Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.     

First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.     

The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m² range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.     

Analysis of Divertor Heat Flux with Infrared Thermography During Gas Fuelling in the HL-2A Tokamak

In HL-2A tokamaks, the behavior of heat flux deposited on the divertor targets has been studied during deuterium gas fuelling. The heat flux is reduced significantly after supersonic molecular beam injection （SMBI） fuelling during Ohmic and electron cyclotron resonance heating （ECRH） divertor discharges. The SMBI fuelling causes an increase in the plasma density and this change results in the experienced change of the edge properties. Most of this reduction in divertor target heat flux occurs together with a high plasma radiation region located at near the X-point. The largest reduction in heat flux profiles is observed at the outboard divertor separatrix strike point, while the heat flux far from the strike point remains almost unchanged. In particular, with SMBI multi-pulses gas fuelling, a partially detached divertor regime is observed with a highly radiating region at the X-point. With the onset of the partially detached divertor regime, a sudden drop in both heat flux and power flow on the divertor target is observed. The reduction in power load on the divertor targets is roughly equal to the increase in plasma radiation loss.     

Research plan for divertor simulation making use of a large tandem mirror device

We are planning to start a study of divertor simulation under the closely resemble to actual fusion plasma environment making use of the advantage of open magnetic field configuration and to contribute the solution for realizing the divertor in ITER as a future research plan of Plasma Research Center of the University of Tsukuba. In the research plan, the concepts of two divertor devices are introduced. One has an axi-symmetric divertor configuration with the separatrix which is similar to toroidal divertor of torus systems and the other is a high heat flux divertor simulator by using an end-mirror exit of the existing tandem mirror device. Development of magnetic field configuration for ensuring the MHD stability is under way and a designed example is investigated under the optimal condition for plasma production. Consideration of plasma heating scheme using Fokker-Planck simulation code was successfully performed at both axi-symmetric divertor and end-mirror regions. Preparative experiments using calorimeter, Mach probe and high-speed camera have been started at the end-mirror region and the heat flux density of the level in 1-10 MW m⁻² was achieved in standard hot-ion mode plasma-confining experiments, which gives a clear prospect of generating the required heat flux density for divertor studies.     

Nuclear Characteristics of Gas-Suspended Boron Carbide Cooling Catalyzed D Fusion Reactor Blanket

Transient behaviors of plasma and in-vessel components have been investigated considering the divertor plasma state (detached/attached) transition. The SAFALY code consisting of a zero-dimensional plasma model and a one-dimensional heat transfer model of components has been modified to take account of the divertor plasma state transition on the basis of the updated divertor plasma physics. Several plasma events, i.e., over fueling, sudden auxiliary heating injection and Confinement improvement events which would be expected to result in overpower, were selected for the International Thermonuclear Experimental Reactor (ITER) and the transient behaviors were calculated on the assumption of a combined failure of plasma control and machine interlock in addition with a postulated plasma transient. The results show that plasma burning passively terminates due to sublimated impurity penetration from the carbon target surface, but there are possibilities of dry out of the coolant for the high heat flux in sudden attached state transition under the multifailure of plasma control. However, effects by the aggravating failure of the divertor are expected to be safely terminated by the confinement boundary, the vacuum vessel and its pressure suppression system.     

Engineering studies for the installation of an axi-symmetric metallic divertor in Tore Supra

Tore Supra (TS) has been designed to operate using technologies that allow long plasma operation (a few minutes), by means of superconducting magnets and actively-cooled high heat flux plasma facing components (PFCs). Actively cooled tungsten PFC will be used in the baffle area of the first ITER divertor. In order to validate such a technology fully (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axi-symmetric divertor in the tokamak Tore Supra has been studied [1]. With this second major upgrade, Tore Supra should be able to address the problematic of long plasma discharges with a metallic divertor.The proposed divertor is made up of two stainless steel casings containing a copper coil winding located at the top and bottom area of the vacuum vessel. These casings are firmly maintained by connection beams and protected by PFC. This paper describes the mechanical design of this major component and its integration in TS, the associated electromagnetic and thermomechanical analysis, the manufacturing issues and finally the integration of ITER representative PFCs.     

Efforts Made to Handling Steady-State High Heat Flux in EAST

There have been three generations divertor designed for EAST to handle steady-state high heat flux form plasma. The first generation divertor was used on the initial phase of the plasma burning. The first generation divertor was just stainless plate 5 mm in thickness bolted on supports which had been applied since 2006–2007. From 2008 to 2013 the second generation divertor has been used. The second generation divertor was graphite divertor that consisted of graphite tiles, heat sink (CuCrZr) and supports (316L). The third generation divertor was tungsten divertor with ITER like design that had been used science 2014. Now days the upper divertor is tungsten divertor (80 modules) and the lower divertor is graphite divertor (16 modules) in EAST. Tungsten divertor is able to withstand 10 MW/m² heat flux on its strike point and graphite divertor can bear 2 MW/m² under same conditions. It is very important to make every efforts to improve thermal extraction technology of divertor by comparing and practice different designs. Such efforts made in EAST can bring experiences and answers for ITER or any next divertor fusion device on nuclear phase.     

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m² plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.     

Active radiative liquid lithium divertor concept

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept [1]. In the RLLD, Li is evaporated from the liquid lithium (LL) coated divertor strike point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal. The modeling results indicated that the Li radiation can be quite strong, so that only a small amount of Li (∼a few mol/s) is needed to significantly reduce the divertor peak heat flux for typical reactor parameters. In this paper, we examine an active version of the RLLD, which we term ARLLD, where LL is injected in the upstream region of divertor. We find that the ARLLD has similar effectiveness in reducing the divertor heat flux as the RLLD, again requiring only a few mol/s of LL to significantly reduce the divertor peak heat flux for a reactor. An advantage of the ARLLD is that one can inject LL proactively even in a feedback mode to insure the divertor peak heat flux remains below an acceptable level, providing the first line of defense against excessive divertor heat loads which could result in damage to divertor PFCs. Moreover, the low confinement property of the divertor (i.e., <1 ms for Li particle confinement time) makes the ARLLD response fast enough to mitigate the effects of possible transient events such as large ELMs.     

Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear.Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.     

Impurity Transport in a Simulated Gas Target Divertor

Future generation fusion reactors and tokamaks will require dissipative divertors to handle the high particle and heat loads leaving the core plasma (100–400 MW/m² in ITER). A radiative divertor is proposed as a possible scenario, utilizing a hydrogen target gas to disperse the plasma momentum and trace impurity radiation to dissipate the plasma heat flux. Introducing an impurity into the target hydrogen gas enhances the radiative power loss but may lead to a significant impurity backflow to the main plasma. Thus, impurity flow control represents a crucial design concern. Such impurity flows are studied experimentally in this thesis. The PISCES-A linear plasma device (n 3 × 10¹⁹ m^–3, kT _e 20 eV) has been used to simulate a gas target divertor. To study the transport of impurities, a trace amount of impurity gas (i.e., neon and argon) is puffed near the target plate along with the hydrogen gas. Varying the hydrogen gas puffing rate permits us to study the effects of various background plasma conditions on the transport of impurities. A 1-1/2-D fluid code has been developed to solve the continuity and momentum equations for a neutral and singly ionized impurity in a hydrogen background plasma. The results indicate an axial reduction in the impurity concentration upstream from the impurity puffing source. Impurity entrainment is more effective for higher hydrogen target pressures (and for higher hydrogen plasma densities). However, if there is a reversal of the background plasma flow, impurity particles can propagate past the plasma flow reversal point and are then no longer entrained.     

Reduction of heat flux on divertor plates by remote radiative cooling in doublet III

Using a single null divertor configuration, heat flux intensity and its profile on the divertor plates as a function of plasma current and density were measured with an infrared camera and thermocouples. The vertical width of the heat flux on the divertor plates 2λ is ≈ 10 cm at the lower separatrix and is ≈ 5.5 cm at the upper separatrix. A diffusion coefficient D which is obtained from the measurement of the diffusion length across the scrape-off field lines is roughly proportional to and its magnitude is on the order of Bohm diffusion. The heat flux on the plates decreases by more than a factor of 5 with increasing electron density in the main plasma and is much smaller than that on the limiters in non-diverted plasmas. Only 3% of ohmic input power goes into the divertor plates at high density of the main plasma, while ≈ 20% goes in at low density. The decrease of heat flux is in good agreement with the increase of radiation loss in the divertor region. The heat flux on the divertor plates can be reduced by remote radiative cooling in high density discharges.     

Infrared thermography inspection of the ITER vertical target qualification prototypes manufactured by European industry using SATIR

High heat flux loaded components which will be installed in the ITER Divertor require a heat flux removal capability in the range 5–10 MW/m² at steady-state and up to 20 MW/m² in transients. Within the ITER plasma facing components procurement context, each party should demonstrate its technical capability to carry out the manufacturing with the required quality. This is achieved through the successful manufacturing and testing of medium-size qualification prototypes. Each Qualification Prototype consists of three high heat flux units mounted onto an actively cooled supporting structure. Currently, the SATIR method has been identified by the ITER Organization as the basic test to decide upon the final acceptance of the ITER Divertor components. SATIR testing was performed on each CFC part of European HHF units prior to the insertion of the twisted tape and prior to assembling the units onto the steel support structure. The paper deals with SATIR results of all qualification prototypes manufactured by European industry.     

Status of the IDTF high-heat-flux test facility

The ITER Divertor Test Facility (IDTF) was designed for the high heat flux tests of outer vertical targets, inner vertical targets and domes of the ITER divertor. This facility was created in the Efremov Institute under the Procurement Arrangement 1.7.P2D.RF (high heat flux tests of the plasma facing units of the ITER divertor).The heat flux is generated by an electron-beam system (EBS), 800 kW power and 60 kV maximum accelerating voltage. The component to be tested is mounted on a manipulator in the vacuum chamber capable of testing objects up to 2.5 m long and 1.5 m wide. The pressure in the vacuum chamber is about 3*10⁻³ Pa. The parameters of the cooling system and the water quality (deionized water) are similar to the cooling conditions of the ITER divertor. The integrated control system regulates all IDTF subsystems and data acquisition from all diagnostic devices, such as pyrometers, IR-cameras, video cameras, flow, pressure and temperature sensors.Started in 2008, the IDTF was commissioned in 2012 with the testing the outer vertical full-scale prototypes and the completion of the PA 1.7.P2D.RF task. This paper details the main characteristics of the IDTF.     

Technological review of the HRP manufacturing process R&D activity

ENEA and Ansaldo Nucleare S.p.A. have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities for the manufacturing of high heat flux plasma-facing components (HHFC), and in particular for the inner vertical target (IVT) of the ITER divertor.This component has to be manufactured by using both armour and structural materials whose properties are defined by ITER. Their physical properties prevent the use of standard joining techniques. The reference armour materials are tungsten and carbon/carbon fibre composite (CFC). The cooling pipe is made of copper alloy (CuCrZr-IG).During the last years ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components of different length, geometry and materials, by using innovative processes: HRP (hot radial pressing) and PBC (pre-brazed casting).The history of the technical issues solved during the R&D phase and the improvements implemented to the assembling tools and equipments are reviewed in the paper together with the testing results.The optimization of the processes started from the successful manufacturing of both W and CFC armoured small scale mockups thermal fatigue tested in the worst ITER operating condition (20 MW/m²) through the achievement of record performances obtained from a monoblock medium scale mockup.On the base of these results ENEA-ANSALDO participated to the European programme for the qualification of the manufacturing technology to be used for the procurement of the ITER divertor IVT, according to the F4E specifications. A divertor inner vertical target prototype (400 mm total length) with three plasma facing component units, was successfully tested at ITER relevant thermal heat fluxes.Now, ANSALDO and ENEA are ready to face the challenge of the ITER inner vertical target production, transferring to an industrial production line the experience gained in the development, optimization and qualification of the PBC and HRP processes.     

On the carbon and tungsten sputtering rate in a magnetron discharge

Magnetron discharge as sputtering source can serve as an alternative tool for the study of the plasma-wall interaction, with applications for ITER divertor. The present work reports on the influence of the target power density and the nature of the projectile on the erosion of C and W targets. The experimental results concern the sputtering rate of carbon and tungsten targets of a d.c. magnetron discharge in argon and helium atmosphere, at different gas pressures in the range of 10-100 mTorr and discharge power densities up to 40 W cm⁻² while the discharge current intensity was used as control parameter. In this investigation, carbon and tungsten sputtering rates were measured using two conventional methods based on gravimetric mass loss and profilometry. Target erosion profiles were compared with the profiles of the ion energy flux bombarding the target, calculated from a 2D fluid model.     

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.     

Effects of Low Energy and High Flux Helium／Hydrogen Plasma Irradiation on Tungsten as Plasma Facing Material

The High-Z material tungsten (W) has been considered as a plasma facing material in the divertor region of ITER (International Thermonuclear Experimental Reactor). In ITER, the divertor is expected to operate under high particle fluxes (> 1023 m-2s-1) from the plasma as well as from intrinsic impurities with a very low energy (< 200 eV). During the past dacade, the effects of plasma irradiation on tungsten have been studied extensively as functions of the ion energy, fluence and surface temperature in the burning plasma conditions. In this paper, recent results concerning blister and bubble formations on the tungsten surface under low energy (< 100 eV) and high flux (> 1021 m-2s-1) He/H plasma irradiation are reviewed to gain a better understanding of the performance of tungsten as a plasma facing material under the burning plasma conditions.     

Geometry and expected performance of the solid tungsten outer divertor row in JET

At JET new plasma-facing components for the main chamber wall and the divertor are being designed and built to mimic the expected ITER plasma wall conditions in the deuterium-tritium operation phase. The main wall elements at JET will be made of beryllium and the divertor plasma-facing surface will be made of tungsten. Most of the divertor tiles will consist of tungsten-coated Carbon Fibre Composite (CFC) material. However one toroidal row in the outer divertor will be made of solid, inertially cooled tungsten. The geometry of these solid tungsten divertor components is optimized within the boundary conditions of the interfaces and the constraints given by the electrodynamical forces. Shadowing calculations as well as rough field line penetration analysis is used to define the geometry of the tungsten lamella stacks. These calculations are based on a set of magnetic equilibria reflecting the operation domain of current JET plasma scenarios. All edges in poloidal and toroidal direction are shadowed to exclude near perpendicular field line impact. In addition, the geometry of the divertor structure is being optimized so that the fraction of the plasma wetted surface is maximised. On the basis of the optimized divertor geometry, performance calculations are done with the help of ANSYS to assess the maximum power exhaust possible with this inertially cooled divertor row.