R.A. Anderl

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Steam-chemical reactivity for irradiated beryllium

This paper reports experimental results concerning the influence of neutron irradiation effects and annealing on the chemical reactivity of beryllium exposed to steam. The work entailed: (1) measurements of swelling, porosity and specific surface area for irradiated Be annealed at temperatures ranging from 700°C to 1200°C and (2) measurements of hydrogen generation rates for unirradiated Be, irradiated Be and irradiated-annealed Be exposed to steam at elevated temperatures. For irradiated Be, volumetric swelling increased from 14% at a 700°C anneal to about 56% at a 1200°C anneal. Gas-release measurements during annealing indicated the development of a surface-connected porosity network. Specific surface areas for irradiated-annealed Be increased with the anneal temperature. Steam-chemical reactivity was similar for irradiated and unirradiated Be at temperatures between 450°C and 600°C. For irradiated Be exposed to steam at 700°C, the reactivity accelerated rapidly and the specimen experienced a temperature excursion. Irradiated-annealed Be showed enhanced chemical reactivity related to its higher specific surface area.

Characterization and analysis of dusts produced in three experimental tokamaks : TFTR, DIII-D, and Alcator C-Mod

Abstract Particulate, referred to as ‘dust’, produced during operation of tokamak systems can be a large source of activated material in a D-T fusion machine. Particulate less than 10 μm in size is easily mobilized both during an accident as well as during routine maintenance activities. We have collected and analyzed dust samples from three tokamaks to determine the potential contribution to accident source terms in future fusion power plants. We have obtained dust samples from Princeton Plasma Physics Laboratorys TFTR (prior to the final run period in 1997), from MITs Alcator C-Mod (during March and April of 1998), and from General Atomics’ DIII-D (August of 1998). This paper presents the results of our analyses for particle size distribution, specific surface area, and dust composition, including the tritium content of TFTR dust.

Hydrogen transport behavior of beryllium

Beryllium is being evaluated for use as a plasma-facing material in the International Thermonuclear Experimental Reactor (ITER). One concern in the evaluation is the retention and permeation of tritium implanted into the plasma-facing surface. We performed laboratory-scale studies to investigate mechanisms that influence hydrogen transport and retention in beryllium foil specimens of rolled powder metallurgy product and rolled ingot cast beryllium. Specimen characterization was accomplished using scanning electron microscopy, Auger electron spectroscopy, and Rutherford backscattering spectrometry (RBS) techniques. Hydrogen transport was investigated using ion-beam permeation experiments and nuclear reaction analysis (NRA). Results indicate that trapping plays a significant role in permeation, re-emission, and retention, and that surface processes at both upstream and downstream surfaces are also important.

On the mechanisms associated with the chemical reactivity of Be in steam

One safety concern surrounding beryllium as a plasma-facing material in a water-cooled Tokamak is steam interactions with hot beryllium and the production of significant quantities of hydrogen. We have tested several different product forms of Be with different densities and levels of porosity. Oxidation kinetics has been determined by measurements of hydrogen release with a mass spectrometer, volumetric measurements of the product gas and weight change of the sample. We discuss and compare with the literature the fundamental mechanisms and kinetics controlling the oxidation of Be in steam. Fully dense beryllium exhibits parabolic, linear and accelerating modes of oxidation as temperature increases from 450°C to 1200°C. The oxidation mechanisms and temperature trends are similar for other product forms. Oxidation rates are higher, however, when processing or annealing significantly increases the extent of interconnected porosity and consequently the effective surface area. The effective surface area as measured by BET surface analyses is a key parameter when comparing oxidation rates.

Hydrogen permeation properties of plasma-sprayed tungsten☆

Abstract Tungsten has been proposed as a plasma-facing component material for advanced fusion facilities. This paper reports on laboratory-scale studies that were done to assess the hydrogen permeation properties of plasma-sprayed tungsten for such applications. The work entailed deuterium permeation measurements for plasma-sprayed (PS) tungsten coatings, sputter-deposited (SP) tungsten coatings, and steel substrate material using a mass-analyzed, 3 keV D3+ ion beam with fluxes of ∼6.5 × 1019 D/m2 s. Extensive characterization analyses for the plasma-sprayed tungsten coatings were made using Auger spectrometry and scanning electron microscopy (SEM). Observed permeation rates through composite PS-tungsten/steel specimens were several orders of magnitude below the permeation levels observed for SP-tungsten/steel composite specimens and pure steel specimens. Characterization analyses indicated that the plasma-sprayed tungsten coating had a nonhomogeneous microstructure that consisted of splats with columnar solidification, partially-melted particles with grain boundaries, and void regions. Reduced permeation levels can be attributed to the complex microstructure and a substantial surface-connected porosity.

Deuterium retention in W, W1%La, C-coated W and W2C ☆

This paper reports the results of a systematic investigation into retention of deuterium implanted into various forms of W including: reduction-rolled, powder-metallurgy foil; discs of W1%La alloy; W and W 2 C prepared by chemical vapor deposition (CVD) and annealed C-coated W discs. Deuterium was implanted at energies of 0.5 keV/D with fluxes of ∼3 × 10 19 D/m 2 s and fluences of 3 x 10 23 D/m 2 , for samples at temperatures from 23°C to 400°C. Retained deuterium quantities were measured using thermal desorption spectroscopy (TDS). Retention in annealed CVD-W and W1%La, for implantation temperatures less than 200°C, is below that in annealed W foil, indicating that trapping may be affected by the different material defect structures. For implantation temperatures less than 300°C, retention in CVD-W 2 C is somewhat higher than that in CVD-W, indicating trapping could be enhanced by trace carbon impurities, differences in W 2 C material structure and recoil carbon-induced material damage. Implantation into C-coated W resulted in orders of magnitude more retention than in uncoated material, because of trapping in the carbon coating.

Tritium saturation in plasma-facing materials surfaces

Plasma-facing components in the International Thermonuclear Experimental Reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10 20 -10 23 particles/m 2 s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments.

Initial studies of tritium behavior in flibe and flibe-facing material

Abstract Flibe–tritium experiment in the Japan–US joint project (JUPITER-II) was initiated in 2001. H/D isotopic exchange experiments were conducted to select a Flibe-facing material. Because of hydrogen isotope interactions with carbon, Ni crucibles were selected for Flibe/tritium behavior experiments. A Flibe–tritium pot with two Ni (or Cu) permeable probes was designed. The rate of the overall tritium permeation through the Flibe-facing Ni or Cu was estimated by numerical simulation using TMAP4 code. Diffusion in bulk Flibe was found to be the rate-determining step for purified Flibe.

The Safety and Tritium Applied Research (STAR) Facility: Status-2004

The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems.