Concept design of CFETR superconducting magnet system based on different maintenance ports

CFETR which stands for “China Fusion Engineering Test Reactor” is a new tokamak device. Its magnet system includes the Toroidal Field (TF) winding, Center solenoid winding (CS) and Poloidal Field (PF) winding. The main goal of the project is to build a fusion engineering Tokamak reactor with its fusion power is 50–200 MW and should be self-sufficiency by blanket.In order to ensure the maintenance ports design and maintenance method, this article discussed the concept design of the magnet system based on different maintenance port cases. The paper detailed studied the magnet system of CFETR including the electromagnetic analysis and parameters for TF (CS)PF. Besides, the volt-seconds of ohmic field are presented as detailed as possible in this paper. In addition, the calculations and optimizations of equilibrium field which should guarantee the plasma discharge of single null shape is carried out. The design work reported here illustrates that the present maintenance ports will not have a great impact on the design of the magnet system. The concept design of the magnet system can meet the requirement of the physical target.     

Proposal of an MFE-IFE cooperative system for a steady-state tritium balance

A new concept of a fusion reactor system, MFE-IFE cooperative system, is proposed. This concept combines the merits of a small-size MFE reactor and a dry-wall IFE reactor and aims at sufficient amount of tritium production and electricity generation without advanced technology. Design window analysis shows a NIF-scale (5 m chamber radius) dry-wall laser fusion reactor with a ～1 GWth fusion output and net tritium breeding ratio (TBR) of 1.74 can sustain an MFE power plant with a fusion power of 3 GWth and net TBR of 0.96. Although more detailed quantitative analyses are required, this concept can be a possible solution for a simultaneous achievement of tritium self-sufficiency and significant net electricity generation.     

System analysis study for Korean fusion DEMO reactor

A conceptual design study for a steady-state Korean fusion DEMO reactor (K-DEMO) has been initiated. Two peculiar features need to be noted. First, the major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8 T is expected to be achieved by using current state-of-the-art high performance Nb₃Sn strand technology. Second, a two-stage development plan is being considered. In the first stage, K-DEMO will demonstrate a net electricity generation but will also act as a component test facility. Then, after a major upgrade, K-DEMO is expected to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). Feasibility of such a practical, near-future demonstration reactor is studied in this paper, based on a zero dimensional system analysis code study. It was shown that a net electric generation on the order of 300 MWe can be achieved below the optimistic β_N limit of 5. The elongation of K-DEMO is around 1.8 with single null configuration. Detailed optimization process and the resultant various plasma parameters are described.     

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.     

Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing.Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin.The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force.Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios.The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ? 1 MW m^?2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ～ R^?2/7. While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system utilizing a high-field-side launch position.     

A preliminary conceptual design study for Korean fusion DEMO reactor

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb₃Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.     

Design of the dummy coil for magnet power supply

It is necessary to test it on a dummy coil, before using a magnet power supply (MPS) to energize a Poloidal Field (PF) coil in the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The dummy coil should accept the same large current from the MPS as the PF coil and be within the capability of the utilities located at the KSTAR site. Therefore a coil design based on the characteristics of the MPS and other restrictive conditions needed to be made. There are three requirements to be met in the design: an electrical requirement, a structural requirement, and a water cooling requirement. The electrical requirement was that the coil should have an inductance of 40 mH. For the structural requirement, the material should be non magnetic. The coil support structure and water cooling manifold were made of SUS 304. The water cooling requirement was that there should be sufficient flow rate so that the temperature rise ΔT should not exceed 12 °C for operation at 12.5 kA for 5 min. Square cross-section hollow conductor with dimensions of 38.1 mm × 38.1 mm was used with a 25.4 mm center hole for cooling water. However, as a result of tests, it was found that the electrical and structural requirements were satisfied but that the water cooling was over designed. It is imperative that the verification will be redone for a test with 12.5 kA for 5 min.     

Proto-CIRCUS tilted-coil tokamak–torsatron hybrid: Design and construction

We present the field-line modeling, design, and construction of a prototype circular-coil tokamak–torsatron hybrid called Proto-CIRCUS. The device has a major radius R = 16 cm and minor radius a < 5 cm. The six “toroidal field” coils are planar as in a tokamak, but they are tilted. This, combined with induced or driven plasma current, is expected to generate rotational transform, as seen in field-line tracing and equilibrium calculations. The device is expected to operate at lower plasma current than a tokamak of comparable size and magnetic field, which might have interesting implications for disruptions and steady-state operation. Additionally, the toroidal magnetic ripple is less pronounced than in an equivalent tokamak in which the coils are not tilted. The tilted coils are interlocked, resulting in a relatively low aspect ratio, and can be moved, both radially and in tilt angle, between discharges. This capability will be exploited for detailed comparisons between calculations and field-line mapping measurements. Such comparisons will reveal whether this relatively simple concept can generate the expected rotational transform.     

Research and development status on fusion DEMO reactor design under the Broader Approach

The main objective of DEMO design activity under the Broader Approach is to develop pre-conceptual design of DEMO options by addressing key design issues on physics, technology and system engineering. This paper describes the latest results of the design activity, including DEMO parameter study, divertor and remote maintenance. DEMO parameter study has recently started with “pulsed” DEMO having a major radius (R_p) of 9 m, and “steady state” DEMO of R_p = 8.2 m or more. Divertor design study has focused on a computer simulation of fully detached plasma under DEMO divertor conditions and the assessment of advanced divertor configuration such as super-X. Comparative study of various maintenance schemes for DEMO and narrowing down the schemes is in progress.     

Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.     

Assessing the feasibility of a high-temperature,helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m^?2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ～1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to be tested with relative ease, providing a new approach to reactor-relevant PMI science.     

Design and fabrication of the KSTAR in-vessel cryo-pump

In-vessel cryo-pump (IVCP) of the Korea Superconducting Tokamak Advanced Research (KSTAR) has been designed, fabricated, and installed in the vacuum vessel for effective particle control by pumping through a divertor gap. For the final engineering design of the IVCP supports to withstand all external forces, a structure analyses were performed for two cases. The first is the thermal stress due to cool-down from room temperature to operating temperature (cryo-panel: 4.4 K, thermal shield: 77 K), and the other is the electro-magnetic stress due to the induced eddy currents during plasma disruptions. When the plasma disrupts, the maximum stress and displacement on the supports were estimated to be 849 MPa and 5.36 mm, respectively. These results were taken into account in the support design. The IVCP system was fabricated in two half-sectors and a pre-assembling test was successfully completed in the factory. Final installation of the IVCP in the vacuum vessel was fulfilled in parallel with a pressurization test (thermal shield: 30 bar, cryo-panel: 10 bar), a helium leak test, and a thermal shock test using liquid nitrogen. As a result, the IVCP system was successfully installed in the vacuum vessel.     

An initial study of demountable high-temperature superconducting toroidal field magnets for the Vulcan tokamak conceptual design

Recent developments have made it possible to consider high-temperature superconductor (HTS) for the design of tokamak toroidal field (TF) magnet systems, potentially influencing the overall design and maintenance scheme of magnetic fusion energy devices. Initial assessments of the engineering challenges and cryogenic-dependent cost and parameters of a demountable, HTS TF magnet system have been carried out using the Vulcan tokamak conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T) as a baseline. Jointed at the midplane to allow vertical removal of the primary vacuum vessel and routine maintenance of core components, structural D-shaped steel support cases provide cryogenic cooling for internally routed YBCO superconducting cables. The cables are constructed by layering ～50 μm thick commercially available YBCO tape, and the interlocking steel support cases self align during assembly to form internal resistive joints between YBCO cables. It is found that designing the TF magnet system for operation between 10 K and 20 K minimizes the total capital and operating cost. Since YBCO is radiation-sensitive, Monte Carlo simulation is used to study advanced shielding materials compatible with the small size of Vulcan. An adequate shield is determined to be 10 cm of zirconium borohydride, which reduces the nuclear heating of the TF coils by a factor of 11.5 and increases the YBCO tape lifetime from two calendar years in the unshielded case to 42 calendar years in the shielded case. Although this initial study presents a plausible conceptual design, future engineering work will be required to develop realistic design solutions for the TF joints, support structure, and cryogenic system.     

Design of FPGA based high-speed data acquisition and real-time data processing system on J-TEXT tokamak

Tokamak experiment requires high-speed data acquisition and processing systems. In traditional data acquisition system, the sampling rate, channel numbers and processing speed are limited by bus throughput and CPU speed. This paper presents a data acquisition and processing system based on FPGA. The data can be processed in real-time before it is passed to the CPU. It provides processing ability for more channels with higher sampling rates than the traditional data acquisition system while ensuring deterministic real-time performance. A working prototype is developed for the newly built polarimeter–interferometer diagnostic system on the Joint Texas Experimental Tokamak (J-TEXT). It provides 16 channels with 120 MHz maximum sampling rate and 16 bit resolution. The onboard FPGA is able to calculate the plasma electron density and Faraday rotation angel. A RAID 5 storage device is adopted providing 700 MB/s read–write speed to buffer the data to the hard disk continuously for better performance.     

Spatially-resolved flat-field soft X-ray spectrometer on experimental advanced superconducting tokamak

A space- and time-resolved flat-field soft X-ray spectrometer with the wavelength range of 1–13 nm has been developed to study impurity behavior on the Experimental Advanced Superconducting Tokamak (EAST). Using an entrance slit, a varied line spacing grating (2400 grooves/mm at the grating center), and a charged coupled device (CCD) system, time evolution of profiles of impurity line emissions were recorded. The spectral resolution of the spectrometer is 0.006 nm at 5 nm when the width of entrance slit is set at 0.03 mm. The best spatial resolution obtained is 24.5 mm with the height of slit at 1.0 mm. The spectrometer is placed 8000 mm away from the plasma center and the observed spatial range covers 0–450 mm from the equatorial plane of EAST. The first experimental results were obtained from the recent EAST campaign. The system was shown to be capable of observing spectral lines from both intrinsic low-Z impurities (C, O, et al.) and highly ionized medium- and high-Z impurities (Fe, Cr, Ni, Cu, et al.). Spectral lines from the full wavelength range (1–13 nm) can be obtained by moving the position of the CCD. Spectra with the wavelength intervals of 1–2 nm show strong metal lines for H-mode discharges. Time evolutions of C VI (3.373 nm) and O VIII (1.897 nm) lines are presented and detail analysis is performed combining electron density intensity, D_α and soft X-ray and extreme ultraviolet (XUV) radiation intensities. Evolutions of profiles of C VI (3.373 nm) and O VIII (1.897 nm) at core plasma were also shown, indicating that the spectrometer can be applied for impurity transport studies,     

JT-60SA Toroidal Field Coils test cryostat development

Within the Broader Approach Agreement, Fusion for Energy will deliver to the Japanese Atomic Energy Association, amongst other components, the 18 Toroidal Field Coils (TFCs) for the superconducting Tokamak JT-60SA [1]. These coils will be individually tested at cryogenic temperatures and at the nominal current in a test cryostat. This cryostat is provided as an in-kind contribution by Belgium and is being developed jointly with CEA-Saclay/France.The vessel is large, oval shaped with an overall length of 11 m, a width of 7.2 m and a height of 6.5 m. To reduce the heat load to the coils the cryostat is covered by LN₂ cooled thermal shields. In addition to the cryostat, three test frames for the coils, the valve box vessel and the insulation vacuum system are also provided by Belgium. The Belgian contribution is design, manufacturing, assembly and test of the vacuum chamber, thermal shield and test frames by the Belgian company Ateliers de la Meuse (ALM), with the support of Centre Spatial de Liège (CSL). The TF coil test facility is assembled and the coil tests are performed by CEA/Saclay.The Belgian contribution, namely the design, manufacturing, assembly and test of the vacuum vessel, the thermal shields, and the test frames as well as of the vacuum pumping system are described in the presentation.     

Design and test program of a simplified divertor dummy coil structure for the WEST project

In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s).The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature.In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress.The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also addressed.