Fusion Reactor Research Articles

Demonstration of aneutronic p-11B reaction in a magnetic confinement device

Abstract Aneutronic fusion using commonly available fuel such as hydrogen and boron 11 (11B) is one of the most attractive potential energy sources. On the other hand, it requires 30 times higher temperature than deuterium-tritium fusion in a thermonuclear fusion reactor condition. Development of techniques to realize its potential for the experimental capability to produce proton-boron 11 (p-11B) fusion in the magnetically confined fusion device using neutral beam injection is desired. Here we report clear experimental exploration and measurements of p-11B fusion reactions supported by intense hydrogen beams and impurity powder dropper installed in the magnetic confinement plasma Large Helical Device. We measured a significant amount of fusion alpha particle emission using a custom designed alpha particle detector based on a passivated implanted planar silicon detector. Intense negative-ion-based hydrogen beam injectors created a large population of up to 160 keV energetic protons to react with the boron-injected plasma. The p-11B alpha particles having MeV energy were measured with the alpha particle detector which gave a fusion rate in a good agreement with the global p-11B alpha emission rate calculated based on classical confinement of energetic proton, using experimentally obtained plasma parameters.

J-TEXT achievements in turbulence and transport in support of future device/reactor

Abstract Following the reconstruction of the TEXT tokamak in China at Huazhong University of Science and Technology, renamed as J-TEXT, a plethora of experimental and theoretical investigations have been conducted to elucidate the intricacies of turbulent transport within the tokamak configuration. These endeavors encompass not only the J-TEXT device's experimental advancements but also delve into critical issues pertinent to the optimization of future fusion devices and reactors. The researches include the topics about the suppression of turbulence, flow drive and damping, density limit, nonlocal transport, intrinsic toroidal flow, turbulence and flow with magnetic island, turbulent transport in stochastic layer, turbulence and zonal flow with energetic particles or helium ash. Several important achievements have been got in the last few years, which will be further elaborated upon in this comprehensive review.

Holistic Hydraulic Simulation for Pebble Bed Using Porous Media Approach

The porous media approach is regarded as an appropriate methodology for hydraulic simulations of complex pebble beds in fusion reactors. In order to determine the parameters (permeability α and inertial loss coefficient C) of the porous media accurately, two methods are proposed: the correction method and the fitting method. In this paper, a single-channel model with sequentially packed pebbles is constructed in order to obtain the pressure drop gradient against superficial velocities. Two methods, the correction method and fitting method, are employed to determine the permeability and inertial loss coefficient, and the results are evaluated with comparisons. Based on the results, both the correction method and fitting method are deemed feasible for the parameter determinations. In consideration of the consumption of resources and time for simulation, the fitting method is recommended during the preliminary design phase to shorten the duration of design, while the correction method is suggested to obtain precise results when the design is accomplished. Both of the methods would be evaluated with the data obtained from experiments in the future.

Evaluating hybrid nuclear-thermal power systems for reduced carbon footprint: A comparative analysis

The article provides an estimated substantiation of the concept of expanding the fuel and energy base and reducing the volume of greenhouse gas emissions with the joint use of organic and nuclear fuel in the structure of hybrid nuclear-thermal power plants. It is shown that the technological scheme (the production of saturated steam in a nuclear steam-producing plant - NPPU, and then overheating to the maximum achieved parameters in the thermal power industry due to organic fuel) leads to a decrease in the specific consumption of gas fuel for the production of electrical energy by 25–30 %, emissions of combustion products by 2.5 times (compared to thermal power plants), consumption of industrial water (compared to nuclear power plants) - by 1.5 times. It is indicated that it is possible to expand the fuel base of modern nuclear energy with thermal neutron reactors through the use of technologies mastered (breeder reactors) and promising (hybrid thermonuclear reactors). The fundamental advantage of the considered complex - JPPP-TPP - its feasibility in a short period of time with a significant reduction in the carbon footprint per unit of energy produced, which is an important condition for the sustainable growth of energy production while maintaining the ecological balance in the geosphere.

A Generic Implementation of the D1S Methodology for Dose Rate Assessment by Monte Carlo Codes in Fusion Applications

Dose rate assessment is of utmost importance in fusion reactors because of the maintenance operations that these facilities will require. The standard way to perform this assessment in ITER is use of the Direct 1-Step (D1S) methodology, which consists of performing neutron transport simulation, material activation, and decay photon transport simulation in one step and thus in one simulation only. Usually, implementation of the D1S methodology requires changing the source files in a reference Monte Carlo code. The purpose of the present work is to develop an easy-to-implement method for Monte Carlo codes to help calculate the dose rate in fusion reactors. To do so, the proposal is to act on nuclear data only and not on source files of the simulation codes. This is done by replacing prompt photon production in evaluation files by suitable decay photon production, taking into account radioactive decays of radionuclides and irradiation history. This study was to be applied first on the TRIPOLI-4® Monte Carlo code for the sake of simplicity. TRIPOLI-4 is the reference code for particle transport at CEA. The verification and validation process relies first on a comparison to the reference Rigorous 2-Step (R2S) methodology and then on an experiment, the so-called Frascati Neutron Generator (FNG) dose experiment. The nuclear database to be changed is of the ENDF-6 format, a recurrent format in neutronics studies. The analysis studied both neutron and photon responses to check if the simulation was performing normally in a physical way and to compare the results with references provided by simulations based on the R2S methodology or by the FNG dose experiment. The simulations have proven to be in good agreement with the experimental results.

High-temperature corrosion testing of titanium beryllides in the presence of water vapor and oxygen

Beryllium-based intermetallics are promising materials for the blankets of future fusion reactors and have potential applications in other areas of the nuclear industry, such as fission reactor reflectors and space technology. Understanding the high-temperature corrosion behavior of these materials in a noble gas medium containing chemically active impurities is essential for evaluating their suitability and guiding their application.This study investigates the high-temperature corrosion of titanium beryllide Be12Ti in the form of plate and grinded samples, produced by JSC “Ulba Metallurgical Plant” (Ust-Kamenogorsk, Kazakhstan). The corrosion tests were conducted under non-isothermal vapor-gas mixture (Ar + D2O or Ar + H2O flowing atmospheres) purging conditions using thermogravimetric (TG) analysis, differential scanning calorimetry (DSC), and mass-spectrometry of the gas phase.As a result of the corrosion tests, new experimental data on thermal effects have been obtained, describing the corrosion processes of Be12Ti samples across a wide range of temperatures and various heating rates in the presence of water vapor in the purge gas. The dependencies of sample mass change under heating conditions have been determined, and the characterization results of the samples before and after high-temperature corrosion tests are presented.Corrosion of titanium beryllides, both for Be12Ti plate and grinded samples, follows similar mechanisms. At around 500 °C, the mass of the samples begins to increase, and hydrogen isotopes are released. The test results indicate that corrosion of titanium beryllides with varying surface inhomogeneities proceeds similarly within the temperature range of 500–900 °C, showing a linear dependence on temperature.The results revealed significant insights into the oxidation mechanisms and the formation of corrosion products, which are crucial for optimizing the material's performance in fusion reactor environments.

CIGALE: An innovative gas neutralizer based high efficiency neutral beam injector concept for future fusion reactors

Abstract This paper outlines the main features of a new high efficiency > 62%) high power (~20 MW D0) Neutral Beam concept based on pragmatic solutions suitable with the reactor requirements. The injector is modular (several beamlines in parallel) with independent ion sources referenced to the ground potential and gas neutralizers held at +1 MV. This topology leads to numerous simplifications; it overcomes the main issues of conventional NB systems, such as the complex 1MV electrical setup, the difficult ion source remote maintenance, the high caesium consumption. The other key parameter is the gas neutralization concept which minimizes the amount of gas by operating at a low gas target and low neutralizer duct conductance. The implementation of an energy recovery system for the residual 1 MeV D- is essential to attain a high wall-plug efficiency. These specific features require thin laminar D- beams provided by a pre-acceleration up to 100 keV in slotted grid apertures to form thin blade-like beamlets, followed by the post-acceleration to 1 MeV by merging the beamlets in a single beam in five gaps (+200 kV per gap). All these specific aspects minimize the beams losses and thermal loads along the beamline and enhance the injector reliability and availability.

Facile Synthesis of U2Ti Intermetallic by Direct Electrochemical Reduction of UO2-TiO2 Composite in LiCl-Li2O Melt

Abstract Alloys of U with Ti are of importance in nuclear industry as fuel for Gen IV fast reactors, hydrogen isotope storage medium for fusion reactors, super conductors, and also as corrosion resistant material for use in various applications. Here, preparation of U2Ti intermetallic compound was investigated by the direct electrochemical reduction of mixed oxide of UO2-TiO2 in LiCl-0.5% Li2O molten salt at 650oC. The mixed oxide pellet of UO2-TiO2 sintered at 1500oC was found to be a mixture of UTi2O6 and UO2 as evidenced by X-ray diffraction (XRD) analysis. Direct electro-lithiothermic reduction of UO2, TiO2, and a mixture of sintered UO2-TiO2 and UTi2O6 coupled with cyclic voltammetry of these oxides in the melt was performed to understand the electro-reduction mechanism. Potentials of reduction of these oxides in the melt, w.r.t Ni/NiO reference electrode, obtained by analysis of CV data of these oxides contained in metallic cavity electrodes and XRD analysis of partially electro-reduced oxides were used to arrive at the electro-reduction mechanism. Results indicate that U2Ti can be prepared conveniently by the electro-lithiothermic reduction of sintered pellet of UO2-TiO2 cathode by constant current electrolysis in a two-electrode cell.

An extensive analysis of SOL properties in high-δ plasmas in ASDEX Upgrade

A set of dedicated H-mode discharges with constant heating power combining Neutral Beam Injection and Electron Cyclotron Resonance Heating have been executed at the ASDEX Upgrade tokamak using a high triangularity magnetic geometry in order to investigate the impact of filamentary transport to divertor and non-divertor components. The evolution of upstream scrape-off layer (SOL) profiles have been correlated with dedicated separatrix quantities, mostly with the turbulence control parameter αt (Eich and Manz 2021 Nucl. Fusion 61 086016) describing the turbulence level at the separatrix. With increasing αt , a broadening of the upstream density profiles in the near-SOL together with the formation of a density shoulder in the far-SOL have been observed. This phenomenon is associated with an enhanced filamentary transport dominating the radial turbulent transport in the far-SOL and confirmed by means of the cooling water calorimetry on non-divertor components. The probe measurements conducted with the ball-pen probe-head mounted on the midplane manipulator and a retarding-field analyzer close to the limiter surface indicate that the key mechanism increasing the radial filamentary transport to the first wall is an increase of the particle flux Γr,fil , caused primarily by the packing fraction fPF,fil and the filament density ne,fil . At the same time, the electron temperature Te and ion temperature Ti measured close to the limiter surface show only small variations above αt > 0.5. Both the filamentary heat flux and the gross erosion derived from the first wall probe measurements reach a magnitude that should be considered in the design of future fusion reactors.

Deuterium permeation studies through bare and Er2O3 coated SS 316L

Hydrogen isotope permeation through the structural walls of a fusion reactor poses concern for both fuel loss and safety. A primary focus in the International Thermonuclear Experimental Reactor (ITER) is the reduction of hydrogen isotopes, particularly tritium permeation through structural materials. In this work, an experimental setup has been developed to study the permeation of deuterium through thin samples such as bare and erbia (Er2O3) coated 100 µm thick stainless steel (SS 316 L). Permeation results through a bare SS 316 L sample yielded deuterium diffusivity and permeability of SS 316 L. Erbia coating on SS 316 L was developed using dip coating technique and its performance was tested as a permeation reduction barrier (PRB). A permeation reduction factor of ∼87 and 359 have been achieved with coating thickness ∼ 100 and ∼200 nm respectively. No measureable permeation flux was obtained for coating thickness of 492 nm. These results indicate the effectiveness of erbia coating in reducing deuterium permeation through SS 316 L.

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.

Investigation of MHD flow balancing in electrically conductive manifolds of liquid metal blankets

In the liquid metal (LM) blanket design for magnetic confinement fusion reactors, manifolds including main and branch ducts are commonly used to direct fluids to various locations of the blanket to cool its components. The primary objective of this structural configuration is to achieve flow balancing within each branch duct. However, magnetohydrodynamic effects will cause uneven flow distribution, consequently, uncontrollable overheating phenomena in certain blanket segments may be caused by this flow distribution. To assess its impact, the influences of external magnetic fields, inlet velocities and coupling wall thicknesses on the flow partitioning in conductive manifolds are investigated in this study. The results indicate that the flow distribution among the branch ducts becomes less uniform as the external magnetic field strength increases, while varying the inlet velocity has minimal impact. In addition, increasing the coupling wall thickness leads to more even flow distribution when the rectangular branch ducts are perpendicular with the magnetic field direction. Based on above physical understanding, three controlling methods for the flow partitioning in conductive manifolds are proposed to achieve flow balancing in each branch duct and then validated through numerical simulations. The results indicate that flow balancing can be achieved by adjusting the coupling wall thickness when the rectangular branch ducts are aligned with the magnetic field direction, however, it is challenging when they are perpendicular. Furthermore, the flow balancing can be realized by a symmetrical design of the manifold with circular branch pipes and it is independent of external magnetic fields and inlet velocities. These findings will enhance the physical understanding of flow distribution patterns in conductive manifolds under strong magnetic fields, which is important for the flow control and design of manifolds in LM blankets.

Observation of alpha-particles in recent D–T experiments on JET

The fusion reaction between deuterium and tritium, D(T,n)4 He is the main source of energy in future thermonuclear reactors. Alpha-particles (4 He-ions) born with an average energy of 3.5 MeV transferring energy to the thermal plasma during their slowing down, should provide the self-sustained D–T plasma burn. The adequate confinement of α-particles is essential to provide efficient heating of the bulk plasma and steady burning of a reactor plasma. That is why the fusion-born α-particle studies have been a priority task in the second D–T experiments (DTE2) on the Joint European Torus (JET) to understand the main mechanisms of their slowing down, redistribution and losses and to develop optimal plasma scenarios. JET with Be-wall and W-divertor, enhanced auxiliary heating systems and improved energetic-particle diagnostic capabilities, producing significant population of α-particles, provided the possibility for comprehensive studying of the α-particle behaviour. Selected results of the confined and lost α-particle measurements, evidence of α-particle self-heating and assessments of the fusion performance are presented in this paper giving an opportunity for further modelling and extrapolation to the International Thermonuclear Experimental Reactor and burning plasma reactors.

Dimensional effects in analysis of laser-induced-desorption diagnostics data

The accurate assessment of the local tritium concentration in the tokamak first wall by means of the laser-induced desorption (LID) diagnostics is sought as one the key solutions to monitoring the local radioactive tritium content in the first wall of the fusion reactor ITER. Numerical models of gas desorption from solids used for LID simulation are usually closed with the one-dimensional transport models. In this study, the temperature and particle dynamics in the target irradiated by a short laser pulse during LID are analyzed by means of the two-dimensional model to assess the validity of using one-dimensional approximation for recovering the diagnostics signal. The quantitative estimates for the parameters governing the heat and particle transfer are presented. The analytical expressions for the sample spatiotemporal temperature profiles driven by the target irradiation with a Gaussian laser beam with the trapezoid temporal shape are derived. The obtained relations are used to simulate tritium desorption from a tungsten sample driven by pulsed heating. It is shown that depending on the ratio between the laser spot radius and the heat diffusion length, the one-dimensional approach can noticeably overestimate the sample temperature in the limit of small laser spot radius (estimated for tungsten as ∼0.5–1.0 mm), resulting in more than 100% larger amounts of tritium desorbed from the target, compared to the two-dimensional approximation. In the limit of large laser spot radius (≥1.5 mm), both approaches yield comparable amounts of tritium desorbed from the sample.

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.

Turbulence link to L-mode, I-mode, and H-mode confinement in the DIII-D tokamak

Understanding the physics of low-confinement (L-), improved-confinement (I-), and high-confinement (H-) modes is critical for fusion reactors. The finding herein reports observations of two types of turbulence coexisting near the L-mode edge, one magnetohydrodynamic (MHD)-like and another micro-tearing mode (MTM)-like, linked to the H-mode and I-mode confinement in the DIII-D tokamak. Ion-scale magnetic and density turbulence is measured using a Faraday-effect radial-interferometer-polarimeter and beam-emission-spectroscopy (BES). Broadband turbulence spectra of up to ∼600 kHz are observed in two discharges where transitions between L-mode, I-mode, and H-mode occurs. Turbulence is found to be inversely correlated with confinement, meaning lower turbulence power at higher confinement. Distinctively, the high-frequency (HF, >∼100 kHz) magnetic turbulence power changes by the most (55%) during transitions primarily involving energy confinement change, whereas the low-frequency (LF, <∼100 kHz) magnetic and density turbulence power changes by the most (80%) during transitions primarily involving particle confinement change. The LF turbulence amplitude oscillates with and leads to deuterium-alpha emission oscillations before an H-mode. These results imply that HF turbulence mainly affects energy confinement whereas LF turbulence can affect particle confinement. The magnetic and density turbulence exhibits coherence up to 0.6 and cross-phase magnitude close to π/2 in most cases, suggesting they have a common origin in both the LF and HF ranges. BES suggests that LF turbulence resides at the edge ( ρ=0.95 ) and HF turbulence can be at the outer core ( ρ=0.8 ) or edge ( ρ=0.95 ). Comparisons of measurements, theory, and gyrokinetic simulations suggest that HF turbulence is MTM-like in all cases, whereas LF turbulence is more consistent with MHD-like modes and the exact instability might change during transitions—except that a drift-wave origin is possible in a low collisionality H-mode. These results suggest that the H-mode involves suppressed MHD-like turbulence, whereas the I-mode mitigates MTM-like turbulence along with largely unchanged MHD-like turbulence.

Hydrogen extraction from 0.1 % H2–He mixture: The interplay phenomena of electrode, temperature, and voltage in BZYN & BZCYYb

Hydrogen pumps, crafted from proton-conductive ceramic electrolyte and paired with either oxide or metallic electrodes, have been designed for the extraction of hydrogen isotopes from helium gas mixtures containing 0.1 % hydrogen isotopes, particularly within the context of TES for nuclear fusion reactors. This study presents the electrochemical hydrogen permeation research conducted on the perovskite proton conductor materials BaZr0.1Ce0.7Y0.1Yb0.1O3-δ (BZCYYb) and BaZr0.8Y0.16Ni0.04O3-δ (BZYN) under these operational conditions. It was observed that nickel electrodes provided superior performance over SrFe0.8Mo0.2O3-δ (SFM) electrodes in terms of hydrogen extraction efficiency. Hydrogen pumps that integrated BZCYYb as the electrolyte with nickel electrodes showed enhanced efficiency, while those utilizing BZYN as the electrolyte coupled with nickel electrodes demonstrated greater stability. Furthermore, the study explored the voltammetric nonlinearity at low hydrogen concentrations and the dependency of concentration polarization efficiency on both voltage and temperature, aiming to establish optimal conditions that balance stability and efficiency for both types of hydrogen pumps.

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.

Improvement in thermal stability of ODS-W alloy through formation of complex oxide dispersoids

Thermal stability is a crucial issue for engineering application of tungsten, especially as plasma facing material in future fusion reactor. Two kinds of ODS-W alloys of W-0.5Y2O3 (WY) and W-0.5Y2O3-0.25Ti (WYT) were fabricated by ball milling and spark plasma sintering aiming at formation of oxide dispersoids with different crystallographic structures. Thermal stability of both ODS-W alloys was evaluated in terms of microhardness and microstructure evolution during isothermal aging at 1100 °C. Dispersoids of sesquioxide Y2O3 were identified in sintered WY, while pyrochlore Y2Ti2O7 with finer size and higher number density were developed in sintered WYT, resulting in higher bending strength of 1020 MPa and microhardness of 880 HV. Compared with sesquioxide Y2O3, complex dispersoids of pyrochlore Y2Ti2O7 exhibited much better coarsening resistance during isothermal aging at 1100 °C up to 100 h, where the intracrystalline and GB dispersoids maintained around 10 nm and 30 nm, respectively. The excellent coarsening resistance of complex Y2Ti2O7 with pyrochlore structure contributed to grain refinement and improved microstructure stability of WYT due to the effect of Zener drag.

Synthesis and radiation damage tolerance of Mo0.75W0.25AlB solid solution for nuclear fusion reactor applications

MAB phases (M = transition metal, A = aluminium (Al), B = boron (B)), a new family of nano-laminated ternary transition metal borides with an excellent combination of functional and structural properties, are currently being investigated as potential radiation shielding materials in nuclear fusion reactors. Here, we report the fabrication of a solid solution MAB phase Mo0.75W0.25AlB (Mo - molybdenum, W - tungsten), in the form of dense and high purity pellets, synthesised for the first time using a reactive hot-pressing method. The shielding and irradiation damage properties of the material were also evaluated. Monte Carlo N-Particle (MCNP) calculations showed that W doping in the MoAlB phase is likely to enhance the shielding capacity against neutrons and gamma rays of different energies. The experimental damage tolerance of Mo0.75W0.25AlB to silicon ion (Si+) irradiation indicated that this solid solution compound has a high resistance to crack formation and other radiation damage effects compared to that of pure MoAlB. Thermal annealing of the amorphous Mo0.75W0.25AlB layer formed by irradiation showed that the damaged structure easily recrystallizes at high temperatures, completely returning to the virgin state. The results show that this solid solution should be considered as a promising candidate material for radiation shielding components.