The design of actively cooled plasma-facing components

In future fusion devices, like in the stellarator Wendelstein 7-X. the target plates of the divertor will be exposed to heat loads up to power densities of 10 MW/m(2) for 1000 s. For this purpose actively cooled target elements with an internal coolant flow return, made of 2-D CFC armor tiles brazed onto a two tube cooling structure were developed and manufactured at the Forschungszentrum Julich. Individual bent- and coolant flow reversal elements were used to achieve a high flexibility in the shape of the target elements. A special brazing technology, using a thin layer of plasma-are deposited titanium was used for the bonding of the cooling structure to the plasma facing armor (PFA). FEM-simulations of the thermal and mechanical behavior show that a detachment of about 25% of the bonded area between the copper tubes and the PFA can be tolerated, without exceeding the critical heat flux at 15 MW/m(2) or a surface temperature of 1400 degreesC at 10 MW/m(2) by using twisted tape inserts with a twist ratio of 2 at a cooling water velocity of 10 m/s. Thermal cycling tests in an electron beam facility up to a power density level 10.5 MW/m(2) show a very good behavior of parts of the target elements, which confirms the performance under fusion relevant conditions. Even defected parts in the bonding interface of the target elements, known from ultrasonic inspections before, show no change in the thermal performance under cycling, which confirms also the structural integrity of partly defected regions

Allowable heat load on the edge of the ITER first wall panel beryllium flat tiles

Plasma facing components are usually qualified to a given heat load density applied at the top face of the armour tiles with normal incidence angle. When employed in tokamak fusion machines, heat loading on the tile sides is possible due to optimised shaping, that doesn't provide edge shadowing for all design situations. An edge heat load may occur both at the tile and component scales. The edge load needs to be controlled and quantified. The adequate control of edge heat loads is especially critical for water cooled components that uses armour tiles which are bonded to the heat sink, for ensuring the long-term integrity of the tile bonding. An edge heat load allowance criterion of 10% of the top heat load is proposed. The 10% criterion is supported by experimental heat flux tests

High heat flux testing of first wall mock-ups with and without neutron irradiation

Beryllium as the plasma facing material for the first wall of ITER will be exposed to thermal, particle and neutron loads. In the frame of the European qualification program for ITER, two HIPped beryllium small scale flat-tile mock-ups consisting of a steel support structure, a CuCrZr/Cu heat sink and two beryllium tiles on top were manufactured by CEA. One mock-up was exposed to neutron irradiation up to 0.75dpa in beryllium in the RBT-6 fission reactor at Dimitrovgrad, Russia, while the other one was kept as reference. Furthermore, an identical mock-up was produced in Russia by manufacturing via electron beam induced rapid brazing and also exposed to the same neutron irradiation conditions. For qualification, all three flat-tile mock-ups were exposed to cyclic steady state heat loads in the electron beam facility JUDITH-1 up to a maximum of 3.0MW/m2. Thereby, each tile was loaded individually as the full loading area exceeds the limits of the facility

High heat flux performance of neutron irradiated plasma facing components

During the operation of ITER, the plasma facing components will undergo different kinds of thermal loadings like thermal fatigue, vertical displacement events and disruptions. In addition degradation effects due to neutron irradiations may play an important role. The electron beam facilities JUDITH and OHBIS have been designed to carry out ITER relevant simulation experiments on neutron irradiated materials and components. Carbon fiber reinforced carbon materials (CFC), Be and W alloys have been tested in thermal shock experiments. Thermal fatigue experiments have been performed with joints of these materials to Cu alloys. In thermal fatigue experiments no influence on the quality of the joints was observed whatever the testing facilities or materials combinations. But for CFC mock-LIPS the surface temperature is significantly increased due to the reduction in thermal conductivity. During experiments at high power densities annealing effects could be observed. Thermal shock tests show a higher erosion after neutron irradiation. The tests described above are not able to simulate the superposition of nuclear and thermal loads. In order to study the synergistic effects, in-pile thermal fatigue experiments with two CFC/Cu mock-ups and a Be-Cu mock-up have been performed in the SM-2 fission reactor in Dimitrovgrad (Russia). A first evaluation showed good performance of all three mock-ups. (C) 2002 Elsevier Science B.V. All rights reserved