Development Of Fusion Reactors Research Articles

Broad Beam-Induced Fragmentation and Joining of Tungsten Oxide Nanorods: Implications for Nanodevice Fabrication and the Development of Fusion Reactors

In this work, for the first time, fragmentation and joining of tungsten oxide (WO3) nanorods induced by a broad ion beam are reported. Although at low energy (5 keV) and moderate ion fluence, nanorods fragment into smaller pieces along the length, at higher ion energies (50–100 keV), a contrary process occurs, which leads to the joining of the nanorods. A state-of-the-art ion–solid interaction simulation, namely, TRI3DYN, has been invoked to explore the possible mechanisms that reveal subtle contributions of surface defects, ion-beam mixing, and sputtering. High-resolution electron microscopy, photoluminescence study, and X-ray photoelectron spectroscopy support the observed results and proposed mechanisms. Such modifications have interesting effects on the electrical conductivity of the nanorod assembly. The change in sample color upon ion irradiation from initial white to yellow, light blue, deep blue, light green, and cyan shows an excellent and reversible chromatic response of tungsten oxide nanorods to irradiation. Such a property can be exploited to fabricate radiation sensors. The fragmentation and joining at different energy scales have essential implications in nanodevice fabrication through the bottom-up approach as well as for the development of fusion reactors.

The effect of W fuzz growth on D retention in polycrystalline W

The retention of hydrogen isotopes in W is one of the main limiting factors affecting the design and development of next-step fusion reactors, such as the international thermonuclear experimental reactor (ITER). Here, we report on D retention behavior under the steady-state He/D ion irradiations of polycrystalline W. We find that the D retention behavior in the W bulk is completely suppressed by the growth of high surface-to-volume W fuzz layer (1.1–4.0 μm thick) over the surface. Our analyses show that due to their frequent collisions in the W fuzz layer, the dose of D ions greatly decreases when they collide into the W bulk beneath the fuzz layer. The W fuzz layer with a high surface-to-volume ratio becomes an effective shield against the ion driven diffusion processes.

Design and development of an Allison type emittance scanner for the SPIDER ion source.

Low divergence negative ion beams are crucial for the development of ITER-like fusion reactors. SPIDER is the prototype beam source of the ITER heating neutral beam injector, and it recently started beam acceleration, up to a voltage of 30 kV. The main diagnostics used to measure beamlet divergence are a movable diagnostic calorimeter (STRIKE), which gives the thermal footprint of the beamlets; beam emission spectroscopy; and visible imaging. These systems do not allow a direct measurement of single beamlet phase-space distribution, which is useful for comparison with numerical simulations and to estimate accelerator performances. To this purpose, a movable Allison type emittance scanner for the SPIDER negative ion beam was developed and proposed for the installation on the STRIKE supporting structure. This paper describes the numerical analyses performed to dimension the mechanical and electrical components, such as the Faraday cup and the slits. An analytical approach based on the integration of an arbitrary phase-space distribution was adopted in order to simulate the device performances. The constraints due to the operation in a high heat load environment are discussed.

Influence of the inverse sheath on divertor plasma performance in tokamak edge plasma simulations

AbstractRecently, it was shown that strong electron thermionic emission from material walls could result in the formation of an “inverse sheath,” which prevents the flow of cold ions to the wall.[1–3] Such regimes look very favourably from the point of view of plasma–material interactions at the edge of magnetic fusion devices, where the problem of the erosion of plasma‐facing components under ion irradiation is one of the key issues for the development of future magnetic fusion reactors. However, it is not clear whether such regimes are compatible with edge plasma parameters and heat removal requirements in fusion reactors. To address the issue of practicality of the “inverse sheath” regime for edge tokamak plasma conditions, we perform a set of numerical simulations with 2D edge plasma transport code UEDGE[4] for a DIII‐D‐like geometry and magnetic configuration. To describe both “standard” and “inverse sheath” conditions within the framework of the UEDGE code (which does not consider the sheath region per se), at the material surfaces, we apply effective boundary conditions that emulate both “standard” and “inverse sheath” regimes. We demonstrate that, for the same input parameters, spatial distributions of edge plasma parameters corresponding to detached divertor and “inverse sheath” regimes are similar, with only a few minor differences. We discuss the compatibility of “inverse sheath” regimes with core plasma parameters.

Experimental study of liquid metal heat transfer in a vertical duct affected by coplanar magnetic field: Downward flow

Liquid metal flows under a high thermal load and strong magnetic field are of great interest in the context of the development of fusion reactor cooling systems. In this study, a downward flow of mercury in a heated duct, affected by a coplanar magnetic field (directed transverse to the main flow and along the long side of the duct), is studied experimentally for symmetrical (two-sided) and asymmetrical (one-sided) duct heating, showing different flow behaviors in the magnetic field. We show that in spite of the qualitatively different structure of the downward flow in the duct with different heating modes, the applied coplanar magnetic field provides a rather general reaction, which includes two scenarios determined by the Richardson number. At high Richardson numbers (Ri≳1), the Nusselt number tends toward the turbulent limit, as expected in the channel without a magnetic field, and at low Richardson numbers (Ri≲1), the Nusselt number tends to its laminar limit.

Progress Toward a Compact Fusion Reactor Using the Sheared-Flow-Stabilized Z-Pinch

The sheared-flow-stabilized (SFS) Z-pinch is a promising confinement concept for the development of a compact fusion reactor. The Z-pinch has been theoretically and experimentally shown to be stable to magnetohydrodynamic modes when sufficient radial shear of the axial flow is present. At the University of Washington, the Fusion Z-pinch Experiment (FuZE) research project examines scaling the SFS Z-pinch toward fusion conditions. The FuZE device produces long-duration, 50-cm-long pinches with measured ion and electron temperatures over 1 keV and number densities greater than cm. Plasma properties are measured with a diagnostic suite that includes magnetic field probes, heterodyne quadrature interferometry, digital holographic interferometry, ion-Doppler spectroscopy, and fast framing photography. Neutrons are produced in the FuZE device when deuterium is injected along with the normal hydrogen or helium fueling species. Neutron generation is diagnosed using plastic scintillator detectors. The neutron production is sustained for 5 to 8 μs, thousands of times longer than the static Z-pinch instability growth time. Measured neutron production is consistent with calculated theoretical values for thermonuclear yield at the observed plasma temperatures and scales with the square of the deuterium concentration. A preliminary reactor concept is designed to incorporate flowing liquid metal walls, which would serve as an electrode, a heat transfer fluid, a radiological shield, and a breeding blanket. Using a liquid metal wall could address several unresolved material and technology issues in existing fusion reactor designs.

The evolution of He nanobubbles in tungsten under fusion-relevant He ion irradiation conditions

He-induced W nanofuzz growth over the W divertor target is one of the main limiting factors affecting the current design and development of fusion reactors. In this paper, based on He reaction rate model in W, we simulate the growth and evolution of He nanobubbles during W nanofuzz formation under fusion-relevant He+ irradiation conditions. Our modeling unveils the existence of He nanobubble-enriched W surface layer (<10 nm), formed due to the He diffusion in W crystal into defect sites. At an elevated temperature, the growth of He bubbles in the W surface layer prevents He atoms diffusing into the deep layer (>10 nm). The formation of W nanofuzz at the surface is attributed to surface bursting of high-density He bubbles with their radius of ~4 nm, and an increase in the surface area of irradiated W. Our findings have been well confirmed by the experimental measurements.

Development of a SIMMER\\RELAP5 coupling tool

The In-Box Loss Of Coolant (LOCA) postulated accident is considered a major concern for the safety connected with the development of EU-DEMO fusion reactor. Relating to the renewed interest in the Water-Cooled Lithium Led blanket concept, an innovative experimental campaign is under development at ENEA Brasimone laboratories aiming at investigating the consequences related to the In-Box LOCA applied to the WCLL breeding blanket. In this frame, a new coupling tool between the SIMMER-III (modified version to implement the PbLi/water chemical interaction) and the RELAP5/Mod3.3 codes (modified version to implement PbLi thermo-physical properties) has been developed together with its preliminary application to simple test cases with water as working fluid. The coupling procedure can be defined as a “two-way”, “non-overlapping”, “online” technique aiming at investigating multi-physics and multi-scales phenomena in support of the development of fusion reactor technologies.

Hot crack susceptibility in multi-pass welds of reduced activation ferritic/martensitic steel F82H

The research and development of nuclear fusion reactors are going on conducting to acquire a next-generation energy source. It is supposed to be that the reduced activation ferritic/martensitic (RAFM) steel F82H is adopted as a structural material of the blanket module, which is the device set on the inner wall of fusion reactors. As welding for thicker plates of F82H, multi-pass welding will be adopted to the joints. In the case of multi-pass welding, hot cracking is one of serious defects in welds and is concerned in welds reheated by following weld passes. Therefore, the objectives of this study are to evaluate hot crack susceptibility in multi-pass welds of F82H and to clarify the cause of hot cracking in multi-pass welds of F82H. The hot crack susceptibility in multi-pass welds is evaluated by the longitudinal Varestraint test with double-bead and triple-bead types. From the observation of fractured surface occurred in welds of F82H after the Varestraitnt test, the crack is identified as ductility-dip crack. The cause of hot crack in multi-pass welds of F82H is clarified by Vickers hardness test, SEM microstructure observations and X-ray diffraction pattern analyses. Based on these tests, when multi-pass welding is conducted for the F82H steels with high Ta contents, it should be considered to control welding conditions for prevention of hot cracking in weld metal as well as high strength martensitic steels. These results suggest that the cause of the crack is intragranular hardening by the precipitation of TaN during welding thermal cycles.

Power accounting of plasma discharges in the linear device Proto-MPEX**This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of

Plasma material interaction (PMI) studies are crucial to the successful development of future fusion reactors. Prototype Material Plasma Exposure eXperiment (Proto-MPEX) is a prototype design for the MPEX, a steady-state linear device being developed to study PMI. The primary purpose of Proto-MPEX is developing the plasma heating source concepts for MPEX. A power accounting study of Proto-MPEX works to identify machine operating parameters that could improve its performance, thereby increasing its PMI research capabilities, potentially impacting the MPEX design concept. To build a comprehensive power balance, an analysis of the helicon region has been performed implementing a diagnostic suite and software modeling to identify mechanisms and locations of heat loss from the main plasma. Of the 106.3 kW of input power, up to 90.5% of the power has been accounted for in the helicon region. When the analysis was extended to encompass the device to its end plates, 49.2% of the input power was accounted for and verified diagnostically. Areas requiring further diagnostic analysis are identified. The required improvements will be implemented in future work. The data acquisition and analysis processes will be streamlined to form a working model for future power balance studies of Proto-MPEX.

Simulation of loss-of-flow transient in water cooled fission blanket of Fusion-fission Hybrid Energy Reactor (FFHER)

Fusion-fission hybrid reactor takes advantage of the best of both domains, fusion and fission, without needing a full development of fusion reactor, and without critical safety issues of fission reactor. However, decay heat still exists in fission blanket of hybrid rector in shutdown state, which will menace the safety of rector if it cannot be smoothly carried out. To testify the passive safety characteristics of fission blanket, the reactor coolant pump shaft seizure accident in fission blanket of FFHER, a fusion-fission hybrid reactor, is simulated. The model of passive residual heat removal system, as well as reactor coolant system, are built using software SCDAP/RELAP5 4.0. The analysis shows that the simulated system response is reasonable. In the SG domination stage, the peak fuel temperature keeps always below the limiting value. In the PRHR domination stage, three cases are simulated and compared to choose appropriate parameter. It is found that PRHR circulation can be rapidly established and transfer enough heat from blanket. The transition from SG domination to PRHR domination is smooth. From the simulation results, it can be concluded that PRHRS can provide sufficient blanket cooling capability in LOFA.

Effect of grain size on the behavior of hydrogen/helium retention in tungsten: a cluster dynamics modeling

Reducing ion retention in materials is a key factor in the management of tritium inventory, the selection of compatible plasma-facing materials (PFMs), and thus the future development of fusion reactors. In this work, by introducing the cellular sink strength of grain boundaries (GBs) into the cluster dynamics model, the behavior of hydrogen (H) and helium (He) retention in W with different grain sizes is studied under various irradiation conditions systematically. It is found that the H/He retention increases dramatically with decreasing grain size at typical service temperatures, due to the enhancement of H/He capture ratio by GBs. Generally, He retention exists in three forms: He in GBs, in dislocations and in clusters (HemVn, Hen and HenI). Our further study shows that, under the irradiation of low energy and low fluence ions, the contribution of He in clusters is negligible. The total He retention is thus dominated by the competing absorption of GBs and dislocations, that is, changing from the dislocation-based to grain boundary-based retention with decreasing grain size. H retention also presents the same behavior. In view of these grain size-related behaviors of H/He retention in W, it is suggested that coarse-grained crystals should be selected for W-based PFMs in practice.

Monte Carlo simulation of ion–material interactions in nuclear fusion devices

ABSTRACTOne of the key aspects regarding the technological development of nuclear fusion reactors is the understanding of the interaction between high-energy ions coming from the confined plasma and the materials that the plasma-facing components are made of. Among the multiple issues important to plasma–wall interactions in fusion devices, physical erosion and composition changes induced by energetic particle bombardment are considered critical due to possible material flaking, changes to surface roughness, impurity transport and the alteration of physicochemical properties of the near surface region due to phenomena such as redeposition or implantation. A Monte Carlo code named MATILDA (Modeling of Atomic Transport in Layered Dynamic Arrays) has been developed over the years to study phenomena related to ion beam bombardment such as erosion rate, composition changes, interphase mixing and material redeposition, which are relevant issues to plasma-aided manufacturing of microelectronics, components on object exposed to intense solar wind, fusion reactor technology and other important industrial fields. In the present work, the code is applied to study three cases of plasma material interactions relevant to fusion devices in order to highlight the code’s capabilities: (1) the Be redeposition process on the ITER divertor, (2) physical erosion enhancement in castellated surfaces and (3) damage to multilayer mirrors used on EUV diagnostics in fusion devices due to particle bombardment.

Thought Experiment to Examine Benchmark Performance for Fusion Nuclear Data

There are many benchmark experiments carried out so far with DT neutrons especially aiming at fusion reactor development. These integral experiments seemed vaguely to validate the nuclear data below 14 MeV. However, no precise studies exist now. The author’s group thus started to examine how well benchmark experiments with DT neutrons can play a benchmarking role for energies below 14 MeV. Recently, as a next phase, to generalize the above discussion, the energy range was expanded to the entire region. In this study, thought experiments with finer energy bins have thus been conducted to discuss how to generally estimate performance of benchmark experiments. As a result of thought experiments with a point detector, the sensitivity for a discrepancy appearing in the benchmark analysis is “equally” due not only to contribution directly conveyed to the deterctor, but also due to indirect contribution of neutrons (named (A)) making neutrons conveying the contribution, indirect controbution of neutrons (B) making the neutrons (A) and so on. From this concept, it would become clear from a sensitivity analysis in advance how well and which energy nuclear data could be benchmarked with a benchmark experiment.

State-space Coil Modelling in Plasma Magnetic Confinement Devices

The need of robust and optimal control schemes is a key factor for the development of future fusion reactors. This paper has dealt with the state-space modelling of the Ultra-Low Iota Super Elongated Stellarator of the UPV/EHU, using a physical lumped parameter equivalent circuit approach. The model obtained has been validated by means of experimental output data showing an excellent matching with the real system. Besides, it has been designed a MPC scheme that has been successfully implemented both in simulation and experimentally using a real-time control platform.

Review of LIBS application in nuclear fusion technology

Nuclear fusion has enormous potential to greatly affect global energy production. The next-generation tokamak ITER, which is aimed at demonstrating the feasibility of energy production from fusion on a commercial scale, is under construction. Wall erosion, material transport, and fuel retention are known factors that shorten the lifetime of ITER during tokamak operation and give rise to safety issues. These factors, which must be understood and solved early in the process of fusion reactor design and development, are among the most important concerns for the community of plasma–wall interaction researchers. To date, laser techniques are among the most promising methods that can solve these open ITER issues, and laser-induced breakdown spectroscopy (LIBS) is an ideal candidate for online monitoring of the walls of current and next-generation (such as ITER) fusion devices. LIBS is a widely used technique for various applications. It has been considered recently as a promising tool for analyzing plasma-facing components in fusion devices in situ. This article reviews the experiments that have been performed by many research groups to assess the feasibility of LIBS for this purpose.

Modelling the thermomechanical behaviour of the tungsten first wall in HiPER laser fusion scenarios

The behaviour of a tungsten first wall is studied under the irradiation conditions predicted for the different operational scenarios of the European laser fusion project HiPER, which is based on direct drive targets and an evacuated dry wall chamber. The scenarios correspond to different stages in the development of a nuclear fusion reactor, from proof of principle (bunch mode facility) to economic feasibility (pre-commercial power plant). This work constitutes a quantitative study to evaluate first wall performance under realistic irradiation conditions in the different scenarios. We calculated the radiation fluxes assuming the geometrical configurations reported so far for HiPER. Then, we calculated the irradiation-induced evolution of first wall temperature and the thermomechanical response of the material. The results indicate that the first wall will plastically deform up to a few microns underneath the surface. Continuous operation in a power plant leads to fatigue failure with crack generation and growth. Finally, crack propagation and the minimum tungsten thickness required to fulfil the first wall protection role is studied. The response of tungsten as a first wall material as well as its main limitations will be discussed for the HiPER scenarios.

A LIBS method for simultaneous monitoring of the impurities and the hydrogenic composition present in the wall of the TJ-II stellarator.

The study of plasma-wall interactions and impurity transport in the plasma fusion devices is critical for the development of future fusion reactors. An experiment to perform laser induced breakdown spectroscopy, using minor modifications of our existing laser blow-off impurity injection system, has been set up thus making both experiments compatible. The radiation produced by the laser pulse focused at the TJ-II wall evaporates a surface layer of deposited impurities and the subsequent radiation produced by the laser-produced plasma is collected by two separate lens and fiber combinations into two spectrometers. The first spectrometer, with low spectral resolution, records a spectrum from 200 to 900 nm to give a survey of impurities present in the wall. The second one, with high resolution, is tuned to the wavelengths of the Hα and Dα lines in order to resolve them and quantify the hydrogen isotopic ratio present on the surface of the wall. The alignment, calibration, and spectral analysis method will be described in detail. First experimental results obtained with this setup will be shown and its relevance for the TJ-II experimental program discussed.

Fusion Studies in Japan

A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

Circumferential tensile test method for mechanical property evaluation of SiC/SiC tube

SiC fiber reinforced/SiC matrix (SiC/SiC) composite is expected to be a candidate material for the first-wall, components in the blanket and divertor of fusion reactors in future. In such components, SiC/SiC composites need to be formed to be various shapes. SiC/SiC tubes has been expected to be employed for blanket and divertor after DEMO reactor, but there is not established mechanical investigation technique. Recent progress of SiC/SiC processing techniques is likely to realize strong, having gas tightness SiC/SiC tubes which will contribute for the development of fusion reactors. This research studies the relationship between deformation and strain of SiC/SiC tube using a circumferential tensile test method to establish a mechanical property investigation method of SiC/SiC tubes.