Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

As an important component of tokamaks, the divertor is mainly responsible for extracting heat and helium ash, and the targets of the divertor need to withstand high heat flux of 10 MW m⁻² for steady-state operation. In this study, we proposed a new strategy, using microchannel cooling technology to remove high heat load on the targets of the divertor. The results demonstrated that the microchannel-based W/Cu flat-type mock-up successfully withstood the thermal fatigue test of 1000 cycles at 10 MW m⁻² with cooling water of 26 l min⁻¹, 30 °C (inlet), 0.8 MPa (inlet), 15 s power on and 15 s dwell time; the maximum temperature on the heat-loaded surface (W surface) of the mock-up was 493 °C, which is much lower than the recrystallization temperature of W (1200 °C). Moreover, no occurrence of macrocrack and 'hot spot' at the W surface, as well as no detachment of W/Cu tiles were observed during the thermal fatigue testing. These results indicate that microchannel cooling technology is an efficient method for removing the heat load of the divertor at a low flow rate. The present study offers a promising solution to replace the monoblock design for the EAST divertor     

Realization of minute-long steady-state H-mode discharges on EAST

In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current (Ip ～ 0.45 MA) was fully-noninductively driven (Vloop ＜ 0.0 V) by a combination of ～2.5 MW LHW,～0.4 MW ECH and ～0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.     

Predictive Modeling for Performance Assessment of ITER-Like Divertor in China Fusion Engineering Testing Reactor

To facilitate the design of the China Fusion Engineering Testing Reactor (CFETR), predictive modeling for the assessment and optimization of the divertor performances is an indispensable approach. This paper presents the modeling of the edge plasma behaviors as well as the W erosion and transport properties in CFETR with ITER-like divertor by using the B2-Eirene/SOLPS 5.0 code package together with the Monte Carlo impurity transport code DIVIMP. As expected, SOLPS modeling of divertor-SOL plasmas finds that the peak heat flux onto the divertor targets greatly exceeds 10 MW/m², an engineering limit posed to the steady-state and/or long-pulse operation of the next-step fusion devices, for a wide range of plasma conditions, and thus modeling of Ar puffing by scanning the puffing rate for radiative divertor is performed. As the increase of the Ar puffing rate, the peak target heat fluxes and plasma temperature decreases exponentially,reflecting that Ar puffing is highly effective at power exhausting. Based on the ion fluxes from SOLPS, the W erosion is calculated by taking into consideration the bombardment of both D and Ar ions, and then the W plasma concentrations are calculated based on the W erosion fluxes using DIVIMP. The calculations show that if the Ar puffing only being used to reduce the divertor heat load, the W plasma contamination in the core plasma exceeds the tolerable value (<10^?5), which demonstrates that some further upgrading of the divertor geometry is still needed.     

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal–mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor con?guration is under construction in SWIP, where ITER-like ?at-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak.     

Integrated Operating Scenario to Achieve 100-Second,High Electron Temperature Discharge on EAST

Stationary long pulse plasma of high electron temperature was produced on EAST for the first time through an integrated control of plasma shape,divertor heat flux,particle exhaust,wall conditioning,impurity management,and the coupling of multiple heating and current drive power.A discharge with a lower single null divertor configuration was maintained for 103 s at a plasma current of 0.4 MA,q_(95)≈7.0,a peak electron temperature of 4.5 keV,and a central density n_e(0)~2.5×10~(19) m~(-3).The plasma current was nearly non-inductive(V_(loop) 0.05 V,poloidal beta ~0.9) driven by a combination of 0.6 MW lower hybrid wave at 2.45 GHz,1.4 MW lower hybrid wave at 4.6 GHz,0.5 MW electron cyclotron heating at 140 GHz,and 0.4 MW modulated neutral deuterium beam injected at 60 kV.This progress demonstrated strong synergy of electron cyclotron and lower hybrid electron heating,current drive,and energy confinement of stationary plasma on EAST.It further introduced an example of integrated "hybrid" operating scenario of interest to ITER and CFETR.     

Design and evaluation of an optimized W/Cu interlayer for W monoblock components

Divertor plasma-facing components of future fusion reactors should be able to withstand heat fluxes of 10-20 MW/m² in stationary operation. Tungsten blocks with an inner cooling tube made of CuCr1Zr, so-called monoblocks, are potential candidates for such water-cooled components. To increase the strength and reliability of the interface between the W and the cooling tube of a Cu-based alloy (CuCr1Zr), a novel advanced W-fibre/Cu metal matrix composite (MMC) was developed for operation temperatures up to 550 °C. Based on optimization results to enhance the adhesion between fibre and matrix, W fibres (W_f) were chemically etched, coated by physical vapour deposition with a continuously graded W/Cu_PVD interlayer and then heated to 800 °C. The W_f/Cu MMC was implemented by hot-isostatic pressing and brazing process in monoblock mock-ups reinforcing the interface between the plasma-facing material and the cooling channel. The suitability of the MMC as an efficient heat sink interface for water-cooled divertor components was tested in the high heat flux (HHF) facility GLADIS. Predictions from finite element simulations of the thermal behaviour of the component under loading conditions were confirmed by the HHF tests. The W_f/Cu MMC interlayer of the mock-ups survived cyclic heat loads above 10 MW/m² without any damage. One W block of each tested mock-up showed stable thermal behaviour at heat fluxes of up to 10.5 MW/m².     

Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m² with the cooling water of 2 m/s, 20 °C, 0.2 MPa.     

Testing of high heat flux components manufactured by ENEA for ITER divertor

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested.A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process.The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France).The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m², 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m², 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m² Critical Heat Flux was measured at relevant thermal–hydraulics conditions at the end of the testing campaign.This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up.These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).     

Thermal Fatigue Study on the Divertor Plate Materials

Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate.Taking tungsten as surface material,a small-mock-up divertor plate was made by hot isostatic press welding (HIP),A thermal cycling experiment for divertor mock-up was carried out in the vacuum,where a high-heat-flux electronic gun was used as the thermal source,A cyclic heat flux of 9MW/m^2 was loaded onto the mock-up,a heating duration of 20s was selcted,the cooling water flow rate was 80ml/s.After 1000 Cycles,the surface and the W/Cu joint of the mock-up did not show any damage,The SEM was used to analyze the microstructure of the welding joint,where no cracks were found also.     

Thermo-mechanical analysis of Wendelstein 7-X plasma facing components

The stellarator experiment Wendelstein 7-X (W7-X) is designed for stationary plasma operation (30 min). Plasma facing components (PFCs) such as the divertor targets, baffles, heat shields and wall panels are being installed in the plasma vessel (PV) in order to protect it and other in-vessel components. The different PFCs will be exposed to different magnitude of heat loads in the range of 100 kW/m²–10 MW/m² during plasma operation. An important issue concerning the design of these PFCs is the thermo-mechanical analysis to verify their suitability for the specified operation phases. A series of finite element (FE) simulations has been performed to achieve this goal. Previous studies focused on the test divertor unit (TDU) and high heat flux (HHF) target elements. The paper presents detailed FE thermo-mechanical analyses of a prototype HHF target module, baffles, heat shields and wall panels, as well as benchmarking against tests.     

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak.The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916.The real-time feed-back control system for the plasma displacement was employed.Modeling results of the evolution of the poloidal field coil currents,the plasma current,the major radius,the plasma configuration all show agreement with experimental measurements.Results from the simulation show that during disruption,heat flux about 8 MW m-2 flows to the upper divertor target plate and about 6 MW m-2 flows to the lower divertor target plate.Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient.This shows that TSC has high flexibility and predictability.     

Design and technological solutions for the plasma facing components of WENDELSTEIN 7-X

The operation of W7-X stellarator for pulse length up to 30 min with 10 MW input power requires a full set of actively water-cooled plasma facing components. From the lower thermally loaded area of the wall protection system designed for an averaged load of 100 kW/m² to the higher loaded area of the divertor up to 10 MW/m², various design and technological solutions have been developed meeting the high load requirements and coping with the restricted available space and the particular 3D-shaped geometry of the plasma vessel. 80 ports are dedicated alone to the water-cooling of plasma facing components and a complex networking of kilometers of pipework will be installed in the plasma vessel to connect all components to the cooling system. An advanced technology was developed in collaboration with industry for the target elements of the high heat flux (HHF) divertor, the so-called “bi-layer” technology for the bonding of flat tiles made from CFC NB31 onto the CuCrZr cooling structure. The design, R&D and the adopted technological solutions of plasma facing components are presented. At present, except the HHF divertor, most of plasma facing components has been already manufactured.     

Progress of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO [1] has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10 MW/m² anticipated for DEMO. The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10 MPa, 600 °C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200 kW, 40 keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600 °C, a pressure of 10 MPa, and a maximal pressure head of 0.5 MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10 MW/m² unscathed.     

Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

After an extensive R&D development program, a full-scale divertor target prototype, manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full monoblock geometry. The lower part (CFC armour) and the upper part (W armour) of each monoblock were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. This paper summarises and analyses the main test results obtained on this prototype.     

Infrared thermography inspection for monoblock divertor target in JT-60SA

Carbon fiber composite (CFC) monoblock divertor target is required for power handling in JT-60SA. Quality of the targets depends on a joining technology in manufacturing process. To inspect the quality of more than 900 target pieces, efficient non-destructive inspection is needed. An infrared thermography inspection (IR inspection), has been proposed by ITER and IRFM, where the quality between CFC and a cooling tube is examined by a use of transient thermal response at a rapid switch from hot to cold water flow. In JT-60SA divertor target, a screw tube will be employed to obtain high heat transfer efficiency with simple structure. Since the time response of the screw tube is much faster than that of smooth tube, it is required to confirm the feasibility of this IR inspection. Thus, the effect of joining defects on transient thermal response of the targets has been investigated experimentally by using the mock-up targets containing defects which are artificially made. It was found that the IR inspection can detect the defects. Moreover, screening criteria of IR inspection for acceptable monoblock target is discussed.     

Manufacturing of the Wendelstein 7-X divertor and wall protection

The in-vessel components of Wendelstein 7-X (W7-X) with a total surface of 265 m² comprise the divertor and the wall protection. The high heat flux (HHF) and lower heat flux (LHF) target, the baffle, the end plates closing the divertor chamber, a cryo vacuum pump (CVP) and a control coil form one divertor unit. Steel panels and the graphite heat shield protect the wall, including the ports. The HHF target elements, the steel panels and the control coils are manufactured by industry. The remaining components will be manufactured by the Max-Planck-Institute für Plasmaphysik (IPP) at its Garching workshops. For all components the final acceptance tests will be performed by IPP. This paper summarizes the main aspects for manufacturing, the preceding development and qualification tests as well as the final acceptance tests for the in-vessel components.     

Current status of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO is being investigated at the Forschungszentrum Karlsruhe within the framework of the EU power plant conceptual study. The design goal is to resist a heat flux of 10 MW/m² at least. The major R&D areas are design, analyses, fabrication technology, and experimental design verification. A modular design is preferred for thermal stress reduction. The HEMJ (He-cooled modular divertor with multiple-jet cooling) was chosen as reference concept. It employs small tiles made of tungsten, which are brazed to a thimble made of tungsten alloy W-1%La₂O₃. The W finger units are connected to the main structure of ODS Eurofer steel by means of a copper casting with mechanical interlock. The divertor modules are cooled by helium jets (10 MPa, 600 °C) impinging onto the heated inner surface of the thimble.In cooperation with the Efremov Institute a combined helium loop & electron beam facility (60 kW, 27 keV) was built in St. Petersburg, Russia, for experimental verification of the design. It enables mock-up testing at a nominal helium inlet temperature of 600 °C, an internal pressure of 10 MPa, and a pressure difference in the mock-up of up to 0.5 MPa. Technological studies were performed on manufacturing of the W finger mock-ups. Several high heat flux tests were successfully performed till now. Post-examination and characterisation of the mock-ups subjected to the high heat flux tests were performed in collaboration with Forschungszentrum Jülich. Altogether, the test results confirm the divertor performance required. The helium-cooled divertor concept was demonstrated to be feasible. The knowledge gained from these experiments and some aspects on the design improvement are discussed in this contribution.     

Thermo-mechanical analysis of the Wendelstein 7-X divertor

The stellarator Wendelstein 7-X (W7-X) has a divertor consisting of 10 units installed inside the plasma vessel (PV). It was decided not to install the long pulse high-heat flux (HHF) divertor targets at the first two years stage of W7-X operation and to start with an adiabatically cooled test divertor unit (TDU) and shorter plasma pulses operation. This allows to accumulate operation experience with much simpler components, and as a result to adjust accurately the actively cooled HHF divertor which replaces the TDU for the stationary operation. Finite element (FE) analyses have been performed for better understanding of thermo-mechanical problems of divertor targets, and to guide the design of the TDU and HHF divertors. This paper presents the detailed results of the temperature response, the deformation and thermal stress of the divertor components.     

The WEST project: PFC shaping solutions investigated for the ITER-like W divertor

Among major issues for PFCs design, the impact of leading edges (exposed surface) which would be directly intersected by particles following magnetic field lines at glancing incident angles in the high heat flux areas is much discussed. This paper presents the key outcome of a thermal analysis performed on different shaping solutions for the ITER-like W monoblocks occurred for the components of the WEST (W Environment for Steady state Tokamak) divertor which could shadow any direct leading edge and to counteract a potential misalignment due to assembly tolerance. The results, in terms of surface temperature rise and wall heat flux into the cooling channel, are discussed for magnetic field lines incident at glancing angles expected in the higher heat flux regions of divertor (i.e. close to the strike point regions) and for perpendicular incident heat flux up to 20 MW/m².     

Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads

High-confinement mode is a very prominent operation style for future fusion device due to its unique advantages. However, the conjuncted edge localized modes (ELMs) are very difficult to control so that divertor plates are very prone to suffer both stationary high heat flux (HHF) loads of long-pulse operating mode and transient shock loads of ELMs. Most previous researches focus on degradation of plasma facing material (PFM), however, as a layer joining PFM and cooling tube, the soft copper interlayer suffers concentrated thermal stress loads due to mismatched thermal expansion of PFM and cooling tube. Its thermal fatigue behavior under such coupled loads is also of great significance to structural safety of divertor component. With such a motivation, the reduction effects on fatigue life time of a typical interlayer of monoblock divertor under series of coupled HHF and ELMs shock loading conditions are investigated. It is found that: (1) The transient shock feature of ELMs loading is propagated into interlayer with less sharp pattern. The increase of damage induced by coupled ELMs loading is limited in single cycle, while the accumulated damage of multiple consecutive coupled loading cycles is increased nonlinearly. (2) Under the coupled HHF and ELMs loading, the fatigue life time of interlayer is generally decreasing. The magnitude of decrease is increasing nonlinearly with the magnitude of ELMs peak and averaged heat flux. (3) For three characteristic parameters of ELMs shock loading such as frequency, duration and peak heat flux, the peak heat flux and frequency are two parameters more sensitive to determine coupled reduction effects on fatigue lifetime of the interlayer, while for high frequency case, time averaged heat flux takes the lead.