M. Rödig

Performance of different tungsten grades under transient thermal loads

Plasma facing components in future thermonuclear fusion devices will be subjected to intense transient thermal loads due to type I edge localized modes (ELMs), plasma disruptions, etc. To exclude irreversible damage to the divertor targets, local energy deposition must remain below the damage threshold for the selected wall materials. For monolithic tungsten (pure tungsten and tungsten alloys) power densities above ?0.3?GW?m?2 with 1?ms duration result in the formation of a dense crack network. Thin tungsten coatings for the so-called ITER-like wall in JET, which have been deposited on a two-directional carbon?fibre composite (CFC) material, are even less resistant to thermal shock damage; here the threshold values are by a factor of 2 lower. First ELM-simulation experiments with high cycle numbers up to 104 cycles on actively cooled bulk tungsten targets do not reveal any cracks for absorbed power densities up to 0.2?GW?m?2 and ELM-durations in the sub-millisecond range (0.8?ms); at somewhat higher power densities (0.27?GW?m?2, ?t = 0.5?ms) cracks have been detected for 106 cycles.

Neutron irradiation effects on plasma facing materials

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

PERFORMANCE OF PLASMA-FACING MATERIALS UNDER INTENSE THERMAL LOADS IN TOKAMAKS AND STELLARATORS

Abstract Beside quasi-stationary plasma operation, short transient thermal pulses with deposited energy densities on the order of several tens of MJ/m2 are a serious concern for next-step devices, in particular, for tokamak devices such as ITER. The most serious of these transient events are plasma disruptions. Here, a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone region. The timescale of these events is typically on the order of 1 ms. In spite of the fact that a dense cloud of ablation vapor will form above the strike zone, only partial shielding of the divertor armor from incident plasma particles will occur. As a consequence, thermal shock–induced crack formation, vaporization, surface melting, melt layer ejection, and particle emission induced by brittle destruction processes will limit the lifetime of the components. In addition, dust particles (neutron-activated metals or tritium-enriched carbon) are a serious concern from a safety point of view. Other transient heat loads that occasionally occur in magnetic confinement experiments such as instabilities in the plasma positioning (vertical displacement events) also may cause irreversible damage to plasma-facing components (PFCs), particularly to metals such as beryllium and tungsten. Other serious damage to PFCs is due to intense fluxes of 14-MeV neutrons in D-T burning plasma devices. Integrated neutron fluence of several tens of displacements per atom in future thermonuclear fusion reactors will degrade essential physical properties of the components (e.g., thermal conductivity). Another serious concern is the embrittlement of the heat sink and the plasma-facing materials (PFMs). To investigate the performance of carbon-based and metallic PFMs under the aforementioned thermal loads, simulation experiments have been performed in highly specialized high-heat-flux test facilities. The neutron-induced degradation of materials and components was investigated on selected test samples that were irradiated in high-flux material test reactors.

Carbon fiber composites application in ITER plasma facing components

Abstract Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

Overview of the EU Small Scale Mock-up Tests for ITER High Heat Flux Components

Abstract This task within the EU R&D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m−2. With flat armour tiles rapid joint failures occurred at 5–16 MW m−2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered.

Neutron-irradiation effects on high heat flux components – examination of plasma-facing materials and their joints

The neutron-irradiation experiments PARIDE 1 and PARIDE 2 have been performed at 350°C and 700°C with fluences of 0.35 dpa. The major part of the post-irradiation tests are high heat flux simulation experiments carried out in the electron beam facility JUDITH. These tests cover thermal fatigue experiments with small-scale high heat flux components, and on the other hand, thermal shock tests on the plasma-facing materials. Actively cooled samples were made from CFC, or beryllium as plasma-facing materials and copper alloys as heat sink materials. Different designs (flat tile, monoblock) and joining techniques (brazing, welding) were used. Best performance was found for CFC/Cu monoblock mock-ups, but also the brazed Be/Cu flat tile mock-ups fulfill the operational requirements for first wall components. Thermal shock experiments show a higher erosion after neutron irradiation. This degradation is either due to a reduced thermal conductivity (carbon) or to a decreased ductility after irradiation (beryllium).

Brittle destruction of carbon-based materials in transient heat load tests

Abstract To investigate the phenomenon of brittle destruction (BD), isotropic fine-grain graphites and carbon fiber composites (CFCs) with and without silicon doping have been exposed to intense transient thermal loads in the electron beam test facility JUDITH. For different pulse durations (1 to 5, 100, 5000 ms) the deposited energy density has been increased stepwise to determine the threshold for the BD process. Particle emission was diagnosed using time-resolved measurements of the absorbed current and by digital photography. A clear correlation of the absorbed current and the onset of the particle emission processes has been established.

Performance of beryllium, carbon, and tungsten under intense thermal fluxes

Abstract Transient heat loads on a millisecond timescale with deposited energy densities beyond 1 MJ m −2 have been simulated in a plasma accelerator facility (VIKA) and in two high power electron beam teststands (JUDITH, JEBIS). Main objective of these experiments was to study and to compare the behaviour of different plasma facing materials (Be, CFC, W) under heat loads which occur during disruptions in future thermonuclear fusion reactors such as ITER. In these tests special attention was paid to the thermal shock resistance, the processes during melt layer formation, and the resulting material erosion. To perform these tests specific loading techniques and diagnostics have been developed and applied. Among these are high heat flux loading experiments at elevated temperatures ( T > DBTT ) of the test coupons, fast surface pyrometry, and reliable techniques for the quantification of the absorbed energy.

Reference testing of actively cooled mock-ups for the neutron-irradiation experiments PARIDE 3 and 4

A neutron irradiation campaign has been performed with new designs of high heat flux components. In parallel to this irradiation, reference tests have been carried out with un-irradiated samples of the same type. This paper reports on the testing of un-irradiated divertor mock-ups from tungsten and CFC attached to water-cooled heat sinks from CuCrZr.

European development of prototypes for ITER high heat flux components

The extensive EU research and development, on international thermonuclear experimental reactor (ITER) high heat flux (HHF) components aims at the demonstration of prototypes for the divertor and baffle with challenging operating requirements. The recent progress of this development is summarised in the paper, particularly concerning the manufacture and testing of mock-ups and prototypes. The available results demonstrate the feasibility of robust solutions with carbon and tungsten armour.