Fusion Reactor Materials Research Articles

Superconducting materials for fusion reactors: A brief review

The introduction of high-temperature superconductors and, in particular, cuprate oxides such as REBa2Cu3O7–δ (where RE = Y, lanthanides) has led to developments in applied physics and technology. These materials have high technological potential for more economical, smaller magnetically confined fusion reactors due to the high field strengths. The introduction of high-temperature superconductors in the form of tapes in fusion reactors is not technologically trivial as these reactors eject high-energy neutrons that can damage the material and hinder its operation. In the present review, we consider the technological issues that need to be addressed for the implementation of REBa2Cu3O7–δ in fusion reactors.

The effect of CD4 injection on WD molecule sputtering at the divertor target in EAST

Tungsten (W) is one of the preferred plasma-facing materials for future fusion reactors because of the advantages of high melting point, low physical sputtering rate and low fuel retention. The lifetime and service performance of W materials are limited by erosion processes. In recent years, tungsten deuteride molecular (WD) sputtering has been discovered in tokamak. Studying WD sputtering helps to better understand the erosion of tungsten materials in tokamak. EAST adopts the method of injecting impurity gas in the divertor to alleviate the particle flux and heat flux bombarding the tungsten divertor, and CD4 is one of the impurities used. CD4 injection experiments on EAST show that after CD4 is injected from the upper outer (UO) divertor target, despite that the electron temperature (Te) around the strike point on the UO divertor target drops to about 8 eV, strong WD sputtering occurs in the far scrape-off layer (SOL) region. The divertor probe data shows that after CD4 injection, the particle flux from the upstream plasma onto the divertor split into two bands, creating a new particle flux strike point in the far SOL region. The energy of incident particles at the original strike point is significantly reduced, while the energy of incident particles at the far SOL region is enhanced. The edge electron density profile at mid-plane indicates that CD4 injection causes an increase of density in the SOL, which could improve the coupling of Lower-Hybrid Wave (LHW) with the edge plasma as evidenced by the decreasing reflected power of LHW. The increase in absorbed power of LHW could change the particle and energy transport in the edge plasma, thus plays an important role in the enhanced WD sputtering in the far SOL region after CD4 injection.

Research on the characteristics of the Helium plasma beam in HIT-PSI

As a high heat flux linear plasma device designed for studying divertor materials in future fusion reactors, HIT-PSI(Plasma Surface Interaction device at Harbin Institute of Technology) has been successfully constructed and has maintained stable operation since its completion. The characteristics of He plasma beams in HIT-PSI are investigated by emission spectroscopy and an infrared camera, with preliminary irradiation experiments conducted by bombarding tungsten with the beam. For relatively conservative discharge parameters, HIT-PSI achieved a steady-state heat flux capacity of ∼40 MW/m2 using infrared measurements, with the full width at half maximum (FWHM) of the heat flux beam reaching 4 mm. These characteristics make HIT-PSI an advanced platform for testing divertor materials and plasma-facing components, providing essential experimental supports for research and development of high-performance divertor materials.

Assessment of the impact behavior of CLF-1 steel with Charpy V-notch testing and miniature Charpy V-notch testing

China Low-activation Ferrite steel (CLF-1), as one of the structural materials developed in China for fusion reactors, is a Reduced activation Ferritic/Martensitic (RAFM) steel modeled after the traditional T91 steel (9Cr-1Mo-0.2V-0.08Nb). For fusion reactor materials, the Charpy impact testing is an important method to assess their notch sensitivity. In this study, Charpy impact tests were conducted on CLF-1 steel using Charpy V-notch specimens (CVN) and Miniature Charpy V-notch specimens (Kleinstprobe, KLST) to assess the impact behavior of CLF-1 steel, and data normalization was performed on the KLST specimen. The results show that as the specimen size of CLF-1 steel decreased, the curves of impact absorbed energy, lateral expansion, and shear fracture appearance shifted towards lower temperatures, and the impact absorbed energy significantly decreased. Both the macroscopic and microscopic characteristics of the specimens obtained from the experiments indicated that KLST specimens could achieve the same fracture characteristics as CVN specimens. Additionally, at the same experimental temperature, KLST specimens exhibited a higher proportion of ductile fracture regions.

Inhomogeneous distribution of oxides in various 9Cr ODS alloys and their mechanical performance at room temperature and 650 °C

Oxide dispersion strengthened (ODS) steels have been considered as promising structural materials for fusion reactors. In fabricating ODS steels, homogenous distribution of nano-sized oxides is always expected. In this work, Ti and C were respectively added into the 9Cr alloy with 2 wt% Mn, and the effects of individual Ti addition on microstructure and mechanical performance of ODS alloys are studied. Regularly distributed oxides were revealed in the 9Cr alloys fabricated by hot isostatic pressing (HIP), irrespective of alloying compositions, which indicates the strong interaction between oxides and phase interface. Alloying elements within the matrix can be incorporated into these regularly distributed oxides. Oxide nanoparticles will affect the phase transformation behaviors and the microstructure of as-HIPed 9Cr alloys without oxides is different from that of 9Cr ODS alloys. By performing post-HIP heat treatments, the matrix conditions of 9Cr ODS alloys can be adjusted, and ferritic and martensitic 9Cr ODS alloys are respectively obtained. With tensile tests at room temperature and 650 °C, effects of oxides and matrix conditions on 9Cr alloys were evaluated. It is found that individual Ti addition has a softening effect on the 9Cr alloys, despite that whether oxides are added. Simultaneous addition of Ti and C can improve the strength of 9Cr alloys evidently. Poor ductility at 650 °C is a critical issue for 9Cr ODS alloys, while the 9Cr alloys without oxides have favorable combination of strength and ductility at both room temperature and 650 °C.

Variable-energy cocktail beam technology for investigating synergistic damage in nuclear materials on LEAF platform

This study addresses the critical issue of synergistic radiation damage in structural materials of fusion reactors, focusing on the interaction between the displacement defects and transmutation-produced hydrogen and helium. These effects are simulated and investigated by employing the advanced multi-beam ion implantation capabilities of the LEAF (Low Energy high intensity highly charged ion Accelerator Facility) platform. Firstly, high-intensity cocktail beams, such as “4He+ and 56Fe14+" and “4He+ and 58Ni15+", are generated and characterized successfully. Then, a complex radiation environment is mimicked within the fusion reactors by applying variable-energy irradiation. Secondly, similar penetration depths for different ions, which are crucial for studying synergistic effects, are obtained by precisely controlling the energy of the cocktail beams through the innovative energy modulation system of the LEAF platform. Finally, the post-irradiation analyses, performed by using the transmission electron microscopy (TEM) and nanoindentation, revealed distinct microstructural changes and alterations in material properties, providing insights into the degradation mechanisms under irradiation. This work not only generates diverse and high-intensity “cocktail” ion beams but also achieves rapid energy switching of the beams. Further, the work is expected to pave the way for the implementation of a novel multi-beam irradiation technique in advanced heavy-ion linear accelerators, and also to provide innovative experimental methods and technical support for studying the synergistic effects of nuclear materials.

Influence of temperature gradients on helium diffusion and clustering in tungsten: A molecular dynamics study

Tungsten and its alloys, used as plasma-facing materials in fusion reactors, endure high-flux, low-energy helium ion irradiation and temperature gradients induced by thermal load and shocks. This study utilizes molecular dynamics (MD) simulations to explore the diffusion and clustering behavior of helium (He) in tungsten (W) under temperature gradients. The migration behavior of isolated helium atoms, small helium clusters, and helium self-interstitial atom (He-SIA) clusters within tungsten is analyzed. It is revealed that both helium and He-SIA clusters tend to migrate towards higher temperature regions, exhibiting negative thermophoresis. The nucleation of helium clusters, the formation of Frenkel pairs, and the interactions between self-interstitial clusters and helium clusters are also investigated. The results indicate that helium concentration significantly impacts He nucleation behavior. Directional diffusion may lead to areas of high concentration over time, potentially promoting the formation of He clusters and bubbles. These insights enhance the understanding of tungsten's performance and durability as a plasma-facing component in fusion reactors.

Laser-Induced Ablation and Desorption of Deuterium-Containing Tungsten Films

The laser-induced desorption (LID) and laser-induced ablation (LIA) methods are compared with each other regarding the possibility of measurements an absolute quantitative analysis of hydrogen isotopes content in first wall materials of fusion reactors. Deuterium containing tungsten films with a thickness of 300–400 nm on a silicon substrate were used as model samples. To implement the LID, the samples were irradiated with laser pulses with a duration of 200 microseconds and an energy density of 50–150 J/cm2, for LIA – 12 ns and 5–15 J/cm2. The registration of residual gases was carried out by quadrupole mass spectrometry. Computer simulation of laser pulse heating was performed for the LID process. The simulation results and experimental data showed that heating at an energy density of 100–150 J/cm2 is sufficient to degas tungsten films of the studied thickness. A comparison of the amount of desorbed deuterium in the LID (150 J/cm2) and LIA (15 J/cm2) modes shows that it is identical within the measurement error and is equal to 4.15±0.15·1014 cm-2.

Evaluation of localized mechanical properties of modified N50 welded joints at cryogenic temperature through a digital image correlation technique

The modified N50 austenitic stainless steel has been determined to be applied as the superconducting magnet supporting material of the Chinese Fusion Engineering Test Reactor (CFETR), especially the jacket material of the CICCs of both the TF coils and the CS coils and the case material of the TF coils, which inevitably requires welding connection. The welding process generally results in welding defects as well as degradation of cryogenic mechanical properties of metallurgical inhomogeneous zones. However, limited research has been reported to demonstrate the distinct localized performance of these zones at cryogenic temperature. The present paper presents an approach to characterize localized mechanical properties of the modified N50 welds at both room temperature (300 K) and cryogenic temperature (6 K). We identified the welding zone (WZ) through metallographic analysis and then the boundaries of the heat-affected zone (HAZ) were recognized by hardness measurements at room temperature. Finally, the localized optical digital image correlation (DIC) technique was used to obtain the localized yield strength Rp0.2 and percentage extension e of the WZ, the HAZ, and the base metal (BM) at both room and cryogenic temperature, of which the observed trend is verified through comparison with that of hardness measurements at room temperature. The technique proposed provides the localized measurement of cryogenic tensile properties of weld metals and also presents experimental data on different zones of the modified N50 welded joints as a jacket material for the next-generation fusion reactors.

Influence of suppressed blistering by heavy ion pre-damage on deuterium retention in tungsten

Surface damage and fuel retention are one of the major threats to the performance of plasma-facing materials (PFMs) in ITER and future fusion reactors. This work aims to investigate the influence of suppressed blistering by heavy ion pre-damage on deuterium (D) retention in tungsten (W). Recrystallized W samples were irradiated with 3.5 MeV iron (Fe13+) ions at room temperature to create displacement damage with a peak damage level of 0.1 dpa. Afterwards, a series of low-energy (38 eV) D plasma exposures were performed at 500 K. Three exposure fluences below and above the blistering threshold (5 × 1024 ∼ 3 × 1025 D m−2) of the pre-damaged W are selected to decouple and couple the damage-induced defects and blistering-induced defects, respectively. Surface observations show that no blisters are formed in un-damaged W after low-fluence D exposure (5 × 1024 D m−2), whereas severe blistering (surface coverage ratio: 34.2 %) occurs in the high-fluence case (3 × 1025 D m−2). In contrast, only a small number of blisters (4.2 %) are formed in Fe-damaged W when D fluence reaches 3 × 1025 D m−2. Moreover, Fe pre-damage increases D retention by a factor of 3.33 and 1.20 at the low-fluence (5 × 1024 D m−2) and medium-fluence (1 × 1025 D m−2) D exposure, respectively. While in the high-fluence case (3 × 1025 D m−2), the enhancement effect of D retention in Fe-damaged W is significantly weakened, such that retention is lower for damaged W, probably due to the significant suppression effect on surface blistering and its accompanying defect formation. This work highlights the suppressed blistering-induced D retention by pre-existing damage in W.

Evaluating the characteristics of functionally graded plate: A Review study

Functionally graded materials (FGMs) were developed as heat barrier materials for aircraft structural applications and fusion reactors. They are now being designed for broad usage as structural components in highly harsh settings. FGMs are one of the most efficient and effective materials for achieving long-term industrial development. The present article examines FGMs in general, their applications, and existing processing approaches. It also demonstrates the behavior of FGM-generated plates under five distinct conditions: buckling, bending, vibration, theories, and loadings.

On the fabrication of ultrafine-grained potassium-doped tungsten: Mechanical milling and spark plasma sintering of K-doped W powder prepared by evaporation-condensation method

Bubble strengthened materials have received lots of attention in recent years, and among them, potassium-doped tungsten (W-K) is extremely desirable for very high-temperature applications such as plasma-facing materials in fusion reactors. However, it is difficult to fabricate ultrafine-grained W-K bulk materials by traditional methods. In this study, a novel preparation method of ultra-fine-grained bubble-strengthened material is presented. Unlike traditional routes, W-K mixed powder was prepared by evaporating and condensing pure K on the surface of nanoscale W powder. Two W-K mixing methods using this evaporation-condensation route were developed. Mechanical milling process combined with regular SPS sintering were carried out on the W-K mixed powder, which successfully manufactured ultrafine-grained (0.9 μm) W-K bulk material with a high density (96.4%). This study can provide important reference for the fabrication of other bubble-strengthened materials.

Influence of dimensional deviation and thickness non-uniformity on the small-size specimen tensile results of CLF-1 steel

Tensile testing of small-size specimens has been widely used to evaluate the influence of neutron irradiation on the mechanical properties for structural materials of fusion reactors. In recent years, researchers have been trying to standardize small-size specimen tensile test method. In the present paper, we have investigated the influence of dimensional deviation and thickness non-uniformity on the tensile results of small-size specimen by both experiments and finite element simulations, based on one of the candidate structural materials of fusion reactor, i.e. CLF-1 steel. It was found that under the conditions of small grain size, small metallurgical defect size and suitable specimen preparation, the tensile results of SS-J3 specimen could be close to the results of large-size specimen. Slightly higher yield and ultimate strengths for small specimens may be caused by the surface treatment of the wheel grinding. The repeatability of ultimate strength was better than that of yield strength, and the repeatability of elongation was worse than that of strength. For tensile tests on eight SS-J3 specimens, the maximum difference in total elongation was about 3 %. With ±0.1 mm deviation of thickness and parallel width, the maximum variation in total elongation was about 1.8 %. The ductility properties were sensitive to specimen's thickness non-uniformity, the measured uniform and total elongations would obviously decrease even with 0.01 mm thickness non-uniformity.

The diffusion and desorption of tritium on Li2TiO3 (-133) surface from first-principles calculations

Li2TiO3 is one of the most promising solid breeder materials for nuclear fusion reactor. Diffusion on the surface is a vital step of the tritium release process. In this work, the diffusion behavior of tritium atoms on the (-133) surface and the desorption of T2 molecules from the (-133) surface are systematically investigated. Possible adsorption sites, local diffusion pathways, and their corresponding activation energies have been identified using first-principles calculations and the climbing image nudged elastic band (CI-NEB) method. The diffusivity and effective activation energy for tritium diffusion on the (-133) surface have been determined using kinetic Monte Carlo (KMC) simulations. The activation energy for T2 desorption from the (-133) surface has also been investigated. The effective activation energy for tritium diffusion on the (-133) surface is found to be 1.06 eV, whereas the activation energy for T2 molecule desorption is 1.46 eV. The (-133) surface exhibits a higher activation energy for tritium diffusion but a lower activation energy for T2 desorption compared to the (001) surface.

Innovative construction of high-strength tungsten-copper joint through laser texture and nano-activation

Tungsten-copper (W-Cu) joints hold immense promise as plasma-facing materials in fusion reactors. However, the inherent immiscibility of W-Cu poses significant challenges in joint fabrication. Here, we introduce an innovative methodology that incorporates laser texture, W surface nano-activation, and subsequent diffusion bonding to fabricate W-Cu joints. Remarkably, the joints achieved exhibit unparalleled mechanical properties, with a peak tensile strength of 201 MPa and a shear strength of 141 MPa, surpassing previously reported W-Cu joints. To gain insights into the underlying mechanisms, we conducted a multiscale analysis utilizing scanning electron microscopy (SEM), high-resolution transmission electron microscopy (HRTEM), and density-functional theory (DFT) calculations. Our findings reveal a unique embedded structure and a metallurgically bonded interface at the W-Cu junction. Furthermore, the diffusion zone at the interface exhibits a fascinating hybrid crystal structure, maintaining a body-centered cubic (BCC) structure in certain regions while displaying a tetragonal crystal structure (with lattice parameters a=b=2.8617, c=3.44) in others. This tetragonal crystal structure formation within the W-Cu diffusion zone remains unexplored in previous literature. In summary, this novel W-Cu bonding approach not only offers a cutting-edge solution for modern manufacturing and fusion energy applications but also lays a solid theoretical foundation for understanding the intricate microstructure-property relationships in W-Cu systems.

A study on neutron emission for proton-induced reactions in 90Zr, 91Zr, and 115In isotopes

Fusion energy heralds the potential of a transformative era, offering a significant solution to global challenges such as climate change, ozone depletion and environmental pollution. Despite its promising prospects, the commercialization of fusion faces several challenges, including high temperature, pressure, plasma stability, fuel supply, costs, etc. It is important to effectively analyze material behavior under plasma conditions, especially in environments where fusion reactions produce high-energy particles such as neutrons. This study investigates the angle-dependent neutron production mechanisms of proton-induced reactions involving the isotopes 90Zr, 91Zr and 115In, which are widely used in fusion reactor materials. Using the Monte Carlo codes PHITS 3.32 and FLUKA, as well as the TALYS 1.96 code, double differential cross-section calculations for neutron emission were performed considering various angles. The research contributes to a broader understanding of fusion processes by providing insights into the behavior of these isotopes under proton-induced reactions.

Sputtering fabrication of isolated W nanocolumns: A possible alternative as plasma facing material for nuclear fusion reactors

Isolated tungsten nanocolumns (W-NCs) have been reported to exhibit a higher radiation resistance than coarse grained W samples, under radiation conditions similar to those that plasma facing materials would face in both magnetic and inertial confinement nuclear fusion approaches (MCF and ICF, respectively). This is so, because their ability to release He via the free surfaces and, the strong reduction of the sputtering yield under KeV Ar and D irradiation as well as, flattening of its angular dependence. The latter is very important for the divertor location in MCF.In this work, we investigate the capabilities of sputtering (an easy control, environmentally friendly, versatile, scalable and low-cost technique) to fabricate isolated W nanocolumns. We study the influence of sputtering parameters (plasma power and deposition angle) on the morphology, density and microstructure of the deposited coatings. Results show that stable α-phase 3D isolated W-NCs are produced at low plasma power (<100 W) and high deposition angles (θ ≥ 75°).

Lithium Neutron Cross Sections During the Manhattan Project and the Quest for the H-Bomb

Neutron cross sections of the stable lithium isotopes 6Li and 7Li were of interest in the 1940s and 1950s in part because of their reactions, which form tritium using moderated neutrons on 6Li and higher-energy neutrons on either isotope. Lithium remains of interest today for use as a blanket and shielding material in fusion reactors, where it can be used to breed tritium for a self-sustaining fuel cycle. During the Manhattan Project, the resonance in the 6Li(n,t) reaction was discovered and later became important for enhancing tritium production for nuclear technologies. The dominant natural isotope 7Li was and remains of interest because of the expense of enriching 6Li. It has been oft reported that the 1954 Castle Bravo nuclear test had a yield twice as large as expected because the nuclear explosive device designers had not properly accounted for the benefits from the 7Li isotope in the fuel; we note that this explanation is false.

The Irradiation Effects in Ferritic, Ferritic-Martensitic and Austenitic Oxide Dispersion Strengthened Alloys: A Review.

High-performance structural materials (HPSMs) are needed for the successful and safe design of fission and fusion reactors. Their operation is associated with unprecedented fluxes of high-energy neutrons and thermomechanical loadings. In fission reactors, HPSMs are used, e.g., for fuel claddings, core internal structural components and reactor pressure vessels. Even stronger requirements are expected for fourth-generation supercritical water fission reactors, with a particular focus on the HPSM's corrosion resistance. The first wall and blanket structural materials in fusion reactors are subjected not only to high energy neutron irradiation, but also to strong mechanical, heat and electromagnetic loadings. This paper presents a historical and state-of-the-art summary focused on the properties and application potential of irradiation-resistant alloys predominantly strengthened by an oxide dispersion. These alloys are categorized according to their matrix as ferritic, ferritic-martensitic and austenitic. Low void swelling, high-temperature He embrittlement, thermal and irradiation hardening and creep are typical phenomena most usually studied in ferritic and ferritic martensitic oxide dispersion strengthened (ODS) alloys. In contrast, austenitic ODS alloys exhibit an increased corrosion and oxidation resistance and a higher creep resistance at elevated temperatures. This is why the advantages and drawbacks of each matrix-type ODS are discussed in this paper.

Effect of boron on the microstructure and performance of HIP-prepared high boron steels

High boron steels containing 2 wt% and 3.3 wt% boron were fabricated using hot isostatic pressing (HIP). The enhanced shielding performance of the steel with higher boron content was quantitatively evaluated through Monte Carlo simulations using software Fluka. The steels were studied with XRD and the phase ratio were calculated using Rietveld refinement method. The microstructure of the materials was investigated using electron backscatter diffraction (EBSD). Comprehensive thermal property measurements, complemented by Hamilton thermal conductivity model analysis, were conducted on high boron steels. The results revealed that the decreased overall thermal conductivity in these materials is primarily attributed to two factors: the increased volume fraction of the boride phase and the enhanced interconnectivity among boride particles. These findings provide crucial insights into the thermal behavior of high boron steels. Room temperature tensile tests indicate that high boron content can cause the fracture mode of the material to transition from ductile to brittle. This study provides a detailed analysis of the effects of boron content on various properties of high boron steels, which can serve as shielding materials in fusion reactors. Advances were proposed for optimizing the thermal performance of high boron steels in this study.