Status of design and experimental activity on module of lithium divertor for KTM tokamak

The projects of ITER and DEMO reactors showed that there are serious difficulties with solving the issues of plasma facing elements (PFE) based on the solid materials. Problems of PFE can be overcome by the use of liquid lithium. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material, in which liquid lithium fills a solid matrix from porous material. The progress in development of lithium technology and also lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, LTX, HT-7 and stellarator TJ II is a good basis for development of the project of steady-state operating lithium divertor module for Kazakhstan tokamak. At present the lithium divertor module for KTM tokamak is development and manufacturing. The paper describes main design features of the module of lithium divertor (MLD). The first step of the hydraulic tests of MLD with fully assembled external thermo-stabilization system, which was connected to in-vessel lithium unit, were performed using ethanol as a model heat transfer media. Test results of MLD have shown that operating parameters of designed and manufactured system for thermo-stabilization are sufficient for proper operation; basic hydraulic characteristics of the system are close to expected values.     

Realization of divertor configuration discharge in J-TEXT tokamak

To extend the operation region of the Joint-Texas Experimental tokamak (J-TEXT) to the divertor configuration and even the H-mode, the divertor configuration discharge has been realized for the first time in the J-TEXT tokamak. Along with the establishment of a power supply for the divertor configuration, the construction of relevant diagnostics, and the installation of the divertor target on the high-field side, divertor discharge has been tested. Through the equilibrium calculation and position stability analysis, the control strategy has evolved to be more stable. High-density experiments and auxiliary heating experiments have been carried out on the divertor configuration. The special midplane single-null (MSN) divertor configuration is shown to be more stable than the limiter configuration in the density limit condition and can reach a higher density in the experiment. In the ECRH experiment, the power injection enhances the electron temperature and density, while more heat outflux is loaded on the divertor target tiles and causes more intensive recycling and impurity release. The future plan for the divertor configuration operation in the J-TEXT tokamak is also included.     

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.     

Towards advanced divertor configurations on the J-TEXT tokamak

Developing advanced magnetic divertor configurations to address the coupling of heat and particle exhaust with impurity control is one of the major challenges currently constraining the further development of fusion research. It has therefore become the focus of extensive attention in recent years. In J-TEXT, several new divertor configurations, including the high-field-side single-null poloidal divertor and the island divertor, as well as their associated fundamental edge divertor plasma physics, have recently been investigated. The purpose of this paper is to briefly summarize the latest progress and achievements in this relevant research field on J-TEXT from the past few years.     

Thermal control of the liquid lithium divertor for NSTX

The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with 0.25 mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500 W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.     

Ion beam micro analysis of deposits at tokamak divertor surfaces

Cross sections of deposited layers in the Joint European Torus (JET) were analysed using the nuclear micro analysis at the Tandem Laboratory, Uppsala University. For deuterium and beryllium the nuclear reactions ²D(³He,p)⁴He and ⁹Be(³He,p_n)¹¹B were exploited for analysis. Typically the analyses have been made with 10 μm spatial resolution and a sensitivity of better than one atomic percent for beryllium or deuterium in carbon matrix. Comparing several different surface treatment techniques shows that polishing the sample surface give very good optical surface information but that some amount of deuterium and beryllium probably is removed. For good quantitative results the measurement can either be done on a rough surface or the top of the polished surface can be cut off.     

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.     

Space-resolved vacuum-ultraviolet spectroscopy for measuring impurity emission from divertor region of EAST tokamak

The measurement of impurity distribution in the divertor region of tokamaks is key to studying edge impurity transport. Therefore, a space-resolved vacuum-ultraviolet (VUV) spectrometer is designed to measure impurity emission in the divertor region on EAST. For good spectral resolution, an eagle-type VUV spectrometer with 1 m long focal length with spherical holograph grating is used in the system. For light collection, a collimating mirror is installed between the EAST plasma and the VUV spectrometer to extend the observing range to cover the upper divertor region. Two types of detectors, i.e. a back-illuminated charge-coupled device detector and a photomultiplier-tube detector, are adopted for the spectral measurement and high-frequency intensity measurement for feedback control, respectively. The angle between the entrance and exit optical axis is fixed at 15°. The detector can be moved along the exit axis to maintain a good focusing position when the wavelength is scanned by rotating the grating. The profile of impurity emissions is projected through the space-resolved slit, which is set horizontally. The spectrometer is equipped with two gratings with 2400 grooves/mm and 2160 grooves/mm, respectively. The overall aberration of the system is reduced by accurate detector positioning. As a result, the total spectral broadening can be reduced to about 0.013 nm. The simulated performance of the system is found to satisfy the requirement of measurement of impurity emissions from the divertor area of the EAST tokamak.     

Preliminary study of divertor particle exhaust in the EAST superconducting tokamak

The particle exhaust of the upper tungsten and lower carbon divertors in EAST has been preliminarily studied during the 2016 experimental campaign. The density decay time during terminating gas puffing has been employed as a key parameter to evaluate the divertor particle exhaust performance. Comparative plasma discharges have been carried out on the particle exhaust performance between two toroidal field directions in the upper single null and lower single null divertor configurations. This work has enhanced the understanding of the effects of the in–out asymmetry and divertor geometry on the efficiency of the divertor particle exhaust. In addition, the sensitivity of the particle exhaust capability on different strike point locations has been analyzed. The experimental results are expected to provide important information on the future upgrade of EAST bottom divertor and facilitate the realization of longer pulse operation.     

Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R&D area in fusion research. Pre-qualification tests for brazed joints between W–CuCrZr and C–CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m² uniform heat flux. Details about experimental and computational work are presented here.     

Design of flowing liquid lithium device for HT-7 tokamak

Lithium is a very attractive element due to its very low radiation power, strong H retention as well as strong O getter activity. Flowing liquid lithium (FLiLi) device, to be used as a plasma-facing limiters, has been designed and will be tested in HT-7 tokamak. It is mainly composed of distributor, guide plate, collector, and heater as well as cooling loop. The heater uses heater strip and cooling loop design, to control the temperature of lithium on the guide plate ranging from 200 °C to 400 °C. The distributor attached to feeding pipe, distributes liquid lithium (LiLi) flowing on the guide plate. The collector was designed to reclaim the superfluous LiLi and transport it out of device.The paper focuses on the design of flowing liquid lithium device. In addition to the process of design, thermal analysis has been carried out using finite element method (FEM) for optimizing the structure of heater and cooling loop and results of analysis are presented.     

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n_e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n_e/n_GW  1), to enable n_e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n_e/n_GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m²) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.     

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

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

Preliminary 2D code simulation of the quasi-snowflake divertor configuration in the FAST tokamak

The behaviour of the SOL in FAST for the new quasi snowflake configuration of the divertor is compared with that of the conventional single null case for three main scenarios: reference, advanced and extreme H-mode. The flexible, quick and versatile 2D code TECXY is used. The main physics processes occurring at the edge are carefully taken into account but the neutral dynamics is simplified. Even though the status close to detachment conditions cannot be detailed, new phenomena generated by the configuration are clearly highlighted. The heat load is strongly mitigated to a level that is much easier to tackle with the present technology. Mitigation is always stronger than expected from the magnetic topology only, especially in the high density regimes. This is attributed to the much longer time that the SOL particles spend in the quite cold region of the X point where mostly the connection length of the magnetic lines is strongly increased and hence to the much more intense interaction of the plasma particle with the background neutrals.     

Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.     

Simulation of neutral gas flow in a tokamak divertor using the Direct Simulation Monte Carlo method

This paper presents a new innovative scientific and engineering approach for describing sub-divertor gas flows of fusion devices by coupling the B2-EIRENE (SOLPS) code and the Direct Simulation Monte Carlo (DSMC) method. The present study exemplifies this with a computational investigation of neutral gas flow in the ITER's sub-divertor region. The numerical results include the flow fields and contours of the overall quantities of practical interest such as the pressure, the temperature and the bulk velocity assuming helium as model gas. Moreover, the study unravels the gas recirculation effect located behind the vertical targets, viz. neutral particles flowing towards the plasma chamber. Comparison between calculations performed by the DSMC method and the ITERVAC code reveals a very good agreement along the main sub-divertor ducts.     

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.     

Achieving temporary divertor plasma detachment with MARFE events by pellet injection in the EAST superconducting tokamak

A new pellet injection system has been equipped on the experimental advanced superconducting tokamak(EAST) in the 2012 campaign,with a pellet size of Ф 2 mm?×?2 mm,a frequency of1 Hz–10 Hz and velocity of 150 m s~(-1)–300 m s~(-1).The deuterium pellet is well-known for plasma fuelling as well as for? triggering the edge localized mode(ELM).In the 2012 campaign,pellet injection experiments were successfully carried out on EAST.Temporary plasma detachment achieved by deuterium pellets has been observed in a double null(DN) divertor configuration,with multi-pellet injections at a repetition frequency of 2 Hz.The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection,which have a duration of 30–60 ms with the maximum degree of detachment(DOD) reaching 3.5 and 37,respectively.Meanwhile,the multifaceted asymmetric radiation from the edge(MARFE) phenomena was also observed at the high field side(HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%–70%,which may be the main cause of divertor plasma detachment.The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.     

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT.The experiments show an increase in the highest central channel line-averaged density from 2.73 x 1019 m-3 to 6.49 x 1019 m-3,while the X-point moves away from the target by increasing the divertor coil current.The corresponding Greenwald fraction rises from 0.50 to 0.79.For the impurity transport,the density normalized radiation intensity(absolute extreme ultraviolet and soft x-ray)of the central channel density decreased significantly(＞50％)with an increase in the plasma density.To better understand the underlying physics mechanisms,the 3D edge Monte Carlo code coupled with EIRENE(EMC3-EIRENE)has been implemented for the first time on J-TEXT.The simulation results show good agreement with the experimental findings.As the X-point moves away from the target,the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.