Revisiting the poloidal magnetic confinement

This note proposes a closed poloidal magnetic configuration with an in-vessel coil system held by shielded supports. A dipole field is bounded by external coils and constrained into a hollow torus aiming at uniform intensity. In the horizontal mid-plane region the external coils and the dipole outer coils are broken in four arcs and bridged by couple of straight branches. Arcs and straight branches build a set of four side coils. In the clearance between their straight branches four tunnels in the poloidal magnetic field are achieved, to pass the supports and the feeders of the in-vessel coil system.A poloidal machine with a plasma thick as those of present large experiments is outlined. The dipole radius is 5.4 m, the plasma about it has a constant poloidal cross-section about 40 m², a volume about 1300 m³ and a minimum thickness 1 m in the outboard. The magnetic field ranges from 1.4 to 1.8 T.     

Gravitational flow of a thin film of liquid metal in a strong magnetic field

The influence of a poloidal magnetic field of the spherical Tokamak on super thin (h ≈ 0.1 mm) film flow of liquid metal driven by gravity over the surface of the cooled divertor plate is addressed. The experimental setup developed at the Institute of Physics, University of Latvia (IPUL) is described, which makes it possible to drive and visualize such liquid metal flows in the solenoid of the superconducting magnet “Magdalena”. As applied to the above setup, the magnetic field effect on the operation of the capillary system of liquid metal flow distribution (CSFD) is evaluated by using molten metal (lithium or eutectic InGaSn alloy) with a very small linear flowrate q ≤ 1 mm²/s, spread uniformly across the substrate. The magnetic field effect on the main parameters of the fully developed film flow is estimated for the above-mentioned liquid metals.An approximation technique has been proposed to calculate the development of the gravitational film flow. A non-linear differential second order equation has been derived, which describes the variation of the film flow thickness over the substrate length versus the flowrate q, magnetic field B and the substrate sloping α.Results of InGaSn film flow observations in a strong (B = 4 T) poloidal magnetic field are presented. Analysis of the video records evidences of experimental realization of a stable stationary film flow at width-uniform supply of InGaSn.     

Initial plasma start-up using partial solenoid coils in Versatile Experiment Spherical Torus (VEST)

Initial plasma start-up experiments based on ohmic discharge using partial solenoid coils located at both vertical ends of a center stack have been carried out in Versatile Experiment Spherical Torus (VEST) at Seoul National University. Ohmic discharges with the help of microwave pre-ionization have been performed according to the pre-programed start-up scenario which was experimentally verified by a series of vacuum field measurements using an internal magnetic probe array. A plasma current of around 0.4 kA has been achieved by ohmic discharge using partial solenoid coils, under the toroidal magnetic field of 0.1 T. The vacuum field calculation and fast camera image have revealed that the small plasma current even with significant amount of loop voltage up to 9.7 V is attributed to the imbalance of poloidal field for equilibrium. Modification of the start-up scenario and upgrade of power supplies are proposed to be carried out in order to achieve higher plasma current in the future experiments.     

Effect of magnetic field on the growth of Be films prepared by thermal evaporation

Grain refinement of beryllium deposits is studied as a significant subject for beryllium capsule in the Inertial Confinement Fusion project. The Be films were prepared on the Si (1 0 0) substrates by thermal evaporation with and without a magnetic field, respectively. The two separate groups of prepared Be films were characterized. The results showed the grain diameter in the Be film transited from 300 nm to 18 nm and the surface roughness of the Be film decreased from 61 nm to 3 nm by application of the magnetic field during the deposition process of Be coating. However, the Be film grown with the magnetic field was easily oxidized in comparison with that grown without magnetic field due to the refined grains, and the oxidation was gradually decreased with the increase of etching depth in the Be film. The reason for grain refinement of Be film was also qualitatively described.     

Further finite element structural analysis of FAST Load Assembly

The FAST (Fusion Advanced Study Torus) machine is a compact high magnetic field tokamak, that will allow to study in an integrated way the main operational issues relating to plasma-wall interaction, plasma operation and burning plasma physics in conditions relevant for ITER and DEMO. The present work deals with the structural analysis of the machine Load Assembly for a proposed new plasma scenario (10 MA – 8.5 T), aimed to increase the operational limits of the machine.A previous paper has dealt with an integrated set of finite element models (regarding a former reference scenario: 6.5 MA – 7.5 T) of the load assembly, including the Toroidal and Poloidal Field Coils and the supporting structure. This set of models has regarded the evaluation of magnetic field values, the evaluation of the electromagnetic forces and the temperatures in all the current-carrying conductors: these analysis have been a preparatory step for the evaluation of the stresses of the main structural components.The previous models have been analyzed further on and improved in some details, including the computation of the out-of-plane electromagnetic forces coming from the interaction between the poloidal magnetic field and the current flowing in the toroidal magnets.After this updating, the structural analysis has been executed, where all forces and temperatures, coming from the formerly mentioned most demanding scenario (10 MA – 8.5 T) and acting on the tokamak's main components, have been considered. The two sets of analysis regarding the reference scenario and the extreme one have been executed and a useful comparison has been carried on.The analyses were carried out by using the FEM code Ansys rel. 13.     

AC loss measurements on Cable in Conduit Conductor with random-changing magnetic fields

Beside the generation of the poloidal component of the field, the main function of the poloidal field coils in a tokamak is the control of the shape and the position of the plasma, according to the chosen plasma scenario. A plasma scenario, namely a sequence of plasma shapes, is obtained and controlled by varying the current in the PF coils. The control currents create magnetic fields having complex trends, being almost random waveforms with frequencies in the range between 0.1 and 10 Hz. As a consequence, AC loss is generated in the superconducting coils exposed to those signals, and the feasibility of a plasma scenario is strictly related to the ability to withstand and remove the heat coming from the AC loss.In order to study what the behavior of the loss is in random magnetic fields, namely similar to the control fields, a SULTAN sample is tested under two kinds of random field signals. The first signal is obtained by summing several harmonic frequency components, in the range between 0.2 and 6 Hz, having random amplitude. The second waveform is generated by a random function generator and it has a much broader spectrum of frequencies. The tests are carried out by varying also the maximum amplitude of the signals. The results are here discussed and compared to the results of the single frequency AC loss tests, and a correlation between them is studied.     

ICRF physics aspects of wall conditioning plasma characterization in TEXTOR

This paper focuses on encouraging results obtained on the characterization of RF produced plasmas during pulsed-mode wall conditioning discharges in ion cyclotron resonance frequency (ICRF) regime in the limiter tokamak TEXTOR. Recent Ion Cyclotron Wall Conditioning (ICWC) experiment carried out in TEXTOR tokamak, lead to the identification of various dependences of the antenna-plasma coupling efficiency on the plasma parameters for possible ICWC-discharge cleaning in ITER at half field. Our ICWC experiments emphasize on (i) study of antenna coupling during the mode conversion scenario, (ii) reproducible generation of ICRF plasmas for wall conditioning, by coupling RF power from one or two ICRF antennas and (iii) effect of application of an additional (along with toroidal magnetic field) stationary vertical (B_V ? B_T) or oscillating poloidal magnetic field (B_p ? B_T) on antenna coupling and relevant plasma parameters.     

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

Numerical study of corrosion of ferritic/martensitic steels in the flowing PbLi with and without a magnetic field

A computational suite called TRANSMAG has been developed to address corrosion of ferritic/martensitic steels and associated transport of corrosion products in the eutectic alloy PbLi as applied to blankets of a fusion power reactor. The computational approach is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. First, the new tool is applied to solve an inverse mass transfer problem, where the saturation concentration of iron in PbLi at temperatures up to 550 °C is reconstructed from the experimental data on corrosion in turbulent flows without a magnetic field. As a result, a new correlation for the saturation concentration C^S has been obtained in the form C^S = e^{13.604–12975/T}, where T is the temperature of PbLi in K and C^S is in wppm. Second, the new correlation is used in the computations of corrosion in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely hydrodynamic flows. In addition, the corrosion behavior was found to be different between the side wall of the duct (parallel to the magnetic field) and the Hartmann wall (perpendicular to the magnetic field) due to formation of high-velocity jets at the side walls. The side walls experience a stronger corrosion attack demonstrating a mass loss up to 2–3 times higher compared to the Hartmann walls. Also, computations of the mass loss are performed to characterize the effect of the temperature (400–550 °C) and the flow velocity (0.1–1 m/s) on corrosion in the presence of a strong 5 T magnetic field prototypic to the outboard blanket conditions.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current I_P = 10 MA, toroidal field B_T = 8.5 T, with a q₉₅ ≈ 2.3 that would correspond to I_P ≈ 20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated.     

Erosion and re-deposition processes in JET tiles studied with ion beams

Plasma facing components in fusion reactor chambers will operate under extreme conditions. Among the processes with implications on the material lifetime are erosion and re-deposition due to plasma interactions.This work will address the behaviour of both JET divertor and outer poloidal limiters (OPL) under plasma irradiation. The limiters comprise about 50 pairs of tiles in a poloidal stack, each of which has a plasma facing surface about 25 mm (poloidal) by 350 mm (toroidal) and is about 50 mm thick. The divertor tiles are located at the bottom of the chamber and withstand high fluxes of radiation and heat. Standard carbon-fibre composite (CFC) tiles coated with a thin layer of W overlaid with a 10 μm layer of C were studied with RBS/PIXE to understand the erosion/re-deposition processes occurring in these regions of the reactor chamber. High resolution surface morphology was assessed through SEM with and without tilting of the sample. The retention of hydrogen isotopes in the tiles were studied combining NRA and ERDA techniques – this is mostly ²H from the fuelling gas, but ³H is also present as a result of ²H–²H fusion reactions, and ¹H coming from the atmospheric exposure.     

Tiles chamfering and power handling of the MK II HD divertor

The JET high triangularity (δ, HD) divertor is an upgrade of the present JET divertor consisting of two modified toroidal segments which are: a new load bearing septum replacement plate (LB-SRP) tile located in the center of the divertor and a high field gap closure (HFGC) tile protecting inboard diagnostic cabling. The aim of the upgrade is to allow high power operation and a wider range of plasma triangularities at the divertor poloidal null. This paper describes the optimisation of the tile chamfering (including edge shadowing) and the power handling evaluation for a set of 12 planned plasma configurations given by the JET team and on two sets of mechanical tile tolerances issued by the JET drawing office. The PROTEUS code (magnetic equilibrium by finite element) is used to calculate the various field line angles, which are inputs for the chamfering angle calculation process. After calculating the chamfering angle values of each face, a checking exercise has been realised on the 3D CATIA models of the tiles by putting them at their extreme tolerance positions and validating if the shadowing is ensured for a angle calculated to take into account the worst possibilities. With the final chamfering angle value for each face, the power handling of the tiles has been estimated with finite element calculations. Power handling is given either by the critical time to reach 1800 °C at the tile surface for a total injected power of 40 MW or by the maximum total injected power allowable for a 10 s power pulse without exceeding 1800 °C. The estimated power handling gives promising results in regard to the JET EP project objectives.     

10 years of engineering and physics achievements by the Large Helical Device project

This article reviews 10 years of engineering and physics achievements by the Large Helical Device (LHD) project with emphasis on the latest results. The LHD is the largest magnetic confinement device among diversified helical systems and employs the world's largest superconducting coils. The cryogenic system has been operated for 50,000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of NBI, 2.9 MW of ICRF and 2.1 MW of ECH. Negative-ion-based ion sources with the accelerating voltage of 180 keV are used for a tangential NBI with the power of 16 MW. The ICRF system has full steady-state operational capability with 1.6 MW. In these 10 years, operational experience as well as a physics database have been accumulated and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantage of helical systems in LHD. Highlighted physical achievements are high beta (5% at the magnetic field of 0.425 T), high density (1.1 × 10²¹ m^?3 at the central temperature of 0.4 keV), high ion temperature (T_i of 5.2 keV at 1.5 × 10¹⁹ m^?3), and steady-state operation (3200 s with 490 kW). These physical parameters have elucidated the potential of net-current free helical plasmas for an attractive fusion reactor. It also should be pointed out that a major part of these engineering and physics achievements is complementary to the tokamak approach and even contributes directly to ITER.     

Direct numerical simulation of turbulent mixed convection in a vertical channel in a wall-normal magnetic field

Liquid metal coolants have a significant role in the design of advanced fusion reactors. There is a need for an investigation of the thermal behavior of the liquid metal in working reactor environment, such as when fluid flow at low Prandtl number (Pr) with a buoyancy effect, is subjected to a magnetic field. In the present study, a direct numerical simulation (DNS) for a low Pr number fluid flow resulting in turbulent heat transfer with buoyancy effect under a magnetic field has been carried out between two vertical plates kept at different temperatures. In this simulation, the values of the Hartmann number (Ha) were 0 and 6, Pr number was 0.06 and Grashof numbers were 6.4 × 10⁵, 9.6 × 10⁵, and 1.6 × 10⁶. The turbulent quantities of the parameters such as the mean temperature, turbulent heat flux, and temperature variance were obtained by direct numerical simulation (DNS). The Reynolds number (Re) for channel flow based on friction velocity averaged by both walls, viscosity, and channel half-width was set to be constant as Re_τ* = 150. A uniform magnetic field was applied in a direction perpendicular to the walls of the channel. The profiles of mean velocity and velocity fluctuations became asymmetric, and the tendency was enhanced with the increasing buoyancy effect. However, by the application of a magnetic field the tendency decreased. In other words, thermal transport between the walls became weak due to the magnetic effect.     

Test of piezo-ceramic motor technology in ITER relevant high magnetic fields

In the framework of a Fusion for Energy (F4E) grant, a test campaign started in 2012 in order to assess the performance of the in-vessel viewing system (IVVS) probe concept and to verify its compatibility when exposed to ITER typical working conditions. ENEA laboratories went through with several tests simulating high magnetic fields, high temperature, high vacuum, gamma radiation and neutron radiation.A customized motor has been adopted to study the performances of ultrasonic piezo motors technology in high magnetic field conditions. This paper reports on the testing activity performed on the motor in a multi Tesla magnetic field. The job was carried out in a test facility of ENEA laboratories able to achieve 14 T. A maximum field of 10 T, fully compliant with ITER requirements (8 T), was applied. A specific mechanical assembly has been designed and manufactured to hold the motor in the region with high homogeneity of the field. Results obtained so far indicate that the motor is compatible with high magnetic fields, and are presented in the paper.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.     

On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios.The research campaign has been carried out following a theoretical–computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal–radial region has been developed, including one breeder cell in the toroidal direction and all the five cells in the poloidal one. No Breeder Units have been modeled, their presence being simulated by effective thermo-mechanical loads.Two sets of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 8 MPa coolant pressure on its cooling channel walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 8 MPa coolant pressure on its internal walls while operating at normal operation thermal level. Results obtained are presented and critically discussed according to the SDC IC code.