J.G. van der Laan

Ceramic breeder research and development: progress and focus

Abstract The world-wide efforts on ceramic breeder materials in the last two years concerned Li 2 O, Li 4 SiO 4 , Li 2 TiO 3 and Li 2 ZrO 3 , with a clear emphasis on the development of Li 2 TiO 3 . Pebble-manufacturing processes have been developed up to a 10 kg scale. Characterisation of materials has advanced. A jump-wise progress is observed in the characterisation of pebble-beds, in particular of their thermo-mechanical behaviour. Thermal property data are still limited. A number of breeder materials have been or are being irradiated in material test reactors like HFR and JMTR. The EXOTIC-8 series of in-pile experiments is a major source of tritium release data. This paper discusses the technical advancements and proposes a focus for further research and development (RD tritium release and retention properties; determination of the key factors limiting blanket life.

Simulation of disruptions on coatings and bulk materials

An important issue for the next step thermonuclear fusion devices is the development of a reliable engineering solution for the first wall and in particular for the divertor. Besides severe mechanical loads, these plasma facing components (PFC) will be subjected to energetic pulses during plasma disruptions. To evaluate the performance of plasma facing materials and thus to predict the lifetime of the PFCs, simulation experiments have been performed in the JAERI electron beam irradiation stand (JEBIS). Here different candidate plasma facing materials were evaluated in short beam pulses (1.2–10 ms) with an energy deposition of 2–9 MJ m −2 . The response of the individual materials to single and multiple shot beam pulses was investigated, and in these analyses special attention was given to the material erosion (melting, sublimation, particle emission). The quantification of these effects was done by weight loss measurements and by optical profilometry on the ablation craters. In addition, microstructural and morphological changes in the loaded surface were investigated.

Effects of pulsed-laser radiation on first-wall materials

Abstract The first wall in NET is made of alloys which melt locally, when hit by plasma disruptions or run away electrons. Therefore several ceramic and carbon materials are candidates as protection coatings or tiles. Results of disruption simulation with a multi-mode Nd: YAG pulsed laser are reported. Pulse durations are in the range of 0.2–20 ms and energy densities are up to 10 MJ/m 2 . These quantities are in line with predictions for NET disruptions. Austenitic stainless steel AISI 316, a graphite material and titanium carbide, materials relevant to the NET design, are studied. The phenomena due to the laser energy deposition are described and quantified in terms of melting depth and erosion loss. These quantities are related to predictions obtained by numerical calculations.

Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

Abstract An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating parameters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in thermal quench phase.

Effects of short pulse high heat fluxes on carbon base plasma facing materials for ITER

New experimental results from disruption heat flux simulations in the millisecond range with laser and electron beams are discussed. For a number of graphites, boronated graphites and carbon fiber composites, the effective enthalpy of ablation is determined as 30±3 MJ/kg, using laser pulses of about 0.3 ms. No effect of boron doping on the ablation enthalpy is found. A simple parametric study of vapour shielding is presented, with shield transmission factors from 5 to 100% and compared with recently published results from pulsed plasma devices in Russia. Estimates for the disruption erosion lifetime of plasma facing components in ITER are briefly discussed.

EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup

Abstract The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li 2 ZrO 3 , LiAlO 2 and Li 8 ZrO 6 and pebbles of Li 4 SiO 4 and Li 2 ZrO 3 , with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li 4 SiO 4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented.

Plasma-sprayed boron carbide coatings for first-wall protection

Abstract Plasma-sprayed boron carbide coatings have been manufactured by different suppliers onto substrates of type 316L stainless steel. The coating thickness ranges from 0.3 to 2.0 mm. The larger thicknesses could only be achieved by application of an adaptive or gradient bond-layer between substrate and the boron carbide top coating. Measurements of the thermal diffusivity of coating materials are reported. Several high heat flux facilities have been used to study the thermal shock and erosion behaviour of the coated samples. A supporting numerical analysis of the thermal behaviour of the coating under normal and off-normal heat loads is presented, focussing on the differences between electron beam and laser beam tests due to volumetric energy deposition. Some aspects of the applicability of plasma sprayed B 4 C coatings for first-wall protection in a next step device are discussed.

Properties of lithium metatitanate pebbles produced by a wet process

Lithium metatitanate (Li2TiO3) is considered as one of the candidate materials for the ceramic breeder in both the ITER Breeding Blanket and the European Helium Cooled Pebble-Bed Blanket for DEMO. A wet process based on powder-gelation for the manufacture of Li2TiO3-pebbles is described. Pebble characteristics and its basic properties are given along with results of out-of-pile tritium release experiments.

Simulation and analysis of the response of carbon materials to off-normal heat loads accompanying plasma disruptions

Abstract The effects of off-normal heat loads accompanying plasmadisruptions have been studied by both experimental and numerical simulations. Experiments have been performed on a range of polygranular graphites, pyrolytic graphite and carbon composite material. The measured erosion is compared with numerical predictions by a transient heat load code taking a revised value for the vaporization enthalpy of carbon. The effect of variations in thermo-physical material parameters on thermal erosion behaviour is discussed.

In-pile performance of a double-walled tube and a tritium permeation barrier

Abstract In two recent irradiation experiments in the HFR Petten, tritium permeation rates through representative materials to be used as cooling tubes of the water-cooled lithium-lead blanket have been measured in-pile. These latest experiments in the EXOTIC 8 series (E 8.9 and E 8.10) are made of a double wall tube (DWT) and a T91 tube with an Fe–Al/Al 2 O 3 layer acting as tritium permeation barrier (TPB). These tubes contain annular pebble beds of ceramic breeder materials for the helium-cooled pebble bed concept blanket as tritium breeding material. Both experiments are built up of two concentric and independently purged containments allowing on-line tritium release rate and permeation rate measurements. In-pile operation has ended in March 2001 after 450 full power days and resulted in an irradiation damage of approximately 2.6 and 3.2 dpa, respectively in T91 steel. This paper reports on the experimental results obtained for in-pile tritium permeation and discusses the influence of purge gas compositions, temperature and irradiation on tritium permeation through the DWT and TPB.