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The adoption of cover crop production systems is lagging in the mid-southern USA due to concerns over yield stability and on-farm profitability. This research was conducted to determine if the inclusion of a cover crop in conservation tillage systems improves yield, profitability, and water use efficiency. The effects of replacing subsoiling with a cereal rye (Secale cereale L.) or tillage radish (Raphanus sativus L. var. longipinnatus) cover crop on soybean [Glycine max (L.) Merr.] grain yield, net returns above specified costs, and water use efficiency were evaluated in a conservation tillage system, i.e., surface residue = 30% at planting, on a Dubbs silt loam (Fine-silty, mixed, active, thermic Typic Hapludalfs) from 2016 to 2018 near Stoneville, MS. Relative to the conservation tillage system with subsoiling, the replacement of subsoiling with a tillage radish cover crop reduced soybean grain yield, net returns above specified costs, and water use efficiency by up to 41% (P = 0.0266). Conversely, the replacement of subsoiling with a cereal rye cover crop had no effect on soybean grain yield or water use efficiency but reduced net returns above specified costs by 28% (P = 0.0266). In the mid-southern USA, a cereal rye cover crop can maintain soybean grain yield and water use efficiency relative to the regional standard, but widespread adoption of this production system is unlikely due to reduced profitability associated with additional seed and planting costs.

In this work, we examine the response of an ultra-fine grained (UFG) tungsten material to high-flux deuterium plasma exposure. UFG tungsten has received considerable interest as a possible plasma-facing material in magnetic confinement fusion devices, in large part because of its improved resistance to neutron damage. However, optimization of the material in this manner may lead to trade-offs in other properties. We address two aspects of the problem in this work: (a) how high-flux plasmas modify the structure of the exposed surface, and (b) how hydrogen isotopes become trapped within the material. The specific UFG tungsten considered here contains 100 nm-width Ti dispersoids (1 wt%) that limit the growth of the W grains to a median size of 960 nm. Metal impurities (Fe, Cr) as well as O were identified within the dispersoids; these species were absent from the W matrix. To simulate relevant particle bombardment conditions, we exposed specimens of the W-Ti material to low energy (100 eV), high-flux (> 1022 m− 2 s− 1) deuterium plasmas in the PISCES-A facility at the University of California, San Diego. To explore different temperature-dependent trapping mechanisms, we considered a range of exposure temperatures between 200 °C and 500 °C. For comparison, we also exposed reference specimens of conventional powder metallurgy warm-rolled and ITER-grade tungsten at 300 °C. Post-mortem focused ion beam profiling and atomic force microscopy of the UFG tungsten revealed no evidence of near-surface bubbles containing high pressure D2 gas, a common surface degradation mechanism associated with plasma exposure. Thermal desorption spectrometry indicated moderately higher trapping of D in the material compared with the reference specimens, though still within the spread of values for different tungsten grades found in the literature database. For the criteria considered here, these results do not indicate any significant obstacles to the potential use of UFG tungsten as a plasma-facing material, although further experimental work is needed to assess material response to transient events and high plasma fluence.

The tungsten ITER divertor will be operated at temperatures above 1000 K. Most of the laboratory experiments on hydrogen isotope retention in tungsten have been performed at lower temperatures where the hydrogen is retained as both atoms and molecules. At higher temperatures, atomic trapping plays a smaller role. The purpose of this paper is to see if hydrogen is trapped at internal voids at elevated temperatures, and to see if gas-filled cavities can be formed at high fluences. Additionally, this paper examines the effect of helium bubbles and radiation damage on trapping.

The Sustained Spheromak Physics Experiment (SSPX) will examine the confinement properties of spheromak plasmas sustained by DC helicity injection. Understanding the plasma-surface interactions is an important component of the experimental program since the spheromak plasma is in close contact with a stabilizing wall (flux conserver) and is maintained by a high current discharge in the coaxial injector region. Peak electron temperatures in the range of 400 eV are expected, so the copper plasma facing surfaces in SSPX have been coated with tungsten to minimize sputtering and plasma contamination. Here, we report on the characterization and conditioning of these surfaces used for the initial studies of spheromak formation in SSPX. The high pressure plasma-sprayed tungsten facing the SSPX plasma was characterized in situ using β-backscattering and ex situ using laboratory measurements on similarly prepared samples. Measurements showed that water can be desorbed effectively through baking while the removal rates of volatile impurity gases during glow discharge and shot conditioning indicated a large source of carbon and oxygen in the porous coating.

Since the introduction of new graphite tiles on the divertor regions and the center post of DIII-D, beta backscatter measurements have been made of the metals codeposited on these surfaces at three different times. At the first vent (June 1988), the upper and lower portions of the machine appeared very similar, with the average amount of metal on a surface increasing with increasing radius and indications that local tile alignment had an effect on the amount of metal deposited. Measurements made during subsequent vents show that there has been much more metal deposited on the lower divertor than on the upper. Metal seems to be concentrated at radii both less than and greater than that expected for the strike zones. From measurements made during the latest vent (December 1989) there is clear indication that material has been eroded from a position of the outer strike zone, probably to be redeposited at a larger radius.

Transient release immediately following implantation has been observed when large 5.1 cm diameter Be disks were implanted with deuterium ions at a flux of 7.4 × 10 16 D/cm 2 s. A series of experiments have been performed at temperatures ranging from 333 to 800 K. The release characteristics vary depending on whether the beryllium surface is clean or is covered with a thick oxide layer. When the surface is clean, the integrated release (D/cm 2 ) increases monotonically with temperature and then levels off at 700 K. This behavior cannot be described with a diffusion-limited release but is consistent with a recombination-limited release model. Furthermore, the release rate exhibits approximately a t −1 time dependence, in good agreement with the results calculated using the model. The recombination coefficient has to be sufficiently small such that the ratio of the diffusivity to the recombination coefficient D/K T ≽ 1 × 10 14 cm −2 at 600 K and must rise sharply with temperature. When the surface of the beryllium samples has a thick oxide layer present the release rate significantly changes and follows a t −0.3 dependence.

With the recent successful operation in JET, beryllium must be considered a leading plasma-facing material. However, little is known about beryllium's hydrogen isotope retention and migration properties. This paper presents the results of a coordinated experimental program performed to characterize the hydrogen isotope retention and release in beryllium. In the first set of experiments, measurements of the plasma-driven permeation of deuterium through beryllium membranes were conducted. No detectable permeation was obtained for temperatures below 670 K, and the higher temperature results were strongly affected by the oxide layers on the samples. In the second set of experiments, samples were exposed to a D-T plasma in the Tritium Plasma Experiment. The samples were subsequently removed for dissolution tritium counting. The retention was seen to initially decrease with increasing exposure temperature, but reached a local maximum at 770 K.

During 1989, following a brief all carbon phase, JET was operated with an evaporated beryllium film on all the plasma facing surfaces and then, in a third phase, with bulk beryllium used for the belt limiters and one set of antenna protection tiles. Subsequent analysis of wall components and long term samples (LTS) using Beta Backscatter and Ion Beam Analysis has been conducted to determine the condition of the First Wall and its deuterium inventory after beryllium operation. Ex-situ analysis of components where deposition occurs during operation shows mixed carbon and beryllium layers with an approximate ratio 1:1 and some localised higher Z components. The deuterium inventory situation is littled changed by the introduction of beryllium. Similar surface levels of deuterium, ~1 × 1022 atoms m−2, have been seen for both carbon and beryllium belt limiter tiles.

Graphite tiles from both the ALT-II and inner-bumper limiters were removed from TEXTOR and subjected to materials analysis. Scanning-electron microscopy and energy dispersive X-ray analysis were performed at the Institut fur Reaktorwerkstoffe, Forschungszentrum Julich. Deuterium profiles and metallic contamination were examined using external ion beam analysis at Sandia National Laboratory-Albuquerque. The erosion and hydrogen recycling of the tiles, while subjected to plasma bombardment, were studied at University of California, Los Angeles. In-situ analysis of the inner-bumper limiter tiles was performed by Sandia National Laboratory-Livermore using beta backscattering. Results indicate low metallic impurity concentration on the surfaces of both types of tiles. Increased metallic concentration coincides with regions of increased plasma flux to the surface. The ALT-II tiles exhibit a uniformly eroded surface. The inner-bumper limiter tiles show both eroded and redeposited regions, in agreement with power deposition measurements to the tiles in TEXTOR. The redeposited regions show enhanced erosion and recycling when exposed to controlled plasma bombardment.

The plasma–material interaction and high heat flux properties of beryllium are reviewed to determine its suitability as a plasma‐facing component in magnetic fusion energy reactors. Consideration is given to beryllium’s outgassing, erosion, and hydrogen retention characteristics. Its responses to normal and off‐normal high heat fluxes are compared to graphite in both the as‐received and the neutron‐irradiated states. Beryllium’s performance in present‐day devices is assessed, and its expected behavior in future reactors is summarized. It is concluded that beryllium is potentially a better plasma‐facing material than graphite and that more development and testing is warranted.

This study demonstrates the feasibility of using supercritical water oxidation to destroy ammonium picrate. Analyses of reactor effluent composition at various temperatures, residence times, and oxidant concentrations were used to design an improved reactor configuration for achieving destruction with minimum corrosion. The engineering evaluation reactor, a room-sized laboratory scale reactor, was reconfigured to incorporate this design change. Destruction of ammonium picrate with minimized corrosion was demonstrated on this reconfigured reactor. Factors that must be considered in scaling up to pilot plant size are discussed.

Carbon, nitrogen and oxygen K-shell correlation-state spectra give detailed qualitative confirmations of predictions by Basch. This prototypical study, with 1s holes on three different elements, shows that π-π * excitations are differentially stabilized according to the location of the 1s hole.

Mirrors with up to eight months exposure in heliostat modules in the field and with extensive corrosion have been studied to determine the nature of the corrosion. Auger electron spectroscopy of exposed surfaces displaying corrosion indicates that a copper/oxygen species is the first decomposition product. This product tends to form or precipitate in spots or rings similar to those which are eventually visible from the front surface. The corrosion may or may not be accompanied by small amounts of chlorine or sulfur. These two elements are superficial, however, and easily removed by sputtering. Further corrosion produces a corrosion product which includes silver, copper, oxygen and carbon. The corroded mirrors have been compared with an uncorroded mirror which was studied after near the glass-silver interface. Tin and chlorine remain in this region from processing; the tin with the glass and the chlorine with the metal layer.

The Tokamak Fusion Test Reactor (TFTR) is scheduled to begin DT operation in 1990 with the on-site tritium inventory limited to 5 grams. The physics and chemistry of the in-vessel tritium inventory will impact safety concerns, and also the entire operating schedule of the tokamak. We have investigated plasma-material interaction processes that will affect this first tritium-fueled tokamak. Tritium inventory estimates for TFTR are derived from: (1) laboratory simulation, (2) in-situ plasma measurements, (3) post-run surface analysis, and (4) modeling. This paper presents the results of these investigations, the derivation of a tritium inventory estimate and its uncertainties, and a discussion of its impact. A particular discharge-by-discharge operating schedule has been developed and evaluated. The major source of in-vessel tritium inventory will be codeposition of tritium and eroded carbon onto surfaces. We find that the on-site limit may be approached unless specific inventory reduction techniques are invoked, e.g., discharge cleaning.

The chemical erosion behavior of graphite materials preexposed in the Tokamak Fusion Test Reactor (TFTR) as the bumper limiter has been investigated spectroscopically under deuterium plasma bombardment in the Plasma Interactive Surface Component Experimental Station‐A (PISCES‐A) facility. The deuterium plasma bombardment conditions are ion bombarding energy of 300 eV, ion flux of 1.7×1018 ions s−1 cm−2, plasma density of 1.4×1012 cm−3, electron temperature of 11 eV, and neutral pressure of 3×10−4 Torr. The chemical erosion yield is measured with calibrated CD‐band spectroscopy during the temperature ramp from 100 to 900 °C at an average rate of ∼5 K/s. The materials used include virgin POCO graphite and graphite tile pieces from the redeposited and eroded areas of the bumper limiter of TFTR. The deuterocarbon formation rate from TFTR redeposits maximizes at ∼500 °C. Essentially the same chemical erosion behavior is observed for TFTR‐eroded and virgin graphites and is characterized by the compound peak, in...

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.

Following a two‐year operational period the Tokamak Fusion Test Reactor (TFTR) graphite fixed bumper limiter has been examined by a variety of methods. The areal density of metals was mapped in situ by beta backscattering. Several tiles were examined in detail by nuclear‐reaction analysis, Rutherford backscattering, and proton‐induced x‐ray emission to measure areal densities of deuterium and impurities. Some areas of the limiter were found to be covered by deposited material several microns thick. Other areas where the incident plasma flux is higher were much cleaner. Long‐term collection coupons were also examined to characterize deposition on the wall. From these results the total amount of deuterium on the limiter and wall in TFTR is estimated to be ∼1.3×1024 atoms. The implications of this for wall‐pumping effects and future tritium inventory are discussed.

Deuterium permeabilities were determined for a series of 19 samples of Fecralloy A (an iron-chromium-aluminum alloy) oxidized in air at temperatures from 900°C to 1150°C for times from 4 hours to 192 hours. Permeability reductions of 10 to 500 times were observed with practically all of the samples falling in the 10 to 100 range. By depositing palladium over some of the oxidized samples it was determined that a significant part of the observed permeability reduction was due to surface effects related to defects in the oxide and the rest was due to protection of a large fraction of the surface by an essentially impermeable layer of aluminum oxide.

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.

Bare POCO AXF‐5Q graphite tiles were installed as the plasma‐facing surface of the Tokamak Fusion Test Reactor (TFTR) movable limiter for a series of ∼2700 high power discharges (600 with up to 6 MW neutral beams). During this operating phase, erosion and deposition processes modified the surface of the limiter. In the regions of the most intense plasma contact, which reached temperatures over 2400 °C, only small amounts of metallic impurities (