High temperatures inside tokamak for fusion research is achieved from auxiliary heating systems like neutral beam injectors (NBI), or RF heating devices, viz., ion cyclotron (IC), electron cyclotron and lower hybrid systems where High Voltage Power Supply (HVPS) is an essential requirement. HVPS based on pulse step modulation (PSM) topology has already demonstrated its ability for broadcast...
One of the key challenges for a fusion power plant is the need to increase the tritium burn-up fraction significantly from the values of only some 0.1% which result from extrapolation (at least above 5%). For a DEMO reference fusion power of 2 GW the fuelling rate necessary to replenish the burnt fuel is rather small (~ 2.7 Pa-m3/s) and the fuel burn-up fraction equal to the ratio of the burnt...
Tokamak reactors used in fusion plants usually require a high current supply. Different sets of coils are installed in order to confine the plasma current within the vessel, one of which is the vertical stabilization (VS) coil. The
current going through the coil will produce a magnetic field, which can control the position of plasma. For example, the VS coil in ITER requires a periodic pulse...
Ampegon has been working with the Karlsruhe Institute of Technology (KIT) to develop a novel design of 10MW power supply for multistage depressed collector gyrotron tubes. These new tubes offer potential for greatly improved control leading to greater output efficiency when coupled with a capable power supply. Ampegon’s new EPSM topology provides optimised control in two modes:
• HVDCPS...
Chinese Fusion Engineering Test Reactor (CFETR), which is under conceptual design to bridge gaps between ITER and DEMO, is envisioned to produce a fusion power (50-200 MW for phase I and up to 1GW for phase II ) with tritium breeding ratio (TBR)≥1.0 and a duty cycle time of approximately 0.3-0.5. This presentation will introduce the current work for the conceptual design of CFETR diagnostic...
This paper studies an inductor-capacitor-inductor(LCL)voltage-source converter(VSC)which can be implemented by zero current switches(ZCS). The ZCS has potential applications in improving the efficiency of high-voltage high-current system, such as servo power supplies at over 10 MW for the compressed plasma suggested in[Li, G. High-gain high-field fusion plasma,Scientific Reports, 2015, 5]....
J.M. Canik1, Z. Sun2, J.S. Hu2, R. Maingi3, R. Lunsford3, G.Z. Zuo2, W. Xu2, M. Huang2, X. C. Meng4, A. Diallo3, D. Mansfield3, T. Osborne4, K. Tritz5 and EAST Team
1Oak Ridge National Laboratory, Oak Ridge, TN, 37831
2Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031, China
3Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451
3Department of Applied...
Machine learning techniques, specifically neural networks (NN), are used with sufficient internal complexity to develop an empirically weighted relationship between a set of filtered X-ray emission measurements and the electron temperature (Te) profile for a specific class of discharges on NSTX. The NN response matrix is used to calculate the Te profile directly from the filtered X-ray diode...
Due to safety and reliability concerns, fast switch is required for the J-TEXT ECRH system to cut off the 100 kV power supply within 20 us when fault occurs. For its high switching speed, excellent controllability and small size, IGBT is an ideal choice as the basic component of the switch to satisfy the fast response requirement. Obviously, connecting multiple IGBTs in series is an essential...
One of the engineering challenges of magnetic confinement fusion is handling the high particle and heat fluxes incident on the divertor. Solid plasma facing materials must withstand these difficult conditions with a minimum of erosion, melting, and cracking. Liquid metals as plasma facing materials potentially alleviate all of these concerns, and are therefore being actively investigated for...
NSTX-U will, because of its low aspect ratio and unique capabilities, be a critical element in worldwide magnetic fusion energy research. It has 6MW of high harmonic fast wave heating at 30MHz, 10MW of neutral beam heating, lithium evaporation capability for wall conditioning, and coils for control of resistive wall modes.
An extensive set of diagnostics is planned for multiple purposes: to...
TRIPOLI-4 Monte Carlo transport code, developed by CEA, has been widely used on fission reactor physics and also can be used on fusion device neutronics. In order to verify the calculation features of TRIPOLI-4 code, a simple dogleg duct model was built to simulate the 14 MeV neutron transport based on a SINBAD fusion benchmark, called Dogleg Duct Streaming Experiment. The reaction rates in...
It is well known that the nuclear elastic scattering (NES) contributes to the slowing-down of suprathermal ions in thermonuclear plasmas [1]. So far several calculations have predicted that ion heating by energetic ions, i.e. transferred power from energetic to bulk ions, is enhanced due to NES [2,3]. NES can also modify the fusion reaction rates [4]. It is important to experimentally...
Assuring self-sufficient tritium production is of critical importance to prospective Deuterium Tritium (DT) fusion power plants.
Achieving a tritium breeding ratio in excess of 1.1 has been identified as a key requirement for future DT fusion power plants.
Tritium breeding ratio values are typically calculated via computationally expensive neutronics simulations as an integral stage in the...
The configuration of the Early Neutron Source (ENS) is the IFMIF-DONES (DEMO Oriented Neutron Source) approach, based on an IFMIF-type neutron source. It aims providing an intense fusion-like neutron spectrum with the objective to qualify on an accelerated time scale structural materials to be used in the future DEMO fusion reactor. IFMIF-DONES is based on the interaction of single 40MeV 125mA...
Boron carbide was proved as a practicable material of in-situ protecting coating for tungsten tiles of Tokamak divertor, which is also expected to be presented towards the other plasma facing materials (PFM) in fusion device. In the work, B4C coating on tungsten substrates by means of inductively coupled plasma (ICP) thermal spraying technique is studied, which is driven by a 24-60 MHz RF...
During the first operation phase OP1.1 of Wendelstein 7-X (W7-X) the magnet systems were not operated up to the maximum current. During the next operation phase OP1.2 the next step in the direction to a full current operation will be taken. Based on lessons learned during the first phase the necessary improvements have been worked out to deal with the challenges in OP1.2.
The superconducting...
Test Blanket Module (TBM) system in ITER facilities shall be designed, fabricated, and installed according to the construction code appropriate to the component class of TBM system. It is essential to properly define the component class of the system considering safety, quality, seismic and etc. Current construction codes have been well established and applied to nuclear power plants. However,...
The developed concept has long history of the development. The first work on this subject was published in 1993 [1]. Initially the concept developed as the tool of the system analysis of safety of the ITER reactor project, at this time on the basis of the conceptual project of the ITER reactor was developed multilevel (up to the 19th level of hierarchy) structurally functional hierarchical...
A control system based on compact reconfigurable I/O (cRIO)-9068 platform of National Instruments has been designed for the electron cyclotron resonance heating system on J-TEXT. The control system is mainly used for monitoring, timing, fast protection and slow protection of the ECRH system. The response time of fast protection is less than 10 μs based on voltage comparators and field...
The China Fusion Engineering Testing Reactor (CFETR) aims at bridging the gap between ITER and DEMO. Its scientific mission is to produce fusion power of 200 MW with tritium self-sustention and duty cycle of 0.3-0.5. The big fusion power and the auxiliary heating power of 100-140 MW, makes the design of CFETR divertor challenging. Previous work focuses on the plasma configuration and the first...
A helium cooled liquid lithium (or lead lithium) concept has been developed to design a liquid breeder blanket in Korea. Ferritic-martensitic steel (FMS) was selected as a structural material for fusion reactors, and a commercial-scale Advanced Reduced Activation Alloy (ARAA) has been developed. An Experimental Loop for a Liquid breeder (ELLI), which we designed and fabricated ourselves, was...
The China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak proposed by the China National Integration Design Group. The missions of CFETR are achieving 50~200MW fusion power, and steady-state operation with the duty circle between 0.3 and 0.5.Breeding blanket, a core components of fusion reactor, takes the role of tritium breeding and energy conversion. It located inside...
In order to carry out long pulse and high power NBI heating experiment on HL-2M tokomak, the high pumping speed pump is very necessary, which can rapidly remove the gas load from vacuum vessel, reduce the affect of the background gas in NBI injector on the tokomak plasma and ensure that the re-ionization loss of neutral beam in drift duct is less than 5%, so that a cryopump with large area...
The European DEMO programme is running an activity which aims to develop a self-consistent and fully integrated design of the Tritium, Matter Injection and Vacuum Systems, supporting a tokamak operation with a high burn-up fraction. The architecture of the DEMO fuel cycle is mainly driven by the need to reduce the tritium inventory in the systems to an absolute minimum. This requires the...
Neutral beam injection(NBI) is an essential plasma heating tool for the China Fusion Engineering Test Reactor(CFETR),that is under engineering conceptual design.The CFETR NBI of energy higher than 500keV is needed. Because of neutralization efficiency of negative ions is higher than that of postive ions under the high energy,the negative ion source is required for NBI,which ask for higher...
Abstract: To simulate the magnetic field environment in EAST Tokamak, and to study the flowing of liquid metal driven by electromagnetic force, and then to guide the liquid lithium limiter experiment in EAST preferably, it’s necessary to develop a magnet system. In order to ensure the experiments are carried out smoothly, the running time of magnet system should be more than 1000s, the...
In order to achieve high requirement of EAST Tokamak, ion cyclotron range of frequency(ICRF) heating is utilized as one of the main auxiliary heating methods, which plays an important role in the coupling RF power to the plasma. During the operation of ICRF, there will be a large amount of heat flux on the surface of antenna and consequently the structural stability and reliability of the...
Three concepts of tritium breeding blanket have been proposed for Chinese Fusion Engineering Test Reactor (CFETR). One of the concepts is helium-cooled ceramic breeder (HCCB) blanket. The HCCB blanket have the S-type cooling pipes in breeding unit (BU), and the BU consists of lithium ceramic pebble beds and beryllium pebble beds. The breeding material and the multiplier material separated by...
PRIMA (Padova Research on ITER Megavolt Accelerator) is a large experimental facility under construction in Padova, Italy, aimed at the development and test of the full scale prototype of Neutral Beam Injectors (NBI), called MITICA, for ITER.
MITICA is designed to accelerate a beam of 40 A of negative deuterium ions up to 1 MV, in order to deliver a power of about 17 MW to the plasma with a...
Chinese Fusion Engineering Test Reactor (CFETR) is under design, which will be operated in two phases [1]. In phase I, CFETR is envisioned to provide 200 MW fusion power Pf and its designed main parameters are R=6.6 m, a=1.8 m, BT=6-7 T, IP=10 MA. However, in phase II which aims to DEMO validation, the Pf is over 1 GW and the IP increases to 11 MA. Considering the large Pf, it will be a...
In 2018, EAST will be operated with a full tungsten (W) divertor in both the upper and lower divertors. Tungsten is a shiny refractory metal; as such, its emissivity in the infrared (IR) range is low. In addition film formation on the tungsten alters the emissivity, which makes precise surface temperature measurements difficult for conventional single-band IR cameras. To resolve this problem,...
A current-pulsed power supply (CPPS) with rapid rising and falling edges, which is used for tearing mode (TM) feedback control, has been developed for magnetic perturbation coils on the J-TEXT tokamak. A bleeder resistor ranging from 200 mΩ to 1000 mΩ is required in the CPPS. When CPPS works regularly in 0.3 seconds with frequency ranging from 1 kHz to 3 kHz, the bleeder resistor will generate...
As one of the core equipment of ITER DC steady-state test platform, high precision power supply is a large capacity AC/DC/AC single-phase inverter with current source characteristics, which provides ± 2000V / 500A output. The difficulties of high precision power supply are high voltage and large current. To meet requirement of high precision output and fast response, the scheme of cascaded...
Alfvén Eigenmodes are instabilities which are considered to be excited by high-energy particles in tokamak plasma. In the future fusion reactor, Alfvén Eigenmodes may change the distribution and transportation of alpha particles. Therefore, the study of Alfvén Eigenmodes is very significant. In present tokamaks with relatively low parameters, external antennas are often used to excite the...
Abstract: The vacuum vessel (VV) of Chinese Fusion Engineering Testing Reactor (CFETR) is a D-shape, double-layer and toroidal structure, which have a high precision requirement, a large-size and too weight, the challenge can be foreseen in the VV manufacturing. As an important control means of the quality, the hyperboloid surface of VV should be inspected to ensure qualified manufacture such...
The development of radio frequency (RF) negative ion sources for neutral beam systems requires knowledge of the plasma parameters. Optical emission spectroscopy (OES) is a non-invasive and in situ diagnostic tool, so optical emission spectroscopy diagnostic system are designed to be applied to the measurements of the RF negative ion source, and diagnostic principle and simplified analysis...
Fusion reactors are complex machines in which many systems operate in concert to achieve the required behaviour. Controlled fusion is dependent on a fine balance between these systems, whose functions often overlap or conflict. A tokamak’s functions can be described by simple terms like fuelling, heating, current generation or plasma stabilisation, but their technological realisation can be...
Reduced activation martensitic/ferritic steel has been selected as the first wall material of ITER testing blanket modules (TBM). The first wall is subjected to hydrogen isotope permeation by the two mechanisms: one is plasma-driven and the other is gas-driven, which may result in tritium safety and extraction issues. Meanwhile, to evaluate hydrogen isotope permeation and inventory in the first...
Tungsten (W) is considered to be the most viable armor material for the plasma-facing components (PFC) of a fusion reactor [1]. The work under fusion plasma will lead to modification of W that would change, in turn, its erosion properties, subsequent redeposition on surface, and would influence gas inventory (tritium, T) in material. Hydrogen (H) in W easily diffuses deep into W bulk even from...
Power and particle handling in the plasma edge region is one of the critical issues, affecting the successful operation of a steady state magnetic fusion power reactor. Tungsten has widely been employed for plasma-facing components in existing fusion experiments and is envisaged to be used for the ITER divertor [1]. Unfortunately, conventionally available tungsten is known to suffer from...
Developing the Simulation of Spectra Code Based on HL-2A tokamak Motional Stark Effect and Beam Emission Spectroscopy
J.WU1, Y.C.CHEN1, P.CHEN1, L.M.YAO1, H.Y.ZHOU2, Y.Liu2, J.L.FU1
1. School of physical electronics, university of electronic science and technology of china
610054, Chengdu, China
2. Southwestern Institute of Physics, 610041
Abstract: Beam emission spectroscopy (BES) diagnostic...
A new technology for developing fusion energy is to use hydrogen isotopes i.e., deuterium (D) and tritium (T). It is a combine effort for building up of International Thermonuclear Reactor (ITER) named as Tokamak, which will come into operation in 2020. Handsome amount of work has already been done by many researchers contingent with plasma shape, halo current and plasma equilibrium properties...
The main I&C (Instrumentation and Control) functions of the ITER VUV (Vacuum Ultraviolet) spectrometer have been prototyped based on ITER CCS(CODAC Core System) and tested at KSTAR. The ITER VUV spectrometer consists of VUV core, divertor VUV and VUV edge, which are expected to use a common VUV detector model, Andor BI(back-illuminated)-CCD(Charge-Coupled Device). While many other auxiliary...
Present 6 MW NBI system is one of the important heating device in KSTAR. In 2016 campaign, the system contributed to high betaN, and almost steady-state operation of 78.3 seconds. In addition to the present system, the KSTAR has plans to enhance heating and current drive. For this plan, one more 6 MW NBI system will be installed in 2017, and will be operated from 2018. Two of three ion sources...
The ELM-like transient high-heat flux generates melt-layer formation of tungsten (W), melt motion and droplet ejection, leading to surface erosion of plasma facing components in large fusion devices such as ITER. This paper will present the experimental investigations of dynamics of W droplet splashing with including the stabilization effects of the magnetic field, which have been performed by...
Water cooled ceramic breeder (WCCB) blanket is being developed for China Fusion Engineering Test Reactor. The water with inlet temperature of 285℃ and pressure of 15.5MPa is adopted to remove the considerable heat in the blanket, which may boil under the accident condition of loss of coolant. Periodic flow pulsation may be generated in the parallel coolant channels of the blanket. As a result,...
A scaled mock-up of the China Helium-Cooled Ceramic Breeder Test Blanket Module (CN HCCB-TBM) was installed to the J-TEXT in order to study eddy current distribution, electromagnetic load and thermal load on the TBM during plasma disruption. J-TEXT TBM mock-up using reduced activation ferritic/martensitic (RAFM) steel as structural material. The measurement experiments investigated the effects...
China Fusion Engineering Test Reactor (CFETR) integration design platform, which is intended to provide a unified environment to integrate physical and engineering design for future reactor-level fusion device, is now under development. It includes a physical design platform and various engineering design modules, such as vacuum vessel, divertor, toroidal/poloidal field coil, blanket, thermal...
As of this day, two major magnetic fusion research projects are held at the Plasma Laboratory for Fusion Energy and Applications at Instituto Tecnológico de Costa Rica (Costa Rica Institute of Technology. The current status of both devices is summarized.
On June 29, 2016, the Stellarator of Costa Rica 1 (SCR-1) produced its first hydrogen plasma, becoming the first Stellarator of Latin...
Deuterium plasma experiments in the Large Helical Device (LHD) will begin in March 2017. In LHD, neutrons are mainly generated by interaction between bulk plasmas and beam ions. Therefore, neutron emission profile measurement plays an important role in the understanding of confined beam-ion behavior.
The vertical neutron camera (VNC) has been developed to measure neutron emission profile in...
Demonstration of tritium self-sustainability is among the key targets of the China Fusion Engineering Test Reactor (CFETR). The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for CFETR. However, tritium retention and permeation in blanket systems can become a bottleneck for the design and operation of future fusion devices with respect to economic and safety...
SPIDER experiment, currently under construction at the Neutral Beam Test Facility (NBTF) in Padua, Italy, is a full-size prototype of the ion source for the ITER Neutral Beam (NB) injectors part of the ITER project.
The Ion Source and Extraction Power Supplies (ISEPS) for SPIDER are supplied by OCEM Energy Technology s.r.l. (OCEM) under a procurement contract with Fusion for Energy (F4E)...
Lithium wall conditioning on EAST have been employed since 2009. The high performance plasma (H-mode) has been recently successfully obtained with the help of lithium coating since the autumn campaign of 2010[1]. And stationary H-mode plasmas over 30s was obtained in 2012. Lithium evaporators and real-time Li power/granules injection are used for the stationary H-mode plasma and ELMs control....
To achieve real-time control of tokamak plasmas, the equilibrium reconstruction have to be completed rapidly enough. For EAST experiment case, real-time equilibrium reconstruction is generally required to provide results within 1ms. A GPU parallel Grad-Shafranov solver is developed in P-EFIT code[1], which is built with the CUDA™ architecture to takes advantage of massively parallel Graphical...
The paper describes the high priority testing activities supporting the ITER Radial Neutron Camera (RNC) design, performed by a consortium of European institutes within a framework contract placed by Fusion For Energy (F4E), the ITER European Domestic Agency.
The main role of the RNC is to measure the uncollided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium...
Boron carbide (B4C) is low-Z material with good chemical stability and effective neutron absorption, so it has received attention for application in nuclear fusion reactors and plasma facing material in fusion devices. B4C coatings are successfully deposited by inductively coupled plasma (ICP) torch, and the results indicate that plasma gas composition has great affection on melting process of...
Understanding of tritium behavior in the plasma facing wall of a fusion reactor is important from viewpoints of fuel control and tritium safety. Tungsten is a primary candidate of plasma facing material because of low sputtering rate and low tritium solubility. However, even if tungsten is used on plasma facing wall, formation of deposited layer cannot be avoided in long-term operation of the...
Tungsten (W) has been proposed as the candidate plasma-facing material for the divertor of International Thermonuclear Experimental Reactor (ITER) because of its beneficial properties such as high melting point, high thermal conductivity and low sputtering yield [1]. For a DEMO reactor, surface coatings made of W are necessary to protect the plasma-facing wall made of reduced activation...
ABSTRACT
As a fully superconducting tokamak device, the Experimental Advanced Superconducting Tokamak (EAST) bears the mission to demonstrate high-power, long-pulse plasma operation with flexible plasma configurations [1]. In 2014, an ITER-like actively water-cooled tungsten (W) divertor was installed in EAST. The castellated monoblock structure in the W divertor facilitates the power...
An improved inboard divertor has been designed for NSTX-U with higher divertor heat flux capability, robustness to halo current strikes and improved bakeout temperatures. The ATJ graphite inboard divertor tiles of the original NSTX-U design may have been subject to radial forces from halo currents which were not effectively resisted by the t-slot tile clamping. The clamping force could have...
Improvements of Heating & Current Drive (H&CD) systems are being investigated for a demonstration fusion power plant DEMO to deliver net electricity for the grid around 2050 [1]. Compared to ITER, which has to show the generation of 500 MW thermal power, the target of DEMO is the successful production of 300 to 500 MW electrical power to the grid and to aim for a self-sufficient Tritium fuel...
First wall panels (FWP), which adjoins along the inner wall of the vacuum vessel (VV) of Tokamak device, are multilayer structures different materials welded by solid welding technique to perform heat exchange enhancement, VV protection and tritium breeding functions. In order to implement online inspection of the delamination defect of FWPs nondestructively, a NDT method capable to make...
The contact resistance effect in the interface between pebble beds and the structure was studied. The lithium ceramics is used as breeder with the form of sphere-shaped pebbles for the extraction of the tritium in some TBM candidates of ITER. It could act as a thermal resistance in the interface and affect the pebbles and structural material temperature. Some models related to the contact...
China Fusion Engineering Test Reactor (CFETR), designed as a bridge connecting ITER and DEMO, was proposed to achieve long-term stable operation with 30–50% duty time factor at low fusion power (50–200 MW). First wall of CFETR services on the conditions with high surface heat flux and intense neutron irradiation. The existing structural design rules for first wall mainly involved stability...
In the fusion reactor, there are large heat loads in the first wall of facing high temperature plasma to be large temperature differences in the cladding walls to form natural convection and there is a magnetic field to damp out or to stabilize the fluid flow of the liquid metal. Natural convection under a magnetic field is different from the general fluid and in-depth study has important...
Between inner and outer shells of vacuum vessel, numerous in-wall shielding (IWS) blocks are installed to provide neutron shielding. Uniquely, the IWS ribs in sector 1 and 6 are manufactured by welding and its manufacturing design shall be secured for not only manufacturability but safety point of views.
For design of IWS ribs, complex loads for multiple directional electromagnetic force,...
In China Fusion Engineering Testing Reactor (CFETR) research, the blanket neutronics experiment is essential in validating the neutronics codes and tools used in blanket. The neutron activation method, supported by neutron transport calculations, is particularly useful in the estimation of the neutron intensity in the blanket, which based on the estimation of the activation reaction rate.
...
Two sets of three coils are positioned near the upper and lower divertor in NSTX Upgrade. These are collectively called the inner PF coils and consist of PF1a,b,c upper and lower. These are used in strike point position control, advanced divertor configuration studies, and Coaxial Helicity Injection (CHI) experiments. The NSTX Upgrade Inner PF coils have low Lorentz force derived stresses...
Modeling of pre-Thermal Quench and Thermal Quench stages
of disruption induced by Massive Gas Injection in ITER
V. Leonov, S. Konovalov, V. Zhogolev
NRC "Kurchatov Institute", Moscow, Russian Federation
Abstract
To reduce energy loads on the first wall and divertor during disruption in ITER it is necessary to re-radiate more 90% of the thermal energy using Disruption Mitigation System...
In fusion application, helium embrittlement is a key inducement to deteriorate mechanical performance of structural steels. To elucidate the mechanisms of helium induced embrittlement of grain boundaries (GBs), molecular dynamics (MD) was used to simulate GB tensile under the effect of helium bubbles in bcc iron at atomic level. Stress-strain curves and snapshots of configuration during...
The purpose of the effort described here is to model, and monitor the insulation shear bonds between the 3 conductors of the TF outer legs. Mechanical failure of the insulation could be a precursor to an electrical failure that could damage the more difficult to repair TF inner leg central column. The shear stress in these bonded layers is proportional to the TF outer-leg out-of-plane (OOP) ...
Several novel design solutions for high performance cooling systems have been developed and realized by Consorzio RFX, permitting to experimentally simulate the challenging heat transfer conditions foreseen in the future fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), has the triple objective to verify the present solution applied inside the...
The fundamental parameters calculations addressing tritium breeding ratio (TBR), neutron wall loading (NWL) and nuclear power generation on a Chinese Fusion Engineering Testing Reactor (CFETR) neutronic analysis model were performed using MCNP code to investigate the feasibility of the helium cooled pebble bed breeder blanket. The neutronic model was created as a 11.25°torus sector of tokamak...
China Fusion Engineering Test Reactor (CFETR) is a tokmak fusion experimental device under design to bridge the R&D gaps between ITER and DEMO. Helium Cooled Ceramic Breeder (HCCB) blanket is one of the candidate blanket concepts for CFETR. Blanket with S-shaped lithium zone and cooling pipes reduces the space of helium manifold and is conducive to tritium breeding. The neutronics analyses of...
Plasma disruption would induce large eddy current in the first wall (FW) and other in-vessel components of the Tokamak system. With the huge confinement magnetic field in the Tokamak structure, huge electromagnetic force may generate in the in-vessel components. The study on the relationship between the plasma disruption and mechanical stress and strain in the FW and other in-vessel components...
The knock-on tail formed by nuclear elastic scattering (NES)[1] due to high energy particles gives various effects to fusion plasma. So far, observation experiment of knock-on tail was only conducted by measuring knock-on tail due to NES caused by alpha-particles using deuterium-tritium plasma at JET [2]. In this experiment, however, quantitative estimation of NES effect is insufficient. We...
Tungsten (W) is the most promising plasma facing material for fusion devices, while copper (Cu) has been proposed as the heat sink material behind plasma facing material[1]. Nevertheless, because of the large difference of coefficient of thermal expansion (CTE) between W and Cu, the joining of these two dissimilar materials causes the high thermal stress concentration at the interface when...
A part of the current fusion mission is to demonstrate that fusion experiments and fusion power plants can be operated in a safe manner for the workers, the local population, and the environment. This paper describes some of the present issues in personnel safety in magnetic fusion environments, including the present-day personnel protection limits. The historical trends of these protection...
A medium-size High toroidal magnetic Field Ultra Low Aspect Ratio Tokamak (HF-ULART) is proposed. The major objective of this is to explore the highest beta limit possible under the maximum toroidal field(TF) to have also high plasma pressure, using present day technology and achievements of tokamak fusion research.
This is the right pathway scenario to initiate studies for a potential...
One aspect of fusion power plant engineering with little insight is the plasma control for electric power generation. At the moment, no model can simulate the behavior of the fusion power plant from the plasma to the turbine generator. Whether the future fusion power plant operation would be load-limit or load-follow, or pulsed or steady state, it is very important to study the relation of the...
Preliminary Design of Laser-Induced Breakdown Spectroscopy Diagnostic for Divertor Analysis in EAST
Cong Li a,b,*, Dongye Zhao a, Zhenhua Hu c, Niels Gierse b, Ping Liu a, Ran Hai a, Fang Ding c, Sebastijan Brezinsek b, Guang-Nan Luo c, Jiansheng Hu c, Liang Wang c, Junling Chen c, Yunfeng Liang b, Christian Linsmeier b, Hongbin Ding a and EAST Team
*a Key Laboratory of Materials...
To develop the concept of China Fusion Engineering Test Reactor (CFETR), a system code for integrated simulation and optimization is being developed. It is a platform which provides the tools of Tokamak conceptual design and engineering analysis. Meanwhile, the detailed design models and corresponding analysis results would be stored in the data management system of this platform. The relevant...
The ITER cryostat—the largest stainless steel vacuum pressure chamber ever built which provides the vacuum environment for components operating in the range from 4.5k to 80k like ITER vacuum vessel and the superconducting magnets. The Cryostat is divided into four section, of which, Base section is most complex because of its web shaped structure sandwiched between two 60mm thick plates with...
The hard core component (HCC) is defined for each undesirable situation with cliff-edge effects, defined as:
1) Dose to population above 10 mSv
2) Contamination of the ground water
3) High radiation field which avois long term human intervention on the site.
Such structure system components (SCCs) are designed to prevent these situations, as well as to return to and maintain a safe state in...
Two blackbody sources permanently located within the ITER Diagnostic Shield Module (DSM) at equatorial port 9 will operate in conjunction with two remotely retractable mirrors to generate and direct blackbody radiation to calibrate the radial and oblique views of the ITER Electron Cyclotron Emission (ECE) diagnostic. The main calibration requirements include a high-emissivity surface heated to...
The system of Correction Coils (CC) is a component of the ITER Magnet system, required to correct toroidal asymmetries and reduce error magnetic fields detrimental for physical processes in the plasma. Coil terminals will be connected to feeder terminals using twin-box joints. Qualification of the manufacturing procedure of the coil terminals is achieved by performing DC tests of prototype...
The outer vessel steady-state sensors (OVSS), a subsystem of the ITER magnetic diagnostics, will contribute to the measurement of the plasma current, plasma-wall clearance, and local perturbations of the magnetic flux surfaces near the wall. The diagnostic consists of a poloidal array of sixty sensors welded to the vacuum vessel outer surface. Each OVSS contains a pair of bismuth Hall sensors...
Abstract: The cryogenic axial insulation breaks (IBs) are key components of the superconducting magnet system for the EAST device, which play an important role of the liquid nitrogen and liquid helium insulation channel . In order to ensure the safe operation of the IBs during the working life of the EAST device, experimental research and analysis on electrical performance should be carried...
PCS is one of the key systems in Tokamak. PCS contains many subsystems, which can be used to control the different plasma parameters quickly and effectively. There is a need for real-time communication between the subsystems of PCS, and it has high requirements on the delay and stability of data transmission.
In this paper, the overall design of PCS Synchronous Data Network is given. The...
Electron Cyclotron Resonance Heating (ECRH) system is an important auxiliary heating method which is wildly used in magnetic confinement fusion. For the ECRH system on J-TEXT, signals should be transmitted to the control system for monitoring and protection, and also to the data acquisition system. Considering the high voltage and harsh Electro Magnetic Interference (EMI) environment, ...
This paper presents the results of the steady state and transient thermal analysis of the updated helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor (CFETR). The updated design of the helium cooled solid breeder blanket for CFETR has been described. The commercial finite element method code ANSYS is used for the thermal analysis in this work. Steady state thermal...
The use of efficient heating and current drive systems is an important research priority for DEMO. One such system in consideration for the European DEMO is the ion cyclotron range of frequency (ICRF) heating system. Extensive operational experience on several existing fusion facilities, a relatively low cost and high plug to power efficiencies motivates the consideration of ICRF for the...
The structural concept design of the central solenoid model coil of China Fusion Engineering Test Reactor has been carried out by ASIPP. The CS Model Coil shall produce a 12 Tesla peak field in the bore of the magnet, and the largest magnetic field change rate is 1.5T/S, and operating current is 47.65 KA. CFETR CS Model coil Structural concept design mainly including coil winding design,...
The European demonstration nuclear fusion power plant (DEMO) is under conceptual design phase within the EUROfusion consortium. The most conservative design options in terms of science and technological developments with respect to ITER: with a Q=10, an operation with pulses 2 hours long lasting and the production of a net electricity power of 500 MW.
The toroidal magnetic field at the plasma...
In-vessel magnetic sensors will be drastically limited, if present at all, in DEMO, due to the pulse duration and the neutron flux, so alternative diagnostics systems based on in-vessel non-magnetic sensors and ex-vessel sensors are under investigation to meet the requirements of a safe and reliable operation of the machine. In particular, the signals provided by a microwave reflectometry...
The search for alternative means of power source to meet the need of ever increase power demand is eternal, given the present global scenario a clean and sustainable form of energy with sufficiency to meet the demand is being investigated. One such form of energy is Nuclear fusion reaction. On the onus of having a harsh condition and a requirement of high pumping speed Cryosorption pumps have...
Neutron irradiation-induced defects will degrade the mechanical properties of future fusion reactor structural materials. The understanding of the mechanisms for the interaction between gliding edge dislocation and irradiation-induced defects, such as voids and helium bubbles, is of vital importance. In this presentations, the interaction between an edge dislocation and helium bubbles with...
Neutral Beam Injection (NBI) auxiliary heating system for Experimental Advanced Superconducting Tokamak (EAST) designed with the design of the 180 degree magnetic field deflection to deflect the un-neutralized particles during beam transmission. In order to protect against the divergent beam, the copper pole shields are placed on both sides of the neutral beam in front of each deflection...
A pilot plant for tritium removal from tritiated water is in operation at ICSI Ramnicu Valcea and is based on catalytic isotopic exchange between tritiated water and hydrogen/deuterium followed by cryogenic distillation aiming to recover tritium. A cryogenic distillation cascade consisting of four distillation columns is in operation and significant effort is required in various batch mode...
Chinese Fusion Engineering Testing Reactor (CFETR) is an ITER-like superconducting TOKAMAK aiming to bridge the gap between ITER and future fusion power plant. Superconducting coils of CFETR provide high-intensity magnetic field to confine the core plasma. Ports are used for RH (Remote Handing) maintenance, plasma diagnose and other measuring equipment. Neutrons leaking from the ports will...
The high confinement mode have been considered as a major operation mode of ITER because of it’s many advantages. However, in this case there will be severe shocks at the plasma boundary, called the edge localized modes (ELMs). It is significant and necessary to research the coupled mechanism of fatigue by both transient and periodic heat loads on PFC components. A FE model of typical...
The International Fusion Materials Irradiation Facility (IFMIF) aims to provide an accelerator-based, D-Li neutron source to produce high energy neutrons at sufficient intensity and irradiation volume for DEMO materials qualification. LIPAc is a 125 mA 9 MeV continuous wave deuteron accelerator whose components are under construction mainly in Europe, which is being installed in Rokkasho...
For the purpose of plasma physics research on tokamak experimental, many diagnostics have been developed for Experimental Advanced Superconducting Tokamak (EAST). A distributed and continuous data acquisition system has been implemented for the diagnostic system. At present, there are more than 60 data acquisition units and more than 2500 raw signals including scientific data, video data, and...
Since the construction of previous generations of fusion reactors, new technology has emerged that can enhance the performance of the high power, power supplies (HPPS) used in fusion energy. For example, wide bandgap (WBG) based switching devices have emerged with ratings suitable for some elements of power conversion. Furthermore, modular multilevel converters (MMC) are a topology that have...
The ITER Neutral Beam Injector (NBI) is designed to deliver 16.5 MW of additional heating power to the plasma, accelerating Deuterium or Hydrogen negative ions up to -1 MV with a current as high as 46A (for H2) . To prove the feasibility of the NBI system and demonstrate the achievement of the very demanding performance, a dedicated test facility is under construction in Padova, Italy, named...
JT-60SA is a Superconducting Tokamak in the framework of the Broader Approach Agreement between Europe and Japan. For this International Project among its various procurements, the Italian National Agency for New Technology Energy and Sustainable Economic Development (ENEA) is providing: four AC/DC converters for the central solenoid superconducting magnets (CS1, CS2, CS3 and CS4 PSs rated ±20...
Use Spectrum Simulation Code SOS to test the performance of the Fast ion D-alpha spectrum on HL-2A
P. Chen1, L. M. Yao1, J. Wu1*, Y.C. Chen1, H. Y. Zhou2, Y. Liu2
1 School of Physical Electronics, University of Electronic and Science Technology of China，610054
2 Southwestern Institute of Physics, 610041
ABSTRACT: In magnetic confined fusion devices, the fast ion is usually generated in heating...
An efficient remote handling (RH) and maintenance scheme is the immediate requirement to ensure the maximum availability of the tokamak devices for the plasma operations. Virtual and augmented reality provides resourceful data to the RH operators for achieving the accurate control over the RH equipment and helps in time optimization of the RH operations.
The VARID facility established at...
Wendelstein 7-X (W7-X), the first fully-optimized stellarator experiment, started its operation in December 2015. W7-X research aims for good plasma confinement and the demonstration of steady-state operation. This could make the stellarator a serious option for a future fusion power plant. Magnetohydrodynamic (MHD) equilibrium data is needed at W7-X for data analysis and plasma operation...
The central solenoid (CS) model coil for China Fusion Engineering Test Reactor (CFETR) is being developed in Hefei, China. The design value of the highest magnetic field for CS model coil is 12T, which is made of Nb3Sn and NbTi CIC conductor hybrid superconducting magnet. The structural parameter of all windings has been confirmed based on the electromagnetic design and optimization of coil....
It is a constant research interest in finding reliable solutions for the long term storage of hydrogen isotopes that integrates both the safety matters and its easy recovery. Thus, various methods have been investigated so far, namely gaseous storage in high pressure gas cylinders, liquid storage in cryogenic tanks or under solid state form. Considering as option the storage on solid...
