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.     

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.     

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.     

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.     

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.     

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).     

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.     

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.     

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.     

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.     

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.     

Novel design of a personnel safety system for Hefei Light Source-Ⅱ

The Hefei Light Source-Ⅱ(HLS-Ⅱ) is a vacuum ultraviolet synchrotron light source. The personnel safety system(PSS), which is a personnel access control system,is a crucial part of the HLS-Ⅱ, as it protects the staff and users at HLS-Ⅱ from radiation damages. The prior version of HLS-Ⅱ PSS was based on an access control system called SiPass. This lacked the personnel management function. Meanwhile, as the prior PSS is a turn-key system,it was not effective for sharing information. To overcome these drawbacks, the novel design of PSS for HLS-Ⅱ is proposed based on the Siemens redundant programmable logic controller under the Experimental Physics and Industrial Control System. The proposed PSS consists of a safety interlock system, access control system, and a radiation monitoring system. The safety interlock system is used to define the interlock logic. The access control system is designed to restrict the access of staff and users at HLS-Ⅱ, and to provide a personnel management function.The radiation monitoring system is used to monitor the radiation dose rate in both the light source and the surrounding areas. This paper details the architecture and the specific design of the novel PSS. The off-line test results demonstrate that the proposed system has achieved the design objectives.     

Advances in optical thermometry for the ITER divertor

Thermography will be an important diagnostic on the ITER tokamak, but the inclusion of reflective materials such as tungsten in the design for ITER's first wall and divertor region presents problems for optical temperature measurement. The ongoing testing of ITER plasma facing components (PFCs) provides an excellent opportunity to resolve such problems. This has focused on the variation of PFC emissivity with temperature and time, as well as environmental influence on thermography. The sensitivity of these systems to ambient temperature, due primarily to modification of the transmission of the optical path, has been established and minimised. The accuracy of the system is then sufficient to measure the variation of emissivity in heated material samples, by comparing its front-face luminance measured with an infrared camera to the temperature given by an implanted thermocouple. Measurements on both tungsten and carbon fibre composite are in broad agreement with theory, and thus give the material's function of emissivity with temperature at the start of its life. To determine its evolution, a bicolour pyroreflectometer was then installed. This uses two lasers to measure the reflectivity in addition to the luminance at two wavelengths, and thus the true temperature can be calculated. This was validated against the instrumented sample, then used along with the camera to observe an ITER mock-up during ∼50,000 s of 5 MW/m² testing. Emissivity was seen to vary little in the 500 °C region. Higher temperature tests are ongoing.     

Calculation of the heat flux in the lower divertor target plate using an infrared camera diagnostic system on the experimental advanced superconducting tokamak

During the discharging of Tokamak devices,interactions between the core plasma and plasma-facing components(PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFCs materials. An infrared(IR) diagnostic system consisting of an IR camera and an endoscope was installed on an Experimental Advanced Superconducting Tokamak(EAST) to monitor the surface temperature of the lower divertor target plate(LDTP) and to calculate the corresponding heat flux based on its surface temperature and physical structure, via the finite element method. First, the temperature obtained by the IR camera was calibrated against the temperature measured by the built-in thermocouple of EAST under baking conditions to determine the true temperature of the LDTP. Next, based on the finite element method, a target plate model was built and a discretization of the modeling domain was carried out. Then, a heat conduction equation and boundary conditions were determined. Finally, the heat flux was calculated. The new numerical tool provided results similar to those for DFLUX; this is important for future work on related physical processes and heat flux control.     

Poloidal divertor for the Wisconsin Tokamak Reactor Design

A poloidal divertor has been incorporated in the Wisconsin Tokamak Reactor Design (UWMAK-I). Its main features include a double null design using superconducting divertor coils located outside the main toroidal magnets and liquid lithium particle collectors located inside the blanket and shield.     

Performance of electro-plated and joined components for divertor application

A general challenge in divertor development, independently of design type and cooling medium water or helium, is the reliable and adapted joining of components. Depending on the design variants, the characteristics of the joints will be focused on functional or structural behavior to guarantee e.g. good thermal conductivity and sufficient mechanical strength. All variants will have in common that tungsten is the plasma facing material. Thus, material combinations to be joined will range from Cu base over steel to tungsten. Especially tungsten shows lacks in adapted joining due to its metallurgical behavior ranging from immiscibility over bad wetting up to brittle intermetallic phase formation. Joining assisted by electro-chemical deposition of functional and filler layers showed that encouraging progress was achieved in wetting applying nickel interlayers. Nickel proved to be a good reference material but alternative elements (e.g. Pd, Fe) may be more attractive in fusion to manufacture suitable joints.Replacing of Ni as activator element by Pd for W/W or W/steel joints was achieved and joining with Cu-filler was successfully performed. Manufactured joints were characterized applying metallurgical testing and SEM/EDX analyses demonstrating the applicability of Pd activator. First shear tests showed that the joints exhibit mechanical stability sufficient for technical application.     

Liquid lithium surface control and its effect on plasma performance in the HT-7 tokamak

Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.