R Duwe

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).

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.

Mechanical properties of pure tantalum after 800 MeV proton irradiation

Abstract Specimens prepared from a spent tantalum target of the pulsed spallation source ISIS, irradiated with 800 MeV proton to a maximum fluence of 1.7×10 25 p m −2 at temperatures lower than 200 °C, were investigated by micro-hardness, three-point bending and tensile tests at room temperature (RT) and at 250 °C. All three types of mechanical measurements consistently showed irradiation hardening. Furthermore, the tensile tests showed that the increase in yield strength is accompanied by a reduction of the strain-to-necking at both test temperatures. For instance, at RT, the strain-to-necking was reduced by irradiation from initially 30% to about 10%. This drop in ductility occurred at doses below 0.6 dpa, whereas afterwards the strain-to-necking remained constant up to the maximum dose of about 11 dpa. The results of tests at 250 °C were similar to those at RT. SEM investigation revealed typical ductile fracture surfaces even for the highest doses. Optical micrography after bending tests showed no cracks even after bending of a 2 mm thick specimen end to end, indicating that pure Ta retained very high ductility after proton irradiation.

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.

Material degradation and particle formation under transient thermal loads

Abstract Carbon-based materials and metals have been exposed to fusion relevant thermal loads in an electron beam test facility to simulate off-normal plasma conditions such as disruptions or vertical displacement events (VDEs). The erosion process in carbon-based materials is dominated by brittle destruction, a process which is associated with the formation of carbon dust; this process becomes essential at a threshold value of approx. 200 MW m −2 . In metals the dominating processes are melting, crack formation in the recrystallized material, and – at higher thermal loads – splashing and boiling of the melt layer. Additional material degradation due to neutron irradiation (up to 0.35 dpa at 350°C and 700°C) and its influence on the high heat flux performance have been investigated.

Manufacture and Testing of Small-Scale Moch-ups for the ITER Divertor

This task within the EU R&D for ITER is primarily aimed at the development of solutions for the divertor target which has to be designed for up to 1000 off-normal transients at 20 MW/m2. Representative small scale mock-ups with carbon and tungsten armour have been manufactured using a similar technology. First tests support the analysis which indicates the best high heat flux capability for carbon mono-blocks.

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

Abstract 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.

High heat flux components with Be armour before and after neutron irradiation

Abstract Beryllium/copper mock-ups produced by different joining techniques have been tested in the electron beam facility JUDITH (Julich Di vertor Ṯest Facility in Hot Cells) at Forschungszentrum Julich. The experiments described in this paper represent the conclusive part of a test program started in 1994. The properties of non-irradiated Be/Cu joints have been characterised in a previous test campaign. Post-irradiation tests are now being carried out to investigate the neutron damage on the joints. The neutron irradiation on selected mock-ups has been carried out in the High Flux Reactor (HFR) at Petten (The Netherlands). Parametric finite element thermal analyses have been carried out to establish the allowable heat flux value to be applied during the tests. Screening tests up to power densities of ∼7 MW/m2 and thermal fatigue tests up to 1000 cycles have been performed. None of these mock-ups showed any indication of failure. Post-mortem analyses (metallography, SEM) have also been conducted.

Heat Shock Tests on Beryllium Samples before and after Neutron Irradiation

Abstract Beside carbon materials and tungsten, beryllium will play an important role as plasma facing material (PFM) in the International Thermonuclear Experimental Reactor (ITER). It will mainly be used for the primary wall, the limiter and the upper baffle. During off normal operation the surface of Be may be loaded by severe thermal shocks, caused by plasma disruptions with energies of several ten MJ/m2 within tens of milliseconds. The influence of high heat fluxes on several un-irradiated Be grade have been investigated before. During the operation of ITER the material will suffer irradiation with 14 MeV neutrons generated in the fusion process. In order to study the material degradation caused by fast neutrons, different samples have been neutron irradiated in the High Flux Reactor (HFR) at Petten. The thermal shock behaviour of the different beryllium grade before and after neutron irradiation is now compared.

HEAT FLUX EXPERIMENTS ON HEAT PIPES FOR PLASMA FACING APPLICATIONS

The heat removal from the leading edge of limiter blades is a critical issue for the technical feasibility of the pump limiter concept. The aim of the present work was to investigate the capability of heat pipes to remove concentrated local heat fluxes. Tubular and flat heat pipes were subjected to local surface heat loads in the JUDITH electron beam facility. The heat pipes were tested until failure or until the operational limit of the component was reached. The absorbed heat fluxes at this point were of the order of several hundred W/cm 2 .