A.E. Pontau

Ion microbeam tomography

Abstract Proton beams with energies of 5 and 7 MeV are focused to 5 μm and used to produce tomograms of capillary tubes and low-density foams. In this energy range, proton energy loss is primarily due to interactions with electrons. Therefore, by measuring the residual energy of protons transmitted through samples in a manner similar to that used for Scanning Transmission Ion Microscopy (STIM), and reconstructing a cross-sectional image from multiple projections, we can map out spatial variations in electron density due to sample geometry and composition. In our experimental arrangement, the sample is translated and rotated in a stationary proton beam. Transmitted proton energies are measured using a silicon surface barrier detector. Tomographie reconstructions are produced from the calculated line-average densities using a procedure based on a filtered backprojection algorithm developed for X-ray computed tomography (CT) systems. The technique is especially useful in characterizing samples where large variations in Z or low total density limit the applicability of X-ray CT analysis.

Deuterium trapping and release in titanium-based coatings for TFTR☆

Abstract We have measured the deuterium trapping and thermal release characteristics of three candidate armor and limiter materials for TFTR: Ti explosively bonded onto Cu; TiB 2 and TiC chemically vapor deposited on C. The re-emission rate of deuterium was monitored during bombardment with 10 keV D + 3 and the post-implantation retention was measured with D( 3 He, α)H nuclear reaction profiling and linear ramp thermal desorption. Titanium claddings showed no thermal release of deuterium for fluences of 10 19 D/cm 2 at 375 K; at 775 K, only 65% of the fluence was released during implantation to 10 18 D/cm 2 . This high retention is shown to result from titanium deutende precipitation and the formation of a bulk deuterium solid solution. In contrast, the TiB 2 and TiC coatings showed a rapid saturation in deuterium retention of 17 D/cm 2 .

The temperature dependence of deuterium trapping in fusion reactor materials

Abstract Measurements were made on the deuterium retention characteristics of various metal alloys implanted with deuterium ions to a standard fluence of 10 22 D / m 2 at 3.3 keV/atom. Gas re-emission, post-implant thermal desorption, and D( 3 He, α) 1 H nuclear reaction techniques were used to assess the temperature dependence of deuterium trapping in four candidate first wall alloys. For 304 Stainless Steel, Inconel 625, and TZM, less than ten percent of the implanted fluence was retained after implantation at 375 K. In contrast, Ti-6Al-4V had ≳90 percent of the implanted fluence retained at implant temperatures of ~675 K, and only limited release at higher temperature. The high trapping in this latter alloy may obviate its use as a first wall material in direct contact with the plasma.

Erosion and redeposition experiments in the PISCES facility

The modification of surfaces during exposure to plasma bombardment is a critical issue in the development of limiter and wall materials for fusion confinement experiments. Controlled studies of the erosion and redeposition of materials during high flux and fluence plasma exposure are now possible in the PISCES facility. PISCES is a continuously operating plasma device which has achieved hydrogen plasma densities of over 1013 cm−3 and electron temperatures of 5 to 24 eV over large areas. Ion fluxes of 1017 to 1019 cm−2 s−1 and fluences of up to 1023 cm−2 have been used to bombard biased samples inserted into th plasma. The plasma parameters can be selected to produce simple sputtering, or redeposition by the ionization and recycling of the sputtered target materials. Collaborative studies on the performance of Cu and Cu-Li alloys (with ANL), stainless steel (with SNLL), and graphite (with IPP at Garching, and SNLL) have been undertaken. Surface topography modification is always observed after a sufficient fluence is achieved. The net erosion rate is significantly lower during redeposition than one would expect from classical sputtering yields. The transport and deposition of different materials by the plasma to the samples during redeposition conditions results in greatly modified surface composition and morphology. Chemical sputtering of graphite during low energy, high flux (>1018 cm−2 s−1) plasma bombardment is observed. Chemically formed hydrocarbons are relatively easily redeposited compared to sputtered carbon. The performance of these materials, the surface morphology evolution, and the characteristics of the redeposited materials are discussed.

Characterization of deposition and erosion of the TFTR bumper limiter and wall

Abstract To understand material transport by the plasma in the TFTR tokamak, graphite bumper limiter tiles and metal surfaces have been studied. Detailed measurements of the TFTR inner bumper limiter POCO ™ AXF-5Q graphite tiles indicate areas of net erosion and areas of net deposition. These areas are poloidally asymmetric and on the scale of a bay (l/20th of the torus) repeat regularly toroidally. Finer scale measurements indicate that there are subtle variations in the interaction of the plasma with the wall. Furthermore, a study of the correlation of limiter deposits with the type of discharges in TFTR indicates that the material composition depth distribution was determined by the tokamak operational history. In particular, the relative amounts of carbon, hydrogen, oxygen, and metal in the deposits changed over time, reflecting plasma impurity levels. The outer wall of the vessel was not exposed to direct plasma flux and does not show evidence of erosion.

Impurity deposition on ALT-I limiter heads☆

Abstract Detailed spatial profiles of impurities deposited on the surface of the ALT-I pumped limiter heads used in TEXTOR have been measured using proton induced X-ray emission in the external ion beam analysis system. The TiC coated, ATJ graphite ALT-I limiter head was analyzed after each of two plasma exposure periods. Measurements were also made on the uncoated ATJ limiter head. For limiter operation at plasma radii ranging from 40 to 46 cm we find large accumulations of metal impurities on the ion and electron sides of the limiter. The composition implies that the source of these impurities is predominantly the stainless steel main limiters in TEXTOR. Surface concentrations of iron as high as 10 18 /cm 2 have been measured. On each of the limiter front faces a relative maximum in metal concentration occurs near the tangency point and relative minima are found near the toroidal extremities. An analysis of impurity transport in the edge region of TEXTOR using Redep is also presented. The calculations predict impurity atom migration patterns that strongly resemble the measured metal atom deposition profiles. The analysis provides considerable insight into impurity transport in the tokamak edge. Benchmark calculations of this type are valuable for the design of future limiter or divertor systems.

STIM tomography: a three-dimensional high resolution imaging tool

Abstract The ion microprobes now found in many accelerator laboratories were developed to perform quantitative elemental microanalysis with high sensitivity. The rapid evolution on this instrument of new computer based techniques has led to the development of high resolution quantitative 3D ion microtomography. This technique offers unique opportunities to examine internal structure of microscopic specimens.

Some practical considerations for ion microtomography

Abstract Spatial broadening and energy straggling of the analysis beam can dictate the choice of ion species and required fluence for given spatial and density resolution in ion microtomography (IMT). For extended objects, fine spatial resolution also implies huge data sets. Data acquisition, reduction, and reconstruction rates must be maximized to achieve optimal materials analysis times. In this paper, we discuss these topics and current efforts in this field, present examples of IMT, and consider future directions.

Material behavior and materials problems in TFTR

Abstract This paper reviews the experience with first-wall materials in TFTR over a 20 month period of operatiori during 1985–1987. Experience with the axisymmetric inner wall limiter, constructed of graphite tiles, is described, including the necessary conditioning procedures needed for impurity and particle control of high power (≤ 20 MW) neutral injection experiments. The thermal effects in disruptions have been quantified and no significant damage to the bumper limiter has occurred as a result of disruptions. Carbon and metal impurity redeposition effects have been quantified through surface analysis of wall samples. Estimates of the tritium retention in the graphite limiter tiles and redeposited carbon films have been made, based on analysis of deuterium retention in removed graphite tiles and wall samples. New limiter structures have been designed using a 2D carbon/carbon ( C C ) composite material for RF antenna protection. Laboratory tests of the important thermal, mechanical, and vacuum properties of C C materials are described. Finally, the last series of experiments in TFTR with in-situ Zr Al surface pumps are discussed. Problems with Zr Al embrittlement have led to the removal of the getter material from the in-torus environment.

Hydrogen and deuterium in graphite limiters from TFTR, PDX and PLT☆

Abstract Retention of hydrogen isotopes in graphite limiters impacts plasma fueling, isotope changeover rates and tritium inventory in fusion reactors. This paper describes measurements of the amount of deuterium (D) in graphite tiles from the TFTR limiter, and in graphite limiters from PDX and PLT, after extended exposure to high power D plasmas. D concentrations decreased abruptly over the first micron and fell to the bulk value of 4 to 13 atomic ppm beyond 40 microns from the exposed surface. The average near-surface areal densities of D on the TFTR tiles was 10 17 D/cm 2 . The total amount of near-surface D in TFTR limiters estimated using this value is larger than the amount of D in a plasma but small compared to the allowed tritium inventory. The D concentrations found in the bulk are large enough to be an important factor for tritium inventory. The 1 at.% of hydrogen (H) found in the limiters and in graphite not exposed to plasma may be an important source of hydrogen to the plasma. We discuss the influence of the H and D concentrations found in the bulk of graphite on tritium inventory estimates and hydrogen isotope changeover rates.