G.R. Smolik

Oxidation and volatilization of TZM alloy in air

The excellent high temperature strength and thermal conductivity of molybdenum-base alloys provide attractive features for components in advanced magnetic and inertial fusion devices. Refractory metal base alloys react readily with oxygen and other gases, and molybdenum alloys are susceptible to losses from highly volatile molybdenum trioxide (MoO{sub 3}) species. Transport of radioactivity by the volatilization, migration, and re-deposition of MoO{sub 3} during a potential accident involving a loss of vacuum or inert environment represents a safety issue. The authors have experimentally measured the oxidation, volatilization and re-deposition of molybdenum from TZM in flowing air between 400 and 800 C. Calculations using chemical thermodynamic data for vapor pressures over pure MoO{sub 3} and a vaporization mass transfer model correlate well with experimental data between 600 and 800 C. Partial saturation of MoO{sub 3} gas species account for influences of flow rate at 700 C. Some anomalies in oxidation rate below 650 C, suggesting that other phases, e.g., MoO{sub 2} or other non-stoichiometric oxides may influence oxidation and volatilization processes under some limited conditions.

The safety implications of tokamak dust size and surface area

Tokamak dust, accumulated primarily from sputtering and disruptions, has important safety issues associated with it. The dust may contain tritium, it may be activated, and may be chemically toxic, and chemically reactive. The size of the dust and the surface area are important parameters in determining potential hydrogen production and transport of activated/chemically toxic dust. In this paper we examine the effect of the size distribution of the dust on these safety issues.

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.

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.

Steam-Chemical Reactivity Studies for Irradiated Beryllium

This paper reports the results of an experimental study to determine the influence of neutron irradiation effects on the chemical reactivity of beryllium exposed to steam. The study entailed measurements of the following: (1) swelling of irradiated Be specimens annealed at temperatures ranging from 450°C to 1200°C, (2) hydrogen generation rates for unirradiated Be control specimens exposed to steam at temperatures from 450°C to 1200°C, and (3) hydrogen generation rates and tritium mobilization rates for irradiated Be exposed to steam at temperatures from 450°C to 700°C. For irradiated Be, swelling occurred at temperatures above 600°C and it increased to about 56% for an anneal temperature of 1200°C. Tritium and 4He were released concurrently from specimens that were annealed at 800°C and above. Steam-Be reactivity measurements for the control specimens were consistent with previous work at temperatures above 700°C, and the new measurements extended the reactivity database down to 450°C. Steam-reactivity measurements for irradiated Be were comparable to control specimens for 600°C and below, but, they indicated a significant enhancement in the chemical reactivity at 700°C.

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.

Steam Chemical Reactivity of Be Pebbles and Be Powder

Abstract This paper reports the results of chemical reactivity experiments for Be pebbles (2-mm and 0.2-mm diameter) and Be powder (14–31 μm diameter) exposed to steam at elevated temperatures, 350 to 900°C for pebbles and 400 to 500°C for powders. We measured BET specific surface areas of 0.12 m2/g for 2-mm pebbles, 0.24 m2/g for 0.2-mm pebbles and 0.66 to 1.21 m2/g for Be powder samples. These experiments showed a complex reactivity behavior for the material, dependent primarily on the test temperature. Average H2 generation rates for powder samples, based on measured BET surface areas, were in good agreement with previous measurements for fully-dense CPM-Be. Rates for the Be pebbles, based on measured BET surface areas, were systematically lower than the CPM-Be rates, possibly because of different surface and bulk features for the pebbles, especially surface-layer impurities, that contribute to the measured BET surface area and influence the oxidation process at the material surface.

Chemical reactivity and mobilization of beryllium exposed to steam

Abstract Beryllium is used in many fusion reactor designs as either an armor for plasma facing surfaces, or as a neutron multiplier in the blanket. Beryllium used in a water-cooled design poses important safety issues related to the chemical reactivity of beryllium in steam and its toxicity. The Fusion Safety Program at the Idaho National Engineering and Environmental Laboratory has been investigating experimentally the chemical reactivity and mobilization of various forms of beryllium for the past 6 years. In this paper we present a summary of this work, including results from fully dense (irradiated and non-irradiated), plasma-sprayed, and 88% dense beryllium. Assembling this data helps us to assess where further testing is needed. Our data help guide designs such that accident temperatures stay below values necessary to ensure beryllium release limits and hydrogen generation limits are met.

FLIBE-D2 permeation experiment and analysis

Abstract Experiment of D2 permeation through Ni facing with purified Flibe is being carried out under the Japan-US joint research project (JUPITER-II). The experiment is proceeding in the following phases; (i) fabrication and assembly of a dual-probe permeation apparatus, (ii) a single-probe Ni/D2 permeation experiment without Flibe, (iii) a dual-probe Ni/D2 permeation experiment without Flibe, (iv) Flibe chemical purification by HF/H2 gas bubbling, (v) physical purification by Flibe transport through a porous Ni filter, (vi) Ni/Flibe/D2 permeation experiment, and (vii) Ni/Flibe/HT permeation experiment. The present paper describes results of the single and dual Ni/D2 permeation experiments in detail.