In order to study the boundary conditions for the occurrence of ratcheting, a testing rig was erected. In this rig, small-scale mock-ups of the first wall were tested under mechanical conditions ...similar to those expected during disruptions. These mock-ups were made of stainless steel AISI 316. They were heated on the surface and water cooled from inside. Disruption forces were simulated by a high-speed hydraulic bending system by impact loads. The mock-ups were heated by a high-frequency generator and cooled by water through two cooling channels.
In addition to the experimental work, simplified analytical analysis and numerical calculations were carried out using the finite element program
abaques. In these calculations, the temperature and stress fields in the mock-ups were investigated and the influence of the different constitutive equations was studied.
The experimental results confirm the ratcheting behaviour predicted by the calculations. It has been demonstrated that the ratcheting behaviour for constant mechanical and cyclic thermal loads is much stronger than in the case when both loads are cyclic in nature.
The plasma facing materials and components in existing and future fusion devices are strongly affected by plasma wall interaction processes. These components, in particular the first wall (FW), the ...limiters and the divertor are subject to intense quasi-stationary thermal loads during plasma operation. While the resulting thermal loads to the first wall will remain below 1 MW·m
−2
, special attention has to be paid to high heat flux components like limiters and the divertor. Here the expected power densities will be at least one order of magnitude above the ones at the FW, with expected peak heat fluxes of up to 20 MW·m
−2
for future magnetic confinement devices. Beside quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJ·m
−2
are another serious concern for next step tokamak devices, in particular for ITER. The most serious events are plasma disruptions, vertical displacement events, and Edge Localized Modes (so-called ELMs). These requirements make high demands on the selection of qualified materials and reliable fabrication processes for actively cooled plasma facing components. High heat flux test facilities based on intense electron and ion beams have been utilized successfully to assess the efficiency and the fatigue life time of different material solutions and design concepts. Modeling and experiments with both normal operation scenarios and transient events, are being performed to evaluate and to quantify the resulting material erosion or damage and thus to assess the life time of the components. Additional research activities are focused on the degradation of materials and joints due to energetic neutrons. In order to investigate irradiation induced property changes, materials samples and actively cooled plasma facing components have been irradiated in fission reactors and tested in thermal load tests. The technical solutions which are considered today are mainly based on beryllium, carbon materials or tungsten as armor materials and copper alloys or stainless steel for the heat sink. Furthermore, the needs for extensive quality control methods and non-destructive analyses during the procurement phase will be highlighted.
An evaluation of the erosion under disruption heat loads is very important to the lifetime prediction of divertor armour tiles of next fusion devices such as ITER. In particular, erosion data on CFCs ...(carbon fiber reinforced composites) and beryllium (Be) as the armour materials is urgently required in the ITER design. For CFCs, high heat flux experiments on the newly developed CFCs with high thermal conductivity have been performed under the heat flux of around 800–2000 MW/m
2 and the pulse length of 2–5 ms in JAERI electron beam irradiation systems (JEBIS). As a result, the weight losses of B
4C doped CFCs after heating were almost same to those of the non doped CFC up to 5 wt% boron content. For Be, we have carried out our first disruption experiments on S65/C grade Be specimens in the Juelich divertor test facility in hot cells (JUDITH) facility as a frame work of the J—EU collaboration. The heating conditions were heat loads of 1250–5000 MW/m
2 for 2–8 ms, and the heated area was 3 × 3 mm
2. As a result, the protuberances of the heated area of Be were observed under the lower heat flux.
Paper introduces thermal fatigue testing devices, developed and operated under fusion-related projects with focus on testing of ITER First Wall (FW) mock-ups. In frame of EFDA tasks, FW mock-up ...testing device was developed and put into operation in Centrum Výzkumu Řež s.r.o. (CV Rez) research centre, and one testing device was modified for thermal fatigue tests in Forsungszentrum Jülich (FZJ) research centre. The FW mock-ups were tested for several parties, where the key role was played by Fusion for Energy (F4E), the European Union's Joint Undertaking for ITER and the Development of Fusion Energy.
Over the last decade alternative technologies for the manufacture of the ITER first wall blanket have been progressively developed. Now, as the build of ITER approaches, the manufacturing route is ...being consolidated around the best solutions found to date. This paper reviews the development of the HIP bonding technologies and discusses the latest results from components produced by AMEC NNC under the auspices of EFDA. The manufacturing stages, non-destructive examination and heat flux test results from the work are presented for the latest first wall prototype components. It is concluded that the technologies developed will allow the production of components which meet the heat flux performance requirements for ITER. As part of this work a low temperature Be/Cu alloy bonding process has been developed which is compatible with both DS-Cu and PH-Cu alloys.
In the present study, the effect of disruptions on beryllium has been studied. Disruptions are simulated in the electron beam facility JUDITH by high energetic pulses of up to 250
MJ/m
2. Under these ...loads, the beryllium surface may roughen combined with the forming of cracks. During the experiments, a special problem arises from the fact that during the neutron irradiation beryllium transmutes to tritium. This tritium is bound in the beryllium matrix, but during the heating of the samples, the tritium may be set free and through the vacuum pump it may be released to the environment. In order to avoid and to quantify this release of tritium, a special tritium trap has been constructed. In this tritium trap the gas is pumped by means of a metal bellows pump through a catalyst tube filled with copper oxide. At a temperature of 300
°C, the tritium is oxidized to HTO. This HTO is lead through gas washing bottles filled with water. Here approximately 98% of the released tritium is caught. The temperatures in the process are controlled by thermocouples, and the tritium content is controlled by a tritium gas monitor and in addition with a liquid scintillation counter (LSC).
The development of plasma facing components for next step fusion devices in Europe is strongly focused to ITER. Here a wide spectrum of different design options for the divertor target and the first ...wall have been investigated with tungsten, CFC, and beryllium armor. Electron beam simulation experiments have been used to determine the performance of high heat flux components under ITER specific thermal loads. Beside thermal fatigue loads with power density levels up to 20 MWm
-2
, off-normal events are a serious concern for the lifetime of plasma facing components. These phenomena are expected to occur on a time scale of a few milliseconds (plasma disruptions) or several hundred milliseconds (vertical displacement events) and have been identified as a major source for the production of neutron activated metallic or tritium enriched carbon dust which is of serious importance from a safety point of view.
The irradiation induced material degradation is another critical concern for future D-T-burning fusion devices. In ITER the integrated neutron fluence to the first wall and the divertor armour will remain in the order of 1 dpa and 0.7 dpa, respectively. This value is low compared to future commercial fusion reactors; nevertheless, a nonnegligible degradation of the materials has been detected, both for mechanical and thermal properties, in particular for the thermal conductivity of carbon based materials. Beside the degradation of individual material properties, the high heat flux performance of actively cooled plasma facing components has been investigated under ITER specific thermal and neutron loads.
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