N. H. Brooks

General Atomics, San Diego, California, United States

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Publications (346)434.49 Total impact

  • [Show abstract] [Hide abstract]
    ABSTRACT: Abstract The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.
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    ABSTRACT: A substantial reduction of net compared to gross erosion of molybdenum and tungsten was observed in experiments conducted in the lower divertor of DIII-D using the divertor material evaluation system. Post-exposure net erosion of molybdenum and tungsten films was measured by Rutherford backscattering (RBS) yielding net erosion rates of 0.4–0.7 nm s−1 for Mo and ~0.14 nm s−1 for W. Gross erosion was estimated using RBS on a 1 mm diameter sample, where re-deposition is negligible. Net erosion on a 1 cm diameter sample was reduced compared to gross erosion by factors of ~2 for Mo and ~3 for W. The experiment was modeled with the REDEP/WBC erosion/re-deposition code package coupled to the Ion Transport in Materials and Compounds—DYNamics mixed-material code, with plasma conditions supplied by the Onion skin modeling + Eirene + Divimp for edGE modeling code with input from divertor Langmuir probes. The code-calculated net/gross erosion rate ratios of 0.46 for Mo and 0.33 for W are in agreement with the experiment.
    Physica Scripta 04/2014; 2014(T159):014030. DOI:10.1088/0031-8949/2014/T159/014030 · 1.13 Impact Factor
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    ABSTRACT: Infrared imaging of hot spots induced by localized magnetic perturbations using the test blanket module (TBM) mock-up on DIII-D is in good agreement with beam-ion loss simulations. The hot spots were seen on the carbon protective tiles surrounding the TBM as they reached temperatures over 1000 °C. The localization of the hot spots on the protective tiles is in fair agreement with fast-ion loss simulations using a range of codes: ASCOT, SPIRAL and OFMCs while the codes predicted peak heat loads that are within 30% of the measured ones. The orbit calculations take into account the birth profile of the beam ions as well as the scattering and slowing down of the ions as they interact with the localized TBM field. The close agreement between orbit calculations and measurements validate the analysis of beam-ion loss calculations for ITER where ferritic material inside the tritium breeding TBMs is expected to produce localized hot spots on the first wall.
    Nuclear Fusion 12/2013; 53(12):3018-. DOI:10.1088/0029-5515/53/12/123018 · 3.06 Impact Factor
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    ABSTRACT: The injection of small deuterium pellets at high repetition rates up to 12× the natural edge localized mode (ELM) frequency has been used to trigger high-frequency ELMs in otherwise low natural ELM frequency H-mode deuterium discharges in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)]. The resulting pellet-triggered ELMs result in up to 12× lower energy and particle fluxes to the divertor than the natural ELMs. The plasma global energy confinement and density are not strongly affected by the pellet perturbations. The plasma core impurity density is strongly reduced with the application of the pellets. These experiments were performed with pellets injected from the low field side pellet in plasmas designed to match the ITER baseline configuration in shape and normalized β operation with input heating power just above the H-mode power threshold. Nonlinear MHD simulations of the injected pellets show that destabilization of ballooning modes by a local pressure perturbation is responsible for the pellet ELM triggering. This strongly reduced ELM intensity shows promise for exploitation in ITER to control ELM size while maintaining high plasma purity and performance.
    Physics of Plasmas 08/2013; 20(8). DOI:10.1063/1.4818772 · 2.14 Impact Factor
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    ABSTRACT: DIII-D experiments on rapid shutdown runaway electron (RE) beams have improved the understanding of the processes involved in RE beam control and dissipation. Improvements in RE beam feedback control have enabled stable confinement of RE beams out to the volt-second limit of the ohmic coil, as well as enabling a ramp down to zero current. Spectroscopic studies of the RE beam have shown that neutrals tend to be excluded from the RE beam centre. Measurements of the RE energy distribution function indicate a broad distribution with mean energy of order several MeV and peak energies of order 30–40 MeV. The distribution function appears more skewed towards low energies than expected from avalanche theory. The RE pitch angle appears fairly directed (θ ~ 0.2) at high energies and more isotropic at lower energies (ε < 100 keV). Collisional dissipation of RE beam current has been studied by massive gas injection of different impurities into RE beams; the equilibrium assimilation of these injected impurities appears to be reasonably well described by radial pressure balance between neutrals and ions. RE current dissipation following massive impurity injection is shown to be more rapid than expected from avalanche theory—this anomalous dissipation may be linked to enhanced radial diffusion caused by the significant quantity of high-Z impurities (typically argon) in the plasma. The final loss of RE beams to the wall has been studied: it was found that conversion of magnetic to kinetic energy is small for RE loss times smaller than the background plasma ohmic decay time of order 1–2 ms.
    Nuclear Fusion 07/2013; 53(8):083004. DOI:10.1088/0029-5515/53/8/083004 · 3.06 Impact Factor
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    ABSTRACT: Experimental observation of net erosion of molybdenum being significantly reduced compared to gross erosion in the divertor of DIII-D is reported for well-controlled plasma conditions. For the first time, gross erosion rates were measured by both spectroscopic and non-spectroscopic methods. In one experiment a net erosion rate of 0.73 ± 0.03 nm/s was measured using ion beam analysis (IBA) of a 1 cm diameter Mo-coated sample. For a 1 mm diameter Mo sample exposed at the same time the net erosion rate was higher at 1.31 nm/s. For the small sample redeposition is expected to be negligible in comparison with the larger sample yielding a net to gross erosion estimate of 0.56 ± 12%. The gross rate was also measured spectroscopically (386 nm MoI line) giving 2.45 nm/s ± factor 2. The experiment was modeled with the REDEP/WBC erosion/redeposition code package coupled to the ITMC–DYN mixed-material code, with plasma conditions supplied by the OEDGE code using Langmuir probe data input. The code-calculated net/gross ratio is =0.46, in good agreement with experiment.
    Journal of Nuclear Materials 07/2013; 438:S309–S312. DOI:10.1016/j.jnucmat.2013.01.052 · 1.87 Impact Factor
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    ABSTRACT: High repetition rate injection of deuterium pellets from the low-field side (LFS) of the DIII-D tokamak is shown to trigger high-frequency edge-localized modes (ELMs) at up to 12× the low natural ELM frequency in H-mode deuterium plasmas designed to match the ITER baseline configuration in shape, normalized beta, and input power just above the H-mode threshold. The pellet size, velocity, and injection location were chosen to limit penetration to the outer 10% of the plasma. The resulting perturbations to the plasma density and energy confinement time are thus minimal (<10%). The triggered ELMs occur at much lower normalized pedestal pressure than the natural ELMs, suggesting that the pellet injection excites a localized high-n instability. Triggered ELMs produce up to 12× lower energy and particle fluxes to the divertor, and result in a strong decrease in plasma core impurity density. These results show for the first time that shallow, LFS pellet injection can dramatically accelerate the ELM cycle and reduce ELM energy fluxes on plasma facing components, and is a viable technique for real-time control of ELMs in ITER.
    Physical Review Letters 06/2013; 110(24). DOI:10.1103/PhysRevLett.110.245001 · 7.51 Impact Factor
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    ABSTRACT: In April 2010, two thermo-oxidation experiments ('O-bakes') were performed in the DIII-D tokamak. Internal surfaces of the tokamak, as well as a number of specimens inserted into the torus, were exposed to a mixture of 20% O2/80% He at a nominal pressure of 9.5 Torr (1.27 kPa) at a temperature of 350–360 °C for a duration of 2 h. Three primary conclusions have been drawn from these experiments: (1) laboratory measurements on the release of deuterium from tokamak codeposits by oxidation have been duplicated in a tokamak environment, (2) no internal tokamak components or systems were adversely affected by the oxidation and (3) the recovery of plasma performance following oxidation was similar to that following regular torus openings.
    Nuclear Fusion 05/2013; 53(7):073008. DOI:10.1088/0029-5515/53/7/073008 · 3.06 Impact Factor
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    ABSTRACT: Understanding of Plasma Surface Interactions (PSI) and the selection of suitable plasma facing materials are critical areas for current tokamak experiments and future D-T burning facilities including ITER and FNSF. In support of PSI studies, DIII-D uses the Divertor Materials Evaluation System (DiMES), which contains a removable probe where material samples can be exposed to as few as a single well-characterized plasma shot. Experiments, consisting of a carbon DiMES probe surface with metal coatings of Be, W, V, Mo or Al, have been exposed to the DIII-D lower divertor strike point plasma for cumulative discharge times of 4-20s. Extensive DIII-D divertor diagnostics provided well-characterized plasmas for modeling efforts. Experimental results were benchmarked with modeling codes to validate and extend the predictive capability of the codes. Reported in this paper are two recent experiments and results. The first is on the net and gross erosion of Mo coatings and the extension of these results to an extrapolated all Mo surface DIII-D machine. The second is on the exposure to vertical displacement discharges and X-point plasma discharges of W-fuzz buttons, which were prepared by the PISCES (UCSD) laboratory. The surprising results are the robustness of the W-fuzz and that W impurity was not detected in the plasma core at the conditions studied.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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    ABSTRACT: The quality of plasma produced in a magnetic confinement fusion device is influenced to a large extent by the neutral gas surrounding the plasma. The plasma is fueled by the ionization of neutrals, and charge exchange interactions between edge neutrals and plasma ions are a sink of energy and momentum. Here we describe a diagnostic capable of measuring the spatial distribution of neutral gas in a magnetically confined fusion plasma. A high intensity (5 MW/cm(2)), narrow bandwidth (0.1 cm(-1)) laser is injected into a hydrogen plasma to excite the Lyman β transition via the simultaneous absorption of two 205 nm photons. The absorption rate, determined by measurement of subsequent Balmer α emission, is proportional to the number of particles with a given velocity. Calibration is performed in situ by filling the chamber to a known pressure of neutral krypton and exciting a transition close in wavelength to that used in hydrogen. We present details of the calibration procedure, including a technique for identifying saturation broadening, measurements of the neutral density profile in a hydrogen helicon plasma, and discuss the application of the diagnostic to plasmas in the DIII-D tokamak.
    The Review of scientific instruments 10/2012; 83(10):10D701. DOI:10.1063/1.4728092 · 1.61 Impact Factor
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    ABSTRACT: Localized hot spots can be created in ITER on the Test Blanket Modules (TBMs) because the ferritic steel of the TBMs distorts the local magnetic field near the modules and alters fast ion confinement. Predicting the TBM heat load levels is important for assessing their effects on the ITER first wall. Experiments in DIII-D were carried out with a mock-up of the ITER TBM ferromagnetic error field to provide data for validation of fast-ion orbit following codes. The front surface temperature of the protective TBM tiles was imaged directly with a calibrated infrared camera and heat loads were extracted. The detailed spot sizes and measured heat loads are compared with results from heat load calculations performed with a suite of orbit following codes. The codes reproduce the hot spots well, thereby validating the codes and giving confidence in predictions for fast-ion heat loads in ITER.
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    ABSTRACT: A high speed, high resolution near infrared (NIR) spectrometer has been installed at DIII-D to make first-of-its-kind observations of the 0.8-2.2 μm region in a tokamak divertor. The goals of this diagnostic are (1) to study Paschen spectra for line-averaged measurement of low temperature plasma parameters, (2) to benchmark the chemical and physically sputtered sources of neutral carbon using the lineshape of the CI, 910 nm multiplet, and (3) to quantify contamination of the 0.75-1.1 μm region where Thomson-shifted laser light is measured by the Thomson scattering diagnostic. Diagnostic capabilities include a 300 mm, f/3.9 design, 300-2400 Gr/mm gratings providing optical resolution of ˜0.65-0.04 nm, and readout at up to 900 frames/second. Data are presented in L-mode plasmas, and in H-mode between ELMs and during the ELM peak. Results acquired by this diagnostic will be applied to design of a proposed divertor Thomson diagnostic for NSTX-U and aid validation of the Thomson system on ITER.
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    ABSTRACT: The OEDGE code is used to assess several methods of determining the upstream separatrix location. The inter-ELM phase of a well-diagnosed ELMing H-mode discharge is being used for this comparison. The OEDGE code utilizes 1D plasma fluid models calculated along the field lines on a 2D computational grid of a poloidal cross-section of the discharge magnetic geometry to produce a 2D model of the background plasma. Langmuir probe data at the targets are used as input to the 1D models. Additional diagnostic measurements, including Thomson scattering, reciprocating probe, divertor spectroscopy and infra-red measurements of target heat flux, may be used to further constrain the plasma background determined by OEDGE. This plasma background thus found, is then used to identify the location of the separatrix in the experimental data by comparing the upstream plasma profiles from OEDGE to the experimental measurements. The OEDGE result is then compared to the separatrix locations predicted using simple pressure balance and power balance considerations.
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    ABSTRACT: Recent DIII-D experiments have investigated the effects of localized magnetic field perturbations, using coils that approximate the magnetization of the test blanket modules (TBMs) in one ITER port. In H-mode discharges, compensation of the TBM field using an applied n=1 field yielded only partial recovery of the plasma rotation, and the compensation field that maximized plasma rotation differed significantly from the field that reduced the resonant magnetic response to a very low value. These results provide insight into the effects of error fields, and suggest an important role for non-resonant magnetic braking. In addition, measurements of localized heat deposition with the TBM field are being compared to orbit following calculations of fast ion loss, and a new fast ion detector has confirmed earlier observations of reduced 1 MeV triton confinement.
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    ABSTRACT: An experiment was conducted in DIII-D to examine carbon deposition when a secondary separatrix is near the wall. The magnetic configuration for this experiment was a biased double-null, similar to that foreseen for ITER. 13C methane was injected toroidally symmetrically near the secondary separatrix into ELMy H-mode deuterium plasmas. The resulting deposition of 13C was determined by nuclear reaction analysis. These results show that very little of the injected 13C was deposited at the primary separatrix, whereas a large fraction of injected 13C was deposited close to the point of injection near the secondary separatrix. Six of the tiles were put back into DIII-D, where they were baked at 350-360 °C for 2 h at ~1 kPa in a 20% O2/80% He gas mixture. Subsequent ion beam analysis of these tiles showed that about 21% of the 13C and 54% of the deuterium were removed by the bake.
    12/2011; T145:4025-. DOI:10.1088/0031-8949/2011/T145/014025
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    ABSTRACT: Deuterium fueling profiles across the separatrix have been calculated with the edge fluid codes UEDGE, SOLPS and EDGE2D/EIRENE for lower single null, ohmic and low-confinement plasmas in DIII-D, ASDEX Upgrade and JET. The fueling profiles generally peak near the divertor x-point, and broader profiles are predicted for the open divertor geometry and horizontal targets in DIII-D than for the more closed geometries and vertical targets in AUG and JET. Significant fueling from the low-field side midplane may also occur when assuming strong radial ion transport in the far scrape-off layer. The dependence of the fueling profiles on upstream density is investigated for all three devices, and between the different codes for a single device. The validity of the predictions is assessed for the DIII-D configuration by comparing the measured ion current to the main chamber walls at the low-field side and divertor targets, and deuterium emission profiles across the divertor legs, and the high-field and low-field side midplane regions to those calculated by UEDGE and SOLPS.
    Plasma Physics and Controlled Fusion 11/2011; 53(12):124017. DOI:10.1088/0741-3335/53/12/124017 · 2.19 Impact Factor
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    ABSTRACT: We apply new atomic modeling techniques to helium and deuterium for diagnostics in the divertor and scrape-off layer regions. Analysis of tomographically inverted images is useful for validating detachment prediction models and power balances in the divertor. We apply tomographic image inversion from fast tangential cameras of helium and Dα emission at the divertor in order to obtain 2D profiles of Te, Ne, and ND (neutral ion density profiles). The accuracy of the atomic models for He I will be cross-checked against Thomson scattering measurements of Te and Ne. This work summarizes several current developments and applications of atomic modeling into diagnostic at the DIII-D tokamak.
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    ABSTRACT: The OEDGE code is used to model the deuterium neutral density and ionization distribution inside the separatrix for an attached L-mode SAPP discharge and an attached ELMy H-mode discharge. The background plasma solution is determined by empirical plasma reconstruction matching as many diagnostic measurements as possible. Recycling fluxes are obtained from measurements by Langmuir probes and spectroscopic measurements of Dα. The relative importance of wall, divertor and recombination sources to core and pedestal fueling are assessed. In addition, the sensitivity of the ionization source location to the details of the plasma solution in the divertor is examined. Several models for plasma-wall contact are used to estimate the strength of the wall recycling source. In the L-mode case, ionization profiles peak at the flux surface ˜1.3 cm inboard of the separatrix (mapped to the outer midplane).
    Journal of Nuclear Materials 11/2011; 438:9024P-. DOI:10.1016/j.jnucmat.2013.01.137 · 1.87 Impact Factor
  • B. L. Dwyer · N. H. Brooks · R. L. Lee ·
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    ABSTRACT: We constructed two devices for the purpose of educational demonstration: a rotating tube containing media of two densities to demonstrate axial confinement and a similar device that uses pressure variation to convert a long plasma glow discharge into a long straight arc [1]. In the first device, the buoyant force is countered by the centripetal force, which confines less dense materials to the center of the column. Similarly, a plasma arc heats the gas through which it passes, creating a hot gaseous bubble that is less dense than the surrounding medium. Rotating its containment envelope stabilizes this gas bubble in an analogous manner to an air bubble in a rotating tube of water. In addition to stabilization, the rotating discharge also exhibits a decrease in buoyancy-driven convection currents. This limits the power loss to the walls, which decreases the field strength requirement for maintaining the arc. These devices demonstrate principles of electrodynamics, plasma physics, and fluid mechanics. They are portable and safe for classroom use. 6pt [1] N.H. Brooks, et al., J. Appl. Phys. 94, 1402 (2003).
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    ABSTRACT: The density profile of hydrogenic neutrals in the edge of DIII-D plays an important role in the problems of momentum transport, pedestal formation, and plasma-wall interaction, but an accurate measurement has proven difficult. A two-photon absorption laser induced fluorescence (TALIF) diagnostic is under construction and is intended to provide temporally and spatially resolved neutral density measurements in the pedestal region. This three-level TALIF scheme offers the advantages of direct excitation of ground state atoms, emission in the visible portion of the spectrum, a high degree of spatial localization, and the potential for a Doppler-free measurement. The large background of Dα emission, the principal challenge of the measurement, can be overcome by the focusing of a high power (1 MW) UV laser. Calculations of the SNR show that densities of 10^15 m-3 or lower can be measured with a spatial resolution of 0.3 mm. We present design details of the proposed laser system, calculations of the expected performance in DIII-D and in a helicon source plasma, and measurements of the HI profile in the helicon plasma.

Publication Stats

3k Citations
434.49 Total Impact Points


  • 1994-2014
    • General Atomics
      San Diego, California, United States
    • Massachusetts Institute of Technology
      Cambridge, Massachusetts, United States
  • 1994-2002
    • Oak Ridge National Laboratory
      • Fusion Energy Division
      Oak Ridge, Florida, United States
  • 1998-2001
    • Lawrence Livermore National Laboratory
      • Physics Division
      Livermore, California, United States
  • 1995
    • Sandia National Laboratories
      • Semiconductor Material and Device Sciences Department
      Albuquerque, New Mexico, United States