Publications

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Carrier transport and recombination are modeled for a heterojunction diode containing irradiation defects. Detailed attention is given to the role of band-to-trap tunneling and how it is affected by band offsets at the junction. Tunneled states are characterized by numerical solution of the one-band effective-mass envelope equation. The interaction with traps is treated assuming capture by the multi-phonon-emission mechanism. It is shown that tunneling can increase carrier recombination at defects by orders of magnitude in the presence of large band offsets. This explains why Npn InGaP/GaAs/GaAs heterojunction bipolar transistors with displacement damage from energetic-particle irradiation are observed to have high carrier recombination in the emitter-base depletion region.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

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The three-dimensional Monte Carlo code ERO has been used to simulate dedicated DIII-D experiments in which Mo and W samples with different sizes were exposed to controlled and well-diagnosed divertor plasma conditions to measure the gross and net erosion rates. Experimentally, the net erosion rate is significantly reduced due to the high local redeposition probability of eroded high-Z materials, which according to the modelling is mainly controlled by the electric field and plasma density within the Chodura sheath. Similar redeposition ratios were obtained from ERO modelling with three different sheath models for small angles between the magnetic field and the material surface, mainly because of their similar mean ionization lengths. The modelled redeposition ratios are close to the measured value. Decreasing the potential drop across the sheath can suppress both gross and net erosion because sputtering yield is decreased due to lower incident energy while the redeposition ratio is not reduced owing to the higher electron density in the Chodura sheath. Taking into account material mixing in the ERO surface model, the net erosion rate of high-Z materials is shown to be strongly dependent on the carbon impurity concentration in the background plasma; higher carbon concentration can suppress net erosion. As a result, the principal experimental results such as net erosion rate and profile and redeposition ratio are well reproduced by the ERO simulations.

Electrical performance and defect characterization of vertical GaN P-i-N diodes before and after irradiation with 2.5 MeV protons and neutrons is investigated. Devices exhibit increase in specific on-resistance following irradiation with protons and neutrons, indicating displacement damage introduces defects into the p-GaN and n- drift regions of the device that impact on-state device performance. The breakdown voltage of these devices, initially above 1700 V, is observed to decrease only slightly for particle fluence < {10{13}} hbox{cm}-2. The unipolar figure of merit for power devices indicates that while the on-resistance and breakdown voltage degrade with irradiation, vertical GaN P-i-Ns remain superior to the performance of the best available, unirradiated silicon devices and on-par with unirradiated modern SiC-based power devices.

This report documents work done for the ITER International Fusion Energy Organization (Sponsor) under a Funds-In Agreement FI 011140916 with Sandia National Laboratories. The work consists of preparing and analyzing samples for an experiment to measure material erosion and deposition in the EAST Tokamak. Sample preparation consisted of depositing thin films of carbon and aluminum onto molybdenum tiles. Analysis consists of measuring the thickness of films before and after exposure to helium plasma in EAST. From these measurements the net erosion and deposition of material will be quantified. Film thickness measurements are made at the Sandia Ion Beam Laboratory using Rutherford backscattering spectrometry and nuclear reaction analysis, as described in this report. This report describes the film deposition and pre-exposure analysis. Results from analysis after plasma exposure will be given in a subsequent report.

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This report examines the temperature dependence of the capture rate of carriers by defects in gallium arsenide and compares two previously published theoretical treatments of this based on multi phonon emission (MPE). The objective is to reduce uncertainty in atomistic simulations of gain degradation in III-V HBTs from neutron irradiation. A major source of uncertainty in those simulations is poor knowledge of carrier capture rates, whose values can differ by several orders of magnitude between various defect types. Most of this variation is due to different dependence on temperature, which is closely related to the relaxation of the defect structure that occurs as a result of the change in charge state of the defect. The uncertainty in capture rate can therefore be greatly reduced by better knowledge of the defect relaxation.

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We analyze a DIII-D tokamak experiment where two tungsten spots on the removable DiMES divertor probe were exposed to 12 s of attached plasma conditions, with moderate strike point temperature and density (~20 eV, ~4.5 × 10¹⁹ m^–3), and 3% carbon impurity content. Both very small (1 mm diameter) and small (1 cm diameter) deposited samples were used for assessing gross and net tungsten sputtering erosion. The analysis uses a 3-D erosion/redeposition code package (REDEP/WBC), with input from a diagnostic-calibrated near-surface plasma code (OEDGE), and with focus on charge state resolved impinging carbon ion flux and energy. The tungsten surfaces are primarily sputtered by the carbon, in charge states +1 to +4. We predict high redeposition (~75%) of sputtered tungsten on the 1 cm spot—with consequent reduced net erosion—and this agrees well with post-exposure DiMES probe RBS analysis data. As a result, this study and recent related work is encouraging for erosion lifetime and non-contamination performance of tokamak reactor high-Z plasma facing components.

The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8 μm is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

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As part of Sandia's program to simulate the effect of displacement damage on operation of heterojunction bipolar transistors (HBTs), we are examining the formulation in 1-D of band-to-band (bb) and band-to-trap (b-t) carrier tunneling.

We studied the desorption of D implanted into Be with a superficial oxide layer. We found that the different oxide thicknesses and implantation at different energies resulted in a strong variation of the fraction stopped within the oxide layer. Thermal desorption of D was subsequently performed, intermitted by nuclear reaction analysis for assessment of the D depth distributions and total retained amounts. Moreover, for the conditions, where part of the D was deposited in the Be substrate, a sharp decrease of the retained amount of D occurs around 200 °C. This is attributed to the release from metallic Be. Correspondingly, the D and O depth profiles show that above 200 °C the remaining D is only retained in the BeO layer. Apparently, the superficial BeO layer does not act as a diffusion barrier for D that is released from the metallic substrate. The retained amount of D deposited within the BeO layer decreases steadily and is not completely released at 350 °C, the foreseen bake-out temperature in ITER.