Magnetic Fusion Systems Research Articles

Evaluation of the Performance of Commercial High Temperature Superconducting Tapes for Dynamo Flux Pump Applications

High Temperature Superconducting (HTS) dynamo flux pumps are a promising alternative to metal current leads for energization and the persistent current mode operation of high current DC superconducting magnet systems for applications in rotating machines, such as Magnetic Resonance Imaging (MRI) or fusion systems. The viability of the flux pump concept has been widely proven by laboratory experiments and research is now in progress for the design and optimization of flux pump devices for practical applications. It has been widely established that the dependence of the critical current density (Jc) on the temperature (T), the magnetic field magnitude (B), and the orientation (θ), has a substantial impact on the overall DC voltage obtained at the terminals, as well as on the current limit and the loss of the flux pump. Since HTS tapes produced by different manufacturers, they show different dependencies of Jc with the amplitude and the orientation of the magnetic field. They also give rise to different outputs when employed in flux pumps. In this paper, we evaluate and compare the performance of several commercial HTS tapes used for flux pumping purposes through numerical simulation. We also investigate the dependence of the flux pump ‘s performance on the operating temperatures. A 2D finite element numerical model is first developed and validated against experimental data at 77 K. Afterward, the same HTS dynamo apparatus used for validation is exploited for the comparison. The Jc(B,θ,T) and n(B,θ,T) relations, which characterize each different tape in the model, are reconstructed via artificial intelligence techniques based on the open-access database of the Robinson Research Institute. It is shown in the paper that certain tapes are more suitable than others for flux pump applications and that the best overall operating temperature is in the vicinity of 77 K.

60 Years of neutral particle analysis: from early tokamaks to ITER

Academician A. D. Sakharov’s idea concerning the emission of atomic flux from hot plasma (1951) inspired scientists of A. F. Ioffe Physico-Technical Institute to create the first in the world instrument called Neutral Atom Analyzer in 1960 and then in 1961 to use it successfully on the Alpha device (USSR, 1958–1963). Now the analysis of fluxes of fast atoms referred to as Neutral Particle Analysis (NPA) is one of the main diagnostic methods for the ion component of plasma in tokamaks, stellarators, and other devices. NPA provides a unique opportunity for studying the ion distribution functions, ion temperatures and hydrogen isotope ratio in hot plasma. Neutral particle analyzers developed at the Ioffe Institute were widely used in the USSR until the late 1970s, and afterwards began to be employed worldwide. Since then, most of the information on the ion distribution functions and the behavior of fast ions in fusion plasma is obtained from NPA measurements on all leading magnetic confinement fusion systems worldwide. The specialized complex of atom analyzers currently being created at the Ioffe Institute is included in the primary list of ITER diagnostics. The integration of this complex on ITER is expected to begin in 2025.

Experiment and measurement with high-energy helium irradiation to tungsten using tandem accelerator for divertor in magnetic confinement fusion system

An experimental system for irradiating high-energy (4 MeV) helium beam was developed to investigate irradiation damage of tungsten by using tandem accelerator. In this system, particle fluence to a tungsten thin plate was achieved to 1022 ions/m2. Electron diffraction images were acquired for damaged tungsten using a transmission electron microscope (TEM) to measure the particle loading by helium irradiation. Crystal structure was qualitatively changed between helium irradiated area and non-irradiated area from diffraction patterns.

Fusion energy in an inertial electrostatic confinement device using a magnetically shielded grid

Theory for a gridded inertial electrostatic confinement (IEC) fusion system is presented, which shows a net energy gain is possible if the grid is magnetically shielded from ion impact. A simplified grid geometry is studied, consisting of two negatively biased coaxial current-carrying rings, oriented such that their opposing magnetic fields produce a spindle cusp. Our analysis indicates that better than break-even performance is possible even in a deuterium-deuterium system at bench-top scales. The proposed device has the unusual property that it can avoid both the cusp losses of traditional magnetic fusion systems and the grid losses of traditional IEC configurations.

Effect of surface segregation and mobility on erosion of plasma-facing materials in magnetic fusion systems

The present work studies the combined effects of collisional sputtering and mixing processes of carbon impurities in tungsten plasma-facing material integrated with thermal processes including surface segregation and diffusion. We used our ITMC-DYN package, which includes description of all collisional and thermal processes, for the analysis of recent experimental results of tungsten erosion and carbon implantation at various target temperatures. Self-consistent integrated modeling predicted thermal processes effects on erosion/deposition dynamics and defined decisive parameters range and their importance. Critical parameters were estimated based on available experimental data. The integrated simulation reproduced the experimental results and predicted the transition from enhanced tungsten erosion to significant carbon coverage on the tungsten surface. These effects for wider range of system conditions with C/H ions irradiation and for reactor conditions can be predicted by including detailed modeling of chemical erosion processes in a self-consistent manner.

Scientific Discovery in Fusion Plasma Turbulence Simulations at Extreme Scale

The primary goal of the extreme-scale plasma turbulence studies described in this article is to gain new insights on confinement scaling in magnetic fusion systems by using powerful, world-class supercomputers to carry out simulations with unprecedented resolution and temporal duration. New insights have been gained on the key question of how the turbulent transport of thermal energy and particles in the plasma and the associated confinement scale from present-generation devices to much larger ITER-size plasmas. In particular, new results from large-scale simulation studies have demonstrated that improvement in confinement as devices grow larger is far more gradual, with significantly lower loss rates than less-powerful computer simulations have indicated in research carried out over the past decade.

Transmutation of Tungsten in Fusion and Fission Nuclear Environments

The amount and type of gaseous and metallic transmutants produced in tungsten (W) when used as a plasma-facing armor in magnetic (MFE) and inertial (IFE) confinement fusion systems were determined and compared to those obtained following irradiation in fission reactors. Up to ∽8% metallic transmutants are generated at the expected lifetime of the fusion blanket. Irradiation in fission reactors to the same fast neutron fluence yields a much larger amount of metallic transmutation products than in fusion systems. While the dominant component in fusion systems is rhenium (Re), osmium (Os) is the main transmutation product in fission reactors. The impact on the W properties needs to be assessed. The results of this work will help guide irradiation experiments in fission reactors to properly simulate the conditions in fusion systems by possible direct implantation of transmutation products in irradiated samples. In addition, the results represent a necessary input for modeling activities aimed at understanding the expected effects on properties.

Transmutation of silicon carbide in fusion nuclear environment

The amount and type of metallic transmutants produced in SiC/SiC when used in magnetic (MFE) and inertial (IFE) confinement fusion systems are determined and compared to those obtained following irradiation in fission reactors. Up to ∼1.3% metallic transmutants are generated at the expected lifetime of the fusion blanket. Irradiation in fission reactors to the same fast neutron fluence produces about an order of magnitude lower metallic transmutation products than in fusion systems. While the dominant component in fusion systems is Mg, P is the main transmutation product in fission reactors. The impact on the SiC/SiC properties is not fully understood. The results of this work will help guide irradiation experiments in fission reactors to properly simulate the conditions in fusion systems by possible ion implantation. In addition, the results represent a necessary input for modeling activities aimed at understanding the expected effects on properties.

Damage parameters of structural materials in fusion environment compared to fission reactor irradiation

Abstract Calculations were performed to quantify the damage parameters in the leading candidate structural and plasma facing materials when used in magnetic and inertial confinement fusion systems and when irradiated in fission reactors. The structural materials considered are ferritic steel, austenitic steel, vanadium alloy and SiC/SiC composite. Plasma facing materials included beryllium, tungsten, and carbon fiber composites. Atomic displacement damage and gas production rates are greatly influenced by the neutron energy spectrum. For the same neutron wall loading, atomic displacement damage is slightly lower in inertial fusion systems than in magnetic fusion systems but gas production is about a factor of 2 lower. In addition, much lower gas production is obtained in samples irradiated in fission reactors. The results help guide irradiation experiments in fission reactors to properly simulate the damage environment in fusion systems and facilitate extrapolating to the expected material performance in fusion systems.

Damage production and accumulation in SiC structures in inertial and magnetic fusion systems

Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is ∼10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.

Diagnostic components in harsh radiation environments: Possible overlap in R&D requirements of inertial confinement and magnetic fusion systems

The next generation of large scale fusion devices--ITER/LMJ/NIF--will require diagnostic components to operate in environments far more severe than those encountered in present facilities. This harsh environment is the result of high fluxes of neutrons, gamma rays, energetic ions, electromagnetic radiation, and in some cases, debris and shrapnel, at levels several orders of magnitude higher than those experienced in today's devices. The similarities and dissimilarities between environmental effects on diagnostic components for the inertial confinement and magnetic confinement fusion fields have been assessed. Areas in which considerable overlap have been identified are optical transmission materials and optical fibers in particular, neutron detection systems and electronics needs. Although both fields extensively use cables in the hostile environment, there is little overlap because the environments and requirements are very different.

Comparison of avalanche-like quenches between NbTi and Nb 3Sn cables

Superconductors carrying currents larger than several 10 kA are required for magnetic confinement fusion systems. A cable-in-conduit-conductor (CICC), which has about 1000 of twisted thin superconducting strands in a ridged conduit, is a promising candidate for such large and high magnetic field superconducting system. A specific phenomenon in the CICC is so called avalanche-like quench (AQ). When some strands become resistive due to current saturation at critical current and/or a mechanical disturbance, the current re-distributes from the strands into neighboring ones, which are electro-magnetically coupled more strongly than the others. This causes current saturation in the neighboring strands. Finally, the whole cable becomes normal. We call this phenomenon an avalanche-like quench. Small multi-strand cables were tested by induced-current methods. Quench behaviors of both Nb 3Sn and NbTi cables are observed and compared. In the experiment, for the Nb 3Sn cable, the critical current measurement was stably carried out at high field without quench. In contrast, for the NbTi cable, we observed a number of AQs during inductive energization. Consequently, we suspected that the reason is based on the difference of critical temperature ( T C) between Nb 3Sn and NbTi. The difference of T C is several Kelvins, however, their macroscopic behaviors differ significantly.

Liquid Vortex Shielding for Fusion Energy Applications

Swirling liquid vortices can be used in fusion chambers to protect their first walls and critical elements from the harmful conditions resulting from fusion reactions. The beam tube structures in heavy ion fusion (HIF) must be shielded from high energy particles, such as neutrons, x-rays and vaporized coolant, that will cause damage. Here an annular wall jet, or vortex tube, is proposed for shielding and is generated by injecting liquid tangent to the inner surface of the tube both azimuthally and axially. Its effectiveness is closely related to the vortex tube flow properties. 3-D particle image velocimetry (PIV) is being conducted to precisely characterize its turbulent structure. The concept of annular vortex flow can be extended to a larger scale to serve as a liquid blanket for other inertial fusion and even magnetic fusion systems. For this purpose a periodic arrangement of injection and suction holes around the chamber circumference are used, generating the layer. Because it is important to match the index of refraction of the fluid with the tube material for optical measurement like PIV, a low viscosity mineral oil was identified and used that can also be employed to do scaled experiments of molten salts at high temperature.

Fusion Simulation Project: Integrated Simulation and Optimization of Magnetic Fusion Systems

This is the final report of a panel established as a subcommittee of the U. S. Department of Energy (DOE) Fusion Energy Sciences Committee (FESAC) on Integrated Simulation and Optimization of Magnetic Fusion Systems (ISOFS). The report was requested by the DOE in February 2002 and the approved report was transmitted to the DOE by the FESAC in December 2002. The report addresses the challenge of how to “develop fully integrated capability for predicting the performance of externally-controlled systems including turbulent transport, macroscopic stability, wave-particle physics, and multi-phase interfaces.”

Engineering, safety, and economic evaluations of aspire

A preconceptual design of a Tokamak fusion reactor concept called ASPIRE (Advanced Save Pool Immersed Reactor) has been developed. This concept provides many of the attractive features that are needed to enhance the capability of fusion to become the power generation technology for the 21st century. Specifically, these features are: inherent safety, low pressure, environmental compatibility, moderate unit size, high availability, high thermal efficiency, simplicity, low radioactive inventory, Class C radioactive waste disposal, and low cost of electricity. We have based ASPIRE on a second stability tokamak. However, the concept is equally applicable to a first stability tokamak or to most other magnetic fusion systems.

U.S. Heavy-Ion Fusion Systems Assessment Project Overview

A multi-institutional study was conducted to evaluate the potential of heavy-ion induction Linacs as inertial confinement fusion (ICF) drivers. This Heavy-Ion Fusion Systems Assessment (HIFSA) study was a U.S. effort to evaluate a wide range of possible system configurations for electric power plants driven by induction Linacs, as opposed to the radio-frequency accelerators used in previous heavy-ion fusion (HIF) power plant conceptual designs. In contrast to these earlier studies, the HIFSA project specifically avoided concentrating on a point design. Rather, cost/performance models of the major systems in an HIF power plant were devised by the institutions with expertise in the applicable technologies (e.g., Lawrence Berkeley Laboratory for induction accelerators and beam transport/focus; McDonnell Douglas Astronautics Company for cost scaling and systems modeling/intergration). (Detailed descriptions of these systems and associated integration/trade-off studies appear in other papers in this special issue.) Some of the key results of the HIFSA study are summarized and their significance assessed. The cardinal conclusions of the study are twofold: (a) Conceptual HIF power plants have estimated cost-of-electricity (COE) values that, at 1 GW (electric), are roughly comparable to those from other ICF and magnetic fusion system studies; and (b) HIF technology is robust in that there existsmore » a large parameter space region in which the COE is close to the minimum; i.e., the minima in COE are broad.« less

Structural Design Considerations for the FED 50-kA Equilibrium Field Coils

The structural support system for two equilibrium field coil conductor concepts is considered for the fusion energy design (FED) 8/10T baseline magnetic fusion system. Both conductor concepts are discussed. Results indicate that regardless of the conductor concept employed, an external support beam/frame structural system is required to equilibrate the accumulative loadings.

Ideal magnetohydrodynamic theory of magnetic fusion systems

Ideal magnetohydrodynamic theory and its application to magnetic fusion systems are reviewed. The review begins with a description and derivation of the model as well as a discussion of the region of validity. Next, the general properties are derived. These are valid for arbitrary geometry and demonstrate the inherently sound physical foundation of the model. The equilibrium behavior of the currently most promising toroidal magnetic fusion concepts are then discussed in detail. Finally, the stability of such equilibria is investigated. Included are discussions of the general stability properties of arbitrary magnetic geometries and of detailed applications to those concepts of current fusion interest.

Neutronics Shielding Analysis of the Last Mirror-Beam Duct System for a Laser Fusion Power Reactor

Three-dimensional Monte Carlo neutronics calculations for the last mirror-beam duct system for a laser-driven fusion power reactor conceptual design, where a three-section duct with focusing mirrors and right angle bends for the last two beam reflections is modeled. The estimate of neutron leakage flux after the second beam reflection is on the order of 10/sup 14/ n/(m/sup 2/ X s). Even though less in magnitude than the flux at the end point of neutral beam injectors in tokamak designs, reported in the range of 10/sup 17/ n/(m/sup 2/ X s), this still leads to a neutron dose rate of 10/sup 6/ rem/h after the second beam bend. Since space is not at a premium as in the case of magnet shields for magnetic confinement systems, minimum-cost lead mortar or concrete can be used as shielding materials instead of other minimum thickness costly materials. An energy leakage rate of 4.6 MW of 14-MeV neutrons per beam precludes the use of concrete behind the last mirror, suggesting the use of a secondary blanket composed of a lead acetate solution with immersed Boral (Al + B/sub 4/C) sheets. Boral as a penetration shielding material is found to reduce the thermal group flux aroundmore » the penetration by two to three orders of magnitude, compared to aluminum, and is recommended as a duct lining material for both inertial and magnetic fusion systems. Neutron heating rates and radiation damage parameters for the duct liner compare to those at the front face of the last mirror and first wall, and will require similar elaborate thermalhydraulic and mechanical designs.« less