Fusion Reactor Components Research Articles

Structural damage in graphene oxide coatings onto Nb substrates upon laser irradiation

In this work graphene oxide (GO) coatings were obtained on Nb substrates by electrophoretic deposition (EPD), using a ramp of stepped potential and variable deposition time, to evaluate the feasibility as a protective coating against the exposition to a 20 W Yb laser (1070 nm). Laser irradiation constitutes a first approximation to the potential damage by local high heat loads that could occur in a Tokamak-type fusion reactor. GO coatings can act as the first barrier against exposure to the fusion reactor plasma, but the coating may experience local high heat loads and simultaneous exposure to energetic particles. The GO/Nb coatings were irradiated with different laser power-frequency settings. With X-ray photoelectron spectroscopy, the bonding evolution of the coatings was followed for each deposition time. Scanning electron microscopy was used to evaluate the coating morphology before and after laser irradiation, whereas Raman spectroscopy was used to evaluate the structural evolution of GO coatings, crystallinity , and disorder in the graphene oxide. An increase of the crystallite size due to sp 2 restoration was observed for the films prepared up to 10 s/V. Atomic force microscopy was used to study the film morphology and to estimate the film thickness by comparing the z-offset between the substrate and coating topography images. Film thickness reduces from ca. 280 nm to 100 nm with the increase in the time at each voltage step. Coating tolerance against laser-induced damage was observed up to 34% of the full laser power, where the coating was damaged and local metal melting was observed. The coating damage occurs between 4.9 × 10 8 MW/m 2 and 2.5 × 10 9 MW/m 2 and higher power levels triggered the occurrence of melting in the metallic substrate. The reduction in laser damage is attributed to the enhanced thermal dissipation by the sp 2 dominion as the crystallite size increases. GO coatings prepared by electrophoresis are shown to be promising to protect nuclear components against damage for high heat loads. • Adherent and stable GO coatings were obtained on Nb substrates by novel EPD process. • GO coatings were irradiated with high-power laser for structural damage evaluation. • Damage threshold for GO coating is 4.9 × 10 8 MW/m 2 and for Nb substrate 2.5 × 10 9 MW/m 2 . • Reduction of GO and sp 2 domain restoration after EPD enhance resistance to radiation. • GO coatings are feasible as plasma facing components in nuclear fusion reactors.

Quantification of the influence of the C, Cr and P contents on the permeability of hydrogen through Fe alloys

Ferritic-martensitic steels are candidate materials for blanket structural components of future thermonuclear fusion reactors. However, the tritium inventory that can be retained in different components of the reactor and its ability to migrate through the walls of any material may affect the correct operation of any fusion device. Therefore, the permeability of hydrogen isotopes through ferritic-martensitic steels, which depends on its metallurgical composition, becomes a key issue.The European Fusion Development Agreement (EFDA) supplied 9 Fe alloys with controlled chemical alloying element contents and microstructure. The main alloying elements were C, Cr and P and they appeared in various concentrations in the 9 alloys. They were experimentally analyzed by means of the gas evolution permeation technique with temperatures ranging from 423 K to 823 K and for high purity hydrogen loading pressures ranging from 5.0 × 102 Pa to 1.5 × 105 Pa. The transport regime turned out to be diffusive and it was studied in depth, so that the permeability of each alloy was characterized by an Arrhenius-type regression for the aforementioned temperature range.This work summarizes all the experimental measurements carried out for the permeability of hydrogen through the 9 alloys. It provides a quantification of the influence of the composition of the alloy on this transport parameter, posing different mathematical expressions for the variation of the permeability as a function of the contents of C, Cr and P.

Computational Investigation of Beryllium and Lithium Performance in Future Fusion Tokamaks

Low-z materials are exemplary candidates in tiling critical plasma-facing components in future fusion reactors due to their low ablation rates under intense high heat fluxes especially during abnormal and hard disruption events. Beryllium and Lithium as low-z materials show good performance as plasma-facing materials in current tokamak. Future tokamaks will exhibit long duration hard disruptions, which in turn requires further investigation of plasma-facing materials, as Li and Be, to judge their performance and evaluate their erosion rates. Electrothermal plasma capillary discharges are used to simulate the high-heat flux deposition on materials to assess their erosion rates. The electrothermal plasma code ETFLOW, which is written for capillary discharges to predict the plasma parameters and erosion rates is used to simulate the high-heat flux conditions similar to expected disruption events for simulated heat fluxes from as low as ~50 to as high as ~290 GW/m2 with a reconnoitering of generating the Be and Li plasmas up to the third ionization (Br+++, Li+++). Performance of Be and Li under the lowest capillary discharge currents (50 kA and 100 kA) is almost identical, however, Li shows sharper increase in the plasma pressure, heat flux, total ablated mass and the exit velocities than Be for higher discharge currents (150, 200 and 250 kA). This huge difference between the performance of Li and Be under low and high heat fluxes can be an important issue for the future magnetic fusion reactors.

The effect of carbon fiber powder reinforced composite coating on mechanical properties of TIG welded black steel pipes

Welding of steel materials is a very significantaction in the development of components for nuclear fusion reactors. The aim of this research is to explore a latest technique for joining pipeline based on composites incorporated with fibers. A systematic experimental programme was carried out to test this new technique in this study. There are three stages to the experimental programme. The samples are developed with various weld process parameters: Current-600 Amperes; Voltage-80 Volts and Temperature- 1200 degreesC. In phase-1 fabrication process two Black Steel (ASTM 795-Grade B) pipes of 3 in.. diameter and 200 mm length are employed for making a butt joint using TIG Welding. In Phase-2 for FRP joining, Carbon is taken as fiber and Epoxy is taken as Matrix. The FRP composite is made by taking 20% carbon as reinforcement, 18% HY955 as hardener and remaining 62% Epoxy as matrix material. The FRP Composite made in different layers is applied on the black steel pipe joint. Four layers of composites are applied successively one after the other on the joint and dried it properly. In phase-3, standard specimens are prepared from each joint of FRP coated black steel pipes for Tensile strength, Flexural Strength and Vickers’s Hardness according to ASTM Standards and mechanical testing is carried out along with microstructural studies. Weld consistency and a basic understanding of weld defect analysis was examined in TIG welded samples with a Black Steel base material.. The results showed that the tensile strength is improved by 26% for the joint with FRP addition, hardness improved by 2% and flexural strength improved by 4.16%. The results inferred that compared to normal TIG Welded joint, the composite material lining on TIG welded parts produce better results.

Neutron irradiation-enhanced grain growth in tungsten and tungsten alloys

To understand the microstructural stability of candidate plasma-facing materials under fusion-relevant environments, neutron irradiation of W and W-3%Re alloys with and without K and La dopants was performed in the mixed-spectrum High Flux Isotope Reactor at nominal temperatures of ~850 °C and ~1100 °C to calculated doses between 0.42 and 0.47 dpa. To the best of our knowledge, this study presents the first experimental evidence of radiation-enhanced recrystallization in W and undoped W–Re alloys at ~850 °C, conditions where thermal annealing does not cause any grain growth in a similar timescale. Potassium- or lanthanum-doped tungsten alloys showed more resistance to radiation-enhanced grain growth. We explain the acceleration of grain growth by analyzing the self-diffusion constant under atomic displacement environments. The microstructural observations of the studied W variants suggest that La doping is more effective than K doping for mitigating recrystallization. This study also found that radiation-enhanced recrystallization is an important consideration when designing and applying W to plasma-facing components in future nuclear fusion reactors.

Magneto-convective fluctuation during downward flow of liquid metal in a heated pipe in a transverse magnetic field

An imposed strong magnetic field suppresses turbulence and profoundly changes the nature of the flow of an electrically conducting fluid. We consider this effect for the case of mixed convection flows in pipes and ducts, in which unique regimes characterized by extreme temperature gradients and high-amplitude fluctuations (the so-called magnetoconvective fluctuations) have been recently discovered. The configuration is directly relevant to the design of the liquid-metal components of future nuclear fusion reactors. This work presents the general picture of the flow transformation emerging from the recent numerical studies (DNS - Direct Numerical Simulation), illustrates the key known facts, and outlines the remaining open questions. Implications for fusion reactor technology and novel experimental and numerical methods are also discussed.

Thermal stability of the microstructure in rolled tungsten for fusion reactors

Plasma-facing components of future fusion reactors will have tungsten-based materials as armor. Annealed pure tungsten is brittle at room temperature restricting its use as plasma-facing material, whereas plastically deformed tungsten behaves in general more ductile at ambient temperatures. During operation as plasma-facing material at high temperatures, the deformation structure induced by plastic deformation becomes unstable. Restoration processes as recovery, recrystallization, and grain growth will alter the microstructure and impair the desired mechanical properties. In particular, recrystallization will reinstate the intrinsic brittleness of tungsten. Achieving a thorough understanding of the occurring restoration mechanisms (for long times at temperatures as close to the desired operation temperatures as possible) and quantifying the temperature-dependent recrystallization kinetics are essential for assessing the materials performance and an informed materials selection. The thermal stability of differently rolled pure tungsten plates is reviewed with the aim of predicting the materials lifetime; the impact of different activation energies on the selection of armor materials highlighted. The concept of a recrystallization temperature constituting a threshold temperature below which recrystallization does not occur is dismissed.

High energy density pulsed argon plasma synthesized nanostructured tungsten for damage mitigation under fusion relevant energetic he ion irradiation

Tungsten (W) is considered as one of the promising candidates for plasma facing components in fusion reactors since it has high threshold energy, high melting point, low threshold shock resistance, and resistance against formation of co-deposits with tritium. Despite these commendable features, helium ions produced during fusion reaction are known to alter the microstructure of tungsten. In this work, dense plasma focus (DPF) device is used to study the effect of helium ion flux on double forged tungsten samples delivering a heat load of 6.27×104J/m2 per shot. The irradiation of virgin W samples is carried out at 5, 10, and 15 DPF helium shots. High heat loads resulted in blisters and micro-cracks on the sample surface. With an increase in the number of shots, the density of the blisters increased and craters on the W surface burst followed by re-solidification of the melted and sputtered surface. Surface nano-structurization, with nanoparticle formation, of W samples (nano-W) is realized by high-energy argon ion exposure of virgin samples in an argon filled DPF device. The average size of nanoparticles is found to increase with the number of shots and also leads to particle agglomeration. At 10 shots, uniformly distributed highly dense nanoparticles of 20–50 nm size have been synthesized. The nano-W samples are then irradiated by instability-accelerated high-energy helium ions in helium filled DPF device for 5, 10, and 15 shots to simulate fusion relevant conditions. The presence of the trapped helium bubbles in virgin-W and nano-W are examined by BSE imaging and XRD, respectively. The nanostructured tungsten shows improved structure and surface properties against Helium ion irradiation.

Thermal change of microstructure and mechanical properties of dispersion strengthened tungsten

Using a mechanical alloying-hot isostatic pressing method, new dispersion strengthened tungsten (DS-W) alloys containing TiO2 particles on grain boundaries have been developed for improvement of plasma facing materials for divertor components of fusion reactors. In this study, effects of the dispersion strengthening on the thermal change of microstructure and the thermal conductivity at high temperature were investigated. Within contrast to pure tungsten, the DS-W did not show significant microstructural change or a drastic decrease of the bending strength after annealing at high temperature.

Development of a solute and defect concentration dependant Ising model for the study of transmutation induced segregation in neutron irradiated W–(Re, Os) systems

In this study, radiation-induced precipitation of transmutation products is addressed via the development of a new solute and vacancy concentration dependant Ising model for the W–Re–Os system. This new model includes interactions between both Os and Re atoms with vacancies, thus facilitating more representative simulations of transmutation in fusion reactor components. Local solute concentration dependencies are introduced for the W–Re, W–Os and Re–Os pair interactions. The model correctly accounts for the repulsion between small clusters of vacancies and the attraction between larger clusters/voids, via the introduction of local vacancy concentration dependant interaction coefficients between pairs of atoms and vacancies. To parameterise the pair interactions between atoms and/or vacancies, the enthalpy of mixing, ΔH mix, for various configurations and solute/defect concentrations, was calculated using density functional theory, within 6 binary systems: W–Re, W–Os, Re–Os, W–vacancy, Re–vacancy and Os–vacancy. The new energy model was implemented into the SPPARKS Monte Carlo code, and successfully used to predict the formation of voids decorated with Re and Os solute atoms. Analysis suggests that there is a strong thermodynamic tendency for Os to bind to these voids with a comparatively weaker binding from Re atoms. The binding energies of various solute/vacancy clusters were calculated and showed that Re and Os solute atoms tend to stabilise small clusters of vacancies, increasing the attractive binding energy between the constituents.

Small-angle neutron scattering from CuCrZr coupons and components.

Small-angle neutron scattering (SANS) is performed to analyse the microstructural state of a reference CuCrZr material with carefully controlled heat treatments, small-scale manufacturing mock-ups of assemblies and high-heat-flux-exposed mock-ups for fusion reactor components. The information derived from the SANS data corresponds well to existing literature data based on microscopic-scale techniques, but is obtained at millimetre scale with minimal surface preparation. The manufacturing method and high-heat-flux testing conditions are confirmed to have little impact on the microstructural properties, demonstrating the validity of these treatments for scaled-up reactor components.

A numerical model for the recrystallization kinetics of tungsten monoblocks under cyclic heat loads

The recrystallization assisted microstructural changes during cyclic heat exposure are surmised detrimental to the thermal fatigue resistance of the tungsten monoblocks that cover the surface of the divertor components of a nuclear fusion reactor. A numerical framework to predict recrystallization in tungsten monoblocks during cyclic heat exposure is presented in this work. The framework is based on a thermal model coupled to a Johnson-Mehl-Avrami-Kolomogorov (JMAK) type recrystallization model. The influence of the initial (starting) microstructural state, i.e. the degree of deformation, on the extent of recrystallization, is considered through the activation energy for recrystallization. For a highly deformed microstructure, significantly more recrystallization is observed compared to self-diffusion driven recrystallization. The activation energy for recrystallization within the range 322 to 350 kJmol−1 during cyclic heat exposure of a monoblock is determined from the simulated recrystallization extent in relation to the experimental observations of differently processed monoblocks. The influence of the heating time per cycle for different activation energies on the recrystallization extent is also investigated. Instead of bulk recrystallization, only localized surface recrystallization is observed for a combination of a shorter heating time per cycle and a moderately deformed initial microstructure, which may contribute to improving the thermal fatigue resistance of the monoblocks.

Microstructural features in additively manufactured EUROFER97 components

By uncoupling the manufacturability from the design process, additive manufacturing of RAFM steels like EUROFER97 can open significant design freedom for components in fusion reactors, e.g., breeding blankets and divertor applications.As mechanical properties of RAFM steels are based on their specific microstructure and high number density of precipitates, good control of the microstructure of additively manufactured parts has to be ensured. As additive manufactured components are known to possess unique microstructures compared to EUROFER97 from standard technologies, the aim of this paper is to investigate additive manufactured EUROFER97 components and the influence of post-processing steps on their microstructure and in particular on the formation of precipitates.This paper covers the technological fabrication process of EUROFER97 by selective laser melting (SLM), including the production and chemical analysis of pre-alloyed EUROFER97 powder. In the initial state after fabrication, SLM-EUROFER97 components exhibit a bimodal, anisotropic microstructure with large ferritic grains instead of the desired homogeneous, fully martensitic microstructure. Further, neither M23C6 nor MX precipitates were detected in the as-built condition.A heat treatment including hot isostatic pressing, austenitization, quenching and tempering, allows achieving a fully martensitic, uniform microstructure as well as the formation of M23C6 and MX precipitates with similar size and volume density as in conventionally produced EUROFER97. In combination with the chemical analysis of the raw material, impurities stemming from the atomization process forming oxides were detected. However, due to their small size, those are not expected to influence the performance of SLM-processed EUROFER97. The comparison of the as-built condition and the condition after hot isostatic pressing and heat treatments reveals the necessity of appropriate post-processing to tailor the microstructure to one comparable to conventional EUROFER97.

New Method for Calculation of Radiation Defect Dipole Tensor and Its Application to Di-Interstitials in Copper

The effect of external and internal elastic strain fields on the anisotropic diffusion of radiation defects (RDs) can be taken into account if one knows the dipole tensor of saddle-point configurations of the diffusing RDs. It is usually calculated by molecular statics, since the insufficient accuracy of the available experimental techniques makes determining it experimentally difficult. However, for an RD with multiple crystallographically non-equivalent metastable and saddle-point configurations (as in the case of di-interstitials), the problem becomes practically unsolvable due to its complexity. In this paper, we used a different approach to solving this problem. The molecular dynamics (MD) method was used to calculate the strain dependences of the RD diffusion tensor for various types of strain states. These dependences were used to determine the dipole tensor of the effective RD saddle-point configuration, which takes into account the contributions of all real saddle-point configurations. The proposed approach was used for studying the diffusion characteristics of RDs, such as di-interstitials in FCC copper (used in plasma-facing components of fusion reactors under development). The effect of the external elastic field on the MD-calculated normalized diffusion tensor (ratio of the diffusion tensor to a third of its trace) of di-interstitials was fully consistent with analytical predictions based on the kinetic theory, the parameters of which were the components of the dipole tensors, including the range of non-linear dependence of the diffusion tensor on strains. The results obtained allowed for one to simulate the anisotropic diffusion of di-interstitials in external and internal elastic fields, and to take into account the contribution of di-interstitials to the radiation deformation of crystals. This contribution can be significant, as MD data on the primary radiation damage in copper shows that ~20% of self-interstitial atoms produced by cascades of atomic collisions are combined into di-interstitials.

Modelling the brittle-to-ductile transition of high-purity tungsten under neutron irradiation

Tungsten becomes significantly more brittle when exposed to neutron irradiation, which can be detrimental for the functionality of plasma facing components in nuclear fusion reactors. A predictive model of the mechanical behaviour of irradiated tungsten is therefore vital in order to assess how the brittle-to-ductile transition temperature evolves under fusion conditions. To this end, a cluster dynamics model is used to describe the defect production in high purity tungsten under neutron irradiation. The resulting defect concentrations provide input for a crystal plasticity model, capable of describing the influence of irradiation induced defects on the mechanical properties of high-purity tungsten. Simulation results reveal a striking match with experimental data in terms of yield stress at various temperatures and irradiation levels. Moreover, also the brittle-to-ductile transition temperature is adequately described by incorporating a probabilistic cleavage model.

Manufacturing and high heat flux testing of flat-type W/Cu/CuCrZr mock-up by HIP assisted brazing process

Divertor is one of the most challenging plasma facing components (PFCs) for fusion reactors both from the design and fabrication technology point of view. In this work, the new flat-type W/Cu/CuCrZr mock-ups were fabricated using hot isostatic pressing (HIP) and optimized brazing techniques. High heat flux (HHF) tests of the mock-up were performed on a comprehensive experimental platform to assess the heat removal capability and thermal fatigue resistance. The results indicate that the mock-up successfully endured the cyclic absorbed heat flux at 20 MW m−2 for 1000 cycles with an inlet pressure of 1 MPa, an inlet water temperature of 20 °C and a velocity of 6.88 m s−1. The peak temperature of the tungsten (W) surface reached 926 °C at the 1000th cycle, which is less than the recrystallization temperature. In addition, no substantial melting at the loaded W surfaces, no detachment of any W/Cu units and no occurrence of any ‘hot spot’ during thermal fatigue cycling were observed. These results indicate that HIP assisted brazing technique is a powerful method for fabricating PFCs employed in HHF tests, and the flat-type W/Cu/CuCrZr structure is very promising to be used to upgrade the lower divertor of Experimental Advanced Superconducting Tokamak(EAST) and manufacture the divertor for China Fusion Engineering Test Reactor(CFETR).

New insights into microstructure of neutron-irradiated tungsten

The development of appropriate materials for fusion reactors that can sustain high neutron fluence at elevated temperatures remains a great challenge. Tungsten is one of the promising candidate materials for plasma-facing components of future fusion reactors, due to several favorable properties as for example a high melting point, a high sputtering resistivity, and a low coefficient of thermal expansion. The microstructural details of a tungsten sample with a 1.25 dpa (displacements per atom) damage dose after neutron irradiation at 800 °C were examined by transmission electron microscopy. Three types of radiation-induced defects were observed, analyzed and characterized: (1) voids with sizes ranging from 10 to 65 nm, (2) dislocation loops with a size of up to 10 nm and (3) W–Re–Os containing σ- and χ-type precipitates. The distribution of voids as well as the nature of the occurring dislocation loops were studied in detail. In addition, nano-chemical analyses revealed that the σ- and χ-type precipitates, which are sometimes attached to voids, are surrounded by a solid solution cloud enriched with Re. For the first time the crystallographic orientation relationship of the σ- and χ-phases to the W-matrix was specified. Furthermore, electron energy-loss spectroscopy could not unambiguously verify the presence of He within individual voids.

Tailoring microstructure in sintered Cu-Cr-Nb-Zr alloys for fusion components

High temperature, creep resistant heat sink materials represent a critical need for plasma facing components in future fusion reactors. In this study, we employ direct current sintering (often referred to as spark plasma sintering) to produce a Cu-Cr-Nb-Zr (CCNZ) alloy from gas-atomized feedstock powder with tailored precipitate distributions for enhanced stability and creep resistance. Microstructure was characterized by synchrotron X-ray diffraction, small angle X-ray scattering, and electron microscopy techniques. We report a multi-modal precipitate distribution containing submicron Cr (~493 nm) and Cr2Nb (~90 nm) precipitates at grain boundaries and a high density of nanoscale Cr (~8 nm) precipitates homogeneously distributed through the Cu matrix. By comparing the as-sintered and aged microstructures, precipitation kinetics are discussed in the context of dislocation networks due to the high sintering pressures biasing precipitate formation and the role of subsequent recovery and recrystallization. Due to the presence of the multi-modal precipitate distribution, the sintered CCNZ alloy exhibited a high hardness of 133.2 HV while retaining an appreciable thermal conductivity of 298.4 W/m·K and electrical conductivity of 74.6% relative to the International Annealed Copper Standard.

Preparation of a water activation experiment at JET to support ITER

Fast neutron induced water activation will result in significant additional nuclear heating loads to cryogenic components in the ITER fusion reactor during operation, which are considered as an important issue in the ITER design. A water activation experiment has been proposed at the Joint European Torus (JET) which would enable measurements of radiation dose rates and 16N concentrations in water, activated in the most representative fusion environment achievable, and extrapolability to ITER conditions. This paper presents a study aimed at establishing the experiment feasibility, in terms of the identification and selection of an experimental location at JET, and the definition of an experimental setup enabling accurate measurements of the quantities of interest. A suitable experimental location in JET was identified in the basement of the JET Torus Hall. The expected 16O(n,p)16N activation rates in the cooling water in the JET tokamak were determined through Monte Carlo particle transport calculations and the induced 16N activity in the water was propagated to the experimental location. Subsequent calculations were performed to determine the expected detector response to 16N γ rays for three different JET pulse scenarios. Two scintillator γ detectors were calibrated with standard γ calibration sources and tested with a 244Cm/13C neutron source, also emitting 6.13 MeV γ rays, corresponding to the 16N γ ray energy. The presented work demonstrates the feasibility of a water activation experiment at JET and provides valuable information for the experiment design.

Recrystallization-mediated crack initiation in tungsten under simultaneous high-flux hydrogen plasma loads and high-cycle transient heating

Tungsten and tungsten-based alloys are the leading material choices for the divertor plasma facing components (PFCs) in future fusion reactors. Recrystallization may occur when they undergo high heat loads, drastically modifying the predesigned grain structures and the associated desired mechanical properties. However, the influence of recrystallization on the thermal fatigue behavior of tungsten PFCs still remains unclear. In this study, ITER-grade tungsten was simultaneously exposed to a high-flux hydrogen plasma (∼5 × 1024 m−2 s−1) and high-cycle (104–105) transient heat loads in the linear plasma device Magnum-PSI. By correlating the surface temperature distribution, obtained by analyzing temperature-, wavelength-, and surface-dependent emissivity, and the surface modifications of the plasma exposed specimens, the crack initiation heat flux factor threshold was found to be ∼2 MW m−2 s0.5 (equivalently, ∼0.07 MJ m−2 for a 1 ms pulse). Based on electron backscatter diffraction analyses of cross-sections near the crack initiation sites, faster recrystallization kinetics near the surface compared to literature was observed and the surface cracks preferentially initiated at high angle grains boundaries (HAGBs). Upon recrystallization, the yield strength decreases which entails increasing cyclic plastic strains. The HAGBs fraction is increased, which constrains the transfer of plastic strains at grain boundaries. The recrystallization decreases the dislocation density, which promotes heterogeneous deformation. All these mechanisms explain the reduced crack initiation threshold of recrystallized tungsten compared to its as-received counterpart. The results provide new insights into the structural failure mechanisms in tungsten PFCs exposed to extreme fusion plasmas.