs to be armoured by a tungsten layer. Joining tungsten and steel is currently a key issue in engineering nuclear fusion reactors be...

Functionally graded (FG) iron/tungsten (Fe/W) composites are considered for stress-relieving interlayers in tungsten-steel joints, required in future fusion reactors. The macroscopic gradation of the two materials allows relaxation of thermally-induced stresses and hence extend the lifetime of the cyclic-loaded dissimilar materials joints. While ma...

Eurofer97, P92 and Fe samples were exposed simultaneously in the linear plasma device PSI-2 to deuterium plasma and with addition of He, Ar or Kr at elevated temperature in the range of 800–950 K. Samples were exposed to plasma with an incident ion energy of 60–80 eV and an incident ion fluence of 5 × 10²⁵ m⁻². Surface morphology investigation of e...

Tungsten-chromium-yttrium (WCrY) smart alloys are foreseen as first wall materials for future fusion devices such as DEMO. While suppressing W oxidation during accidental conditions, they should behave like pure W during plasma operation due to preferential sputtering of the lighter alloying elements Cr and Y causing W enrichment at the surface. Th...

The temperature driven segregation of Cr to the surface of the tungsten-based WCrY alloy is analysed with low energy ion scattering of He+ ions with an energy of 1 keV in the temperature range from room temperature to 1000 K. Due to the high surface sensitivity, these measurements probe only the composition of the outermost monolayer. The surface c...

Fusions reactors have to handle numerous specifications before being able to show viable commercial operation, one of which is to find a proper Plasma Facing Material (PFM) which can withstand the high heat loads of several tens of megawatts per square meters combined with the pulse operation of a tokamak and many other problematics (Brezinsek et a...

ERO2.0 is a recently developed Monte‐Carlo code for modelling global erosion and redeposition in fusion devices. We report here on the code's application to ITER for studying the erosion of the beryllium (Be) first wall armour under burning plasma steady state diverted conditions. An important goal of the study is to provide synthetic signals for t...

Surface morphology of plasma-facing components (PFCs) and its evolution during the plasma irradiation has shown to have a significant effect on the erosion and subsequent transport of sputtered particles in plasma. This in turn can influence the resulting lifetime of PFCs. A model for treatment of the surface roughness effect on the erosion of PFCs...

The temperature driven segregation of Cr to the surface of the tungsten-based WCrY alloy is analysed with low energy ion scattering of He+ ions with an energy of 1 keV in the temperature range from room temperature to 1000 K. Due to the high surface sensitivity, these measurements probe only the composition of the outermost monolayer. The surface c...

An invited talk on development of tungsten - based plasma-facing materials for a future power plant

Poster on technological aspects of industrial manufacturing of smart alloys giving an additional glimpse on what would happen after a 5 year exposure in DEMO. Please have a look :-)

Outlook  Performance of smart alloys in seeded plasmas  Flux dependence of sputtering and elemental composition of smart alloys  Modeling of plasma effects on smart alloys  Industrial upscaling  First wall mockups Summary  "Smart" alloys as a first wall material for DEMO must be compatible with regular operation and provide a passive safety i...

WCrY Smart Alloys are developed as first wall material of future fusion devices such as DEMO. They aim at behaving like pure W during plasma operation due to depletion of the alloying elements Cr and Y. The Cr concentration gradients induced by preferential plasma sputtering cause Cr-diffusion. The exposure of WCrY and W samples to pure D plasma, w...

In this work, we present a new application for the line shapes of emission induced by reflected hydrogen atoms. Optical properties of the solids in contact with the plasma could be effectively measured at the wavelength of Balmer lines: time-resolved measurements of reflectance and polarization properties of mirrors are performed using the waveleng...

As a candidate material for plasma facing material in future fusion reactor, tungsten (W) fiber reinforced tungsten (Wf/W) composite has been recently developed. The crack resistance of Wf/W is proven to be significantly higher compared to normal tungsten. However, the W-fibers used always become embrittlement during the powder metallurgy (PM) proc...

Analysis of elemental distributions in plasma-facing components (PFCs) is vital for the study of material erosion, deposition, and fuel retention in Wendelstein 7-X (W7-X) stellarator operating in full 3D geometry. In this work, we report the results on the application of picosecond laser-induced breakdown spectroscopy (ps-LIBS) combined with laser...

Tungsten is the most promising first wall material for nuclear fusion reactors. One disadvantage, however, is its intrinsic brittleness. Therefore, tungsten fiber reinforced tungsten (Wf/W) is developed for extrinsic toughening. Wf/W can be produced by chemical vapor deposition (CVD), e.g. by reducing WF6 with H2 using heated W-fibers as substrate....

Beryllium will be one of the plasma-facing materials for ITER. It will have to sustain high fluxes of hydrogen isotopes and as a consequence significant amounts of tritium can be retained in the wall. For safety and operational reasons, the deuterium and tritium inventory in the vacuum vessel must be limited. As a consequence, hydrogen diffusion, t...

Tungsten-chromium-yttrium (WCrY) smart alloys are foreseen as the first wall material for future fusion devices such as Demonstration Power Plant (DEMO). While suppressing W oxidation during accidental conditions, they should behave like pure W during plasma operation due to preferential sputtering of the lighter alloying elements Cr, Y, and W enri...

Micro-structured tungsten; an advanced plasma-facing component A. Terra, G. Sergienko, D. Borodin, A. Huber, A. Kreter, Y. Martynova, S. Möller, M. Rasiński, M. Wirtz, Th. Loewenhoff, S. Brezinsek and Ch. Linsmeier Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung, 52425 Jülich, Germany a.terra@fz-juelich.de Tungsten (W) is...

Surface morphology and its evolution during the plasma irradiation is known to have a large influence on the erosion and resulting lifetime of plasma-facing components as well as tritium retention. For instance, surface roughness can affect physical sputtering, re-deposition, as well as angular distributions of the sputtered species. In this study...

The development and application of tritium permeation barriers (TPB) are crucial for safe and economical fusion reactor operation. In order to specify the requirements and important characteristics of TPB, the deuterium permeation flux through two different fusion relevant steels, namely Eurofer97 and 316L(N)-IG, were measured and compared. Further...

For the development of the tritium monitoring system in ITER the hydrogen isotope release by Laser-Induced Desorption (LID) from Be layers is studied to determine the laser parameters for a high desorption efficiency while minimising dust production and surface modifications is also pursued. Be layers of 1 µm thickness with 25–30 at% D and 3 × 1022...

ERO is a 3D Monte-Carlo impurity transport and plasma-surface interaction code. In 2011 it was applied for the ITER first wall (FW) life time predictions [1] (critical blanket module BM11). After that the same code was significantly improved during its application to existing fusion-relevant plasma devices: the tokamak JET equipped with an ITER-lik...

A reaction-diffusion model with surface occupation dependent desorption [D. Matveev et al., Nucl. Instr. Meth. B 430 (2018)23–30]has been updated to handle multiple hydrogen species to simulate hydrogen/deuterium isotope-exchange experiments performed on polycrystalline beryllium samples under ultra-high vacuum laboratory conditions. In the experim...

The Multi-Purpose-Manipulator (MPM) has been operated, as a versatile carrier system for probes, since the first campaign OP. 1.1 in 2015 at Wendelstein 7X. The MPM is mounted at the outboard midplane and is able to perform fast plunges through the entire scrape off layer up to the last closed flux surface. In addition, a system for gas injection w...