Damao Yao

Overview of plasma-facing materials and components for EAST

By the end of September 2006, the engineering commissioning of the Experimental Advanced Superconducting Tokamak (EAST) was completed and the first H2 plasma was achieved with stainless steel as the plasma-facing material (PFM). In the following phases, with gradual increase in the heating power, the EAST divertor targets have to handle expected heat loads up to ~10 MW m−2. The PFMs to be employed include doped graphite and W coatings. The former will be used to cover the whole plasma-facing surface in the first phase, and then the latter applied to replace the graphite tiles gradually in the second phase, and finally a whole W plasma-facing surface will be expected. Graphite has been studied for many years and contributed greatly to the HT-7 to achieve long pulses of some 100 s. The W coatings are being developed and the latest R&D development is reported in the paper.

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

Engineering Studies on the EAST Tungsten Divertor

Experimental advanced superconducting tokamak device is a D-shaped full superconducting tokamak with actively water cooled plasma facing components. To achieve long pulse and high βH-mode plasma, new plasma position and shape are calculated and optimized during the campaign of 2013-2014. New divertors are designed and developed to fit the plasma and endure the heat flux up to 10 MW/m2. The divertor is International Thermonuclear Experimental Reactor-like. It bases on monoblock and cassette technology and is composed of plasma facing component (PFC) units, cassettes, and support systems. Monoblock structures are just employed on the PFC units of target plates of the divertors, thus W flat tiles are used on the baffles and dome. At the end of monoblocks, end boxes are applied. Cassettes act not only as the supports of the PFC units, but also as manifolds for the cooling channels of the units. The support systems consist of inner support rails, outer support rails, and auxiliary braces. The support systems are installed on the vacuum vessel before cassettes are put on. To dock the cassette with the support system, it is just lifted and pushed forward. To verify and optimize the structure, Research and Development work has carried out for the divertor. The prototypes of PFC units and cassettes are fabricated with different manufacture processes. The total assembly process is also simulated with these prototypes on mockup facility and the vacuum vessel. The results show that the divertor system can be manufactured successfully according to the requirements of drawings and installed conveniently and precisely on the tokamak. All these works certify that the design of the divertor is all right. The batch production of the divertors can be started.

Investigation on the Possibility of Tritium Self-Sufficiency for CFETR Using a PWR Water-Cooled Blanket

The neutron wall load (Pn) of Chinese fusion engineering testing reactor (CFETR) will be less than 1 MW/m2. To meet the net tritium breeding ratio (TBR) of the reactor, a new water-cooled blanket concept is considered. The blanket neutronics schemes are performed to explore the local TBR issues in the (Pn) range of 1-5 MW/m2, which aims at the effective design of the blanket concept considering the tritium self-sufficiency. As a result, the calculation results are compared with the local TBR values and the material fraction changes. It is found that the local TBR has the high value at low (Pn) while the blanket size in radial direction is determined. It is mainly because of the total breeding area increasing due to the pipe pitch increasing in the model. This leads to the possibility for CFETR using a simplified blanket interior. In addition, to match the pressurized water reactor (PWR) water-cooled condition, a reduced size of blanket module in toroidal direction is achievable. It can be concluded that a PWR water-cooled blanket has more benefits to CFETR engineering implementation in the future.

The design and R&D work of EAST tungsten divertor

EAST tokamak is one of the advanced full superconducting tokamaks in the world. To achieve better plasma parameters, new plasma position and shape are calculated and optimized in the 2013s update. New divertors should be deigned and developed to endure higher heat flux less than 10MW/m2. The divertor is ITER-like structure. It bases on mono-block and cassette technology. The divertor is composed of PFC units, cassettes and support systems. Mono-block structures are just employed on the PFC units of target plates of the divertors, thus W flat tiles are used on the baffles and dome. Cassettes act as not only the supports of the PFC units, but also the cooling channels of the units. The support systems consist of inner support rails, outer support rails and auxiliary braces. The systems are installed on vacuum vessel before cassettes putting on. When cassette assembly with support system, it is just lifted and pushed forward. To optimize and verify the structure, R&D work has done on the divertor. The prototypes of PFC units in different positions and cassettes with different manufacture processes are fabricated. The assembly process is also simulated with these prototypes on mimic facility and vacuum vessel. The results show that the divertor system can be manufactured successfully according to the drawing requirements. And they can also be installed conveniently and precisely on the tokamak. All R&D work and tests will certify that the design of divertor is all right.

Kinematics Analysis of Double Seal Door for ITER Remote Handling Transfer Cask System

The moving direction of double seal door (DSD) of ITER remote handling transfer cask and the force of hydraulic pole will change significantly at the guide rail inflexion position (GRIP) which is a mutant site, so it is very possible to make the structure damage or the system failure at the GRIP. In this paper, the kinematics simulation and analysis of DSD were done based on special constitution restriction and working process by software ADAMS. The stress distribution of guide rail and hydraulic pole were obtained by the above simulation, at the same time the optimal GRIP was confirmed according to the force analysis result. The above-mentioned analysis process and results not only provide technical data for the optimization design and the prototype manufacture of DSD, but also provide the examples and references of kinematics analysis for other important components of ITER.Copyright

Design of Langmuir probe diagnostic system for the upgraded lower tungsten divertor in EAST tokamak

In order to achieve long-pulse H-mode plasma scenario over 400 s with high heating power in the Experimental Advanced Superconducting Tokamak (EAST) device, the lower graphite divertor will be upgraded into a tungsten (W) divertor with active water cooling, which consists of the W/Cu monoblock units and the W flat-tile units as the divertor plasma facing components. As a fundamental diagnostic tool, the divertor Langmuir probe (Div-LP) diagnostic system will be upgraded accordingly. This paper presents the design of two kinds of new Div-LP systems, which are planned to be utilized on the W/Cu monoblock units and the W flat-tile units for the upgraded lower tungsten divertor, respectively, including their structures and preliminary poloidal and toroidal layouts. The Div-LP diagnostic system can measure the plasma parameters with the schemes of triple-probe, double-probe, and single-probe, to obtain the spatial and temporal distribution of plasma behavior on the divertor targets, which is useful for the discharge control and operation in EAST. In addition, the thermal analysis of the two kinds of probe assemblies is also carried out by using the three-dimensional finite element code ANSYS, which is aimed to get the optimal designs to withstand the long-pulse and high-power operation in EAST future experiments.

The Design and Manufacture of the Neutral Beam Injection Thermal Shield for the EAST Tokamak

Abstract Neutral beam injection (NBI) is a high-power auxiliary heating system for the EAST device. We present a thermal shield (TS) structure to protect the neck pipe of the EAST equatorial port to avoid damage from the NBI beam. Since the EAST port has a big trumpet structure, a straight section, and a small trumpet structure, to accommodate the port structure, a TS concept is put forward including its cooling system. The cooling loops and the sub-branches were designed with interfaces between the inner cooling branches. The heat removal capability is verified by a thermal hydraulics analysis based on ANSYS code. In particular, fabrication is addressed with technical processing technology, especially for the embedded cooling pipes in the heat sinks. The pipes are checked for leaks after bending and the embedding processing. The assembly activities are demonstrated in the spatial space zones of the port before the engineering installation. It is confirmed that the TS structure is safe and will run feasibly in the EAST discharge. It is indicated that the TS structure can provide thermal shielding and remove heat for the NBI device in the port region.

Analysis of EAST’s New Tungsten Divertor and Cooling System During a Disruption with Halo Currents

Abstract Chinese Academy of Sciences Institute of Plasma Physics (ASIPP) Experimental Advanced Superconducting Tokamak (EAST) has designed and built a new outer divertor with an ITER-like cooling system. As part of a joint collaboration, the Plasma Science and Fusion Center at MIT performed analyses on the EAST design to determine loading, stresses and deflections due to the eddy currents and halo currents occurring during a disruption. The analysis was done using the finite element program COMSOL using techniques developed at MIT to recreate actual tokamak discharges from measured data. This technique has been used successfully to recreate discharges from Alcator C-Mod, a high field tokamak with TZM tiles at the Plasma Science Fusion Center at MIT, and allows us to recreate the fields for any disruption from the EAST data base. For the new divertor, an upward moving disruption was chosen as the design scenario. The plasma filament model predicts fields, eddy currents and loads due to a disruption, but the divertor will also be exposed to halo currents. The new EAST divertor borrows its cooling system design from ITER where the plasma facing tungsten tiles are water cooled by a CuCrZr manifold and pipes attached to the tiles. Halo currents traveling down these tubes and crossing the toroidal field will result in large loads in these components, and COMSOL is used to predict the stresses and deflections. The model predicts that the EAST divertor will survive the combined loading due to the eddy and halo currents.

A multi-layer breeding blanket concept for CFETR based on PWR condition

A breeding blanket concept with the multi-layer structure based on the PWR water-cooled condition was presented for CFETR. To explore the feasibility of the blanket scheme, the neutronics and hydraulics programs were carried out. It was found when the Pn is less than 3MW/m2, the local TBR (tritium breeding ratio) would be in range of 1.46-1.7. Thus, the net TBR would be more than 1.05, which meets the tritium self-sustaining requirement for the fusion reactor. Especially, the local TBR is 1.66 at a neutron wall load (Pn) of 0.5 MW/m2 and the corresponding net TBR is 1.21. On the other hand, it also clarified a pipe bore with 7-8 mm at an inlet velocity of 3-4 m/s would be suitable for the heat removal of the blanket module. In addition, the total pressure drop would be under 0.2 MPa in the cooling system. It was concluded that the blanket concept would be more effective and benefit to CFETR in view of its neutron wall load level and the tritium self-sustaining efficiency.