Central Solenoid Research Articles

Progress toward ITER’s First Plasma

ITER reached 50% completion of the work required to achieve First Plasma in November 2017. Progress has been made on ITER infrastructure since the 2016 Fusion Energy Conference, most visibly the construction of many key buildings. The key parts of the tokamak assembly building and the tokamak bioshield have been completed. The tokamak building itself will be ready for equipment in 2020. The cryogenic plant and the magnet power supply buildings are complete, and these systems begin commissioning in 2020. The power conversion and distribution area is complete and in operation, and construction has started on the component cooling water system building. Manufacturing of the basic components of the ITER tokamak is also proceeding well. The base and lower cylinder of the cryostat have been assembled on the ITER site. The first modules of the central solenoid and of the six poloidal field coils have been wound. The first winding packs of the toroidal field magnets are complete, as are the first casings, which have been verified to meet the high tolerances required. The first vacuum vessel sector is near completion and demonstrated to meet strict tolerances. The heating and current drive systems (neutral beams, electron cyclotron heating and ion cyclotron heating) are in the final design phase. The sequence of ITER operation from First Plasma to the achievement of the Q = 10 and Q = 5 project goals has been adapted to the Staged Approach to construction, a stepwise installation of all systems. The ITER Research Plan has been revised in 2017 to be consistent with the systems available in each phase. Physics studies focus on the disruption mitigation system, design of the ITER tungsten divertor, and modelling of ITER plasma scenarios. Modelling concentrates on the initial phases of the Research Plan and on the Q = 10 scenario, especially plasma termination.

A Hybrid Solenoid Coupler for Wireless Charging Applications

This paper introduces a primary pad for wireless charging applications, which is able to transfer high power over large air gaps and is tolerant to lateral displacements suitable for electric vehicle charging applications. The developed pad is an improved solenoid designed specifically to address issues such as flux leakage and loss while retaining good coupling. It is a hybrid structure that consists of a combination of rectangular- or square-shaped coils and a central solenoid, all connected in series. When energized, it produces a single-sided, polarized flux similar to the flux pattern of the solenoid and double D (DDP) pads. The power transfer capability of this pad is investigated through finite-element analysis simulations. The results are compared first to a similar sized solenoid from which it was derived, and also a DDP primary pad and are verified through measurements on prototype systems in the laboratory. The aim of this paper is to provide an improved hybrid solenoid pad showing higher coupling factors and lower stray field emissions with a performance (coupling and stray field emissions) similar to that of DDPs.

Large vacuum flux surfaces generated by tilted planar coils

Helical equilibria can be generated by arrangements of planar coils similar to tokamaks, but without a central solenoid and with the toroidal field (TF) coils tilted with respect to the vertical. This is known from earlier numerical works, e.g. Moroz 1995 Phys. Plasmas 2 4269. However, such concept tends to need large coils but form small plasmas (of large aspect ratio). Here it is numerically shown that larger, more attractive vacuum flux surfaces—relative to the size of the device—can be generated by carefully optimizing the inclination of the TF coils and currents in the various coil-sets. Vacuum configurations of aspect ratios as low as 4.1 are found for six tilted TF circular coils. Higher numbers of TF coils have advantages (smaller effective ripple) and disadvantages (lower rotational transform, smaller plasma). Finally, the aspect ratio A of the vacuum flux surfaces is quantified as a function of the ratio Ac of the coil radius to the radial location of the coil center. It is found that, in order to minimize A, it is beneficial to interlink or marginally interlink the TF coils (Ac ≲ 1).

Type tests of the ITER switching network unit components and protective make switches

In ITER, each circuit of the central solenoid as well as poloidal field coils PF1 and PF6 is provided with a system for plasma initiation, called the Switching Network Unit (SNU), able to supply voltage up to 8.5 kV for the coils. This will be obtained by inserting resistors in series with the pre-energized coils by means of a DC current commutation unit (CCU) composed of connected-in-parallel mechanical bypass switch and a thyristor circuit breaker.After a short description of the design, the paper focuses on the procedures and results of the type tests performed on the full-scale prototypes of CCU switching units, namely, the Bypass Open Switch (BPOS) and Bypass Make Switch (BPMS), which are able to perform commutation of direct current up to 55 kA at a voltage of 10 kV within 2–4 ms. In addition, Protective Make Switch (PMS) is considered, which is similar to BPMS in design. This device is intended to shunt the AC/DC converter supplying power to the coil in case of its failure or in case of a quench, when the energy stored in the magnet system is rerouted into fast discharge resistors (FDR) by a special fast discharge unit (FDU).Various types of electrical, hydraulic, pneumatic tests, performed in 2017–2018 are described as well as EMC tests and functional tests with control system.The successful results of the type tests confirmed the suitability of the design and compliance with the ITER requirements making it possible to start manufacturing of the switches for delivery to the ITER Site.

Safety factor profile control with reduced central solenoid flux consumption during plasma current ramp-up phase using a reinforcement learning technique

Safety factor profile control via active feedback control of electron temperature profile during a plasma current ramp-up phase of a DEMO reactor is investigated to minimize the magnetic flux consumption of a central solenoid (CS) for wide range of q profiles. It is shown that q profiles with positive, weak and reversed magnetic shear can be obtained with the resistive flux consumption less than 60% of the empirical estimation which is calculated using the Ejima constant of 0.45. For the optimization of the target electron temperature profile and feedback gain to control electron heating power, reinforcement learning technique is introduced. One important feature of the system trained by reinforcement learning is that it can optimize the target electron temperature adaptive to the present status of a plasma. This adaptive feature of the reinforcement learning enables to control q profiles even in the case that an effective charge profile is randomly modified and it is not measured directly.

Addressing the feasibility of inboard direct-line injection of high-speed pellets, for core fueling of DEMO

Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along “direct line of sight” (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard “DLS” paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step.

Preliminary electromagnetic loads calculation for the divertor of K-DEMO

The variation of plasma current and magnetic fields generated by superconducting magnet coils causes electromagnetic (EM) loads especially during the abrupt plasma current changes such as major disruption (MD) and the vertical displacement event (VDE) of plasma. The EM loads are one of the most important external loads for in-vessel components like blanket and divertor modules. The aim of this study is to estimate the EM loads on the K-DEMO divertor module under MD by using the commercial finite element software, ANSYS Mechanical/Emag. The superconducting magnet system of K-DEMO includes 16 toroidal field (TF) coils, 8 central solenoid (CS) coils, and 12 poloidal field (PF) coils. The magnetic field at the plasma center is about 7.4 T and the peak field is as high as ˜16 T. The conceptual model of the K-DEMO divertor module including outboard and inboard targets, dome, and the cassette body with connecting supports was developed. The 22.5° sector model applying symmetric boundary condition on cutting surfaces was employed to calculate EM loads efficiently. The reduced activation ferritic martensitic (RAFM) steel has been considered as the structural material of in-vessel components. Based on the current design concept of the K-DEMO, EM loads and eddy current are calculated.

MEST: A new Magnetic Energy Storage and Transfer system for improving the power handling in fusion experiments

A new magnetic energy storage scheme is studied for improving the power handling in fusion experiments: it can be applied both to tokamak or RFP experiments to supply the poloidal superconducting coils and can efficiently support the operation of the Central Solenoid (CS), without the need for resistive switching networks, thus with the advantage of energy dissipation avoidance. The basic idea, presented with reference to the CS circuit, is to provide an additional Superconducting Magnetic Energy Storage coil (TC), pre-charged along with the CS one to the same current value and then to transfer step by step the energy from one to the other and vice versa during the different phases of the plasma pulse, via switched capacitor. The principle is applied to the case of a Fusion-Fission Hybrid Reactor, based on Reversed Field Pinch configuration operated exploiting the flux double swing.

Fatigue Cyclic Testing of the ITER CS Inlets

ITER Central Solenoid (CS) is cooled by injecting supercritical helium at the inner diameter (ID), an area with the highest stress. The jacket near the helium inlet is the weakest structural element of the ITER CS due to the high stress concentration. To verify adequate mechanical performance of the inlets, we made six full-scale specimens and subjected them to relevant cycling loading in liquid nitrogen to assess the operational margin of the inlets.To increase fatigue life of the inlets, we used a treatment called ultrasonic peening (UP). This treatment allows for the creation of a compressive residual stress at the surface of the jacket with the highest stress, which significantly delays initiation of the fatigue crack. For comparison purposes, one of the samples was intentionally not UP treated. Test results showed significant advantages of the UP treatment and demonstrated sufficient life to support the ITER CS mission.

Assessment of Residual Stresses in ITER CS Helium Inlet Welds Fatigue Tested at Cryogenic Temperature

The ITER Central Solenoid (CS) consists of six independent wound modules. The cooling of the cable-in-conduit conductor is assured by a forced flow of supercritical He at 4.5 K supplied by He inlets located at the innermost radius of the coil. The inlets consist of a racetrack-shaped boss welded to the outer conduit wall through a full penetration Tungsten Inert Gas (TIG) weld. They are critical structural elements submitted to severe cyclic stresses due to the electro-magnetic forces acting on the coils. The weld contour is shape-optimised and locally processed by Ultrasonic Shot Peening (USP), conferring large compressive residual stresses on a subsurface layer of several millimetres thickness to improve fatigue strength. The distribution of the residual stresses and the effect of USP on microstructure and mechanical properties is assessed, with reference to the results of a cryogenic fatigue test campaign, performed on peened and as-welded inlets for comparison.

RESULTS OF PLASMA DISCHARGE OBTAINING EXPERIMENTS AT KTM TOKAMAK

Main experimental work results of obtaining plasma discharge at KTM tokamak are outlined in the paper. Results of initial prebreakdown magnetic field configuration scenario calculations by Plasmaless Tokscen numerical code and results of measuring parameters of selected scenario by diagnostic set of KTM are shown. All KTM electromagnetic system power supplies have been used during experiments on KTM tokamak. However, power supplies of central solenoid and toroidal magnetic field coils have been used only on its half project power, because of incompleteness of commissioning works.

AC losses calculation of Bi-2212 Cable-in-Conduit Conductor used in CFETR

Chinese fusion engineering test reactor (CFETR) is designed to bridge gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). Phase II of CFETR aims for DEMO validation with a fusion power over 1 GW, H-mode physics. In order to get this high operation parameters, the superconducting magnet system of CFETR has been designed based on the larger machine with R = 7.2 m, a = 2.2 m, BT = 6.5 T, and the maximum field at CS (Central Solenoid) Coils will be over 15 T. A hybrid superconducting CS magnet employing Bi-2212 round wires and Nb3Sn strands is considered for CFETR. In this paper, a design of 10 kA CICC twisted by Bi-2212 round wires at 15 T field and 4.5 K temperature was introduced, and a method was study to calculate AC losses of Bi-2212 cable in different indentation and strain conditions. In the CS coils, AC losses were calculated in detail, hysteresis loss was sensitive to indentation and strain, however, coupling losses were sensitive to indentation depth only.

AC Loss Evaluation of Poloidal Field for Experimental and Advanced Superconducting Tokamak

Experimental and advanced superconducting Tokamak (EAST) was designed by the Institute of Plasma Physics Chinese Academy of Sciences. The magnet system of EAST includes poloidal field (PF) coils, central solenoid (CS) coils, and toroidal field coils. The heat generated by the ac losses of PF and CS coils increases the burden of the cooling system, which reduces the critical current of the superconducting cable. Once the superconducting magnet quenches, it will pose a serious threat to the stability of the whole device. In this paper, the piecewise linear method and the finite-element analysis were used to analyze the ac loss of the EAST PF and CS coils under the specific operating current. The whole calculation process is described in detail. The final results meet the stability margin of the design of the EAST device. Meanwhile, the ac losses of PF and CS coils for EAST are theoretically evaluated in terms of the structure of magnet, the simplified current, and the distribution of the coil. The results of this paper not only provide data support for the next design and experiment of the EAST device, but provide some references for similar calculation of other large nuclear fusion magnets.

Experimental Evaluation of the NbTi Inter-Winding Joint for CFETR Central Solenoid Model Coil

The Central Solenoid Model Coil (CSMC) project of China Fusion Engineering Test Reactor (CFETR) was begun in 2014 to develop and verify the manufacture technology of large-scale superconducting magnets. The CFETR-CSMC coil is mainly composed of inner Nb3Sn windings and outer NbTi windings. There are 12 joints in total, including six interwinding joints and six transition joints. A simple test facility is required for joint experimental evaluation. In this paper, the test facility, sample assembly, and installation and calibration of Hall sensor are described in detail. The first test results of NbTi inter-winding joint are also given. The resistance of the measured NbTi inter-winding joint is about 0.55 nΩ and changes slightly in the current range from 10 to 20 kA.

Online Heat Load Test and Analysis for the Whole 5-K GA Feeder System

Heat conduction load test is a main and tough issue in the complex and huge general atom (GA) feeder system cold test. The GA feeder system is a superconducting system connects from the room temperature power supply system to the 4.5-K busbar of the magnet, supply 52 kA current and cryogenic provisions to the central solenoid magnet (CSM) coil, used for International Thermonuclear Experimental Reactor CSM magnet test at General Atomics Company, San Diego, CA, USA. This system have been developed and constructed at the Institute of Plasma Physics, Chinese Academy of Sciences, and completed the full cold test before transferred to the USA. In all the cold test items, the requirement of overall heat load 68 W is a big challenge, for its many leakage heat sources and cooling circuits: high temperature superconducting current lead, vacuum barrier, busbars, pipes, supports, etc. In order to get the steady results, many efforts were took to achieve the required value, such as reduce the inlet pressure of the 5-K loop as much as possible, minimize the flow rate of different 5-K cooling circuits, and at the same time, a dedicated program was developed to online calculate the density and the heat load while monitoring and analyzing the heat load. This paper will give a detail description of the 5-K overall heat load measurement.

Design updates of magnet system for Korean fusion demonstration reactor, K-DEMO

A conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. The superconducting magnet system is one of the key components of the K-DEMO and the preliminary study on superconducting magnet system has been done in 2015. The superconducting magnet system of K-DEMO consists of 16 TF (Toroidal Field), 8 CS (Central Solenoid) and 12 PF (Poloidal Field) coils. All of the TF, CS and PF coils use internally-cooled Cable-In-Conduit Conductors (CICC). A high performance Nb3Sn-based conductor, an ITER-type Nb3Sn and NbTi conductor will be used for the magnet. Based on the preliminary design, continuous design study is being performed and the K-DEMO magnet system has been evolved mainly for solving engineering issues. Two issues were considered for the TF, the reinforcement of the low field TF CICC structure and the fabrication stability of the high field TF CICC. In the CS magnet, reducing the number of Helium feed-throughs was a main issue. A small change has been made for the PF magnets for keeping a void fraction of the PF conductor. In this work, the major parameters and design updates of the K-DEMO magnet system is presented.

Control and operation strategy design of ITER coil power supply system

The large-scale ITER coil power supply system will generate huge current for superconducting magnets by plenty of AC/DC converters, which are based on thyristor technology on various circuits. A comprehensive consideration on design, manufacturing, testing and commissioning is the foundation to guarantee the safety operation of the CPSS system. This paper presents the design of converters control and operation strategies according to different typical circuit topologies. As well as illustrate the corresponding operation systems with associated functions to achieve those operation strategies. 2-Quadrant, 4-Quadrant and sequential control operation strategies are applied in corresponding PF (Poloidal Field), CS (Central Solenoid), CC (Correction Coil) and VS1 (Vertical Stability) converters with associated current closed-loop or voltage open-loop control during plasma operation. Three individual operation systems are comprised of TF, PF/CS and CC conventional systems with two architecture hierarchies to perform the control strategies and fulfill the real-time communication with central system. Typical simulation and test results are presented to verify the control strategies in accordance with the manufacturing and factory acceptance test. The converters will be installed in the ITER site soon and all the control strategies will be tested on the following commissioning phase.

A new test method of AC loss assessment for fusion conductors

The coupling currents loss for fusion conductors is frequently assessed applying a sinusoidal field sweep of fixed, small amplitude and variable frequency. From the initial slope of the loss curve, the coupling loss time constant is derived and applied in the loss calculation over the whole range of field transient. In this work, the traditional AC loss assessment is compared to an alternative experimental assessment made on a trapezoidal field change, which mimics the actual field change rate in the central solenoid of tokamaks during the plasma start-up. The comparative loss assessment is made using the test results of the Nb3Sn react&wind conductor developed at SPC and an ITER TF conductor. The proposed AC loss assessment allows an accurate estimate of the deposited energy during plasma start-up for the Central Solenoid.

Automatic identification of the plasma equilibrium operating space in tokamaks

In order to identify the plasma equilibrium operating space for future tokamaks, a new objective function is introduced in the inverse static free-boundary equilibrium code FEEQS.M. This function comprises terms which penalize the violation of the central solenoid and poloidal field coils limitations (currents and forces). The penalization terms do not require any weight tuning. Hence, this new approach automatizes to a large extent the identification of the operating space. As an illustration, the new method is applied on the ITER 15 and 17 MA inductive scenarios, and similar operating spaces compared to previous works are found. These operating spaces are obtained within a few (∼ 3) hours of computing time on a single standard CPU.

Sub-cable Butt Joint Design for CS Magnet of CFETR

The China Fusion Engineering Test Reactor (CFETR) is a new superconducting TOKAMAK fusion reactor. In Central Solenoid (CS) magnet of CFETR, the highest magnetic field in CS magnet will get 17.5 T,...