Lower hybrid heating and current drive system for a tokamak reactor

A lower hybrid heating system has been designed for heating a tokamak reactor to ignition and for sustaining steady-state operation by driving the toroidal plasma current. The power spectrum from an active/passive waveguide grill is computed, and the resulting equilibrium current density profile is computed from a full electromagnetic WKB analysis of wave propagation in a cylinder. The corresponding toroidal current profile is a low-current equilibrium which is stable to various ideal modes at an economically acceptable beta. The electronic circuitry is designed to minimize the electric power required for current drive, and the resulting design appears to provide reliable operation in a reactor environment. The same system can drive current during reactor startup if some of the waveguides are modified slightly. A typical sequence of startup equilibria is calculated.     

Active control system upgrade design for lower hybrid current drive system on Alcator C-Mod

As a part of the scheduled expansion of the Alcator C-Mod lower hybrid current drive (LHCD) system from 12 to 16 klystrons to accommodate installation of a second LH antenna, the active control system (ACS) is being redesigned to accommodate the additional klystrons. Digitizers and output modules will be cPCI modules provided by D-tAcq Solutions. The real-time application will run on a standard PC server running Linux. Initially, the new ACS system will be designed to control 8 klystrons on the second LH antenna and the existing ACS will control the remaining 8 klystrons on the existing LH antenna. Experience gained operating the existing LHCD system has given us insight into the design of a more robust, compact, efficient and simple system for the new ACS. The design upgrade will be patterned on the digital plasma control system (DPCS [1]) in use on C-Mod.     

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, fabrication and testing of UHV compatible high power RF devices for lower hybrid current drive system on SST-1 tokamak

Maintaining plasma current under steady state conditions is one of the most important pre-requisites for a tokamak-based reactor. Lower hybrid current drive (LHCD) system aims to drive tokamak plasma current by means of RF power. The LHCD system on SST-1 tokamak is based on two 500 kW, CW klystrons operating at 3.7 GHz. A waveguide transmission line transmits power from source to the antenna. A phased array waveguide antenna is used to couple power to the plasma. The antenna side of the transmission line is placed inside the tokamak vacuum vessel. The design and fabrication of this In-Vessel system has to satisfy the demands of high power RF as well as ultra high vacuum (UHV) compatibility. This paper describes some of the critical UHV compatible In-Vessel RF devices, their design, fabrication, and test results.     

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.     

Proposed high voltage power supply for the ITER relevant lower hybrid current drive system

In the framework of the EFDA task HCD-08-03-01, the ITER lower hybrid current drive (LHCD) system design has been reviewed. The system aims to generate 24 MW of RF power at 5 GHz, of which 20 MW would be coupled to the plasmas. The present state of the art does not allow envisaging a unitary output of the klystrons exceeding 500 kW, so the project is based on 48 klystron units, leaving some margin when the transmission lines losses are taken into account. A high voltage power supply (HVPS), required to operate the klystrons, is proposed. A single HVPS would be used to feed and operate four klystrons in parallel configuration. Based on the above considerations, it is proposed to design and develop twelve HVPS, based on pulse step modulator (PSM) technology, each rated for 90 kV/90 A. This paper describes in details, the typical electrical requirements and the conceptual design of the proposed HVPS for the ITER LHCD system.     

Numerical analysis for the free-boundary current reversal equilibrium in the AC plasma current operation in a tokamak

In recent decades,tokamak discharges with zero total toroidal current have been reported in tokamak experiments,and this is one of the key problems in alternating current(AC)operations.An efficient free-boundary equilibrium code is developed to investigate such advanced tokamak discharges with current reversal equilibrium configuration.The calculation results show that the reversal current equilibrium can maintain finite pressure and also has considerable effects on the position of the X-point and the magnetic separatrix shape,and hence also on the position of the strike point on the divertor plates,which is extremely useful for magnetic design,MHD stability analysis,and experimental data analysis etc.for the AC plasma current operation on tokamaks.     

Resonant loop system for the KSTAR ICRF current drive

The phased current distribution at current straps for the KSTAR ICRF antenna causes a power imbalance at each strap owing to the mutual couplings between current straps. In order to mitigate the effect of coupling, a decoupler connecting two phased feeding lines are designed based on both a lumped element antenna model and a distributed transmission line model. Though the decoupler parameter is dependent on the loading resistance, which depends on plasma condition, an analysis shows that the decoupling is effective in the wide range of loading resistance assuming the low variation of mutual inductance between straps. A circuit analysis also shows that the RF characteristics of a complex RF transmission system are matched well for the asymmetric antenna current spectrum aiming for a non-inductive current drive of KSTAR. The calibration result of decoupler after installation is also discussed.     

Physics design of current drive and strategy of heating system for EHL-2 spherical torus

ENN He Long-2（EHL-2） is the next-generation large mega-Ampere（MA） spherical torus（ST）proposed and funded by the ENN company.The design parameters are:T_i0> 30 keV,n_e0～1×10²⁰ m^-3,I_p～3 MA,B_t～3 T.One of the biggest challenges of EHL-2 is how to achieve several MA current flat-tops with limited voltage-seconds（Vs） of the center solenoid（CS）coils.In order to minimize the consumption of Vs,a fully non-inductive start-up by electron cyclotron resonance heating（ECRH） will be applied in EHL-2.The ramp-up phase will be accomplished with the synergetic mode between the CS and non-inductive methods.The strategy of non-inductive start-up and ramp-up with synergetic mode has been verified on EXL-50U's experiments.Based on this strategy,numerical simulations indicate the feasibility of EHL-2 achieving 3 MA plasma current.A high-performance steady-state scenario with I_p～1.5 MA is also designed.In this scenario,the bootstrap current fraction f_BS> 70%,the safety factor q at the magnetic axis q₀> 2,the minimum safety factor q_min> 1,the poloidal beta β_p> 3 and normalized beta β_N> 2.3.Each design iteration integrates the validation of physical models with the constraints of engineering implementation,gradually optimizing the performance of the heating and current drive（H&CD） systems.Numerical simulation results for general auxiliary H&CD systems such as neutral beam injection（NBI）,electron cyclotron（EC） wave,ion cyclotron wave（ICW）,and lower hybrid wave（LHW） are presented.These simulation results ensure that the 31 MW H&CD systems comprehensively cover all scenarios while maintaining engineering feasibility.     

The design of the assembly tools for the ITER tokamak

The conceptual design of the purpose-built assembly tools required for ITER tokamak assembly is given. The ITER machine assembly is sub-divided into five major activities: lower cryostat, sector sub-assembly, sector assembly, ex-vessel, and in-vessel [1]. The core components, vacuum vessel (VV) and toroidal field coil (TFC), are assembled from nine 40° sub-assemblies, each comprising a 40° VV sector, two TFCs, and the associated VV thermal shield (VVTS). The lower cryostat activities must be completed prior to sector assembly in pit to prepare the foundations for the core components, and to locate the lower components to be trapped once the core components installation begins. In-vessel and ex-vessel activities follow completion of sector assembly. To perform these assembly activities requires both massive, purpose-built tools, and standard heavy handling and support tools. The tools have the capability of supporting and adjusting the largest of the ITER components; with maximum linear dimension 19 m and mass 1200 tonne, with a precision in the low mm range. Conceptual designs for these tools have been elaborated with the collaboration of the Korean Domestic Agency (KO DA). The structural analysis was performed as well using ANSYS code.     

MDSPLUS在EAST低杂波数据管理系统中的应用

随着EAST放电时间的增长,HT-7上采用的关系型数据库已经不能很好的满足要求,因此在EAST低杂波系统中我们采用了专门为核聚变开发的mdsplus数据库.     

Analysis and design of the Alfven wave antenna system for the SUNIST spherical tokamak

Analysis and design of the Alfven wave antenna system for the SUNIST spherical tokamak are presented. Two candidate antenna concepts, folded and unfolded, are analyzed and compared with each other. In the frequency range of Alfven resonance the impedance spectrums of both two concept antennas for major modes are numerically calculated in a 1-D MHD framework. The folded concept is chosen for engineering design. The antenna system is designed to be simple and requires least modification to the vacuum vessel. The definition of the antenna shape is guided by the analyses with constraints of existing hardware layouts. Each antenna unit consists of two stainless steel straps with a thickness of 1 mm. A number of boron nitride tiles are assembled together as the side limiters for plasma shielding. Estimation shows that the structure is robust enough to withstand the electromagnetic force and the heat load for typical discharge duty cycles.     

Personnel protection during the operation of Thomson scattering laser system on COMPASS tokamak

A recently installed Thomson scattering diagnostic on COMPASS tokamak uses two high power lasers. The presented protection system ensures laser safety of the personnel. Protection covers three areas–laser laboratory, spectrometer laboratory and tokamak hall. Laser protection system inputs are controlling the covering of the laser beam path, the entrance doors, the beam shutters and the laser cooling. Six regimes are defined for the protection system, covering all operation of the laser system, including laser service and low power beam alignment. Hardware implementation of the protection is based on PLC. The system is controlled via PCs with a touch screen. Connection to the COMPASS personnel protection system is described.     

The timing system on the J-TEXT tokamak

This paper describes the timing system designed to control the operation time-sequence and to generate clocks for various sub-systems on J-TEXT tokamak. The J-TEXT timing system is organized as a distributed system which is connected by a tree-structured optical fiber network. It can generate delayed triggers and gate signals (0 μs–4000 s), while providing reference clocks for other sub-systems. Besides, it provides event handling and timestamping functions. It is integrated into the J-TEXT Control, Data Access and Communication (J-TEXT CODAC) system, and it can be monitored and configured by Experimental Physics and Industrial Control System (EPICS). The configuration of this system including tree-structured network is managed in XML files by dedicated management software. This system has already been deployed on J-TEXT tokamak and it is serving J-TEXT in daily experiments.     

Physics of the conceptual design of the ITER plasma control system

The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints.In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma performance must apply multiple control functions simultaneously with a limited number of actuators. A sophisticated shared actuator management system is being designed to prioritize the goals that need to be controlled or weigh the algorithms and actuators in real-time according to dynamic control needs. The underlying architecture will be event-based so that many possible plasma or plant system events or faults could trigger automatic changes in the control algorithms or operational scenario, depending on real-time operating limits and conditions.     

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 of Remote Control System for EAST tokamak

The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.     

ECH system developments including the design of an intelligent fault processor on the DIII-D tokamak

A new generation fault processor is in development which is intended to increase fault handling flexibility and reduce the number of incomplete DIII-D shots due to gyrotron faults. The processor, which is based upon a field programmable gate array device, will analyze signals for aberrant operation and ramp down high voltage to try to avoid hard faults. The processor will then attempt to ramp back up to an attainable operating point. The new generation fault processor will be developed during an expansion of the electron cyclotron heating (ECH) areas that will include the installation of a depressed collector gyrotron and associated equipment. Existing systems will also be upgraded. Testing of real-time control of the ECH launcher poloidal drives by the DIII-D plasma control system will be completed. The ECH control system software will be upgraded for increased scalability and to increase operator productivity. Resources permitting, all systems will receive an extra layer of interlocks for the filament and magnet power supplies, added shielding for the tank electronics, programmable filament boost shape for long pulses, and electronics upgrades for the installation of the advanced fault processor.     

On the safety of conceptual fusion-fission hybrid reactors

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium.As a result of these studies, it appears that highly reliable and even redundant decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered.     

Improving concept design of divertor support system for FAST tokamak using TRIZ theory and AHP approach

The paper focuses on the application of the Theory of Inventive Problem Solving (TRIZ) to divertor Remote Handling (RH) issues in Fusion Advanced Studies Torus (FAST), a satellite tokamak acting as a test bed for the study and the development of innovative technologies oriented to ITER and DEMO programs. The objective of this study consists in generating concepts or solutions able to overcome design and technical weak points in the current maintenance procedure. Two different concepts are designed with the help of a parametric CAD software, CATIA V5, using a top-down modeling approach; kinematic simulations of the remote handling system are performed using Digital Mock-Up (DMU) capabilities of the software. The evaluation of the concepts is carried out involving a group of experts in a participative design approach using virtual reality, classifying the concepts with the help of the Analytical Hierarchy Process (AHP).