CFC Monoblock Research Articles

Design, research and development for plasma facing components in JT-60SA

An overview of plasma facing components (PFCs) in JT-60SA is introduced, including the upper divertor, lower divertor, inboard first wall and outboard first wall. These PFCs are upgraded in four stages during a plasma experiment in JT-60SA. At stage one, only a partial inboard first wall and upper divertor are installed with graphite tiles and inertial cooling. At stage two, the remaining PFCs including the outboard first wall and lower divertor are installed, and the inboard surface is fully covered. At stage three, the vertical targets of the lower divertor are replaced with CFC monoblock targets and water cooling. At stage four, the whole carbon wall is replaced with a tungsten wall, and remote handling for the lower divertor is introduced. The research and development on major issues concerning PFCs are presented. Mock-ups of a CFC monoblock target were fabricated and it showed good heat removal of the expected maximum heat load of 15 MW m−2 without degradation of the performance. Tungsten-coated samples were fabricated on the CFC and they showed no severe damage at surface temperatures of up to 2000 °C. Laser-welded samples showed good appearance and no defects were found by RT.

R&D activities on manufacturing plasma-facing unit for prototype of ITER divertor outer target in JADA

Japan Domestic Agency (JADA) carried out R&Ds activities to improve joining CFC monoblocks onto a CuCrZr cooling tube in PFUs to boost the success rate of joint and to confirm load carrying capability of the monoblock attachments to Steel Support Structure (SSS) against tensile force simulating electromagnetic load to pull PFUs from SSS. In joining the CFC monoblocks to the cooling tube, JADA has adopted brazing by using noble-metal-free filler with the following improvements; (1) metalizing joint surface of CFC using Ti-coating with accurate thickness controlling, (2) Changing buffer layer material from soft pure copper to Cu–W alloy. By using the present improved joint, JADA has manufactured three mock-ups with 5 CFC monoblocks and tested against repetitive high heat loads more than 20MW/m2. All of CFC monoblocks of each mockup can survive the high heat loads throughout 1000 cycles with no degradation of heat removal capability. Regarding the load carrying capability of monoblock attachments to SSS, tensile experiments were carried out using the same geometries of CFC and tungsten monoblocks in PFUs and the results show that both geometries and joints meet the ITER requirements, that is, 3kN and 8kN, respectively.

One-step brazing process for CFC monoblock joints and mechanical testing

A new method of joining CFC to copper (CFC/Cu) and CFC/Cu to CuCrZr alloy (CFC/Cu/CuCrZr) was previously developed for the flat-type configuration. The joining technique foresees a single-step brazing process: the brazing of the three materials (CFC, Cu and CuCrZr) can be performed in a single heat treatment using the same non-active brazing alloy. The composite surface was previously modified by solid state reaction with chromium with the purpose of increasing the wettability of CFC by the brazing alloy. The feasibility of this process also for monoblock geometry is described in this work. The thermal fatigue resistance of the joined samples (quenching from 450 °C to RT; 50 cycles) was tested and the joints were characterized by apparent shear tests before and after thermal fatigue. The apparent shear strength of the CFC/Cu/CuCrZr joined samples was unaffected after these thermal fatigue tests.

Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program “Broader Approach”. The maximum heat flux to JT-60SA divertor is estimated to ∼15 MW/m 2 for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R&D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R&D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m 2 as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

Experimental activity on the definition of acceptance criteria for the ITER divertor plasma facing components

Tens of thousands of armor/heat sink joints will be produced by the industry during the manufacturing of ITER divertor PFC, statistically, there is a probability that joints with defects be delivered. The purpose of this paper is to study the detection and evolution during operation of calibrated defects artificially implemented on samples, as an experimental basis for the definition of acceptance criteria for the bond armor/heat sink in the frame of industrial manufacturing conditions.It was found that current CFC monoblock design option was compatible with the heat loads specified at the lower part of the vertical target (up to 20 MW/m 2), including the presence of armor/heat sink defects (up to 50° extension for a location at 0° or 45°) detectable with NDE techniques developed in Europe (US, SATIR). The current W monoblock design appeared suitable for the upper part of the vertical target with defects extension up to 50°, but is not adapted for heat flux of 20 MW/m 2. The studied W flat tile design proved to be compatible with fluxes of 5 MW/m 2 but unable to sustain cycling fluxes of 10 MW/m 2.

Pulsed thermography of CFC monoblock divertor components

In the frame of the Wendelstein 7-X project, the ARGUS facility of PLANSEE for pulsed thermography testing (PTT) has been developed as an inspection method for divertor components containing a CFC flat tile armor. The benefit is that inspection without a cooling system and without a reference element can be performed. The usability of PTT for monoblock components inspection was investigated on the PLANSEE qualification vertical target prototypes (QVTP) for ITER. The influence of different defects on PTT was analyzed by finite element modeling (FEM). With these results an acceptance criterion for PTT was defined. The size of natural defects detected by PTT was determined by X-ray computed tomography (CT). The defect resolution required by ITER specifications is fulfilled by PTT. The repeatability of the testing procedure is sufficiently high to perform inspections without a reference element. Because of these features thermographic inspection in a commercial manufacturing line becomes feasible.

Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined. US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.

Evaluation of the erosion on the CFC tiles of the ITER divertor by means of FE calculations

In order to evaluate by finite element (FE) calculations the possibility of erosion instability on the CFC tiles of the vertical target of the ITER divertor an original methodology has been developed using the CAST3M CEA finite element code. The used approach provides a series of steady state calculation steps, the mesh being updated at each step of the iteration in order to take into account the erosion between two steps. Both physical and chemical sputtering together with sublimation have been included in the code to describe the loss of material by the thermal and particle loads envisaged for ITER normal operation regime. This model has been validated by comparison with analytical or other code results and then applied to the ITER divertor vertical target (CFC monoblock geometry). The calculations have pointed out the possibility of some erosion instabilities when a monoblock has a neighbour one with reduced conductivity or with interface defects. However it was shown that when applying normal condition loads (10 MW/m 2) for 10,000 s and then off-normal loads (20 MW/m 2) for 10 s the total erosion was 1 mm. Based on these first results no modification of the monoblock acceptance criteria is necessary. This way of calculation can be easily transposed in other FE codes and extensively used for checking various ITER operative scenarios.

Testing of actively cooled mock-ups in several high heat flux facilities—An International Round Robin Test

Several electron beam and ion beam facilities are involved in high heat flux testing of plasma-facing components for next step fusion devices. Up to a certain degree, these machines are comparable, but differences concern, e.g. beam generation, beam sweeping, calibration techniques and diagnostics. In order to get an information if tests in the different facilities are really comparable, a set of actively cooled CFC monoblocks has been heated in four electron beam and one ion beam facility at comparable power densities. The temperature response during these loadings has been registered and used as a criteria for assessment.

Hot radial pressing: An alternative technique for the manufacturing of plasma-facing components

The Hot radial pressing (HRP) manufacturing technique is based on the radial diffusion bonding principle performed between the cooling tube and the armour tile. The bonding is achieved by pressurizing the cooling tube while the joining interface is kept at the vacuum and temperature conditions. This technique has been used for the manufacturing of relevant mock-ups of the ITER divertor vertical target. Tungsten monoblock mock-ups were successfully tested to high heat flux thermal fatigue (20 MW/m 2 of absorbed heat flux for 1000 cycles). After these good results the activity is now focused on the developing of a manufacturing process suitable also for the CFC monoblock mock-ups. A FE calculation was performed to investigate the stress involved in the CFC tiles during the process and to avoid the CFC fracture. The results obtained by the FE calculation and by the test performed in air simulating a HRP manufacturing process for a CFC monoblock mock-ups is reported in the paper.

Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime.

Non-destructive testing of CFC monoblock divertor mock-ups

Non-destructive examination (NDE) methods for joint interfaces between different materials in high heat flux (HHF) components of divertor should be urgently developed to assure quality and reliability of joining techniques. The purpose of this work is to demonstrate the ability of using ultrasonic wave and thermography NDE techniques to detect the defect in the joining interface (joint defect) of divertor mock-ups with carbon-fiber reinforced carbon monoblock armor tiles brazed on a copper cooling tube. The results of both NDEs are benchmarked with HHF tests and cross-sectional observation of the mock-up to correlate the joint defect size detected with NDEs to the thermal response of the mock-up with initial joint defects. From the results of the HHF tests and the cross-sectional observations, it can be concluded that both NDE techniques have sufficient accuracy to predict the surface temperature of the HHF components.

Ultrasonic non-destructive testing on CFC monoblock divertor mock-up

Non-destructive ultrasonic testing has been applied for the characterization of joints by means of a polymer transducer. One of the advantages of the polymer transducer is flexibility in its shape and the possibility to install multiple transducers in one probe, which can reduce the time for inspection. As a first step, the size effect of the transducer on the resolution and sensitivity was examined to detect the joint flaw. Transducers with circumferential angles of 5°, 10° and 30° were tested. For this test a small divertor element with a drilled hole was prepared, which simulates a joint defect. The transducers with angles of 30 could not characterize the size of the artificial joint flaw. On the contrary, the size of the artificial defect was successfully detected with an accuracy of 90% by means of the transducers with angles of 5 and 10°. From the viewpoint of the sensitivity of the detection of the joint flaw, the transducer with the angle of 10° is appropriate because it could detect the largest intensity of the reflected signal caused by the same artificial defect of the joint interface.

Manufacturing and testing of a prototypical divertor vertical target for ITER

After an extensive R&D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m 2, 1000 cycles at 10 MW/m 2 and more then 1000 cycles at 15–20 MW/m 2. The final critical heat flux test reached a value in excess of 30 MW/m 2.

Comparison of electron beam test facilities for testing of high heat flux components

In the last few years, electron beam facilities for the testing of high heat flux components have been erected in Europe, Japan, Russia and in the USA. In principle all the facilities are comparable, but some machine parameters are quite different. These differences include electron beam operation (beam generation, beam diameter, sweeping mode), as well as the temperature measurement devices, calibration techniques and the definition of absorbed power densities. In order to assess the influence of these machine parameters and techniques on the results of high heat flux experiments, a round robin test has been performed in five facilities. In these tests, actively cooled CFC monoblock mock-ups were heated by electron beams using target power densities up to 15 MW/m 2. Mock-up temperatures and their distribution, measured by different methods (IR camera, pyrometer, thermocouples), have been used as criteria for comparison. The evaluation of data from the different facilities shows good agreement for identical target loading conditions.

Testing of Neutron-Irradiated Plasma Facing Components

In order to study degradation effects of neutrons on plasma-facing materials and joints, actively-cooled CFC monoblock mock-ups were irradiated in the High Flux Reactor in Petten up to 0.35 dpa at 350 and 750°C. Later, these samples were tested by means of an electron beam facility under static and cyclic heating conditions. The heat removal efficiency and the thermal fatigue behavior of these samples were compared to those of corresponding non-irradiated samples. A significant increase of surface temperature was observed for all CFC samples, due to a reduced thermal conductivity of the materials after neutron irradiation. This effect is less distinctive for samples irradiated at the higher temperature. Long term fatigue tests with 1000 heating cycles at 15 MW/m2 did not create any failure of the plasma-facing material or the bond layer of the tested mock-ups.

Testing of actively cooled high heat flux mock-ups

Several un-irradiated CFC monoblock mock-ups have been loaded in thermal fatigue tests up to 1000 cycles at power densities <25 MW/m 2. No indication of failure was observed for these loading conditions. Two of the mock-ups were inspected by ultra-sonic methods before thermal cycling. It could be proved that the voids found in the post-mortem metallography existed before and had no effect on the integrity of the mock-up. For the first time, neutron-irradiated CFC monoblock mock-ups have been tested in the electron beam facility JUDITH. These mock-ups had been irradiated before in the High Flux Reactor at Petten up to 0.3 dpa at 320°C and 770°C. All samples showed a significant increase of surface temperature, due to the irradiation induced decrease in thermal conductivity of the CFC materials.

Divertor development for ITER

The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented.

Thermal fatigue tests with actively cooled divertor mock-ups for ITER

Mock-ups for high heat flux components with beryllium and CFC armour materials have been tested by means of the electron beam facility JUDITH. The experiments concerned screening tests to evaluate heat removal efficiency and thermal fatigue tests. CFC monoblocks attached to DS-Cu (Glidcop Al25) and CuCrZr tubes by active metal casting and Ti brazing showed the best thermal fatigue behaviour. They survived more than 1000 cycles at heat loads up to 25 MW m −2 without any indication of failure. Operational limits are given only by the surface temperature on the CFC tiles. Most of the beryllium mock-ups were of the flat tile type. Joining techniques were brazing, hot isostatic pressing (HIP) and diffusion bonding. HIPed and diffusion bonded Be/Cu modules have not yet reached the standards for application in high heat flux components. The limit of this production method is reached for heat loads of approximately 5 MW m −2. Brazing with and without silver seems to be a more robust solution. A flat tile mock-up with CuMnSnCe braze was loaded at 5.4 MW m −2 for 1000 cycles without damage. The first test with a beryllium monoblock joined to a CuCrZr tube by means of Incusil brazing shows promising results; it survived 1000 cycles at 4.5 MW m −2 without failure.