Tungsten and CFC degradation under combined high cycle transient and steady state heat loads

Within the ITER divertor lifetime millions of transient events are expected during H-mode operation due to edge localized modes (type I ELMs). These will deposit their energy on plasma facing materials that are pre-heated to various surface temperatures, depending on the steady state heat load (SSHL) at the respective location, leading to synergistic effects. An electron beam facility was used to simulate ELM-like heat loads with ITER relevant power densities (≈0.5 GW/m²) and pulse duration (0.5 ms). At the same time additional SSHL was applied to obtain different base temperatures. Experiments were performed on actively cooled pure tungsten and the carbon fiber composite (CFC) NB41, applying 10³–10⁶ pulses of 0.5 ms duration with a power density of 0.14–0.55 GW/m² and 0.55–0.68 GW/m² on tungsten and CFC, respectively. Surface temperatures were about 200 °C, 400 °C and 700 °C for tungsten and about 450 °C for CFC. Crack formation in tungsten was preceded by roughening due to plastic deformation. In case of T_surf ≈ 200 °C cracks propagated comparably fast (brittle material), while slow propagation and recrystallization around the crack edges indicated fatigue damage at higher temperatures. Compared to tungsten, CFC showed a higher damage threshold.     

Development and experimental study of beryllium window for ITER radial X-ray camera

Radial X-ray camera (RXC) is a diagnostic device planned to be installed in the ITER Equatorial Port #12. Beryllium window will be installed between the inner and outer camera of RXC, which severs as the transmission photocathode substrate and also the vacuum isolation component. In this paper the design and manufacture process of two types of beryllium windows were introduced. Although 50 μm thickness of beryllium foil is the best choice, the 80 μm one with X-ray threshold of 1.34 keV was selected for safety consideration. Using the intermediate layer (low purity of beryllium) between the beryllium foil and the stainless steel base flange is an effective strategy to limit the welding thermal deformation and thermal stress of the thin foil caused by bonding between different materials. By using ANSYS software, the feasibility of the aperture design was analyzed and validated. Metal sealing ring was applied in the mechanical clamped beryllium window for its good stability under high temperature and neutron radiation. Although both of the hollow metal sealing ring with 0.03 mm silver coating and the pure silver sealing ring can satisfy the sealing requirement, the later one was chosen to produce the final product. Two hours 240 °C high temperature baking test, two hours 3.3 Hz vibration test and fatigue test were performed on the two types of beryllium windows. Based on the tests results, the two types of beryllium windows could stand the high temperature baking during the wall conditioning phase of ITER tokamak and the vibration during transportation without causing large leakage. Both of the two types of beryllium windows could bear impact load (0.1 MPa pressure difference) for many times without failure.     

Korean high heat flux test facility by using electron beam system for ITER first wall semi-prototype

A Korean high heat flux test facility for the semi-prototype (SP) qualification of an ITER first wall (FW) will be constructed to evaluate the fabrication technologies required for the ITER FW, and the acceptance of these developed technologies will be established for the ITER FW manufacturing procedure. Korea participated in this qualification program, and is responsible for suitable arrangements for the heat flux test of our fabricated SPs. Qualification testing can be started provided that adequate quality and control measures are implemented and validated by the ITER Organization (IO). The controlling measures required for all heat flux tests shall be concrete and demonstrate the satisfaction of the IO test programs. Each country shall provide a test plan covering the quality and controlling measures in the high heat flux test facility to be implemented throughout the test program. Korean high heat flux testing for these ITER plasma facing materials will be performed by using a 60 kV electron beam and a power supply system of 300 kW, where the allowable target dimension is 70 cm × 50 cm in a vacuum chamber. In addition, this facility needs a cooling system for a high-temperature target and decontamination system for beryllium filtration.     

Fabrication and high heat flux test of large mockups for ITER first wall semi-prototype

The main topic of an ITER blanket first wall procurement is to qualify whether each party has the key technology needed for the fabrication and joining of first wall components. A semi-prototype qualification project will be released requiring that the single components of a full-scale first wall must be fabricated and successfully pass high heat flux tests using a hypervapotron cooling channel. In this work, various mockup types have been modeled and fabricated to develop the joining technology for a semi-prototype. The semi-prototype, which has three double-fingered panels, is a scaled-down component of a full-size first wall. The standard or slit mockups with a 80 mm × 80 mm single beryllium tile joined to a CuCrZr heat sink were fabricated to qualify our HIP (Hot Isostatic Pressing) technology for the joining of semi-prototype. These standard mockups were installed to perform a high heat flux test in the Korea heat load test facility (KoHLT). For a preliminary test of a semi-prototype, thermo-hydraulic mockups of 710 mm × 100 mm were designed and fabricated to verify the Cu/SS cooling performance, such as hypervapotron. For the high heat flux test in our KoHLT facility, the normal cycle is based on an expected heat flux of 300 s in accordance with the ITER qualification specifications. These tests will be performed to qualify the joining technologies, which is required for an ITER blanket first wall and a semi-prototype.     

Thermo-mechanical analyses and assessment with respect to the design codes and standards of the HCPB-TBM Breeder Unit

The Helium Cooled Pebble Bed-Test Blanket Module (HCPB-TBM) is one of the two breeding blanket concepts currently under development in Europe. Key component of the HCPB-TBM, the Breeder Unit (BU), has entered the detailed engineering design phase. After establishing a base design, thermal and thermo-mechanical analyses have been performed under typical ITER operational conditions: the results are presented and discussed in this paper.A full scaled finite element model of the base design of the HCPB-TBM BU has been built to run thermal analyses of the beryllium and Li₄SO₄ pebble beds and thermo-mechanical analyses of the BU structure, both in steady state and in a typical transient regime during a pulse of the ITER D–T phase. The temperatures reached in the Li₄SO₄ and beryllium pebble beds in the BU base design are 930.8 °C and 712.9 °C, respectively, which are above the recommended values of 920 °C for Li₄SO₄ and 650 °C for the beryllium pebbles. The maximum temperature in the structural steel is 548.4 °C, which remains under the design limit defined for the TBM studies (550 °C). In order to decrease the temperature in the hot spots identified in the pebble beds, a reduction of the Li₄SO₄ and beryllium bed volumes has been adopted. As for the structural material, the thermo-mechanical analyses have been assessed with respect the RCC-MR design code (completed for irradiation damages with ITER SDC-IC). The results reveal some problematic points in the base design, concentrated in the coolant inlet and outlet pipes and in the connection region of the BU cooling plates with the BU backplate. Submodeling technique has been used to improve the design in these regions. An increase in the thickness of the coolant inlet and outlet pipes and a redesign of the BU backplate have led to a fulfillment of the codes and standards. The design modifications of pebble bed region and structural material have been implemented in the final design of the BU that is presently used as reference for the design and test of a BU mockup in KIT.     

Operation of an ITER relevant inspection robot on Tore Supra tokamak

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years R&D programme was jointly conducted by CEA-LIST Interactive Robotics Unit and the Institute for Magnetic Fusion Research (IRFM) in order to demonstrate the feasibility and reliability of an in-vessel inspection robot relevant to ITER requirements.The Articulated Inspection Arm robot (AIA) is an 8-m long multilink carrier with a payload up to 10 kg operable between plasma under tokamak conditioning environment; its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel.Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration.In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10^?6 Pa and 120 °C) after a conditioning phase at 200 °C to avoid outgassing pollution of the chamber.This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.     

Development of water-cooled solid breeder test blanket module in JAEA

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m², 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.     

Erosion of lithium coatings on TZM molybdenum and graphite during high-flux plasma bombardment

The rate at which Li films will erode under plasma bombardment in the NSTX-U divertor is currently unknown. It is important to characterize this erosion rate so that the coatings can be replenished before they are completely depleted. An empirical formula for the Li erosion rate as a function of deuterium ion flux, incident ion energy, and Li temperature was developed based on existing theoretical and experimental work. These predictions were tested on the Magnum-PSI linear plasma device capable of ion fluxes >10²⁴ m⁻² s⁻¹, ion energies of 20 eV and Li temperatures >800 °C. Li-coated graphite and TZM molybdenum samples were exposed to a series of plasma pulses during which neutral Li radiation was measured with a fast camera. The total Li erosion rate was inferred from measurements of Li-I emission. The measured erosion rates are significantly lower than the predictions of the empirical formula. Strong evidence of fast Li diffusion into graphite substrates was also observed.     

Overview of recent developments in pellet injection for ITER

Pellet injection is the primary fueling technique planned for core fueling of ITER burning plasmas. Also, the injection of relatively small pellets to purposely trigger rapid small edge localized modes (ELMs) has been proposed as a possible solution to the heat flux damage from larger natural ELMs likely to be an issue on the ITER divertor surfaces. The ITER pellet injection system is designed to inject pellets into the plasma through both inner and outer wall guide tubes. The inner wall guide tubes will provide high throughput pellet fueling while the outer wall guide tubes will be used primarily to trigger ELMs at a high frequency (>15 Hz). The pellet fueling rate of each injector is to be up to 120 Pa m³/s, which will require the formation of solid D–T at a volumetric rate of ～1500 mm³/s. Two injectors are to be provided for ITER at the startup with a provision for up to six injectors during the D–T phase. The required throughput of each injector is greater than that of any injector built to date, and a novel twin-screw continuous extrusion system is being developed to meet the challenging design parameters. Status of the development activities is presented, highlighting recent progress.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Prototyping of the Blanket Shield Module for the ITER EC H&CD Upper launcher

The design of the ITER Electron Cyclotron Heating and Current Drive (ECH&CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components.The most outstanding structural In-vessel component of an ECH&CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3 MW/m³ at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5 m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5 MW/m² due to plasma radiation must be dissipated from the FWP.To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure.Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy.This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE.     

Diffusion bonding beryllium to Reduced Activation Ferritic Martensitic steel: Development of processes and techniques

Beryllium was successfully bonded to a Reduced Activation Ferritic Martensitic (RAFM) steel with a maximum strength of 150 MPa in tension and 168 MPa in shear. These strengths were achieved using Hot Isostatic Pressing (HIP), at temperatures between 700 °C and 750 °C for 2 h and under a pressure of 103 MPa. To obtain these strengths, 10 μm of titanium and 20 μm of copper were deposited on the beryllium substrate prior to HIP bonding. The copper film acted a bonding aid to the RAFM steel, while the titanium acted as a diffusion barrier between the copper and the beryllium, suppressing the formation of brittle intermetallics that are known to compromise mechanical performance. Slow cooling from the peak HIP temperature along with an imposed hold time at 450 °C further enhanced the final mechanical strength of the bond.     

Operational characteristics of the high flux plasma generator Magnum-PSI

In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10²³–10²⁵ m⁻² s⁻¹ hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.     

Design progress on ITER port plug test facility

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment.The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs.Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 × 10^?5 Pa at 100 °C containing a port plug. The heating system shall provide water at 240 °C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests.This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.     

ASDEX Upgrade enhancements in view of ITER application

The aim of the ASDEX Upgrade (AUG) programme is to support the design, prepare the physics base and develop regimes beyond the baseline of ITER and for DEMO. Its ITER-like geometry, poloidal field system, versatile heating system and power fluxes make AUG particularly suited.After the transition to fully tungsten coated plasma facing components AUG could be operated without prior boronizations and a low permanent deuterium retention was found qualifying W as wall material. ITER-like baseline H-modes (H₉₈ ～ 1, β_N ～ 2) were routinely achieved up to 1.2 MA plasma currents. W concentrations could be kept at an acceptable level of <5 × 10^?5 by central wave heating (enhancing impurity outward transport) and ELM pacing with gas puffing. The compatibility of high performance improved H-modes, the ITER hybrid scenario, with an un-boronized W wall was demonstrated achieving H₉₈ ～ 1.1 and β_N up to 2.6 at modest triangularities δ ≤ 0.3. This performance is reached despite the gas puffing needed for W influx control. Increasing δ to 0.35 allowed at even higher puff rates still a H₉₈ ～ 1.1.Reliable plasma operation in support of ITER comprised the demonstration of ECRF assisted low voltage plasma start-up and current rise at toroidal electric fields below 0.3 V/m resulting in a ITER compatible range of plasma internal inductance of 0.71–0.97. Disruption mitigation is feasible using strong gas puffs, and the achieved electron densities approach values needed for runaway suppression.Present hardware extensions in support of ITER include the upgrading of ECRH by a 4 MW/10 s system with large deposition variability (tuneable frequency between 105 and 140 GHz, real-time steerable mirrors) for central heating and MHD mode control. A powerful system of 24 in-vessel coils produces error fields up to toroidal mode number n = 4 for ELM suppression and mode rotation control. In connection with a close conducting wall they will open up the road for RWM stabilization in advanced scenarios. For those we are considering LHCD for current drive and profile control with up to 500 kA driven current. The tungsten sources are dominated by sputtering from intrinsic light impurities, and the W influx from the outboard limiters are the main source for the core plasma. ICRH induced electric fields accelerate light impurities, restricting the use of ICRH to just after boronization. 4-strap antennas imbedded in extended wall structures might solve this problem. Finally, doubling the plasma volume with plasma currents above 2 MA in AUG could be the solution for a needed ITER satellite.     

Issues from activation/dose rates calculation for full W ITER machine and increased fluence

The paper presents and discusses the results of the activation calculations for ITER to investigate the effect of a possible increase of the neutron fluence on the First Wall (FW) and of the use of tungsten instead of beryllium as inboard and outboard FW protective layer. The new analyses based on the recent ITER design have been performed using the most recent, reliable and validated nuclear data and codes. Three situations have been considered: (1) average FW neutron fluence of 1.0 MWa/m² with beryllium FW protective layers (PLs); (2) average FW neutron fluence of 0.5 MWa/m² with tungsten FW PLs and (3) average FW neutron fluence of 1.0 MWa/m² with tungsten FW PLs.The used approach considers the Scalenea-1 radiation transport sequence for obtaining the 175 groups neutron fluxes in all the materials/zones on the radial direction of the ITER equatorial midplane and the EASY2007 code package for the material activation analysis. For tungsten PLs, calculations have also been performed using a MCNP approach in a 1D geometry. In this way the effect of the multi-group treatment of cross-sections is compared versus a continuous energy treatment and the deriving self-shielding effect is determined in the case of W where many resonances are present.     

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m²) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/m² until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/m²) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.     

Vacuum system of SST-1 Tokamak

Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN₂ cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10^?4 mbar and 1.0 × 10^?5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10^?6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10^?5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ～3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10^?5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10 °C deviation was achieved with a net mass flow rate of 0.8 kg/s at 1.5 bar gauge inlet pressure and supply temperature of 230 °C at the heater end. Also during gas feed system installation, the pressure inside the VV was raised from 3.01 × 10^?5 mbar to 1.72 × 10^?4 mbar by triggering a pulse of lower amplitude of 25 voltage direct current (VDC) for 100 s to piezoelectric valve. This paper describes in detail the design and implementation of the various vacuum subsystems including relevant experimental results.     

Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements.The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented.     

Preliminary design and performance study of EAST liquid lithium limiter based on CPS

Lithium has the ability of H recycling suppression and impurities absorption and it can be used as plasma facing material (PFM) in tokamaks. Lithium conditioning experiments were launched on EAST, HT-7 and some other tokamaks for many years by using the methods of GDC, IRCF and evaporation. Liquid lithium has better performances in effective lifetime and heat removal aspects compared to non-liquid lithium. While, applying liquid lithium in the tokamak would cause the safety problem as the lithium can react with many substances violently and the magnetohydrodynamic behavior is difficult to be handled. EAST liquid lithium limiter (LLL) system is under developing and will be applied in EAST to study the main technologies of the liquid lithium application. The normal operation temperature of the limiter is expected as 230–550 °C under the active cooling of water. Capillary porous system (CPS) is used to prevent the lithium from splashing under large electromagnetic force by increasing the surface tension of the lithium. In order to investigate the cooling performance of the cooling design, the thermal-hydraulic analysis was done which shows that with 3 m/s flowing velocity, the water can keep the limiter under 550 °C all the time if the heat flux is lower than 0.7 MW/m². Under heat flux of 1 MW/m², the limiter should be retreated within 7 s to avoid erosion. The pressure drop of the coolant under 3 m/s is less than 40 kPa with temperature difference nearly 34 °C which meet the design requirements very well. The key manufacture process and technologies like vacuum bonding between the CuCrZr heat sink and 316L guide plate were well studied in the R&D process. The heating test on the test bench showed that the CPS can be heated efficiently by the heaters attached into the heat sink.