Optimization of design point for a fusion energy systems integration test facility FAST

Physics and technology maturity level required for the K-DEMO design points

The Navy has had a long standing interest in power applications of superconductivity. Recent advances in high temperature superconducting (HTS) materials have caused the Navy science and technology community to assess the potential of HTS magnets for Naval applications. A program has begun to test HTS conductors and magnets in order to determine realistic performance expectations from which to determine proper systems integration. This paper presents some early results of industrially produced HTS magnets and discusses possible Naval applications. Speculative application toward magneto-hydrodynamic (MHD) propulsion is also discussed-specifically with respect to HTS materials.

The stored energy and neutron yields of TFTR plasmas have been modelled, with small dispersion, by considering a parameter associated with the peakedness of the beam fuelling profile as an independent variable. Scalings with fixed multiplicative constants and exponential powers determined from the experimental data are applied to deuterium discharges at different major radii in TFTR, and also to deuterium-tritium (DT) plasmas. The domain of the data considered includes the supershot, high poloidal beta, L mode and limiter H mode operational regimes, as well as discharges with a reversed shear (RS) magnetic configuration. The role of plasma parameters, such as plasma current Ip, edge safety factor qa and toroidal field Bt, are considered in this study. Improved energy confinement and neutron production are relatively insensitive to these parameters compared with the beam fuelling peakedness parameter when considering plasmas that are stable to magnetohydrodynamic modes. In contrast, plasma stability is dependent on these quantities and also on parameters describing the peakedness of the equilibrium profiles. On the basis of externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path towards a discharge condition with fusion power gain Q>1 is suggested

Recent U.S. fusion development strategy reports all recommend that the U.S. should pursue innovative science and technology to enable construction of a fusion pilot plant (FPP) that produces net electricity from fusion at low capital cost. Compact tokamaks have been proposed as a means of potentially reducing the capital cost of a FPP. However, compact steady-state tokamak FPPs face the challenge of integrating a high fraction of self-driven current with high core confinement, plasma pressure, and high divertor parallel heat flux. This integration is sufficiently challenging that a dedicated sustained-high-power-density (SHPD) tokamak facility is proposed by the U.S. community as the optimal way to close this integration gap. Performance projections for the steady-state tokamak FPP regime are presented and a preliminary SHPD device with substantial flexibility in lower aspect ratio (A = 2–2.5), shaping, and divertor configuration to narrow gaps to an FPP is described.

Innovation leads the way to attractive inertial fusion energy reactors–Prometheus-L and Prometheus-H

Single-particle orbits in a low aspect ratio axisymmetric configuration have been studied to clarify their role in current initiation using electron cyclotron waves. In this study, electron guiding centre orbits are studied neglecting collisions and charge exchange. Current is generated mainly by the toroidal drift of mirror-trapped electrons. Analytic scaling laws for the current and the poloidal field generation are derived, and they are found to be functions of temperature, poloidal field strength, number of electrons and parameters defining the magnetic and geometrical configuration. The current increases with temperature and decreases with the poloidal field strength as long as the temperature is much smaller than the characteristic temperature, whose thermal velocity is defined by the product of poloidal cyclotron frequency and the major radius.

CFETR which stands for “China Fusion Engineering Test Reactor” is a new tokamak device. The mission and goal of CFETR are as follows: (1) ITER-like; complementary with ITER; (2) Fusion power 50-200 MW; (3) Duty cycle time (or burning time)~(30-50%); (4) Tritium must be self-sufficiency by blanket. The main parameters of low temperature superconducting magnet system are as follows: (1) The central magnetic field in plasma area is designed as 5.0 T. (2) The major and minor radius of plasma is 5.7 m and 1.6m. The main parameters of CFETR's were optimized several times within past year according to the further physical target and engineering. Due to the restrictions of the critical magnetic field strength, low temperature superconducting conductor may difficulty applied in low aspect ratio tokamak device in a limited space. In order to achieve a much higher central magnetic field in plasma area and much higher maximum capacity of the volt seconds provided by center solenoid winding, it is possible to apply hybrid superconducting magnet to the TF coil and CS coil in a low aspect ratio design condition. In this paper, the hybrid magnet winding by using of Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and Bi <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2212</sub> are designed and detailed analyzed which can be selected as a reference for CFETR's magnet system design. The high temperature magnet part is Bi2212 conductor, its dimension is 30 mm×35 mm/turn and current density is 117.46 A/mm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . The low temperature magnet part is Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn conductor which can good work under 11T, its current density is about 80A/mm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . In addition, the hybrid superconducting magnet system will be expected to save space for sufficient tritium blanket due to reduced conductor size.

A spherical torus based volumetric neutron source (ST-VNS) concept has been developed as a possible intermediate step to develop the necessary technology for reactor components of future fusion power plants. Such a VNS would complement ITER in testing, developing and qualifying nuclear technology components. Two recently design concepts for a spherical torus based VNS have been investigated. The initial design operated at 39 MW fusion power with a 0.8 m major radius yielding an average neutron wall loading of 1 MW/m/sup 2/. A higher fusion power (326 MW) ST-VNS design was conceived by increasing the major radius to 1.07 m yielding an average neutron wall loading of 5 MW/m/sup 2/. In this paper, we report the results of the neutronics analyses of this high power design along with several design modifications including the effects of using ITER materials in the key components. The results of a feasibility study of using the ST-VNS with suitably designed blankets to provide an intense neutron source for neutron science applications is also reported.

Configuration optimization is carried out for the heliotron-type fusion energy reactor FFHR. One of the important issues is to find sufficient clearances between the ergodic region outside the nested magnetic surfaces and blankets at the inboard side of the torus so that direct losses of alpha particles are minimized and the heat flux on the first walls is reduced. The latest design has a fairly large major radius Rc ∼ 17 m of the helical coils in order to satisfy this condition. It has been found, as an alternative design, that equivalent clearances are obtained with Rc = 15 m by employing a lower helical pitch parameter and splitting the helical coils in the poloidal cross-section at the outboard side. Furthermore, splitting the helical coils provides another modified configuration at Rc ∼ 17 m that ensures magnetic well formation in the fairly large nested magnetic surfaces with outward shifted configurations. From the engineering viewpoint, we propose that such helical coils be constructed by prefabricating half-pitch segments using high-temperature superconductors; the segments are then to be assembled on site with joints.

The nuclear reaction of deuterium–tritium (D–T) fusion by the usual magnetic or inertial confinement suffers from a number of difficulties and problems caused by tritium handling, neutron damage to materials and neutron-induced radioactivity, etc. The study of the nuclear synthesis reaction of deuterium–helium-3 (D–3He) at low collision energies (below 1 keV) is of interest for its applications in nuclear physics and astrophysics. Spherical tokamak (ST) reactors have a low aspect ratio and can confine plasma with β≈1. These capabilities of ST reactors are due to the use of the alternative D–3He reaction. In this work, the burn condition of D–3He reaction was calculated by using zero-dimensional particles and power equations, and, with the use of the parameters of the ST reactor, the stability limit of D–3He reaction was calculated and then the results were compared with those of D–T reaction. The obtained results show that the burn conditions of D–3He reaction required a higher temperature and had a much more limited temperature range in comparison to those of D–T reaction.

Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

The dispersion and damping rate of the ion acoustic waves (IAWs) in deuterium–tritium (D–T) fusion plasmas with slowing-down distributed α particles are investigated within the framework of kinetic theory. It is shown that the concentration of α particles has an important effect on IAWs. The frequency of IAWs decreases as the α concentration increases. Inversely, the damping rate increases as the α concentration increases. The results for slowing-down distributed α particles are compared with those obtained when α particles are assumed to have an equivalent Maxwellian distribution.

The current International Thermonuclear Experimental Reactor (ITER) design features a 6-m major radius, a 2.15-m minor radius, and a 22-MA plasma current. About 330 V-s in the poloidal field system inductively drive the current for hundreds of seconds. About 125 MW of neutral-beam, lower-hybrid, and electron-cyclotron power are provided for steady-state current drive and heating. Because the ITER design must account for the fact that the tokamak database is not exact, a reasonably cautious approach must be taken. There are some benefits of such an approach: the chances of success are enhanced, ITER traverses most of the distance toward the size and power of a commercial reactor, and it may be capable of more performance than promised. By extending the operating rules to the maximum expected range, the fusion power may be increased such that the machine might be capable of simulating a small net energy condition. If so, then considerable progress could be made toward fusion-power applications. >

This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with βN = 1.8, HH = 1.0, and fnGW = 0.85 had sufficient potential to achieve the electric break-even condition (net electric power ) under the following engineering conditions: machine major radius 6.5 m ⩽ Rp ⩽ 8.5 m, the maximum magnetic field on TF coils Btmax = 16 T, thermal efficiency ηe = 30%, and NBI system efficiency ηNBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are βN and fnGW. βN ⩾ 3.0 is required for with fusion power Pf ∼ 3000 MW. On the other hand, fnGW ⩾ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of Tave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs Pf (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius Rp ∼ 7.5 m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy.

One key quantity to be determined in the design of burning-plasma devices (CIT, ITER, reactors, etc.) is the level of plasma current (I) required to meet the desired plasma performance goals (ignition, high Q, etc.) and device objectives (fusion power, wall loading, current drive power, etc.). It is shown that these goals and objectives can be expressed in terms of the “figure-of-merit” parameter IAα/Rx [~ f(LAy)], where A is the aspect ratio, R is the major radius, L (= R, a) is the characteristic length, B is the toroidal magnetic field on axis, and the exponents α ~ 1 ± 0.5 and x ~ 0–0.5 (y ± 1–2) depend on the confinement assumptions and operational limits. To reach ignition or high Q, the main goal is to optimize IAα/Rx, subject to other engineering design constraints. In a CIT-like device (with R ~ 2 m, κ ~ 2, qΨ ≥ 3), the ignition requirement is I(A/3)α ~ 9–15 MA for “enhanced” L-mode (H-mode) confinement scaling expressions; an ITER-like device (with R ~ 5–6 m, κ ~ 2, qΨ ≥ 3) would require I(A/3)α ~ 15–25 MA. These requirements are embodied in the present CIT (with I ~ 11 MA, A ~ 3.25) and ITER (with I ~ 18–22, A ~ 3.1–2.6) designs.