Analysis and Modeling of DIII-D Experiments With OMFIT and Neural Networks

NASA Astrophysics Data System (ADS)

Meneghini, O.; Luna, C.; Smith, S. P.; Lao, L. L.; GA Theory Team

2013-10-01

The OMFIT integrated modeling framework is designed to facilitate experimental data analysis and enable integrated simulations. This talk introduces this framework and presents a selection of its applications to the DIII-D experiment. Examples include kinetic equilibrium reconstruction analysis; evaluation of MHD stability in the core and in the edge; and self-consistent predictive steady-state transport modeling. The OMFIT framework also provides the platform for an innovative approach based on neural networks to predict electron and ion energy fluxes. In our study a multi-layer feed-forward back-propagation neural network is built and trained over a database of DIII-D data. It is found that given the same parameters that the highest fidelity models use, the neural network model is able to predict to a large degree the heat transport profiles observed in the DIII-D experiments. Once the network is built, the numerical cost of evaluating the transport coefficients is virtually nonexistent, thus making the neural network model particularly well suited for plasma control and quick exploration of operational scenarios. The implementation of the neural network model and benchmark with experimental results and gyro-kinetic models will be discussed. Work supported in part by the US DOE under DE-FG02-95ER54309.

Comparison of a low- to high-confinement transition theory with experimental data from DIII-D.

PubMed

Guzdar, P N; Kleva, R G; Groebner, R J; Gohil, P

2002-12-23

From our recent theory based on the generation of shear flow and field in finite beta plasmas, the criterion for bifurcation from low to high confinement mode yields a critical parameter proportional to T(e)/square root (L(n)), where T(e) is the electron temperature and L(n) is the density scale length. The predicted threshold shows very good agreement with edge measurements on discharges undergoing low-to-high transitions in DIII-D. The observed differences in the transitions with the reversal of the toroidal magnetic field are reconciled in terms of this critical parameter. The theory also provides an explanation for pellet injection H modes in DIII-D, thereby unifying unconnected methods for accomplishing the transition.

Anomalous transport scaling in the DIII-D tokamak matched by supercomputer simulation.

PubMed

Candy, J; Waltz, R E

2003-07-25

Gyrokinetic simulation of tokamak transport has evolved sufficiently to allow direct comparison of numerical results with experimental data. It is to be emphasized that only with the simultaneous inclusion of many distinct and complex effects can this comparison realistically be made. Until now, numerical studies of tokamak microturbulence have been restricted to either (a) flux tubes or (b) electrostatic fluctuations. Using a newly developed global electromagnetic solver, we have been able to recover via direct simulation the Bohm-like scaling observed in DIII-D L-mode discharges. We also match, well within experimental uncertainty, the measured energy diffusivities.

Presentation Stations of the General Atomics Fusion Educational Program

NASA Astrophysics Data System (ADS)

Lee, R. L.; Fusion Group Education Outreach Team

1996-11-01

The General Atomics Fusion Group's Educational Program has been actively promoting fusion science and applications throughout San Diego County's secondary school systems for over three years. The educational program allows many students to learn more about nuclear fusion science, its applications, and what it takes to become an active participant in an important field of study. It also helps educators to better understand how to teach fusion science in their classroom. Tours of the DIII--D facility are a centerpiece of the program. Over 1000 students visited the DIII--D research facility during the 1995--1996 school year for a half-day of presentations, discussions, and hands-on learning. Interactive presentations are provided at six different stations by GA scientists and engineers to small groups of students during the tours. Stations include topics on energy, plasma science, the electromagnetic spectrum, radiation and risk assessment, and data acquisition. Included also is a tour of the DIII--D machine hall and model where students can see and discuss many aspects of the tokamak. Portions of each station will be presented and discussed.

Collegiate Athletics: An Investigation into Athletic Persistence of Freshman Student-Athletes Participating in NCAA Division-III Varsity Athletic Programs

ERIC Educational Resources Information Center

Sombito, Lester Jamili

2015-01-01

Understanding of the persistence of student-athletes in athletics at the D-III level is complex. This research study investigated the issue of student-athlete retention by focusing on Division III (D-III) student-athlete persistence in athletics by asking the following research question, "To what extent do freshman student-athletes persist in…

Nonlinear MHD simulations of Quiescent H-mode plasmas in DIII-D

DOE PAGES

Liu, Feng; Huijsmans, G. T. A.; Loarte, A.; ...

2015-09-04

In the Quiescent H-mode (QH-mode) regime, the edge harmonic oscillation (EHO), thought to be a saturated kink-peeling mode (KPM) driven unstable by current and rotation, is found in experiment to provide sufficient stationary edge particle transport to avoid the periodic expulsion of particles and energy by edge localized modes (ELMs). In this article, both linear and nonlinear MHD modelling of QH-mode plasmas from the DIII-D tokamak have been investigated to understand the mechanism leading to the appearance of the EHO in QH-mode plasmas. For the first time nonlinear MHD simulations with low-n modes both with ideal wall and resistive wallmore » boundary conditions have been carried out with 3-D non-linear MHD code JOREK. The results show, in agreement with the original conjectures, that in the nonlinear phase, kink peeling modes are the main unstable modes in QH-mode plasmas of DIIID and that the kink-peeling modes saturate non-linearly leading to a 3-D stationary state. The characteristics of the kink-peeling modes, in terms of mode structure and associated decrease of the edge plasma density associated with them, are in good agreement with experimental measurements of the EHO in DIII-D. Finally, the effect of plasma resistivity, the role of plasma parallel rotation as well as the effect of the conductivity of the vacuum vessel wall on the destabilization and saturation of kink-peeling modes have been evaluated for experimental QH-mode plasma conditions in DIII-D.« less

Topological mirror superconductivity.

PubMed

Zhang, Fan; Kane, C L; Mele, E J

2013-08-02

We demonstrate the existence of topological superconductors (SCs) protected by mirror and time-reversal symmetries. D-dimensional (D=1, 2, 3) crystalline SCs are characterized by 2(D-1) independent integer topological invariants, which take the form of mirror Berry phases. These invariants determine the distribution of Majorana modes on a mirror symmetric boundary. The parity of total mirror Berry phase is the Z(2) index of a class DIII SC, implying that a DIII topological SC with a mirror line must also be a topological mirror SC but not vice versa and that a DIII SC with a mirror plane is always time-reversal trivial but can be mirror topological. We introduce representative models and suggest experimental signatures in feasible systems. Advances in quantum computing, the case for nodal SCs, the case for class D, and topological SCs protected by rotational symmetries are pointed out.

High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal.

PubMed

Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Haskey, S R; Kaplan, D H

2016-11-01

A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. The unique combination of experimentally measured main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qian, Jinping P.; Garofalo, Andrea M.; Gong, Xianzu Z.

Recent EAST/DIII-D joint experiments on the high poloidal betamore » $${{\\beta}_{\\text{P}}}$$ regime in DIII-D have extended operation with internal transport barriers (ITBs) and excellent energy confinement (H 98y2 ~ 1.6) to higher plasma current, for lower q 95 ≤ 7.0, and more balanced neutral beam injection (NBI) (torque injection < 2 Nm), for lower plasma rotation than previous results. Transport analysis and experimental measurements at low toroidal rotation suggest that the E × B shear effect is not key to the ITB formation in these high $${{\\beta}_{\\text{P}}}$$ discharges. Experiments and TGLF modeling show that the Shafranov shift has a key stabilizing effect on turbulence. Extrapolation of the DIII-D results using a 0D model shows that with the improved confinement, the high bootstrap fraction regime could achieve fusion gain Q = 5 in ITER at $${{\\beta}_{\\text{N}}}$$ ~ 2.9 and q 95 ~ 7. With the optimization of q(0), the required improved confinement is achievable when using 1.5D TGLF-SAT1 for transport simulations. Furthermore, results reported in this paper suggest that the DIII-D high $${{\\beta}_{\\text{P}}}$$ scenario could be a candidate for ITER steady state operation.« less

The DIII-D Map -- An Area-Preserving Map for Trajectories of Magnetic Field Lines in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

2007-11-01

The EFIT data for the DIII-D shot 115467 3000 ms is used to calculate the generating function for an area-preserving map for trajectories of magnetic field lines in the DIII-D. We call this map the DIII-D map. The generating function is a bivariate polynomial in base vectors &1/2circ;cos(θ) and &1/2circ;sin(θ). ψ is toroidal flux and θ is poloidal angle. The generating function is calculated using a canonical transformation from (ψ,θ) to physical coordinates (R,Z) in the DIII-D [1] and nonlinear regression. The equilibrium generating function gives an excellent representation of the equilibrium flux surfaces in the DIII-D. The DIII-D map is then used to calculate effects of the magnetic perturbations in the DIII-D. Preliminary results of the DIII-D map will be presented. This work is supported by US DOE OFES DE-FG02-01ER54624 and DE-FG02-04ER54793. [1] A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys Lett A 364 140--145 (2007).

Destruction of the Last Good Magnetic Surface in Diii-D Usn with Elms and C-Coils Shot 115467 due to ELMs Using Maps

NASA Astrophysics Data System (ADS)

Butcher, L.; Ali, H.; Punjabi, A.

2004-11-01

The unperturbed magnetic topology of DIII-D USN shot 115467 in the absence of ELMs and C-coils is described by the symmetric simple map (SSM) with the map parameter k=0.2623 [1]. For this k, the last good surface passes through x=0 and y=0.9995, q_edge=6.48 if six iterations of the SSM are taken to be equivalent to a single toroidal circuit of DIII-D, and the q_edge equals the q_edge in the DIII-D for shot 115467. The low mn map calculates the effects of the m=1, n=+1,-1 internal magnetic perturbation on the trajectories of field lines. We use the low mn map to represent the effects of the ELMs in the DIII-D. We study the changes in the last good surface and its destruction as the amplitude ɛ in the low mn map is increased. This work is supported by NASA SHARP program and DE-FG02-02ER54673. [1] H. Ali, A. Punjabi, A. Boozer, and T. Evans, presented at the 31st European Physical Society Plasma Physics Meeting, London, UK, June 29, 2004, paper P2-172.

Advances in the high bootstrap fraction regime on DIII-D towards the Q  =  5 mission of ITER steady state

NASA Astrophysics Data System (ADS)

Qian, J. P.; Garofalo, A. M.; Gong, X. Z.; Ren, Q. L.; Ding, S. Y.; Solomon, W. M.; Xu, G. S.; Grierson, B. A.; Guo, W. F.; Holcomb, C. T.; McClenaghan, J.; McKee, G. R.; Pan, C. K.; Huang, J.; Staebler, G. M.; Wan, B. N.

2017-05-01

Recent EAST/DIII-D joint experiments on the high poloidal beta {β\\text{P}} regime in DIII-D have extended operation with internal transport barriers (ITBs) and excellent energy confinement (H 98y2 ~ 1.6) to higher plasma current, for lower q 95  ⩽  7.0, and more balanced neutral beam injection (NBI) (torque injection  <  2 Nm), for lower plasma rotation than previous results (Garofalo et al, IAEA 2014, Gong et al 2014 IAEA Int. Conf. on Fusion Energy). Transport analysis and experimental measurements at low toroidal rotation suggest that the E  ×  B shear effect is not key to the ITB formation in these high {β\\text{P}} discharges. Experiments and TGLF modeling show that the Shafranov shift has a key stabilizing effect on turbulence. Extrapolation of the DIII-D results using a 0D model shows that with the improved confinement, the high bootstrap fraction regime could achieve fusion gain Q  =  5 in ITER at {β\\text{N}} ~ 2.9 and q 95 ~ 7. With the optimization of q(0), the required improved confinement is achievable when using 1.5D TGLF-SAT1 for transport simulations. Results reported in this paper suggest that the DIII-D high {β\\text{P}} scenario could be a candidate for ITER steady state operation.

Disruption avoidance and fast ramp-down techniques for the DIII-D experimental scenarios

NASA Astrophysics Data System (ADS)

Barr, Jayson; Eidietis, N. W.; Humphreys, D. A.; Sammuli, B.; Luce, T.

2017-10-01

Plasma current ramp-down in ITER will continue in H-mode from 15 MA to 10 MA, and will keep a diverted shape until termination. This is in contrast to the limited ramp-down scenarios typically used in DIII-D operations. Additionally, fast emergency ramp-down scenarios for ITER and future reactors are a priority for disruption avoidance. New experiments in DIII-D use the ramp-down phase of a variety of experiments including in the ITER baseline scenario to survey and identify optimized ramp-down scenarios for both scheduled terminations and terminations triggered by off-normal event detection. Systematic scans in current ramp-rate (1-5 MA/s), neutral beam power (including βN feedback) and ramp-down shaping (limited versus continued diverted) have identified fast ramp-down scenarios for Lower Single Null (LSN) and Double Null (DN) plasmas. Scenario-specific methods and their rates of successful termination will be presented and compared relative to a historical data-set of ramp-down programming in the limiter configuration. Locked modes are found to be the most significant challenge to disruption avoidance in diverted ramp-downs. Results for LSN diverted discharges that begin the rampdown with large locked-modes will also be presented. If available, results of similar experiments on EAST will be presented. Work supported by US DOE under DE-FC02-04ER54698 and DE-SC0010685.

Development of a high power Helicon system for DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tooker, Joseph F.; Nagy, Alexander; deGrassie, John

A mechanism for driving current off-axis in high beta tokamaks using fast electromagnetic waves, called Helicons, will be experimentally tested in the DIII-D tokamak. This method is calculated to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and it may be well suited to reactor-like configurations. A low power (100 W) 476 MHz “combline” antenna, consisting of 12 inductively coupled, electrostatically shielded, modular resonators, was recently installed in DIII-D. Initial operation showed that the plasma operating conditions were achieved under which helicon waves can be launched. Plasma operations also showed that the location ofmore » the antenna has not reduced the performance of, or introduced excessive impurities into, the discharges produced in DIII-D. The development of a high power (1 MW) Helicon system is underway. This antenna consists of 30 modules mounted on the inside of the outer wall of the vacuum vessel slightly above the midplane. Carbon tiles around the antenna protect the antenna from thermal plasma streaming along field lines. A 1.2 MW, 476 MHz klystron system will be transferred from the Stanford Linear Accelerator to DIII-D to provide the RF input power to the antenna. Lastly, a description of the design and fabrication of high power antenna and the RF feeds, the klystron and RF distribution systems, and their installation will be presented.« less

Development of a high power Helicon system for DIII-D

DOE PAGES

Tooker, Joseph F.; Nagy, Alexander; deGrassie, John; ...

2017-03-29

A mechanism for driving current off-axis in high beta tokamaks using fast electromagnetic waves, called Helicons, will be experimentally tested in the DIII-D tokamak. This method is calculated to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and it may be well suited to reactor-like configurations. A low power (100 W) 476 MHz “combline” antenna, consisting of 12 inductively coupled, electrostatically shielded, modular resonators, was recently installed in DIII-D. Initial operation showed that the plasma operating conditions were achieved under which helicon waves can be launched. Plasma operations also showed that the location ofmore » the antenna has not reduced the performance of, or introduced excessive impurities into, the discharges produced in DIII-D. The development of a high power (1 MW) Helicon system is underway. This antenna consists of 30 modules mounted on the inside of the outer wall of the vacuum vessel slightly above the midplane. Carbon tiles around the antenna protect the antenna from thermal plasma streaming along field lines. A 1.2 MW, 476 MHz klystron system will be transferred from the Stanford Linear Accelerator to DIII-D to provide the RF input power to the antenna. Lastly, a description of the design and fabrication of high power antenna and the RF feeds, the klystron and RF distribution systems, and their installation will be presented.« less

Overview of the recent DiMES and MiMES experiments in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Wong, C. P. C.; Litnovsky, A.; Wampler, W. R.; Boedo, J. A.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Hollmann, E. M.; Jacob, W.; Krasheninnikov, S. I.; Krieger, K.; Lasnier, C. J.; Leonard, A. W.; McLean, A. G.; Marot, M.; Moyer, R. A.; Petrie, T. W.; Philipps, V.; Smirnov, R. D.; Stangeby, P. C.; Watkins, J. G.; West, W. P.; Yu, J. H.

2009-12-01

Divertor and midplane material evaluation systems (DiMES and MiMES) in the DIII-D tokamak are used to address a variety of plasma-material interaction (PMI) issues relevant to ITER. Among the topics studied are carbon erosion and re-deposition, hydrogenic retention in the gaps between plasma-facing components (PFCs), deterioration of diagnostic mirrors from carbon deposition and techniques to mitigate that deposition, and dynamics and transport of dust. An overview of the recent experimental results is presented.

40 CFR 52.1586 - Section 110(a)(2) infrastructure requirements.

Code of Federal Regulations, 2013 CFR

2013-07-01

... and the 1997 and 2006 PM2.5 NAAQS are disapproved for the following sections 110(a)(2)(C) (PSD program only), (D)(i)(II) (PSD program only), (D)(ii), and (J) (PSD program only). These requirements are being...

40 CFR 52.1586 - Section 110(a)(2) infrastructure requirements.

Code of Federal Regulations, 2014 CFR

2014-07-01

... and the 1997 and 2006 PM2.5 NAAQS are disapproved for the following sections 110(a)(2)(C) (PSD program only), (D)(i)(II) (PSD program only), (D)(ii), and (J) (PSD program only). These requirements are being...

High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal

DOE PAGES

Grierson, B. A.; Burrell, K. H.; Chrystal, C.; ...

2016-09-12

A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. Furthermore, the unique combination of experimentally measuredmore » main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.« less

MHD modeling of a DIII-D low-torque QH-mode discharge and comparison to observations

DOE PAGES

King, Jacob R.; Kruger, S. E.; Burrell, K. H.; ...

2017-03-07

Extended-MHD modeling of DIII-D tokamak quiescent H-mode (QH-mode) discharges with nonlinear NIMROD simulations saturates into a turbulent state but does not saturate when the steady-state flow inferred from measurements is not included. This is consistent with the experimental observations of the quiescent regime on DIII-D. The simulation with flow develops into a saturated turbulent state where the n Φ = 1 and 2 toroidal modes become dominant through an inverse cascade. Each mode in the range of n Φ = 1–5 is dominant at a different time. Consistent with experimental observations during QH-mode, the simulated state leads to large particlemore » transport relative to the thermal transport. Analysis shows that the amplitude and phase of the density and temperature perturbations differ resulting in greater fluctuation-induced convective particle transport relative to the convective thermal transport. As a result, comparison to magnetic-coil measurements shows that rotation frequencies differ between the simulation and experiment, which indicates that more sophisticated extended-MHD two-fluid modeling is required.« less

A New System to Monitor Data Analyses and Results of Physics Data Validation Between Pulses at DIII-D

NASA Astrophysics Data System (ADS)

Flanagan, S.; Schachter, J. M.; Schissel, D. P.

2001-10-01

A Data Analysis Monitoring (DAM) system has been developed to monitor between pulse physics analysis at the DIII-D National Fusion Facility. The system allows for rapid detection of discrepancies in diagnostic measurements or the results from physics analysis codes. This enables problems to be detected and possibly fixed between pulses as opposed to after the experimental run has concluded thus increasing the efficiency of experimental time. An example of a consistency check is comparing the stored energy from integrating the measured kinetic profiles to that calculated from magnetic measurements by EFIT. This new system also tracks the progress of MDSplus dispatching of software for data analysis and the loading of analyzed data into MDSplus. DAM uses a Java Servlet to receive messages, Clips to implement expert system logic, and displays its results to multiple web clients via HTML. If an error is detected by DAM, users can view more detailed information so that steps can be taken to eliminate the error for the next pulse. A demonstration of this system including a simulated DIII-D pulse cycle will be presented.

Study of the L-mode tokamak plasma “shortfall” with local and global nonlinear gyrokinetic δf particle-in-cell simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chowdhury, J.; Wan, Weigang; Chen, Yang

2014-11-15

The δ f particle-in-cell code GEM is used to study the transport “shortfall” problem of gyrokinetic simulations. In local simulations, the GEM results confirm the previously reported simulation results of DIII-D [Holland et al., Phys. Plasmas 16, 052301 (2009)] and Alcator C-Mod [Howard et al., Nucl. Fusion 53, 123011 (2013)] tokamaks with the continuum code GYRO. Namely, for DIII-D the simulations closely predict the ion heat flux at the core, while substantially underpredict transport towards the edge; while for Alcator C-Mod, the simulations show agreement with the experimental values of ion heat flux, at least within the range of experimental error.more » Global simulations are carried out for DIII-D L-mode plasmas to study the effect of edge turbulence on the outer core ion heat transport. The edge turbulence enhances the outer core ion heat transport through turbulence spreading. However, this edge turbulence spreading effect is not enough to explain the transport underprediction.« less

MHD modeling of a DIII-D low-torque QH-mode discharge and comparison to observations

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, Jacob R.; Kruger, S. E.; Burrell, K. H.

Extended-MHD modeling of DIII-D tokamak quiescent H-mode (QH-mode) discharges with nonlinear NIMROD simulations saturates into a turbulent state but does not saturate when the steady-state flow inferred from measurements is not included. This is consistent with the experimental observations of the quiescent regime on DIII-D. The simulation with flow develops into a saturated turbulent state where the n Φ = 1 and 2 toroidal modes become dominant through an inverse cascade. Each mode in the range of n Φ = 1–5 is dominant at a different time. Consistent with experimental observations during QH-mode, the simulated state leads to large particlemore » transport relative to the thermal transport. Analysis shows that the amplitude and phase of the density and temperature perturbations differ resulting in greater fluctuation-induced convective particle transport relative to the convective thermal transport. As a result, comparison to magnetic-coil measurements shows that rotation frequencies differ between the simulation and experiment, which indicates that more sophisticated extended-MHD two-fluid modeling is required.« less

Tokamak Equilibrium Reconstruction with MSE-LS Data in DIII-D

NASA Astrophysics Data System (ADS)

Lao, L.; Grierson, B.; Burrell, K. H.

2016-10-01

Equilibrium analysis of plasmas in DIII-D using EFIT was upgraded to include the internal magnetic field determined from spectroscopic measurements of motional-Stark-effect line-splitting (MSE-LS). MSE-LS provides measurements of the magnitude of the internal magnetic field, rather than the pitch angle as provided by MSE line-polarization (MSE-LP) used in most tokamaks to date. EFIT MSE-LS reconstruction algorithms and verifications are described. The capability of MSE-LS to provide significant constraints on the equilibrium analysis is evaluated. Reconstruction results with both synthetic and experimental MSE-LS data from 10 DIII-D discharges run over a range of conditions show that MSE-LS measurements can contribute to the equilibrium reconstruction of pressure and safety factor profiles. Adequate MSE-LS measurement accuracy and number of spatial locations are necessary. The 7 available experimental measurements provide useful additional constraints when used with other internal measurements. Using MSE-LS as the only internal measurement yields less current profile information. Work supported by the PPPL Subcontract S013769-F and US DOE under DE-FC02-04ER54698.

Tensile and impact properties of General Atomics 832864 heat of V-4Cr-4Ti alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Nowicki, L.J.; Gazda, J.

A 1300-kg heat of V-4Cr-4Ti alloy was procured by General Atomics (GA) for the DIII-D radiative divertor program. To determine the mechanical properties of this alloy, tensile and Charpy tests were conducted on specimens prepared from pieces of 4.8-mm-thick as-rolled plates, a major product form for the DIII-D application. The tensile tests were conducted at three temperatures, 26, 280 and 380 C, the last two being the anticipated peak temperatures during DIII-D boronization and postvent bake-out, respectively. Results from these tests show that the tensile and impact properties of the 832864 heat are comparable to those of the other smallermore » V-(4-5)Cr-(4-5)Ti alloy heats previously developed by the US Fusion Materials Program and that scale-up of vanadium alloy production can be successfully achieved as long as reasonable process control is implemented.« less

Physics-model-based nonlinear actuator trajectory optimization and safety factor profile feedback control for advanced scenario development in DIII-D

DOE PAGES

Barton, Justin E.; Boyer, Mark D.; Shi, Wenyu; ...

2015-07-30

DIII-D experimental results are reported to demonstrate the potential of physics-model-based safety factor profile control for robust and reproducible sustainment of advanced scenarios. In the absence of feedback control, variability in wall conditions and plasma impurities, as well as drifts due to external disturbances, can limit the reproducibility of discharges with simple pre-programmed scenario trajectories. The control architecture utilized is a feedforward + feedback scheme where the feedforward commands are computed off-line and the feedback commands are computed on-line. In this work, firstly a first-principles-driven (FPD), physics-based model of the q profile and normalized beta (β N) dynamics is embeddedmore » into a numerical optimization algorithm to design feedforward actuator trajectories that sheer the plasma through the tokamak operating space to reach a desired stationary target state that is characterized by the achieved q profile and β N. Good agreement between experimental results and simulations demonstrates the accuracy of the models employed for physics-model-based control design. Secondly, a feedback algorithm for q profile control is designed following a FPD approach, and the ability of the controller to achieve and maintain a target q profile evolution is tested in DIII-D high confinement (H-mode) experiments. The controller is shown to be able to effectively control the q profile when β N is relatively close to the target, indicating the need for integrated q profile and β N control to further enhance the ability to achieve robust scenario execution. Furthermore, the ability of an integrated q profile + β N feedback controller to track a desired target is demonstrated through simulation.« less

High Harmonic Fast Wave Damping on an Ion Beam: NSTX and DIII-D Regimes Compared

NASA Astrophysics Data System (ADS)

Pinsker, R. I.; Choi, C. C.; Petty, C. C.; Porkolab, M.; Wilson, J. R.; Murakami, M.; Harvey, R. W.

2004-11-01

Both NSTX and DIII-D use the combination of fast Alfven waves (FW) and neutral beam injection (NBI) for central electron heating and current drive. Damping of the fast wave on the beam ions at moderate to high harmonics (4th--20th) of the beam ion cyclotron frequency represents a loss process. In DIII-D current drive experiments at low density in which 4th and 8th harmonics were compared, damping at the 8th harmonic damping was much weaker than at the 4th [1]. However, recent simulations have predicted that in higher density and higher beam power regimes (of interest to the Advanced Tokamak program) the beam ion absorption will transition to the unmagnetized ion regime, where the damping is significant and essentially independent of harmonic number. In the present work, the transition from magnetized to unmagnetized ion regimes for the NSTX and DIII-D HHFW experiments is studied theoretically, with a combination of simple semi-analytic models and numerical models. \\vspace0.25 em [1] C.C. Petty, et al., Plasma Phys. and Contr. Fusion 43, 1747 (2001).

Destruction of the Last Good Magnetic Surface in Diii-D Usn with Elms and C-Coils Shot 115467 due to Generalized Asymmetries Using Maps

NASA Astrophysics Data System (ADS)

Herrera, A.; Ali, H.; Punjabi, A.

2004-11-01

The unperturbed magnetic topology of DIII-D USN shot 115467 in the absence of ELMs and C-coils is described by the symmetric simple map (SSM) with the map parameter k=0.2623. For this k, the last good surface passes through x=0 and y=0.9995, q_edge=6.48 if six iterations of the SSM are taken to be equivalent to a single toroidal circuit of DIII-D, and the q_edge equals the q_edge in the DIII-D for shot 115467 [1]. The map parameter k represents the effects of the toroidal asymmetries. We study the changes in the last good surface and its destruction as the map parameter k is increased. This work is supported by NASA SHARP program and DE-FG02-02ER54673. [1] H. Ali, A. Punjabi, A. Boozer, and T. Evans, presented at the 31st European Physical Society Plasma Physics Meeting, London, UK, June 29, 2004, paper P2-172.

Confinement improvement in the high poloidal beta regime on DIII-D and application to steady-state H-mode on EAST

NASA Astrophysics Data System (ADS)

Ding, S.; Garofalo, A. M.; Qian, J.; Cui, L.; McClenaghan, J. T.; Pan, C.; Chen, J.; Zhai, X.; McKee, G.; Ren, Q.; Gong, X.; Holcomb, C. T.; Guo, W.; Lao, L.; Ferron, J.; Hyatt, A.; Staebler, G.; Solomon, W.; Du, H.; Zang, Q.; Huang, J.; Wan, B.

2017-05-01

Systematic experimental and modeling investigations on DIII-D show attractive transport properties of fully non-inductive high βp plasmas. Experiments on DIII-D show that the large-radius internal transport barrier (ITB), a key feature providing excellent confinement in the high βp regime, is maintained when the scenario is extended from q95 ˜ 12 to 7 and from rapid to near-zero toroidal rotation. The robustness of confinement versus rotation was predicted by gyrofluid modeling showing dominant neoclassical ion energy transport even without the E × B shear effect. The physics mechanism of turbulence suppression, we found, is the Shafranov shift, which is essential and sets a βp threshold for large-radius ITB formation in the high βp scenario on DIII-D. This is confirmed by two different parameter-scan experiments, one for a βN scan and the other for a q95 scan. They both give the same βp threshold at 1.9 in the experiment. The experimental trend of increasing thermal transport with decreasing βp is consistent with transport modeling. The progress toward the high βp scenario on Experimental Advanced Superconducting Tokamak (EAST) is reported. The very first step of extending the high βp scenario on DIII-D to long pulse on EAST is to establish a long pulse H-mode with ITB on EAST. This paper shows the first 61 s fully non-inductive H-mode with stationary ITB feature and actively cooled ITER-like tungsten divertor in the very recent EAST experiment. The successful use of lower hybrid wave as a key tool to optimize the current profile in the EAST experiment is also introduced. Results show that as the electron density is increased, the fully non-inductive current profile broadens on EAST. The improved understanding and modeling capability are also used to develop advanced scenarios for the China Fusion Engineering Test Reactor. Overall, these results provide encouragement that the high βp regime can be extended to a lower safety factor and very low rotation, providing a potential path to high performance steady state operation in future devices.

Automatic Between-Pulse Analysis of DIII-D Experimental Data Performed Remotely on a Supercomputer at Argonne Leadership Computing Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kostuk, M.; Uram, T. D.; Evans, T.

For the first time, an automatically triggered, between-pulse fusion science analysis code was run on-demand at a remotely located supercomputer at Argonne Leadership Computing Facility (ALCF, Lemont, IL) in support of in-process experiments being performed at DIII-D (San Diego, CA). This represents a new paradigm for combining geographically distant experimental and high performance computing (HPC) facilities to provide enhanced data analysis that is quickly available to researchers. Enhanced analysis improves the understanding of the current pulse, translating into a more efficient use of experimental resources, and to the quality of the resultant science. The analysis code used here, called SURFMN,more » calculates the magnetic structure of the plasma using Fourier transform. Increasing the number of Fourier components provides a more accurate determination of the stochastic boundary layer near the plasma edge by better resolving magnetic islands, but requires 26 minutes to complete using local DIII-D resources, putting it well outside the useful time range for between pulse analysis. These islands relate to confinement and edge localized mode (ELM) suppression, and may be controlled by adjusting coil currents for the next pulse. Argonne has ensured on-demand execution of SURFMN by providing a reserved queue, a specialized service that launches the code after receiving an automatic trigger, and with network access from the worker nodes for data transfer. Runs are executed on 252 cores of ALCF’s Cooley cluster and the data is available locally at DIII-D within three minutes of triggering. The original SURFMN design limits additional improvements with more cores, however our work shows a path forward where codes that benefit from thousands of processors can run between pulses.« less

Automatic Between-Pulse Analysis of DIII-D Experimental Data Performed Remotely on a Supercomputer at Argonne Leadership Computing Facility

DOE PAGES

Kostuk, M.; Uram, T. D.; Evans, T.; ...

2018-02-01

For the first time, an automatically triggered, between-pulse fusion science analysis code was run on-demand at a remotely located supercomputer at Argonne Leadership Computing Facility (ALCF, Lemont, IL) in support of in-process experiments being performed at DIII-D (San Diego, CA). This represents a new paradigm for combining geographically distant experimental and high performance computing (HPC) facilities to provide enhanced data analysis that is quickly available to researchers. Enhanced analysis improves the understanding of the current pulse, translating into a more efficient use of experimental resources, and to the quality of the resultant science. The analysis code used here, called SURFMN,more » calculates the magnetic structure of the plasma using Fourier transform. Increasing the number of Fourier components provides a more accurate determination of the stochastic boundary layer near the plasma edge by better resolving magnetic islands, but requires 26 minutes to complete using local DIII-D resources, putting it well outside the useful time range for between pulse analysis. These islands relate to confinement and edge localized mode (ELM) suppression, and may be controlled by adjusting coil currents for the next pulse. Argonne has ensured on-demand execution of SURFMN by providing a reserved queue, a specialized service that launches the code after receiving an automatic trigger, and with network access from the worker nodes for data transfer. Runs are executed on 252 cores of ALCF’s Cooley cluster and the data is available locally at DIII-D within three minutes of triggering. The original SURFMN design limits additional improvements with more cores, however our work shows a path forward where codes that benefit from thousands of processors can run between pulses.« less

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

NASA Astrophysics Data System (ADS)

Covele, B.; Kotschenreuther, M.; Mahajan, S.; Valanju, P.; Leonard, A.; Watkins, J.; Makowski, M.; Fenstermacher, M.; Si, H.

2017-08-01

The X-divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at 10-20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. However, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. The model also points to carbon radiation as the primary driver of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency for core operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.

Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

NASA Astrophysics Data System (ADS)

Kramer, G. J.; Budny, R.; Nazikian, R.; Heidbrink, W. W.; Kurki-Suonio, T.; Salmi, A.; Schaffer, M. J.; van Zeeland, M. A.; Shinohara, K.; Snipes, J. A.; Spong, D.

2010-11-01

The fast beam-ion confinement in the presence of a scaled mock-up of two Test Blanket Modules (TBM) for ITER was studied in DIII-D. The TBM on DIII-D has four vertically arranged protective carbon tiles with thermocouples placed at the back of each tile. Temperature increases of up to 200^oC were measured for the two tiles closest to the midplane when the TBM fields were present. These measurements agree qualitatively with results from the full orbit-following beam-ion code, SPIRAL, that predict beam-ion losses to be localized on the central two carbon tiles when the TBM fields present. Within the experimental uncertainties no significant change in the fast-ion population was found in the core of these plasmas which is consistent with SPIRAL analysis. These experiments indicate that the TBM fields do not affect the fast-ion confinement in a harmful way which is good news for ITER.

Some Aspects of Advanced Tokamak Modeling in DIII-D

NASA Astrophysics Data System (ADS)

St John, H. E.; Petty, C. C.; Murakami, M.; Kinsey, J. E.

2000-10-01

We extend previous work(M. Murakami, et al., General Atomics Report GA-A23310 (1999).) done on time dependent DIII-D advanced tokamak simulations by introducing theoretical confinement models rather than relying on power balance derived transport coefficients. We explore using NBCD and off axis ECCD together with a self-consistent aligned bootstrap current, driven by the internal transport barrier dynamics generated with the GLF23 confinement model, to shape the hollow current profile and to maintain MHD stable conditions. Our theoretical modeling approach uses measured DIII-D initial conditions to start off the simulations in a smooth consistent manner. This mitigates the troublesome long lived perturbations in the ohmic current profile that is normally caused by inconsistent initial data. To achieve this goal our simulation uses a sequence of time dependent eqdsks generated autonomously by the EFIT MHD equilibrium code in analyzing experimental data to supply the history for the simulation.

Investigation of energy transport in DIII-D high- β P EAST-demonstration discharges with the TGLF turbulent and NEO neoclassical transport models [Investigation of energy transport in DIII-D high- β P EAST-demonstration discharges with turbulent and neoclassical transport models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pan, Chengkang; Staebler, Gary M.; Lao, Lang L.

Here, energy transport analyses of DIII-D high-β P EAST-demonstration discharges have been performed using the TGYRO transport package with TGLF turbulent and NEO neoclassical transport models under the OMFIT integrated modeling framework. Ion energy transport is shown to be dominated by neoclassical transport and ion temperature profiles predicted by TGYRO agree closely with the experimental measured profiles for these high-β P discharges. Ion energy transport is largely insensitive to reductions in the E × B flow shear stabilization. The Shafranov shift is shown to play a role in the suppression of the ion turbulent energy transport below the neoclassical level.more » Electron turbulent energy transport is under-predicted by TGLF and a significant shortfall in the electron energy transport over the whole core plasma is found with TGLF predictions for these high-β P discharges. TGYRO can successfully predict the experimental ion and electron temperature profiles by artificially increasing the saturated turbulence level for ETG driven modes used in TGLF.« less

Investigation of energy transport in DIII-D high- β P EAST-demonstration discharges with the TGLF turbulent and NEO neoclassical transport models [Investigation of energy transport in DIII-D high- β P EAST-demonstration discharges with turbulent and neoclassical transport models

DOE PAGES

Pan, Chengkang; Staebler, Gary M.; Lao, Lang L.; ...

2017-01-11

Here, energy transport analyses of DIII-D high-β P EAST-demonstration discharges have been performed using the TGYRO transport package with TGLF turbulent and NEO neoclassical transport models under the OMFIT integrated modeling framework. Ion energy transport is shown to be dominated by neoclassical transport and ion temperature profiles predicted by TGYRO agree closely with the experimental measured profiles for these high-β P discharges. Ion energy transport is largely insensitive to reductions in the E × B flow shear stabilization. The Shafranov shift is shown to play a role in the suppression of the ion turbulent energy transport below the neoclassical level.more » Electron turbulent energy transport is under-predicted by TGLF and a significant shortfall in the electron energy transport over the whole core plasma is found with TGLF predictions for these high-β P discharges. TGYRO can successfully predict the experimental ion and electron temperature profiles by artificially increasing the saturated turbulence level for ETG driven modes used in TGLF.« less

MHD modeling of a DIII-D low-torque QH-mode discharge and comparison to observations

NASA Astrophysics Data System (ADS)

King, J. R.; Kruger, S. E.; Burrell, K. H.; Chen, X.; Garofalo, A. M.; Groebner, R. J.; Olofsson, K. E. J.; Pankin, A. Y.; Snyder, P. B.

2017-05-01

Extended-MHD modeling of DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] quiescent H-mode (QH-mode) discharges with nonlinear NIMROD [C. R. Sovinec et al., J. Comput. Phys. 195, 355 (2004)] simulations saturates into a turbulent state but does not saturate when the steady-state flow inferred from measurements is not included. This is consistent with the experimental observations of the quiescent regime on DIII-D. The simulation with flow develops into a saturated turbulent state where the nϕ=1 and 2 toroidal modes become dominant through an inverse cascade. Each mode in the range of nϕ=1 -5 is dominant at a different time. Consistent with experimental observations during QH-mode, the simulated state leads to large particle transport relative to the thermal transport. Analysis shows that the amplitude and phase of the density and temperature perturbations differ resulting in greater fluctuation-induced convective particle transport relative to the convective thermal transport. Comparison to magnetic-coil measurements shows that rotation frequencies differ between the simulation and experiment, which indicates that more sophisticated extended-MHD two-fluid modeling is required.

Controlling fusion yield in tokamaks with spin polarized fuel, and feasibility studies on the DIII-D tokamak

DOE PAGES

Pace, D. C.; Lanctot, M. J.; Jackson, G. L.; ...

2015-09-21

The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here asmore » an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. In this study, such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States’ magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress.« less

Destruction of the Last Good Magnetic Surface in Diii-D Usn with Elms and C-Coils Shot 115467 due to C-Coils Using Maps

NASA Astrophysics Data System (ADS)

McCray, A.; Punjabi, A.; Ali, H.

2004-11-01

Unperturbed magnetic topology of DIII-D shot 115467 is described by the symmetric simple map (SSM) with map parameter k=0.2623 [1], then last good surface passes through x=0 and y=0.9995, q_edge=6.48 (same as in shot 115467) if six iterations of SSM are taken to be equivalent to single toroidal circuit of DIII-D. The dipole map (DM) calculates the effects of localized, external high mode numbers magnetic perturbations on motion of field lines. We use dipole map to describe effects of C-coils on field line trajectories in DIII-D. We apply DM after each iteration of SSM, with s=1.0021, x_dipole=1.5617, y_dipole= 0 [1] for shot 115467. We study the changes in the last good surface and its destruction as a function of I_C-coil. This work is supported by NASA SHARP program and DE-FG02-02ER54673. [1] H. Ali, A. Punjabi, A. Boozer, and T. Evans, presented at the 31st European Physical Society Plasma Physics Meeting, London, UK, June 29, 2004, paper P2-172.

Customizable scientific web-portal for DIII-D nuclear fusion experiment

NASA Astrophysics Data System (ADS)

Abla, G.; Kim, E. N.; Schissel, D. P.

2010-04-01

Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

New Frontier Science Campaign on DIII-D launched in 2017

NASA Astrophysics Data System (ADS)

Koepke, M.; Buttery, R.; Carter, T.; Egedal, J.; Forest, C.; Fox, W.; Ji, H.; Howes, G.; Piovesan, P.; Sarff, J.; Skiff, F.; Spong, D.; DIII-D FSE Collaboration Collaboration

2017-10-01

The DIII-D Frontier Science Experiments initiative explores the potential to use the DIII-D tokamak facility to investigate questions of value beyond the usual fusion-energy science mission of DIII-D. The campaign is unique within DOE-SC-FES because the DIII-D tokamak supplied a multi-day-shot platform for non-fusion-energy-motivated research for the first time. All selected FSE campaign projects competed on the basis of potential intellectual impact and on the degree to which the ability to achieve success as a transformational advance relied on the capabilities of DIII-D. The motivation of the following FSE projects, as well as the selection process, will be summarized (1) Self-organization of Unstable Flux Ropes: Universal Structures in Space/Astrophysical Plasmas (2) Impact of Magnetic Perturbations on Turbulence: Zonal Flow Interactions and Saturation (3) Interaction of Alfvén/whistler fluctuations and Runaway Electrons (4) Self-consistent chaos in magnetic field dynamics These basic-plasma experiments, conducted in collaboration with the DIII-D team, were carried out during 5 shot days in FY2017. Additional days are earmarked in FY2018. Future studies with additional FSE-community members are envisioned. Opportunities exist to piggy back with DIII-D research A proper solicitation and peer review would be appropriate going forward if this activity on DIII-D continues Funding from U.S. DOE is gratefully acknowledged.

Gyrokinetic simulation of edge blobs and divertor heat-load footprint

NASA Astrophysics Data System (ADS)

Chang, C. S.; Ku, S.; Hager, R.; Churchill, M.; D'Azevedo, E.; Worley, P.

2015-11-01

Gyrokinetic study of divertor heat-load width Lq has been performed using the edge gyrokinetic code XGC1. Both neoclassical and electrostatic turbulence physics are self-consistently included in the simulation with fully nonlinear Fokker-Planck collision operation and neutral recycling. Gyrokinetic ions and drift kinetic electrons constitute the plasma in realistic magnetic separatrix geometry. The electron density fluctuations from nonlinear turbulence form blobs, as similarly seen in the experiments. DIII-D and NSTX geometries have been used to represent today's conventional and tight aspect ratio tokamaks. XGC1 shows that the ion neoclassical orbit dynamics dominates over the blob physics in setting Lq in the sample DIII-D and NSTX plasmas, re-discovering the experimentally observed 1/Ip type scaling. Magnitude of Lq is in the right ballpark, too, in comparison with experimental data. However, in an ITER standard plasma, XGC1 shows that the negligible neoclassical orbit excursion effect makes the blob dynamics to dominate Lq. Differently from Lq 1mm (when mapped back to outboard midplane) as was predicted by simple-minded extrapolation from the present-day data, XGC1 shows that Lq in ITER is about 1 cm that is somewhat smaller than the average blob size. Supported by US DOE and the INCITE program.

Design and Control of Small Neutral Beam Arc Chamber for Investigations of DIII-D Neutral Beam Failure During Helium Operation

NASA Astrophysics Data System (ADS)

Fremlin, Carl; Beckers, Jasper; Crowley, Brendan; Rauch, Joseph; Scoville, Jim

2017-10-01

The Neutral Beam system on the DIII-D tokamak consists of eight ion sources using the Common Long Pulse Source (CLPS) design. During helium operation, desired for research regarding the ITER pre-nuclear phase, it has been observed that the ion source arc chamber performance steadily deteriorates, eventually failing due to electrical breakdown of the insulation. A significant investment of manpower and time is required for repairs. To study the cause of failure a small analogue of the DIII-D neutral beam arc chamber has been constructed. This poster presents the design and analysis of the arc chamber including the PLC based operational control system for the experiment, analysis of the magnetic confinement and details of the diagnostic suite. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

Investigation of torque generated by Test Blanket Module mock-up in DIII-D

NASA Astrophysics Data System (ADS)

Salmi, A.; Tala, T.; Lanctot, M.; Degrassie, J. S.; Paz-Soldan, C.; Logan, N.; Solomon, W. M.; Grierson, B. A.

2015-11-01

Experiments at DIII-D have investigated the scaling of Test Blanket Module (TBM) torque with plasma pressure and collisionality by performing dimensionless parameter scans. In each configuration, neutral beam torque modulation and TBM torque modulation were sequentially applied to allow experimental characterization of the TBM generated torque and the underlying transport. Calculations of the neoclassical toroidal viscosity (NTV) torque with PENT code of these plasmas find that TBM torque is strongly edge localized while the tentative experimental analysis indicates a more radially broad TBM torque profile. Both the experimental and PENT results will be elaborated and experimental TBM torque scaling with pressure and collisionality presented. Experimental validation of existing plasma response and NTV torque models is an important step toward understanding the impact of magnetic field ripple on plasma rotation, and for predicting the required compensation fields. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC02-09CH11466.

An area-preserving mapping in natural canonical coordinates for magnetic field line trajectories in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh

2009-11-01

The new approach of integrating magnetic field line trajectories in natural canonical coordinates (Punjabi and Ali 2008 Phys. Plasmas 15 122502) in divertor tokamaks is used for the DIII-D tokamak (Luxon and Davis1985 Fusion Technol. 8 441). The equilibrium EFIT data (Evans et al 2004 Phys. Rev. Lett. 92 235003, Lao et al 2005 Fusion Sci. Technol. 48 968) for the DIII-D tokamak shot 115467 at 3000 ms is used to construct the equilibrium generating function (EGF) for the DIII-D in natural canonical coordinates. The EGF gives quite an accurate representation of the closed and open equilibrium magnetic surfaces near the separatrix, the separatrix, the position of the X-point and the poloidal magnetic flux inside the ideal separatrix in the DIII-D. The equilibrium safety factor q from the EGF is somewhat smaller than the DIII-D EFIT q profile. The equilibrium safety factor is calculated from EGF as described in the previous paper (Punjabi and Ali 2008 Phys. Plasmas 15 122502). Here the safety factor for the open surfaces in the DIII-D is calculated. A canonical transformation is used to construct a symplectic mapping for magnetic field line trajectories in the DIII-D in natural canonical coordinates. The map is explored in more detail in this work, and is used to calculate field line trajectories in the DIII-D tokamak. The continuous analogue of the map does not distort the DIII-D magnetic surfaces in different toroidal planes between successive iterations of the map. The map parameter k can represent effects of magnetic asymmetries in the DIII-D. These effects in the DIII-D are illustrated. The DIII-D map is then used to calculate stochastic broadening of the ideal separatrix from the topological noise and field errors, the low mn, the high mn and peeling-ballooning magnetic perturbations in the DIII-D. The width of the stochastic layer scales as 1/2 power of amplitude with a maximum deviation of 6% from the Boozer-Rechester scaling (Boozer and Rechester 1978 Phys. Fluids 21 682). The loss of poloidal flux scales linearly with the amplitude of perturbation with a maximum deviation of 10% from linearity. Perturbations with higher mode numbers result in higher stochasticity. The higher the complexity and coupling in the equilibrium magnetic geometry, the closer is the scaling to the Boozer-Rechester scaling of width. The comparison of the EGF for the simple map (Punjabi et al 1992 Phys. Rev. Lett. 69 3322) with that of the DIII-D shows that the more complex the magnetic geometry and the more coupling of modes in equilibrium, the more robust or resilient is the system against the chaos-inducing, symmetry-breaking perturbations.

Installation and Testing of ITER Integrated Modeling and Analysis Suite (IMAS) on DIII-D

NASA Astrophysics Data System (ADS)

Lao, L.; Kostuk, M.; Meneghini, O.; Smith, S.; Staebler, G.; Kalling, R.; Pinches, S.

2017-10-01

A critical objective of the ITER Integrated Modeling Program is the development of IMAS to support ITER plasma operation and research activities. An IMAS framework has been established based on the earlier work carried out within the EU. It consists of a physics data model and a workflow engine. The data model is capable of representing both simulation and experimental data and is applicable to ITER and other devices. IMAS has been successfully installed on a local DIII-D server using a flexible installer capable of managing the core data access tools (Access Layer and Data Dictionary) and optionally the Kepler workflow engine and coupling tools. A general adaptor for OMFIT (a workflow engine) is being built for adaptation of any analysis code to IMAS using a new IMAS universal access layer (UAL) interface developed from an existing OMFIT EU Integrated Tokamak Modeling UAL. Ongoing work includes development of a general adaptor for EFIT and TGLF based on this new UAL that can be readily extended for other physics codes within OMFIT. Work supported by US DOE under DE-FC02-04ER54698.

Flow reversal, convection, and modeling in the DIII-D divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

1998-12-01

Measurements of the parallel Mach number of background plasma in the DIII-D tokamak divertor [M. A. Mahdavi {ital et al.} in {ital Proceedings, 16th International Conference}, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997) Vol. I, p. 397] were performed using a fast scanning Mach probe. The parallel particle flow shows evidence of complex behavior such as reverse flow, i.e., flow away from the target plate, stagnant flow, and large scale convection. For detached discharges, measurements confirm predictions of convective flow towards the divertor target plate at near sound speed over large regions in the divertor. The resulting convected heatmore » flux is a dominant heat transport mechanism in the divertor. For attached discharges with high recycling, particle flow reversal in a thin region at or near the outer separatrix, thereby confirming the existence of a mechanism by which impurities can be transported away from the divertor target plates. Modeling results from the two-dimensional fluid code UEDGE [G. D. Porter and the DIII-D Team, {open_quotes}Divertor characterization experiments and modelling in DIII-D,{close_quotes} in {ital Proceedings of the 23rd European Conference on Controlled Fusion and Plasma Physics}, 24{endash}28 June 1996, Kiev, Ukraine (European Physical Society, Petit-Lancy, Switzerland, 1996), Vol. 20C, Part II, p. 699] can reproduce the main features of the experimental observations. {copyright} {ital 1998 American Institute of Physics.}« less

Testing an H-mode Pedestal Model Using DIII-D Data

NASA Astrophysics Data System (ADS)

Kritz, A. H.; Onjun, T.; Bateman, G.; Guzdar, P. N.; Mahajan, S. M.; Osborne, T.

2004-11-01

Tests against experimental data are carried out for a model of the pedestal at the edge of H-mode plasmas based on double-Beltrami solutions of the two-fluid Hall-MHD equations for the interaction of the magnetic and velocity fields.(S.M. Mahajan and Z. Yoshida, PRL 81 (1998) 4863, Phys. Plasmas 7 (2000) 635.) The width and height of the pedestal predicted by the model are tested against experimental data from the DIII-D tokamak. The model for the pedestal width, which has a particularly simple form, namely, inversely proportional to the square root of the density, does not appear to capture the parameter dependence of the experimental data. When the model for the pedestal temperature is rescaled to optimize agreement with data, the RMS error is found to be comparable with the RMS error found using other pedestal models.(T. Onjun, G. Bateman, A.H. Kritz, G. Hammett, Phys. Plasmas 9 (2002) 5018.)

Modeling of material erosion and redeposition for dedicated DiMES experiments on DIII-D

NASA Astrophysics Data System (ADS)

Ding, R.; Abrams, T.; Chrobak, C. P.; Guo, H. Y.; Snyder, P. B.; Chan, V. S.; Rudakov, D. L.; Stangeby, P. C.; Elder, J. D.; Tskhakaya, D.; Wampler, W. R.; Kirschner, A.; McLean, A. G.

2015-11-01

Erosion and redeposition of plasma facing materials is a key issue for high-power, long pulse tokamak operation. A series of experiments has been carried out on DIII-D in which well-characterized samples of different materials were exposed to divertor plasma using DiMES. Such experiments provide a good benchmark for PMI codes, such as ERO. It was found that the erosion and redeposition are strongly determined by the impurity content in the plasma and sheath properties near the surface. The principal experimental results (net erosion rate and profile, net/gross erosion ratio) are reproduced by ERO simulations to within the uncertainties, indicating that the controlling physics has likely been identified. New techniques suggested by modeling such as external biasing and local gas injection for suppressing material erosion are planned to be tested in DiMES/DIII-D experiments. Work supported by US DOE DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC52-07NA27344.

Control of plasma stored energy for burn control using DIII-D in-vessel coils

DOE PAGES

Hawryluk, Richard J.; Eidietis, Nicholas W.; Grierson, Brian A.; ...

2015-04-09

A new approach has been experimentally demonstrated to control the stored energy by applying a non-axisymmetric magnetic field using the DIII-D in-vessel coils to modify the energy confinement time. In future burning plasma experiments as well as magnetic fusion energy power plants, various concepts have been proposed to control the fusion power. The fusion power in a power plant operating at high gain can be related to the plasma stored energy and hence, is a strong function of the energy confinement time. Thus, an actuator that modifies the confinement time can be used to adjust the fusion power. In relativelymore » low collisionality DIII-D discharges, the application of nonaxisymmetric magnetic fields results in a decrease in confinement time and density pumpout. Furthermore, gas puffing was used to compensate the density pumpout in the pedestal while control of the stored energy was demonstrated by the application of non-axisymmetric fields.« less

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

DOE PAGES

Covele, Brent; Kotschenreuther, M.; Mahajan, S.; ...

2017-06-23

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Covele, Brent; Kotschenreuther, M.; Mahajan, S.

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Investigation of Neutral Beam Arc Chamber Failure During Helium Operations at DIII-D

NASA Astrophysics Data System (ADS)

Beckers, Jasper; Crowley, Brendan; Scoville, J. T.; Jaspers, Roger; Sobota, Ana

2017-10-01

The Neutral Beam system on the DIII-D tokamak consists of eight ion sources using the Common Long Pulse Source (CLPS) design. During helium operation, desired for research regarding the ITER pre-nuclear phase, it has been observed that the ion source arc chamber performance steadily deteriorates, eventually failing due to electrical breakdown across the insulation. This poster presents the details and preliminary results of an experimental effort to replicate the problem in a bench top ion source with similar plasma parameters. The initial aim of the experiment is to test the hypothesis that during helium operation there is increased tungsten evaporation and sputtering due to ion bombardment of the hot cathodes, leading to the deposition of filament material on the insulation and subsequent short circuits. Ultimately the aim of the experiment is to find methods to ameliorate the problems associated with helium operation at DIII-D. Work supported by U.S. DOE under DE-FC02-04ER54698.

Tangles of the ideal separatrix from low mn perturbation in the DIII-D

NASA Astrophysics Data System (ADS)

Goss, Talisa; Crank, Willie; Ali, Halima; Punjabi, Alkesh

2010-11-01

The equilibrium EFIT data for the DIII-D shot 115467 at 3000 ms is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [A. Punjabi, and H. Ali, Phys. Plasmas 15, 122502 (2008); A. Punjabi, Nucl. Fusion 49, 115020 (2009)]. The generating function represents the axisymmetric magnetic geometry and the topology of the DIII-D shot very accurately. A symplectic map for field line trajectories in the natural canonical coordinates in the DIII-D is constructed. We call this map the DIII-D map. The natural canonical coordinates can be readily inverted to physical coordinates (R,φ,Z). Low mn magnetic perturbation with mode numbers (m,n)=(1,1)+(1,-1) is added to the generating function of the map. The amplitude for the low mn perturbation is chosen to be 6X10-4, which is the expected value of the amplitude in tokamaks. The forward and backward DIII-D maps with low mn perturbation are used to calculate the tangles of the ideal separatrix from low mn perturbation in the DIII-D. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

Comparing Turbulence Simulation with Experiment in DIII-D

NASA Astrophysics Data System (ADS)

Ross, D. W.; Bravenec, R. V.; Dorland, W.; Beer, M. A.; Hammett, G. W.; McKee, G. R.; Murakami, M.; Jackson, G. L.

2000-10-01

Gyrofluid simulations of DIII-D discharges with the GRYFFIN code(D. W. Ross et al.), Transport Task Force Workshop, Burlington, VT, (2000). are compared with transport and fluctuation measurements. The evolution of confinement-improved discharges(G. R. McKee et al.), Phys. Plasmas 7, 1870 (200) is studied at early times following impurity injection, when EXB rotational shear plays a small role. The ion thermal transport predicted by the code is consistent with the experimental values. Experimentally, changes in density profiles resulting from the injection of neon, lead to reduction in fluctuation levels and transport following the injection. This triggers subsequent changes in the shearing rate that further reduce the turbulence.(M. Murakami et al.), European Physical Society, Budapest (2000); M. Murakami et al., this meeting. Estimated uncertainties in the plasma profiles, however, make it difficult to simulate these reductions with the code. These cases will also be studied with the GS2 gyrokinetic code.

2D microwave imaging reflectometer electronics.

PubMed

Spear, A G; Domier, C W; Hu, X; Muscatello, C M; Ren, X; Tobias, B J; Luhmann, N C

2014-11-01

A 2D microwave imaging reflectometer system has been developed to visualize electron density fluctuations on the DIII-D tokamak. Simultaneously illuminated at four probe frequencies, large aperture optics image reflections from four density-dependent cutoff surfaces in the plasma over an extended region of the DIII-D plasma. Localized density fluctuations in the vicinity of the plasma cutoff surfaces modulate the plasma reflections, yielding a 2D image of electron density fluctuations. Details are presented of the receiver down conversion electronics that generate the in-phase (I) and quadrature (Q) reflectometer signals from which 2D density fluctuation data are obtained. Also presented are details on the control system and backplane used to manage the electronics as well as an introduction to the computer based control program.

Destruction of the Last Good Surface in Diii-D Usn with ELMs and C-Coils Shot 115467 due to ELMs and C-Coils Using Maps

NASA Astrophysics Data System (ADS)

Sherrow, K.; Punjabi, A.; Ali, H.

2004-11-01

Unperturbed magnetic topology of DIII-D shot 115467 is described by the symmetric simple map (SSM) with parameter k=0.2623, then q_edge=6.48 (as in shot 115467) if six iterations of SSM are taken to be equivalent to single toroidal circuit of DIII-D [1]. Low mn map (LM) calculates effects of m=1, n=+1,-1 modes on trajectories of field lines. We use LM with amplitude ɛ=6X10-4 (value expected in modern divertor tokamaks) to describe effects of ELMs. With ELMs, last good surface passes through x=0, y=0.98375. We use dipole map (DM) to represent effects of C-coils. We apply DM after each iteration of SSM. We use s=1.0021, x_dipole=1.5617, y_dipole= 0 for DIII-D shot 115467 [1]. We study changes in the last good surface and its destruction as function of I_C-coil with fixed ɛ = 6X10-4. This work is supported by NASA SHARP program and DE-FG02-02ER54673. [1] H. Ali, A. Punjabi, A. Boozer, and T. Evans, 31st EPS Plasma Phys Mtg, London, UK, June 29, 2004, paper P2-172.

Integrated modeling of high βN steady state scenario on DIII-D

DOE PAGES

Park, Jin Myung; Ferron, J. R.; Holcomb, Christopher T.; ...

2018-01-10

Theory-based integrated modeling validated against DIII-D experiments predicts that fully non-inductive DIII-D operation with β N > 4.5 is possible with certain upgrades. IPS-FASTRAN is a new iterative numerical procedure that integrates models of core transport, edge pedestal, equilibrium, stability, heating, and current drive self-consistently to find steady-state ( d/dt = 0) solutions and reproduces most features of DIII-D high β N discharges with a stationary current profile. Projecting forward to scenarios possible on DIII-D with future upgrades, the high q min > 2 scenario achieves stable operation at β N as high as 5 by using a very broadmore » current density profile to improve the ideal-wall stabilization of low- n instabilities along with confinement enhancement from low magnetic shear. This modeling guides the necessary upgrades of the heating and current drive system to realize reactor-relevant high β N steady-state scenarios on DIII-D by simultaneous optimization of the current and pressure profiles.« less

Integrated modeling of high βN steady state scenario on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, Jin Myung; Ferron, J. R.; Holcomb, Christopher T.

Theory-based integrated modeling validated against DIII-D experiments predicts that fully non-inductive DIII-D operation with β N > 4.5 is possible with certain upgrades. IPS-FASTRAN is a new iterative numerical procedure that integrates models of core transport, edge pedestal, equilibrium, stability, heating, and current drive self-consistently to find steady-state ( d/dt = 0) solutions and reproduces most features of DIII-D high β N discharges with a stationary current profile. Projecting forward to scenarios possible on DIII-D with future upgrades, the high q min > 2 scenario achieves stable operation at β N as high as 5 by using a very broadmore » current density profile to improve the ideal-wall stabilization of low- n instabilities along with confinement enhancement from low magnetic shear. This modeling guides the necessary upgrades of the heating and current drive system to realize reactor-relevant high β N steady-state scenarios on DIII-D by simultaneous optimization of the current and pressure profiles.« less

Integrated modeling of high βN steady state scenario on DIII-D

NASA Astrophysics Data System (ADS)

Park, J. M.; Ferron, J. R.; Holcomb, C. T.; Buttery, R. J.; Solomon, W. M.; Batchelor, D. B.; Elwasif, W.; Green, D. L.; Kim, K.; Meneghini, O.; Murakami, M.; Snyder, P. B.

2018-01-01

Theory-based integrated modeling validated against DIII-D experiments predicts that fully non-inductive DIII-D operation with βN > 4.5 is possible with certain upgrades. IPS-FASTRAN is a new iterative numerical procedure that integrates models of core transport, edge pedestal, equilibrium, stability, heating, and current drive self-consistently to find steady-state (d/dt = 0) solutions and reproduces most features of DIII-D high βN discharges with a stationary current profile. Projecting forward to scenarios possible on DIII-D with future upgrades, the high qmin > 2 scenario achieves stable operation at βN as high as 5 by using a very broad current density profile to improve the ideal-wall stabilization of low-n instabilities along with confinement enhancement from low magnetic shear. This modeling guides the necessary upgrades of the heating and current drive system to realize reactor-relevant high βN steady-state scenarios on DIII-D by simultaneous optimization of the current and pressure profiles.

Effect of heating scheme on SOL width in DIII-D and EAST

DOE PAGES

Wang, L.; Makowski, M. A.; Guo, H. Y.; ...

2017-03-10

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broadermore » SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. Finally, the joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas.« less

Effect of heating scheme on SOL width in DIII-D and EAST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, L.; Makowski, M. A.; Guo, H. Y.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broadermore » SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. Finally, the joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas.« less

Shafranov shift bifurcation of turbulent transport in the high βp scenario on DIII-D

NASA Astrophysics Data System (ADS)

McClenaghan, J.; Garofalo, A. M.; Staebler, G. M.; Qian, J.; Gong, X.; Ding, S. Y.

2017-10-01

The Shafranov shift stabilization of turbulence creates a bifurcation in transport leading to formation of a large radius internal transport barrier (ITB) in the high βp scenario on DIII-D. The high βp scenario exhibits high confinement at high βN and high bootstrap fraction in the absence of rapid rotation or negative central shear. Spontaneous formation of an ITB at fixed βN is examined. The energy confinement improves following formation of the ITB. The improvement is associated with a decrease in the minimum mid-radius characteristic turbulence parameter associated with the Shafranov shift: α - s , where α =q2 Rdβ / dρ is a measure of the Shafranov shift, and s is the magnetic shear. After ITB formation, α - s > 0 within region of ITB and α - s < 0 outside the ITB. Before ITB formation, α - s < 0 throughout the entire core. TGLF transport simulations show a bifurcation of the transport depending on the electron pressure gradient scale length. Before ITB formation, the experimental scale length is on the high-transport side of bifurcation. After ITB formation, experimental scale length is on the low-transport side of the bifurcation in the region of the ITB. Work supported in part by the US Department of Energy, Office of Science, Office of Fusion Energy Sciences DE-FC02-04ER54698 (Cooperative Agreement #DE-SC0010685), and by the National Magnetic Confinement Fusion Program of China (No. 2015GB102002, 2015GB10.

Microinstability properties of negative magnetic shear discharges in the Tokamak Fusion Test Reactor and DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rewoldt, G.; Tang, W.M.; Lao, L.L.

1997-03-01

The microinstability properties of discharges with negative (reversed) magnetic shear in the Tokamak Fusion Test Reactor (TFTR) and DIII-D experiments with and without confinement transitions are investigated. A comprehensive kinetic linear eigenmode calculation employing the ballooning representation is employed with experimentally measured profile data, and using the corresponding numerically computed magnetohydrodynamic (MHD) equilibria. The instability considered is the toroidal drift mode (trapped-electron-{eta}{sub i} mode). A variety of physical effects associated with differing q-profiles are explained. In addition, different negative magnetic shear discharges at different times in the discharge for TFTR and DIII-D are analyzed. The effects of sheared toroidal rotation,more » using data from direct spectroscopic measurements for carbon, are analyzed using comparisons with results from a two-dimensional calculation. Comparisons are also made for nonlinear stabilization associated with shear in E{sub r}/RB{sub {theta}}. The relative importance of changes in different profiles (density, temperature, q, rotation, etc.) on the linear growth rates is considered.« less

A novel technique for real-time estimation of edge pedestal density gradients via reflectometer time delay data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeng, L., E-mail: zeng@fusion.gat.com; Doyle, E. J.; Rhodes, T. L.

2016-11-15

A new model-based technique for fast estimation of the pedestal electron density gradient has been developed. The technique uses ordinary mode polarization profile reflectometer time delay data and does not require direct profile inversion. Because of its simple data processing, the technique can be readily implemented via a Field-Programmable Gate Array, so as to provide a real-time density gradient estimate, suitable for use in plasma control systems such as envisioned for ITER, and possibly for DIII-D and Experimental Advanced Superconducting Tokamak. The method is based on a simple edge plasma model with a linear pedestal density gradient and low scrape-off-layermore » density. By measuring reflectometer time delays for three adjacent frequencies, the pedestal density gradient can be estimated analytically via the new approach. Using existing DIII-D profile reflectometer data, the estimated density gradients obtained from the new technique are found to be in good agreement with the actual density gradients for a number of dynamic DIII-D plasma conditions.« less

Modeling of detachment experiments at DIII-D

DOE PAGES

Canik, John M.; Briesemeister, Alexis R.; Lasnier, C. J.; ...

2014-11-26

Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. Lastly, these results suggest that the model is adequate to predict the divertor temperature, provided thatmore » the discrepancy in radiated power level can be resolved.« less

The spectral basis of optimal error field correction on DIII-D

DOE PAGES

Paz-Soldan, Carlos A.; Buttery, Richard J.; Garofalo, Andrea M.; ...

2014-04-28

Here, experimental optimum error field correction (EFC) currents found in a wide breadth of dedicated experiments on DIII-D are shown to be consistent with the currents required to null the poloidal harmonics of the vacuum field which drive the kink mode near the plasma edge. This allows the identification of empirical metrics which predict optimal EFC currents with accuracy comparable to that of first- principles modeling which includes the ideal plasma response. While further metric refinements are desirable, this work suggests optimal EFC currents can be effectively fed-forward based purely on knowledge of the vacuum error field and basic equilibriummore » properties which are routinely calculated in real-time.« less

Symplectic homoclinic tangles of the ideal separatrix of the DIII-D from type I ELMs

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima

2012-10-01

The ideal separatrix of the divertor tokamaks is a degenerate manifold where both the stable and unstable manifolds coincide. Non-axisymmetric magnetic perturbations remove the degeneracy; and split the separatrix manifold. This creates an extremely complex topological structure, called homoclinic tangles. The unstable manifold intersects the stable manifold and creates alternating inner and outer lobes at successive homoclinic points. The Hamiltonian system must preserve the symplectic topological invariance, and this controls the size and radial extent of the lobes. Very recently, lobes near the X-point have been experimentally observed in MAST [A. Kirk et al, PRL 108, 255003 (2012)]. We have used the DIII-D map [A. Punjabi, NF 49, 115020 (2009)] to calculate symplectic homoclinic tangles of the ideal separatrix of the DIII-D from the type I ELMs represented by the peeling-ballooning modes (m,n)=(30,10)+(40,10). The DIII-D map is symplectic, accurate, and is in natural canonical coordinates which are invertible to physical coordinates [A. Punjabi and H. Ali, POP 15, 122502 (2008)]. To our knowledge, we are the first to symplectically calculate these tangles in physical space. Homoclinic tangles of separatrix can cause radial displacement of mobile passing electrons and create sheared radial electric fields and currents, resulting in radial flows, drifts, differential spinning, and reduction in turbulence, and other effects. This work is supported by the grants DE-FG02-01ER54624 and DE-FG02-04ER54793.

Comparison of the numerical modelling and experimental measurements of DIII-D separatrix displacements during H-modes with resonant magnetic perturbations

DOE PAGES

Orlov, Dmitry M.; Moyer, Richard A.; Evans, Todd E.; ...

2014-08-15

Numerical modeling of the plasma boundary position and its displacement due to external magnetic perturbations in DIII-D low-collisionality H-mode discharges is presented. The results of the vacuum model are compared to the experimental measurements for boundary displacements including Thomson scattering electron temperature T e, charge exchange recombination spectroscopy, beam emission spectroscopy, soft x-ray, and divertor Langmuir probe measurements. Magnetically perturbed discharges with toroidal mode number n=2 and n=3 are studied. It is shown that the vacuum model predictions agree well with the measurements above and below the midplane, and disagree at the outer midplane in discharges where significant kink amplificationmore » is present. Lastly, the role of the plasma response is studied using the two-fluid MHD code M3D-C 1, and the results are compared to the vacuum model showing that the plasma response model underestimates the boundary displacements.« less

Analysis of Particle Transport in DIII-D H-mode Plasma with a Generalized Pinch-Diffusion Model

NASA Astrophysics Data System (ADS)

Owen, L. W.; Stacey, W. M.; Groebner, R. J.; Callen, J. D.; Bonnin, X.

2009-11-01

Interpretative analyses of particle transport in the pedestal region of H-mode plasmas typically yield diffusion coefficients that are very small (<0.1 m^2/s) in the steep gradient region when a purely diffusive particle flux is fitted to the experimental density gradients. Previous evaluation of the particle and momentum balance equations using the experimental data indicated that the pedestal profiles are consistent with transport described by a pinch-diffusion particle flux relation [1]. This type of model is used to calculate the diffusion coefficient and pinch velocity in the core for an inter-ELM H-mode plasma in the DIII-D discharge 98889. Full-plasma SOPLS simulations using neutral beam particle and energy sources from ONETWO calculations and the model transport coefficients show good agreement with the measured density pedestal profile. 6pt [1] W.M. Stacey and R.J. Groebner, Phys. Plasmas 12, 042504 (2005).

Pedestal-to-Wall 3D Fluid Transport Simulations on DIII-D

DOE PAGES

Lore, Jeremy D.; Wolfmeister, Alexis Briesemeister; Ferraro, Nathaniel M.; ...

2017-03-30

The 3D fluid-plasma edge transport code EMC3-EIRENE is used to test several magnetic field models with and without plasma response against DIII-D experimental data for even and odd-parity n=3 magnetic field perturbations. The field models include ideal and extended MHD equilibria, and the vacuum approximation. Plasma response is required to reduce the stochasticity in the pedestal region for even-parity fields, however too much screening suppresses the measured splitting of the downstream T e profile. Odd-parity perturbations result in weak tearing and only small additional peaks in the downstream measurements. In this case plasma response is required to increase the sizemore » of the lobe structure. Finally, no single model is able to simultaneously reproduce the upstream and downstream characteristics for both odd and even-parity perturbations.« less

An accurate symplectic calculation of the inboard magnetic footprint from statistical topological noise and field errors in the DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Punjabi, Alkesh; Ali, Halima

2011-02-15

Any canonical transformation of Hamiltonian equations is symplectic, and any area-preserving transformation in 2D is a symplectomorphism. Based on these, a discrete symplectic map and its continuous symplectic analog are derived for forward magnetic field line trajectories in natural canonical coordinates. The unperturbed axisymmetric Hamiltonian for magnetic field lines is constructed from the experimental data in the DIII-D [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The equilibrium Hamiltonian is a highly accurate, analytic, and realistic representation of the magnetic geometry of the DIII-D. These symplectic mathematical maps are used to calculate the magnetic footprint onmore » the inboard collector plate in the DIII-D. Internal statistical topological noise and field errors are irreducible and ubiquitous in magnetic confinement schemes for fusion. It is important to know the stochasticity and magnetic footprint from noise and error fields. The estimates of the spectrum and mode amplitudes of the spatial topological noise and magnetic errors in the DIII-D are used as magnetic perturbation. The discrete and continuous symplectic maps are used to calculate the magnetic footprint on the inboard collector plate of the DIII-D by inverting the natural coordinates to physical coordinates. The combination of highly accurate equilibrium generating function, natural canonical coordinates, symplecticity, and small step-size together gives a very accurate calculation of magnetic footprint. Radial variation of magnetic perturbation and the response of plasma to perturbation are not included. The inboard footprint from noise and errors are dominated by m=3, n=1 mode. The footprint is in the form of a toroidally winding helical strip. The width of stochastic layer scales as (1/2) power of amplitude. The area of footprint scales as first power of amplitude. The physical parameters such as toroidal angle, length, and poloidal angle covered before striking, and the safety factor all have fractal structure. The average field diffusion near the X-point for lines that strike and that do not strike differs by about three to four orders of magnitude. The magnetic footprint gives the maximal bounds on size and heat flux density on collector plate.« less

DIII-D Electron Cyclotron Heating System Status and Upgrades

DOE PAGES

Cengher, Mirela; Lohr, John; Gorelov, Yuri; ...

2016-06-23

The DIII-D Electron Cyclotron Heating (ECH) system consists of six 110 GHz gyrotrons with corrugated coaxial 31.75 mm waveguide transmission lines and steerable launching mirrors. The system has been gradually updated, leading to increased experimental flexibility and a high system reliability of 91% in the past year. Operationally, the gyrotrons can generate up to a total of 4.8 MW of rf power for pulses up to 5 seconds. The maximum ECH energy injected into the DIII-D is 16.6 MJ. The HE1,1 mode content is over 85% for all the lines, and the transmission coefficient is better than -1.1 dB formore » all the transmission lines, close to the theoretical value. A new depressed collector gyrotron was recently installed and was injecting up to 640 kW of power into the plasma during 2014-2015 tokamak operations. Four dual waveguide launchers, which can steer the RF beams ±20 degrees poloidally and toroidally, are used for real-time neoclassical tearing mode control and suppression. The launchers now have increased poloidal scanning speed and beam positioning accuracy of ~±2 mm at the plasma center. Two more gyrotrons are expected to be installed and operational in 2015- 2016. The first is a repaired 110 GHz, 1 MW gyrotron that had a gun failure after more than 11 years of operation at DIII-D. The second is a newly designed depressed collector tube in the 1.5 MW class, operating at 117.5 GHz, manufactured by Communications and Power Industries (CPI).« less

Magnetic shear effects on plasma transport and turbulence at high electron to ion temperature ratio in DIII-D and JT-60U plasmas

NASA Astrophysics Data System (ADS)

Yoshida, M.; McKee, G. R.; Murakami, M.; Grierson, B. A.; Nakata, M.; Davis, E. M.; Marinoni, A.; Ono, M.; Rhodes, T. L.; Sung, C.; Schmitz, L.; Petty, C. C.; Ferron, J. R.; Turco, F.; Garofalo, A. M.; Holcomb, C. T.; Collins, C. M.; Solomon, W. M.

2017-05-01

Negative magnetic shear has been demonstrated in DIII-D and JT-60U to mitigate the confinement degradation typically observed with increasing the electron to ion temperature ratio (T e/T i). In recent experiments in DIII-D negative central magnetic shear (NCS) discharges, the thermal transport in the internal transport barrier formed around the radius of the minimum safety factor (q min) remained almost constant and modestly increased in the region outside of q min compared to the positive shear (PS) case, when T e/T i increased from about 0.8 to 1.1 through electron cyclotron heating (ECH). The benefit of NCS extending into the region outside of q min can be explained by the lower magnetic shear in the NCS plasma over the plasma radius relative to the PS plasma. Reduced confinement degradation at high T e/T i with NCS plasmas was commonly observed in DIII-D and JT-60U. The mechanism of the different transport responses between the NCS and PS plasmas has been assessed in terms of fluctuation measurements and gyrokinetic simulations in DIII-D; NCS gave a smaller rise in the low-wavenumber broadband turbulent fluctuations with the increase in T e/T i compared with the PS case. This is consistent with gyrokinetic simulations, which show a smaller rise in the growth rates of the ion temperature gradient mode in the NCS plasmas, with increasing T e/T i. Gyrokinetic simulations also showed a change in the stability of the electron modes with ECH applied, consistent with higher-wavenumber fluctuation measurements, although more detailed simulations are needed to give a quantitative explanation for the experimental observations. Control of q-profile and magnetic shear will allow confinement improvement in future machines with dominant electron heating.

Extending the validation of multi-mode model for anomalous transport to high beta poloidal tokamak scenario in DIII-D

NASA Astrophysics Data System (ADS)

Pankin, A. Y.; Kritz, A. H.; Rafiq, T.; Garofalo, A. M.; Holod, I.; Weiland, J.

2018-05-01

The Multi-Mode Model (MMM7.1) for anomalous transport is tested in predictive modeling of temperature profiles of a high beta poloidal DIII-D discharge. This new H-mode plasma regime, with high beta poloidal and high bootstrap currents, has been studied in DIII-D tokamak discharges [A. Garofalo et al., Nucl. Fusion 55, 123025 (2015)]. The role of instabilities that can drive the anomalous transport described by MMM7.1 is investigated. The temperature profiles for a high beta poloidal DIII-D discharge are computed using the NCLASS model for the neoclassical transport and the Weiland and Electron Temperature Gradient (ETG) components of the MMM7.1 model for the anomalous transport. The neoclassical transport is found to be the main contributor to the ion thermal transport in the plasma core. The contributions from the ion temperature gradient driven modes are found to be important only outside of the internal transport barrier. The magnitudes of the predicted temperature profiles are found to be in a reasonable agreement with experimental profiles. The simulation results approximately reproduce the internal transport barrier in the ion temperature profile but not in the electron temperature profile due to a weak dependence of the ETG driven transport on the Shafranov shift in the ETG component of MMM7.1. Possible effects that can contribute to stabilization of these modes, for example, effects associated with the large beta poloidal such as the Shafranov shift stabilization in the MMM7.1 model, are discussed. It is demonstrated that the E × B flow shear has a relatively small effect in the formation of the internal transport barrier in the high beta poloidal DIII-D discharge 154406. The Shafranov shift (alpha stabilization) and small or reversed magnetic shear profiles are found to be the primary reasons for quenched anomalous transport in this discharge.

Active control for stabilization of neoclassical tearing modesa)

NASA Astrophysics Data System (ADS)

Humphreys, D. A.; Ferron, J. R.; La Haye, R. J.; Luce, T. C.; Petty, C. C.; Prater, R.; Welander, A. S.

2006-05-01

This work describes active control algorithms used by DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to stabilize and maintain suppression of 3/2 or 2/1 neoclassical tearing modes (NTMs) by application of electron cyclotron current drive (ECCD) at the rational q surface. The DIII-D NTM control system can determine the correct q-surface/ECCD alignment and stabilize existing modes within 100-500ms of activation, or prevent mode growth with preemptive application of ECCD, in both cases enabling stable operation at normalized beta values above 3.5. Because NTMs can limit performance or cause plasma-terminating disruptions in tokamaks, their stabilization is essential to the high performance operation of ITER [R. Aymar et al., ITER Joint Central Team, ITER Home Teams, Nucl. Fusion 41, 1301 (2001)]. The DIII-D NTM control system has demonstrated many elements of an eventual ITER solution, including general algorithms for robust detection of q-surface/ECCD alignment and for real-time maintenance of alignment following the disappearance of the mode. This latter capability, unique to DIII-D, is based on real-time reconstruction of q-surface geometry by a Grad-Shafranov solver using external magnetics and internal motional Stark effect measurements. Alignment is achieved by varying either the plasma major radius (and the rational q surface) or the toroidal field (and the deposition location). The requirement to achieve and maintain q-surface/ECCD alignment with accuracy on the order of 1cm is routinely met by the DIII-D Plasma Control System and these algorithms. We discuss the integrated plasma control design process used for developing these and other general control algorithms, which includes physics-based modeling and testing of the algorithm implementation against simulations of actuator and plasma responses. This systematic design/test method and modeling environment enabled successful mode suppression by the NTM control system upon first-time use in an experimental discharge.

Magnetic shear effects on plasma transport and turbulence at high electron to ion temperature ratio in DIII-D and JT-60U plasmas

DOE PAGES

Yoshida, Maiko; McKee, George R.; Murakami, Masanori; ...

2017-03-30

We demonstrated negative magnetic shear in DIII-D and JT-60U in order to mitigate the confinement degradation typically observed with increasing the electron to ion temperature ratio (T-e/T-i). In recent experiments in DIII-D negative central magnetic shear (NCS) discharges, the thermal transport in the internal transport barrier formed around the radius of the minimum safety factor (q(min)) remained almost constant and modestly increased in the region outside of q(min) compared to the positive shear (PS) case, when T-e/T-i increased from about 0.8 to 1.1 through electron cyclotron heating (ECH). The benefit of NCS extending into the region outside of qmin canmore » be explained by the lower magnetic shear in the NCS plasma over the plasma radius relative to the PS plasma. Reduced confinement degradation at high T-e/T-i with NCS plasmas was commonly observed in DIII-D and JT-60U. Furthermore, the mechanism of the different transport responses between the NCS and PS plasmas has been assessed in terms of fluctuation measurements and gyrokinetic simulations in DIII-D; NCS gave a smaller rise in the low-wavenumber broadband turbulent fluctuations with the increase in T-e/T-i compared with the PS case. This is consistent with gyrokinetic simulations, and this shows a smaller rise in the growth rates of the ion temperature gradient mode in the NCS plasmas, with increasing T-e/T-i. Gyrokinetic simulations also showed a change in the stability of the electron modes with ECH applied, consistent with higher-wavenumber fluctuation measurements, although more detailed simulations are needed to give a quantitative explanation for the experimental observations. Control of q-profile and magnetic shear will allow confinement improvement in future machines with dominant electron heating.« less

MHD modeling of DIII-D QH-mode discharges and comparison to observations

NASA Astrophysics Data System (ADS)

King, Jacob

2016-10-01

MHD modeling of DIII-D QH-mode discharges and comparison to observations Nonlinear NIMROD simulations, initialized from a reconstruction of a DIII-D QH-mode discharge with broadband MHD, saturate into a turbulent state, but do not saturate when flow is not included. This is consistent with the experimental results of the quiescent regime observed on DIII-D with broadband MHD activity [Garofalo et al., PoP (2015) and refs. within]. These ELM-free discharges have the normalized pedestal-plasma confinement necessary for burning-plasma operation on ITER. Relative to QH-mode operation with more coherent MHD activity, operation with broadband MHD tends to occur at higher densities and lower rotation and thus may be more relevant to ITER. The nonlinear NIMROD simulations require highly accurate equilibrium reconstructions. Our equilibrium reconstructions include the scrape-off-layer profiles and the measured toroidal and poloidal rotation profiles. The simulation develops into a saturated turbulent state and the n=1 and 2 modes become dominant through an inverse cascade. Each toroidal mode in the range of n=1-5 is dominant at a different time. The perturbations are advected and sheared apart in the counter-clockwise direction consistent with the direction of the poloidal flow inside the LCFS. Work towards validation through comparison to magnetic coil and Doppler reflectometry measurements is presented. Consistent with experimental observations during QH-mode, the simulated state leads to large particle transport relative to the thermal transport. Analysis shows that the phase of the density and temperature perturbations differ resulting in greater convective particle transport relative to the convective thermal transport. This work supported by the U.S. Department of Energy Office of Science and the SciDAC Center for Extended MHD Modeling under Contract Numbers DE-FC02-06ER54875, DE-FC02-08ER54972 and DE-FC02-04ER54698.

Rotational Shear Effects on Edge Harmonic Oscillations in DIII-D Quiescent H-mode Discharges

NASA Astrophysics Data System (ADS)

Chen, Xi; Burrell, K. H.; Ferraro, N. M.; Osborne, T. H.; Austin, M. E.; Garofalo, A. M.; Groebner, R. J.; Kramer, G. J.; Luhmann, N. C., Jr.; McKee, G. R.; Muscatello, C. M.; Nazikian, R.; Ren, X.; Snyder, P. B.; Solomon, Wm.; Tobias, B. J.; Yan, Z.

2015-11-01

In quiescent H-mode (QH) regime, the edge harmonic oscillations (EHO) play an important role in avoiding the transient ELM power fluxes by providing benign and continuous edge particle transport. A detailed theoretical, experimental and modeling comparison has been made of low-n (n <= 5) EHO in DIII-D QH-mode plasmas. The calculated linear eigenmode structure from the extended MHD code M3D-C1 matches closely the coherent EHO properties from external magnetics data and internal measurements using the ECE, BES, ECE-I and MIR diagnostics, as well as the kink/peeling mode properties of the ideal MHD code ELITE. The numerical investigations indicate that the low-n EHO-like solutions from M3D-C1 are destabilized by the toroidal rotational shear while high-n modes are stabilized. This effect is independent of the rotation direction, suggesting that the low-n EHO can be destabilized in principle with rotation in both directions. These modeling results are consistent with experimental observations of the EHO and support the proposed theory of the EHO as a rotational shear driven kink/peeling mode.

Quantitative comparisons between experimentally measured 2D carbon radiation and Monte Carlo impurity (MCI) code simulations

NASA Astrophysics Data System (ADS)

Evans, T. E.; Finkenthal, D. F.; Fenstermacher, M. E.; Leonard, A. W.; Porter, G. D.; West, W. P.

Experimentally measured carbon line emissions and total radiated power distributions from the DIII-D divertor and scrape-off layer (SOL) are compared to those calculated with the Monte Carlo impurity (MCI) model. A UEDGE [T.D. Rognlien et al., J. Nucl. Mater. 196-198 (1992) 347] background plasma is used in MCI with the Roth and Garcia-Rosales (RG-R) chemical sputtering model [J. Roth, C. García-Rosales, Nucl. Fusion 36 (1992) 196] and/or one of six physical sputtering models. While results from these simulations do not reproduce all of the features seen in the experimentally measured radiation patterns, the total radiated power calculated in MCI is in relatively good agreement with that measured by the DIII-D bolometric system when the Smith78 [D.L. Smith, J. Nucl. Mater. 75 (1978) 20] physical sputtering model is coupled to RG-R chemical sputtering in an unaltered UEDGE plasma. Alternatively, MCI simulations done with UEDGE background ion temperatures along the divertor target plates adjusted to better match those measured in the experiment resulted in three physical sputtering models which when coupled to the RG-R model gave a total radiated power that was within 10% of measured value.

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

DOE PAGES

Okabayashi, M.; Zanca, P.; Strait, E. J.; ...

2016-11-25

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

NASA Astrophysics Data System (ADS)

Okabayashi, M.; Zanca, P.; Strait, E. J.; Garofalo, A. M.; Hanson, J. M.; In, Y.; La Haye, R. J.; Marrelli, L.; Martin, P.; Paccagnella, R.; Paz-Soldan, C.; Piovesan, P.; Piron, C.; Piron, L.; Shiraki, D.; Volpe, F. A.; DIII-D, The; RFX-mod Teams

2017-01-01

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. Here we have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance is provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. Here β N is defined as β N  =  β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. The internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.

ECH Technology Development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Temkin, Richard

2014-12-24

Electron Cyclotron Heating (ECH) is needed for plasma heating, current drive, plasma stability control, and other applications in fusion energy sciences research. The program of fusion energy sciences supported by U. S. DOE, Office of Science, Fusion Energy Sciences relies on the development of ECH technology to meet the needs of several plasma devices working at the frontier of fusion energy sciences research. The largest operating ECH system in the world is at DIII-D, consisting of six 1 MW, 110 GHz gyrotrons capable of ten second pulsed operation, plus two newer gyrotrons. The ECH Technology Development research program investigated themore » options for upgrading the DIII-D 110 GHz ECH system. Options included extending present-day 1 MW technology to 1.3 – 1.5 MW power levels or developing an entirely new approach to achieve up to 2 MW of power per gyrotron. The research consisted of theoretical research and designs conducted by Communication and Power Industries of Palo Alto, CA working with MIT. Results of the study would be validated in a later phase by research on short pulse length gyrotrons at MIT and long pulse / cw gyrotrons in industry. This research follows a highly successful program of development that has led to the highly reliable, six megawatt ECH system at the DIII-D tokamak. Eventually, gyrotrons at the 1.5 megawatt to multi-megawatt power level will be needed for heating and current drive in large scale plasmas including ITER and DEMO.« less

48 CFR 219.202-1 - Encouraging small business participation in acquisitions.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 48 Federal Acquisition Regulations System 3 2012-10-01 2012-10-01 false Encouraging small business... ACQUISITION REGULATIONS SYSTEM, DEPARTMENT OF DEFENSE SOCIOECONOMIC PROGRAMS SMALL BUSINESS PROGRAMS Policies 219.202-1 Encouraging small business participation in acquisitions. See PGI 205.207(d)(iii) for...

Use of reconstructed 3D VMEC equilibria to match effects of toroidally rotating discharges in DIII-D

DOE PAGES

Wingen, Andreas; Wilcox, Robert S.; Cianciosa, Mark R.; ...

2016-10-13

Here, a technique for tokamak equilibrium reconstructions is used for multiple DIII-D discharges, including L-mode and H-mode cases when weakly 3D fieldsmore » $$\\left(\\delta B/B\\sim {{10}^{-3}}\\right)$$ are applied. The technique couples diagnostics to the non-linear, ideal MHD equilibrium solver VMEC, using the V3FIT code, to find the most likely 3D equilibrium based on a suite of measurements. It is demonstrated that V3FIT can be used to find non-linear 3D equilibria that are consistent with experimental measurements of the plasma response to very weak 3D perturbations, as well as with 2D profile measurements. Observations at DIII-D show that plasma rotation larger than 20 krad s –1 changes the relative phase between the applied 3D fields and the measured plasma response. Discharges with low averaged rotation (10 krad s –1) and peaked rotation profiles (40 krad s –1) are reconstructed. Similarities and differences to forward modeled VMEC equilibria, which do not include rotational effects, are shown. Toroidal phase shifts of up to $${{30}^{\\circ}}$$ are found between the measured and forward modeled plasma responses at the highest values of rotation. The plasma response phases of reconstructed equilibra on the other hand match the measured ones. This is the first time V3FIT has been used to reconstruct weakly 3D tokamak equilibria.« less

Performance history and upgrades for the DIII-D gyrotron complex

DOE PAGES

Lohr, J.; Anderson, J. P.; Cengher, M.; ...

2015-03-12

The gyrotron installation on the DIII-D tokamak has been in operation at the second harmonic of the electron cyclotron resonance since the mid-1990s. Prior to that a large installation of ten 60 GHz tubes was operated at the fundamental resonance. The system has been upgraded regularly and is an everyday tool for experiments on DIII-D.

Investigation of island formation due to RMPs in DIII-D plasmas with the SIESTA resistive MHD equilibrium code

NASA Astrophysics Data System (ADS)

Hirshman, S. P.; Shafer, M. W.; Seal, S. K.; Canik, J. M.

2016-04-01

> The SIESTA magnetohydrodynamic (MHD) equilibrium code has been used to compute a sequence of ideally stable equilibria resulting from numerical variation of the helical resonant magnetic perturbation (RMP) applied to an axisymmetric DIII-D plasma equilibrium. Increasing the perturbation strength at the dominant , resonant surface leads to lower MHD energies and increases in the equilibrium island widths at the (and sidebands) surfaces, in agreement with theoretical expectations. Island overlap at large perturbation strengths leads to stochastic magnetic fields which correlate well with the experimentally inferred field structure. The magnitude and spatial phase (around the dominant rational surfaces) of the resonant (shielding) component of the parallel current are shown to change qualitatively with the magnetic island topology.

Location of the first plasma response to resonant magnetic perturbations in DIII-D H-mode plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xiao, W. W.; Evans, T. E.; Tynan, G. R.

2016-04-27

The resonant location of the first plasma response to periodic toroidal phase flips of a Resonant Magnetic Perturbation (RMP) field is experimentally identified in the DIII-D tokamak using phase minima of the modulated plasma density and toroidal rotation relative to the RMP field. Furthermore, the plasma response coincides with the q=3 rational surface and electron fluid velocity null, which is consistent with simulations of the plasma response to the RMP field from a resistive Magnetohydrodynamics modeling. We also observe an asymmetric propagation of the particle and the momentum from the resonant location of the plasma response to the RMP intomore » to core and into the plasma edge.« less

A NEW SYSTEM TO MONITOR DATA ANALYSES AND RESULTS OF PHYSICS DATA VALIDATION BETWEEN PULSES AT DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

FLANAGAN,A; SCHACHTER,J.M; SCHISSEL,D.P

2003-02-01

A Data Analysis Monitoring (DAM) system has been developed to monitor between pulse physics analysis at the DIII-D National Fusion Facility (http://nssrv1.gat.com:8000/dam). The system allows for rapid detection of discrepancies in diagnostic measurements or the results from physics analysis codes. This enables problems to be detected and possibly fixed between pulses as opposed to after the experimental run has concluded thus increasing the efficiency of experimental time. An example of a consistency check is comparing the experimentally measured neutron rate and the expected neutron emission, RDD0D. A significant difference between these two values could indicate a problem with one ormore » more diagnostics, or the presence of unanticipated phenomena in the plasma. This new system also tracks the progress of MDSplus dispatched data analysis software and the loading of analyzed data into MDSplus. DAM uses a Java Servlet to receive messages, CLIPS to implement expert system logic, and displays its results to multiple web clients via HTML. If an error is detected by DAM, users can view more detailed information so that steps can be taken to eliminate the error for the next pulse.« less

System to monitor data analyses and results of physics data validation between pulses at DIII-D

NASA Astrophysics Data System (ADS)

Flanagan, S.; Schachter, J. M.; Schissel, D. P.

2004-06-01

A data analysis monitoring (DAM) system has been developed to monitor between pulse physics analysis at the DIII-D National Fusion Facility (http://nssrv1.gat.com:8000/dam). The system allows for rapid detection of discrepancies in diagnostic measurements or the results from physics analysis codes. This enables problems to be detected and possibly fixed between pulses as opposed to after the experimental run has concluded, thus increasing the efficiency of experimental time. An example of a consistency check is comparing the experimentally measured neutron rate and the expected neutron emission, RDD0D. A significant difference between these two values could indicate a problem with one or more diagnostics, or the presence of unanticipated phenomena in the plasma. This system also tracks the progress of MDSplus dispatched data analysis software and the loading of analyzed data into MDSplus. DAM uses a Java Servlet to receive messages, C Language Integrated Production System to implement expert system logic, and displays its results to multiple web clients via Hypertext Markup Language. If an error is detected by DAM, users can view more detailed information so that steps can be taken to eliminate the error for the next pulse.

DiMES PMI research at DIII-D in support of ITER and beyond

DOE PAGES

Rudakov, Dimitry L.; Abrams, Tyler; Ding, Rui; ...

2017-03-27

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based onmore » experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. Furthermore, these techniques may find applications in the future devices.« less

Performance and data analysis aspects of the new DIII-D monostatic profile reflectometer system

DOE PAGES

Zeng, Lei; Peebles, William A.; Doyle, Edward J.; ...

2014-08-07

A new frequency-modulated (FMCW) profile reflectometer system, featuring a monostatic antenna geometry (using one microwave antenna for both launch and receive), has been installed on the DIII-D tokamak, providing a first experimental test of this measurement approach for profile reflectometry. Significant features of the new system are briefly described in this paper, including the new monostatic arrangement, use of overmoded, broadband transmission waveguide, and dual-polarization combination/demultiplexing. Updated data processing and analysis, and in-service performance aspects of the new monostatic profile reflectometer system are also presented. By using a raytracing code (GENRAY) to determine the approximate trajectory of the probe beam,more » the electron density (n e) profile can be successfully reconstructed with L-mode plasmas vertically shifted by more than 10 cm off the vessel midplane. Specifically, it is demonstrated that the new system has a capability to measure n e profiles with plasma vertical offsets of up to ±17 cm. Furthermore, examples are also presented of accurate, high time and spatial resolution density profile measurements made over a wide range of DIII-D conditions, e.g. the measured temporal evolution of the density profile across an L-H transition.« less

SOL and Edge Flows in DIII-D

NASA Astrophysics Data System (ADS)

Boedo, J. A.; Degrassie, J. S.; Grierson, B. A.; Rudakov, D. A.

2015-11-01

Recent measurements at DIII-D edge plasmas at the outer midplane show that, in the absence of external torque, the edge and near-SOL plasma flow is largely dominated by the intrinsic source of rotation most likely due to thermal ion loss. We also show that when NBI heating is present, the core momentum competes with the edge intrinsic momentum and can overwhelm it, in short, NBI-heated discharges at high power tend to determine edge and near SOL flows. Experiments performed in the DIII-D tokamak with OH heated, ECH-heated and NBI-heated discharges are diagnosed for core plasma flow with CER and edge/SOL plasma flow with Mach probes. We have changed the amount of NBI, OH and ECH heating while scanning the discharge collisionality. We have compared the experimental measurements to two complementary thermal ion loss theories that explain most of the observed features, including a scaling with Ti. One theory considers passing and trapped particles that are lost via a loss cone purely due to drifts and the other considers turbulence-enhanced loss of passing particles. Work supported by the US DOE under DE-FC02-04ER54698, DE-FG02-07ER54917 and DE-AC02-09CH111466.

Simulations of Low Power DIII-D Helicon Antenna Coupling

NASA Astrophysics Data System (ADS)

Smithe, David; Jenkins, Thomas

2017-10-01

We present an overview and initial progress for a new project to model coupling of the DIII-D Helicon Antenna. We lay the necessary computational groundwork for the modeling of both low-power and high power helicon antenna operation, by constructing numerical representations for both the antenna hardware and the DIII-D plasma. CAD files containing the detailed geometry of the low power antenna hardware are imported into the VSim software's FDTD plasma model. The plasma can be represented numerically by importing EQDSK or EFIT files. In addition, approximate analytic forms for the ensuing profiles and fields are constructed to facilitate parameter scans in the various regimes of anticipated antenna operation. To verify the accuracy of the numerical plasma and antenna representations, we will then run baseline simulations of low-power antenna operation, and verify that the predictions for loading, linear coupling, and mode partitioning (i.e. into helicon and slow modes) are consistent with the measurements from the low power helicon antenna experimental campaign, as well as with other independent models. Progress on these baseline simulations will be presented, and any inconsistencies and issues that arise during this process will be identified. Support provided by DOE Grant DE-SC0017843.

Effect of the scrape-off layer in AORSA full wave simulations of fast wave minority, mid/high harmonic, and helicon heating regimes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bertelli, N., E-mail: nbertell@pppl.gov; Gerhardt, S.; Hosea, J. C.

2015-12-10

Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interactions between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for the Nationalmore » Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown, confirming the same behavior found for a single toroidal mode results in Bertelli et al, Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is moved away from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended to “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for Alcator C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime. A substantial discussion of some of the main aspects, such as (i) the pitch angle of the magnetic field; (ii) minority heating vs. mid/high harmonic regimes is presented showing the different behavior of the RF field in the SOL region for NSTX-U scenarios with different plasma current. Finally, the preliminary results of the impact of the SOL region on the evaluation of the helicon current drive efficiency in DIII-D is presented for the first time and briefly compared with the different regimes mentioned above.« less

Effect of the scrape-off layer in AORSA full wave simulations of fast wave minority, mid/high harmonic, and helicon heating regimes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bertelli, Nicola; Jaeger, E. F.; Lau, Cornwall H

2015-01-01

Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interactions between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for the Nationalmore » Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown, confirming the same behavior found for a single toroidal mode results in Bertelli et al, Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is moved away from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended to "conventional" tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for Alcator C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime. A substantial discussion of some of the main aspects, such as (i) the pitch angle of the magnetic field; (ii) minority heating vs. mid/high harmonic regimes is presented showing the different behavior of the RF field in the SOL region for NSTX-U scenarios with different plasma current. Finally, the preliminary results of the impact of the SOL region on the evaluation of the helicon current drive efficiency in DIII-D is presented for the first time and briefly compared with the different regimes mentioned above.« less

Spatial and temporal analysis of DIII-D 3D magnetic diagnostic data

DOE PAGES

Strait, E. J.; King, J. D.; Hanson, J. M.; ...

2016-08-11

An extensive set of magnetic diagnostics in DIII-D is aimed at measuring non-axisymmetric "3D" features of tokamak plasmas, with typical amplitudes ~10 -3 to 10 -5 of the total magnetic field. We describe hardware and software techniques used at DIII-D to condition the individual signals and analysis to estimate the spatial structure from an ensemble of discrete measurements. Lastly, applications of the analysis include detection of non-rotating MHD instabilities, plasma control, and validation of MHD stability and 3D equilibrium models.

Analysis Tools for the Ion Cyclotron Emission Diagnostic on DIII-D

NASA Astrophysics Data System (ADS)

Del Castillo, C. A.; Thome, K. E.; Pinsker, R. I.; Meneghini, O.; Pace, D. C.

2017-10-01

Ion cyclotron emission (ICE) waves are excited by suprathermal particles such as neutral beam particles and fusion products. An ICE diagnostic is in consideration for use at ITER, where it could provide important passive measurement of fast ions location and losses, which are otherwise difficult to determine. Simple ICE data analysis codes had previously been developed, but more sophisticated codes are required to facilitate data analysis. Several terabytes of ICE data were collected on DIII-D during the 2015-2017 campaign. The ICE diagnostic consists of antenna straps and dedicated magnetic probes that are both digitized at 200 MHz. A suite of Python spectral analysis tools within the OMFIT framework is under development to perform the memory-intensive analysis of this data. A fast and optimized analysis allows ready access to data visualizations as spectrograms and as plots of both frequency and time cuts of the data. A database of processed ICE data is being constructed to understand the relationship between the frequency and intensity of ICE and a variety of experimental parameters including neutral beam power and geometry, local and global plasma parameters, magnetic fields, and many others. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

Measured vs. Predicted Pedestal Pressure During RMP ELM Control in DIII-D

NASA Astrophysics Data System (ADS)

Zywicki, Bailey; Fenstermacher, Max; Groebner, Richard; Meneghini, Orso

2017-10-01

From database analysis of DIII-D plasmas with Resonant Magnetic Perturbations (RMPs) for ELM control, we will compare the experimental pedestal pressure (p_ped) to EPED code predictions and present the dependence of any p_ped differences from EPED on RMP parameters not included in the EPED model e.g. RMP field strength, toroidal and poloidal spectrum etc. The EPED code, based on Peeling-Ballooning and Kinetic Ballooning instability constraints, will also be used by ITER to predict the H-mode p_ped without RMPs. ITER plans to use RMPs as an effective ELM control method. The need to control ELMs in ITER is of the utmost priority, as it directly correlates to the lifetime of the plasma facing components. An accurate means of determining the impact of RMP ELM control on the p_ped is needed, because the device fusion power is strongly dependent on p_ped. With this new collection of data, we aim to provide guidance to predictions of the ITER pedestal during RMP ELM control that can be incorporated in a future predictive code. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698, and DE-AC52-07NA27344.

DIII-D Upgrade to Prepare the Basis for Steady-State Burning Plasmas

NASA Astrophysics Data System (ADS)

Buttery, R. J.; Guo, H. Y.; Taylor, T. S.; Wade, M. R.; Hill, D. N.

2014-10-01

Future steady-state burning plasma facilities will access new physics regimes and modes of plasma behavior. It is vital to prepare for this both experimentally using existing facilities, and theoretically in order to develop the tools to project to and optimize these devices. An upgrade to DIII-D is proposed to address the three critical aspects where research must go beyond what we can do now: (i) torque free electron heating to address the energy, particle and momentum transport mechanisms of burning plasmas using electron cyclotron (EC) heating and full power balanced neutral beams; (ii) off-axis heating and current drive to develop the path to true fusion steady state by reorienting neutral beams and deploying EC and helicon current drive; (iii) a new divertor with hot walls and reactor relevant materials to develop the basis for benign detached divertor operation compatible with wall materials and a high performance fusion core. These elements with modest incremental cost and enacted as a user facility for the whole US program will enable the US to lead on ITER and take a decision to proceed with a Fusion Nuclear Science Facility. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

A fresh look at electron cyclotron current drive power requirements for stabilization of tearing modes in ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

La Haye, R. J., E-mail: lahaye@fusion.gat.com

2015-12-10

ITER is an international project to design and build an experimental fusion reactor based on the “tokamak” concept. ITER relies upon localized electron cyclotron current drive (ECCD) at the rational safety factor q=2 to suppress or stabilize the expected poloidal mode m=2, toroidal mode n=1 neoclassical tearing mode (NTM) islands. Such islands if unmitigated degrade energy confinement, lock to the resistive wall (stop rotating), cause loss of “H-mode” and induce disruption. The International Tokamak Physics Activity (ITPA) on MHD, Disruptions and Magnetic Control joint experiment group MDC-8 on Current Drive Prevention/Stabilization of Neoclassical Tearing Modes started in 2005, after whichmore » assessments were made for the requirements for ECCD needed in ITER, particularly that of rf power and alignment on q=2 [1]. Narrow well-aligned rf current parallel to and of order of one percent of the total plasma current is needed to replace the “missing” current in the island O-points and heal or preempt (avoid destabilization by applying ECCD on q=2 in absence of the mode) the island [2-4]. This paper updates the advances in ECCD stabilization on NTMs learned in DIII-D experiments and modeling during the last 5 to 10 years as applies to stabilization by localized ECCD of tearing modes in ITER. This includes the ECCD (inside the q=1 radius) stabilization of the NTM “seeding” instability known as sawteeth (m/n=1/1) [5]. Recent measurements in DIII-D show that the ITER-similar current profile is classically unstable, curvature stabilization must not be neglected, and the small island width stabilization effect from helical ion polarization currents is stronger than was previously thought [6]. The consequences of updated assumptions in ITER modeling of the minimum well-aligned ECCD power needed are all-in-all favorable (and well-within the ITER 24 gyrotron capability) when all effects are included. However, a “wild card” may be broadening of the localized ECCD by the presence of the island; various theories predict broadening could occur and there is experimental evidence for broadening in DIII-D. Wider than now expected ECCD in ITER would make alignment easier to do but weaken the stabilization and thus require more rf power. In addition to updated modeling for ITER, advances in the ITER-relevant DIII-D ECCD gyrotron launch mirror control system hardware and real-time plasma control system have been made [7] and there are plans for application in DIII-D ITER demonstration discharges.« less

A fresh look at electron cyclotron current drive power requirements for stabilization of tearing modes in ITER

NASA Astrophysics Data System (ADS)

La Haye, R. J.

2015-12-01

ITER is an international project to design and build an experimental fusion reactor based on the "tokamak" concept. ITER relies upon localized electron cyclotron current drive (ECCD) at the rational safety factor q=2 to suppress or stabilize the expected poloidal mode m=2, toroidal mode n=1 neoclassical tearing mode (NTM) islands. Such islands if unmitigated degrade energy confinement, lock to the resistive wall (stop rotating), cause loss of "H-mode" and induce disruption. The International Tokamak Physics Activity (ITPA) on MHD, Disruptions and Magnetic Control joint experiment group MDC-8 on Current Drive Prevention/Stabilization of Neoclassical Tearing Modes started in 2005, after which assessments were made for the requirements for ECCD needed in ITER, particularly that of rf power and alignment on q=2 [1]. Narrow well-aligned rf current parallel to and of order of one percent of the total plasma current is needed to replace the "missing" current in the island O-points and heal or preempt (avoid destabilization by applying ECCD on q=2 in absence of the mode) the island [2-4]. This paper updates the advances in ECCD stabilization on NTMs learned in DIII-D experiments and modeling during the last 5 to 10 years as applies to stabilization by localized ECCD of tearing modes in ITER. This includes the ECCD (inside the q=1 radius) stabilization of the NTM "seeding" instability known as sawteeth (m/n=1/1) [5]. Recent measurements in DIII-D show that the ITER-similar current profile is classically unstable, curvature stabilization must not be neglected, and the small island width stabilization effect from helical ion polarization currents is stronger than was previously thought [6]. The consequences of updated assumptions in ITER modeling of the minimum well-aligned ECCD power needed are all-in-all favorable (and well-within the ITER 24 gyrotron capability) when all effects are included. However, a "wild card" may be broadening of the localized ECCD by the presence of the island; various theories predict broadening could occur and there is experimental evidence for broadening in DIII-D. Wider than now expected ECCD in ITER would make alignment easier to do but weaken the stabilization and thus require more rf power. In addition to updated modeling for ITER, advances in the ITER-relevant DIII-D ECCD gyrotron launch mirror control system hardware and real-time plasma control system have been made [7] and there are plans for application in DIII-D ITER demonstration discharges.

Full wave simulations of fast wave efficiency and power losses in the scrape-off layer of tokamak plasmas in mid/high harmonic and minority heating regimes

DOE PAGES

Bertelli, N.; Jaeger, E. F.; Hosea, J. C.; ...

2015-12-17

Here, several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves (HHFW), have found strong interaction between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 2D and 3D AORSAmore » results for the National Spherical Torus eXperiment (NSTX) have shown a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is removed from in front of the antenna by increasing the edge density. Here, full wave simulations have been extended for 'conventional' tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results in HHFW regime show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for C-Mod and EAST, which operate in the minority heating regime.« less

Dimensionless size scaling of intrinsic rotation in DIII-D

DOE PAGES

deGrassie, John S.; Solomon, Wayne M.; Rice, J. E.; ...

2016-08-01

A dimensionless empirical scaling for intrinsic toroidal rotation is given; M A ~β Nρ*, where M A is the toroidal velocity divided by the Alfvén velocity, β N the usual normalized β value, and ρ* is the ion gyroradius divided by the minor radius. This scaling describes well experimental data from DIII-D, and also some published data from C-Mod and JET. The velocity used in this scaling is in an outer location in minor radius, outside of the interior core and inside of the large gradient edge region in H-mode conditions. Furthermore, this scaling establishes the basic magnitude of themore » intrinsic toroidal rotation and its relation to the rich variety of rotation profiles that can be realized for intrinsic conditions is discussed.« less

The DIII-D Plasma Control System as a Scientific Research Tool

NASA Astrophysics Data System (ADS)

Hyatt, A. W.; Ferron, J. R.; Humphreys, D. A.; Leuer, J. A.; Walker, M. L.; Welander, A. S.

2006-10-01

The digital plasma control system (PCS) is an essential element of the DIII-D tokamak as a scientific research instrument, providing experimenters with real-time measurement and control of the plasma equilibrium, heating, current drive, transport, stability, and plasma-wall interactions. A wide range of sensors and actuators allow feedback control not only of global quantities such as discharge shape, plasma energy, and toroidal rotation, but also of non-axisymmetric magnetic fields and features of the internal profiles of temperature and current density. These diverse capabilities of the PCS improve the effectiveness of tokamak operation and enable unique physics experiments. We will present an overview of the PCS and the systems it controls and interacts with, and show examples of various plasma parameters controlled by the PCS and its actuators.

QUANTITATIVE TESTS OF ELMS AS INTERMEDIATE N PEELING-BALOONING MODES

DOE Office of Scientific and Technical Information (OSTI.GOV)

LAO,LL; SNYDER,PB; LEONARD,AW

2003-03-01

A271 QUANTITATIVE TESTS OF ELMS AS INTERMEDIATE N PEELING-BALOONING MODES. Several testable features of the working model of edge localized modes (ELMs) as intermediate toroidal mode number peeling-ballooning modes are evaluated quantitatively using DIII-D and JT-60U experimental data and the ELITE MHD stability code. These include the hypothesis that ELM sizes are related to the radial widths of the unstable MHD modes, the unstable modes have a strong ballooning character localized in the outboard bad curvature region, and ELM size generally becomes smaller at high edge collisionality. ELMs are triggered when the growth rates of the unstable MHD modes becomemore » significantly large. These testable features are consistent with many ELM observations in DIII-D and JT-60U discharges.« less

Experimental tests of linear and nonlinear three-dimensional equilibrium models in DIII-D

DOE PAGES

King, Josh D.; Strait, Edward J.; Lazerson, Samuel A.; ...

2015-07-01

DIII-D experiments using new detailed magnetic diagnostics show that linear, ideal magnetohydrodynamics (MHD) theory quantitatively describes the magnetic structure (as measured externally) of three-dimensional (3D) equilibria resulting from applied fields with toroidal mode number n = 1, while a nonlinear solution to ideal MHD force balance, using the VMEC code, requires the inclusion of n ≥ 1 to achieve similar agreement. Moreover, these tests are carried out near ITER baseline parameters, providing a validated basis on which to exploit 3D fields for plasma control development. We determine scans of the applied poloidal spectrum and edge safety factors which confirm thatmore » low-pressure, n = 1 non-axisymmetric tokamak equilibria are a single, dominant, stable eigenmode. But, at higher beta, near the ideal kink mode stability limit in the absence of a conducting wall, the qualitative features of the 3D structure are observed to vary in a way that is not captured by ideal MHD.« less

Carbon Radiation Studies in the DIII-D Divertor with the Monte Carlo Impurity (MCI) Code

NASA Astrophysics Data System (ADS)

Evans, T. E.; Leonard, A. W.; West, W. P.; Finkenthal, D. F.; Fenstermacher, M. E.; Porter, G. D.; Chu, Y.

1998-11-01

Carbon sputtering and transport are modeled in the DIII--D divertor with the MCI code. Calculated 2-D radiation patterns are compared with measured radiation distributions. The results are particularly sensitive to Ti near the divertor target plates. For example, increasing the ion temperature from 8 eV to 20 eV in MCI raises P_rad^div from 1626 to 2862 kW. Although this presents difficulties in assessing which sputtering model best describes the plasma-surface interaction physics (because of experimental uncertainties in T_i), processes which either produce too much or too little radiated power compared to the measured value of 1718 kW can be eliminated. Based on this, the number of viable sputtering options has been reduced from 12 to 4. For the conditions studied, three of these options involve both physical and chemical sputtering, and one requires only physical sputtering.

Confinement improvement in the high poloidal beta regime on DIII-D and application to steady-state H-mode on EAST

DOE PAGES

Ding, Siye; Garofalo, A. M.; Qian, J.; ...

2017-05-03

Systematic experimental and modeling investigations on DIII-D and EAST show attractive transport properties of fully non-inductive high β p plasmas. Experiments on DIII-D show that the large-radius internal transport barrier (ITB), a key feature providing excellent confinement in the high β p regime, is maintained when the scenario is extended from q 95 ~ 12 to 7 and from rapid to near-zero toroidal rotation. The robustness of confinement versus rotation was predicted by gyro fluid modeling showing dominant neoclassical ion energy transport even without E B shear effect. The physics mechanism of turbulence suppression, we found, is the Shafranov shift,more » which is essential and sets a β p threshold for large-radius ITB formation in the high β p scenario on DIII-D. This is confirmed by two different parameter-scan experiments, one for β N scan and the other for q 95 scan. They both give the same p threshold at 1.9 in the experiment. Furthermore, the experiment trend of increasing thermal transport with decreasing β p is consistent with transport modeling. The very first step of extending high β p scenario on DIII-D to long pulse on EAST is to establish long pulse H-mode with ITB on EAST. Our paper shows the first 61 sec fully non-inductive H-mode with stationary ITB feature and actively cooled ITER-like tungsten divertor in the very recent EAST experiment. The successful use of lower hybrid wave (LWH) as a key tool to optimize current profile in EAST experiment is also introduced. Results show that as the electron density is increased, the fully non-inductive current profile broadens on EAST. The improved understanding and modeling capability is also used to develop advanced scenarios for CFETR. These results provide encouragement that the high β p regime can be extended to lower safety factor and very low rotation, providing a potential path to high performance steady state operation in future devices.« less

Confinement improvement in the high poloidal beta regime on DIII-D and application to steady-state H-mode on EAST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ding, Siye; Garofalo, A. M.; Qian, J.

Systematic experimental and modeling investigations on DIII-D and EAST show attractive transport properties of fully non-inductive high β p plasmas. Experiments on DIII-D show that the large-radius internal transport barrier (ITB), a key feature providing excellent confinement in the high β p regime, is maintained when the scenario is extended from q 95 ~ 12 to 7 and from rapid to near-zero toroidal rotation. The robustness of confinement versus rotation was predicted by gyro fluid modeling showing dominant neoclassical ion energy transport even without E B shear effect. The physics mechanism of turbulence suppression, we found, is the Shafranov shift,more » which is essential and sets a β p threshold for large-radius ITB formation in the high β p scenario on DIII-D. This is confirmed by two different parameter-scan experiments, one for β N scan and the other for q 95 scan. They both give the same p threshold at 1.9 in the experiment. Furthermore, the experiment trend of increasing thermal transport with decreasing β p is consistent with transport modeling. The very first step of extending high β p scenario on DIII-D to long pulse on EAST is to establish long pulse H-mode with ITB on EAST. Our paper shows the first 61 sec fully non-inductive H-mode with stationary ITB feature and actively cooled ITER-like tungsten divertor in the very recent EAST experiment. The successful use of lower hybrid wave (LWH) as a key tool to optimize current profile in EAST experiment is also introduced. Results show that as the electron density is increased, the fully non-inductive current profile broadens on EAST. The improved understanding and modeling capability is also used to develop advanced scenarios for CFETR. These results provide encouragement that the high β p regime can be extended to lower safety factor and very low rotation, providing a potential path to high performance steady state operation in future devices.« less

Error field detection in DIII-D by magnetic steering of locked modes

DOE PAGES

Shiraki, Daisuke; La Haye, Robert J.; Logan, Nikolas C.; ...

2014-02-20

Optimal correction coil currents for the n = 1 intrinsic error field of the DIII-D tokamak are inferred by applying a rotating external magnetic perturbation to steer the phase of a saturated locked mode with poloidal/toroidal mode number m/n = 2/1. The error field is detected non-disruptively in a single discharge, based on the toroidal torque balance of the resonant surface, which is assumed to be dominated by the balance of resonant electromagnetic torques. This is equivalent to the island being locked at all times to the resonant 2/1 component of the total of the applied and intrinsic error fields,more » such that the deviation of the locked mode phase from the applied field phase depends on the existing error field. The optimal set of correction coil currents is determined to be those currents which best cancels the torque from the error field, based on fitting of the torque balance model. The toroidal electromagnetic torques are calculated from experimental data using a simplified approach incorporating realistic DIII-D geometry, and including the effect of the plasma response on island torque balance based on the ideal plasma response to external fields. This method of error field detection is demonstrated in DIII-D discharges, and the results are compared with those based on the onset of low-density locked modes in ohmic plasmas. Furthermore, this magnetic steering technique presents an efficient approach to error field detection and is a promising method for ITER, particularly during initial operation when the lack of auxiliary heating systems makes established techniques based on rotation or plasma amplification unsuitable.« less

Advances in understanding quiescent H-mode plasmas in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burrell, K.H.; West, W.P.; Gohil, P.

2005-05-15

Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasingmore » power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by the peeling-ballooning theory is consistent with the substantial improvement in pedestal pressure achieved by changing the plasma shape from a single null divertor to a high triangularity double null. Detailed ELITE calculations for the high triangularity plasmas have demonstrated that the plasma operating point is marginally stable to peeling-ballooning modes. Comparison of ELMing, coinjected and quiescent, counterinjected discharges with the same shape, current, toroidal field, electron density, and electron temperature indicates that the edge radial electric field or the edge toroidal rotation are also playing a role in edge stability. The EHO produces electron, main ion, and impurity particle transport at the plasma edge which is more rapid than that produced by ELMs under similar conditions. The EHO also decreases the edge rotation while producing little change in the edge electron and ion temperatures. Other edge electromagnetic modes also produce particle transport; this includes the incoherent, broadband activity seen at high triangularity. Pedestal values of {nu}{sub *} and {beta}{sub T} bracketing, those required for International Experimental Thermonuclear Reactor [Nucl. Fusion 39, 2137 (1999)] have been achieved in DIII-D, demonstrating the QH-mode edge densities are sufficient for future devices.« less

Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

NASA Astrophysics Data System (ADS)

Wang, Y. M.; Xu, X. Q.; Yan, Z.; Mckee, G. R.; Grierson, B. A.; Xia, T. Y.; Gao, X.

2018-02-01

A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n  =  30 or k_θρ_i˜0.12 . The ion diamagnetic drift and E× B convection flow are balanced when the radial electric field (E r ) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density n_e˜1.5×1019 m-3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40 kHz and 10 kHz respectively. The poloidal wave number k_θ is about 0.2 cm -1 (k_θρ_i˜0.05 ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are  ˜3.5-6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. The electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.

X-ray and extreme ultraviolet spectroscopy on DIII-D

NASA Astrophysics Data System (ADS)

Victor, B. S.; Allen, S. L.; Beiersdorfer, P.; Magee, E. W.

2017-06-01

Two spectrometers were installed to measure tungsten emission in the core of DIII-D plasmas during a metal rings experimental campaign. The spectral range of the high-resolution (1340 spectral channels), variable-ruled grating X-ray and Extreme Ultraviolet Spectrometer (XEUS) extends from 10-71 dot A. The spectral range of the second spectrometer, the Long-Wavelength Extreme Ultraviolet Spectrometer (LoWEUS), measures between 31-174 dot A. Three groups of tungsten lines were identified with XEUS: W38+-W45+ from 47-63 dot A, W27+-W35+ from 45-55 dot A, and W28+-W33+ from 16-30 dot A. Emission lines from tungsten charge states W28+, W43+, W44+, and W45+ are identified and the line amplitude is presented versus time. Peak emission of W43+-W45+ occurs between core Te=2.5-3 keV, and peak emission of W28+ occurs at core Te<=1.3 keV. One group of tungsten lines, W40+-W45+, between 120-140 dot A, was identified with LoWEUS. W43+-W45+ lines measured with LoWEUS track the sawtooth cycle. Sensitivity to the sawtooth cycle and the correlation of the peak emission with core electron temperature show that these spectrometers track the on-axis tungsten emission of DIII-D plasmas.

Development and validation of a critical gradient energetic particle driven Alfven eigenmode transport model for DIII-D tilted neutral beam experiments

DOE PAGES

Waltz, Ronald E.; Bass, Eric M.; Heidbrink, William W.; ...

2015-10-30

Recent experiments with the DIII-D tilted neutral beam injection (NBI) varying the beam energetic particle (EP) source profiles have provided strong evidence that unstable Alfven eigenmodes (AE) drive stiff EP transport at a critical EP density gradient. Here the critical gradient is identified by the local AE growth rate being equal to the local ITG/TEM growth rate at the same low toroidal mode number. The growth rates are taken from the gyrokinetic code GYRO. Simulation show that the slowing down beam-like EP distribution has a slightly lower critical gradient than the Maxwellian. The ALPHA EP density transport code, used tomore » validate the model, combines the low-n stiff EP critical density gradient AE mid-core transport with the energy independent high-n ITG/TEM density transport model controling the central core EP density profile. For the on-axis NBI heated DIII-D shot 146102, while the net loss to the edge is small, about half the birth fast ions are transported from the central core r/a < 0.5 and the central density is about half the slowing down density. Lastly, these results are in good agreement with experimental fast ion pressure profiles inferred from MSE constrained EFIT equilibria.« less

RMP effects on the W and C erosion/deposition balance on W test samples in DIII-D

NASA Astrophysics Data System (ADS)

Hinson, E. T.; Frerichs, H.; Schmitz, O.; Evans, T. E.; Guo, H. Y.; Thomas, D. M.; Rudakov, D. L.; Abrams, T.; Unterberg, E. A.; Briesemeister, A.; Lasnier, C. J.; McLean, A. G.; Makowski, M.; Wampler, W. R.; Watkins, J. G.; Wang, H. Q.

2016-10-01

Clear evidence for alteration of the W and C erosion by resonant magnetic perturbation (RMP) fields has been obtained in an experiment exposing W-coated DiMES samples in the DIII-D divertor to outer strike point (OSP) sweeps in comparable series of discharges with and without the application of RMP. Gross erosion measurements of W and C during these sweeps using the S/XB method show that the 3-D boundary induced by the RMP significantly alters the erosion rate from DiMES. In particular, application of RMP smooths radial W erosion anisotropy seen for the axisymmetric case, where the W erosion rate for the OSP sweep in the outward direction significantly exceeds the erosion rate observed for the subsequent inward radial sweep over the sample. This finding is likely related to a change in the W/C erosion and redeposition balance in the C-dominated wall environment at DIII-D. Moreover, non-axisymmetric plasma structure on the W sample has to be considered. This challenge will be further examined by comparison of experimental results to EMC3-EIRENE modeling. Work supported by US DOE DE-SC0013911, DE-FC02-04ER54698, DE-FH02-07ER54917, DE-AC05-06OR23100, DE-AC05-00OR22725, DE-AC52-07NA27344, and DE-AC04-94AL85000.

Liquid crystal polymer substrate MMIC receiver modules for the ECE Imaging system on the DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhu, Y.; Ye, Y.; Yu, J-H

A new generation of millimeter-wave heterodyne imaging receiver arrays has been developed and demonstrated on the DIII-D ECEI system. Improved circuit integration, allowing for absolute calibration, improved noise performance, and shielding from out-of-band emission, is made possible by using advanced liquid crystal polymer (LCP) substrates and MMIC (Monolithic Microwave Integrated Circuit) receiver chips. This array exhibits ~ 15 dB additional gain and > 30x reduction in noise temperature compared to the previous generation and provide ECEI capability for absolute 2-D electron temperature profile measurements. Each LCP horn-waveguide module houses a 3x3 mm GaAs MMIC receiver chip, which consists of amore » low noise amplifier (LNA), balanced mixer, local oscillator multiplier chain driven by ~12 GHz input via an RF cable to the enclosure box, and IF amplifier. A proof-of-principle instrument with 5 poloidal channels was installed on DIII-D in 2017. The full proof-of-principle system installation (20 poloidal x 8 radial channels) was commissioned early in 2018. The LCP ECEI system is used for pedestal region measurements, especially focusing on temperature evolution during ELM bursting. The DIII-D ECE Imaging signal has been significantly improved with extremely effective shielding of out-of-band microwave noise which plagued previous ECE Imaging studies on DIII-D. In H-mode ELM bursting, the radial propagation of electron heat flow has been detected on DIII-D. The LCP ECE Imaging is expected to be a valuable diagnostic tool for ELM physics investigations.« less

Investigation of island formation due to RMPs in DIII-D plasmas with the SIESTA resistive MHD equilibrium code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirshman, S. P.; Shafer, M. W.; Seal, S. K.

The SIESTA magnetohydrodynamic (MHD) equilibrium code has been used to compute a sequence of ideally stable equilibria resulting from numerical variation of the helical resonant magnetic perturbation (RMP) applied to an axisymmetric DIII-D plasma equilibrium. Increasing the perturbation strength at the dominant m=2, n=-1 , resonant surface leads to lower MHD energies and increases in the equilibrium island widths at the m=2 (and sidebands) surfaces, in agreement with theoretical expectations. Island overlap at large perturbation strengths leads to stochastic magnetic fields which correlate well with the experimentally inferred field structure. The magnitude and spatial phase (around the dominant rational surfaces)more » of the resonant (shielding) component of the parallel current are shown to change qualitatively with the magnetic island topology.« less

Investigation of island formation due to RMPs in DIII-D plasmas with the SIESTA resistive MHD equilibrium code

DOE PAGES

Hirshman, S. P.; Shafer, M. W.; Seal, S. K.; ...

2016-03-03

The SIESTA magnetohydrodynamic (MHD) equilibrium code has been used to compute a sequence of ideally stable equilibria resulting from numerical variation of the helical resonant magnetic perturbation (RMP) applied to an axisymmetric DIII-D plasma equilibrium. Increasing the perturbation strength at the dominant m=2, n=-1 , resonant surface leads to lower MHD energies and increases in the equilibrium island widths at the m=2 (and sidebands) surfaces, in agreement with theoretical expectations. Island overlap at large perturbation strengths leads to stochastic magnetic fields which correlate well with the experimentally inferred field structure. The magnitude and spatial phase (around the dominant rational surfaces)more » of the resonant (shielding) component of the parallel current are shown to change qualitatively with the magnetic island topology.« less

Mapping and uncertainty analysis of energy and pitch angle phase space in the DIII-D fast ion loss detector.

PubMed

Pace, D C; Pipes, R; Fisher, R K; Van Zeeland, M A

2014-11-01

New phase space mapping and uncertainty analysis of energetic ion loss data in the DIII-D tokamak provides experimental results that serve as valuable constraints in first-principles simulations of energetic ion transport. Beam ion losses are measured by the fast ion loss detector (FILD) diagnostic system consisting of two magnetic spectrometers placed independently along the outer wall. Monte Carlo simulations of mono-energetic and single-pitch ions reaching the FILDs are used to determine the expected uncertainty in the measurements. Modeling shows that the variation in gyrophase of 80 keV beam ions at the FILD aperture can produce an apparent measured energy signature spanning across 50-140 keV. These calculations compare favorably with experiments in which neutral beam prompt loss provides a well known energy and pitch distribution.

Experimentally-based ExB drifts in the DIII-D divertor and SOL calculated from integration of Ohm's law using Thomson scattering measurements of T e and n e

DOE PAGES

Stangeby, Peter C.; Elder, J. David; McLean, Adam G.; ...

2017-03-27

We calculated the 2D spatial distributions of cross field drift velocities from 2D Thomson scattering measurements of T e and n e in the divertor and SOL of DIII-D. In contrast with the method that has been used on DIII-D where the 2D distribution of plasma potential V plasma is obtained from measurements of the probe floating potential of reciprocating probes, the present method does not require insertion of a probe into the plasma and can therefore be used in high power discharges. Furthermore, the 2D spatial distribution of V plasma is calculated from Ohm’s Law for the parallel electricmore » field E || along each flux tube, E || s || = -1.71dT e/ds || - T e/n edn e/ds ||, where the Thomson scattering values of T e and n e are used. To within a constant of integration, V plasma is obtained by integrating E || along the flux-tubes (field lines); the constant is obtained for each flux tube using the sheath drop at the target calculated from the characteristic of Langmuir probes built into the divertor tiles. The 2D distributions of E./01/2 = -dV4/ds./01/2, E452510/2 = -dV4/ds452510/2, v789 452510/2 = E./01/2/B and v789 ./01/2 = E452510/2/B are then calculated as well as the particle drift flux densities Γ789 452510/2 = nv789 452510/2 and Γ789 ./01/2 = nv789 ./01/2 for electrons, fuel ions and impurity ions, using the appropriate values of particle density, n.« less

Experimentally-based ExB drifts in the DIII-D divertor and SOL calculated from integration of Ohm's law using Thomson scattering measurements of T e and n e

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stangeby, Peter C.; Elder, J. David; McLean, Adam G.

We calculated the 2D spatial distributions of cross field drift velocities from 2D Thomson scattering measurements of T e and n e in the divertor and SOL of DIII-D. In contrast with the method that has been used on DIII-D where the 2D distribution of plasma potential V plasma is obtained from measurements of the probe floating potential of reciprocating probes, the present method does not require insertion of a probe into the plasma and can therefore be used in high power discharges. Furthermore, the 2D spatial distribution of V plasma is calculated from Ohm’s Law for the parallel electricmore » field E || along each flux tube, E || s || = -1.71dT e/ds || - T e/n edn e/ds ||, where the Thomson scattering values of T e and n e are used. To within a constant of integration, V plasma is obtained by integrating E || along the flux-tubes (field lines); the constant is obtained for each flux tube using the sheath drop at the target calculated from the characteristic of Langmuir probes built into the divertor tiles. The 2D distributions of E./01/2 = -dV4/ds./01/2, E452510/2 = -dV4/ds452510/2, v789 452510/2 = E./01/2/B and v789 ./01/2 = E452510/2/B are then calculated as well as the particle drift flux densities Γ789 452510/2 = nv789 452510/2 and Γ789 ./01/2 = nv789 ./01/2 for electrons, fuel ions and impurity ions, using the appropriate values of particle density, n.« less

A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roquemore, A; Maingi, R; Lasnier, C

2007-06-19

In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX smallmore » type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.« less

Comparison of measured impurity poloidal rotation in DIII-D with neoclassical predictions under low toroidal field conditions

DOE PAGES

Burrell, Keith H.; Grierson, Brian A.; Solomon, Wayne M.; ...

2014-06-26

Here, predictive understanding of plasma transport is a long-term goal of fusion research. This requires testing models of plasma rotation including poloidal rotation. The present experiment was motivated by recent poloidal rotation measurements on spherical tokamaks (NSTX and MAST) which showed that the poloidal rotation of C +6 is much closer to the neoclassical prediction than reported results in larger aspect ratio machines such as TFTR, DIII-D, JT-60U and JET working at significantly higher toroidal field and ion temperature. We investigated whether the difference in aspect ratio (1.44 on NSTX versus 2.7 on DIII-D) could explain this. We measured Cmore » +6 poloidal rotation in DIII-D under conditions which matched, as best possible, those in the NSTX experiment; we matched plasma current (0.65 MA), on-axis toroidal field (0.55T), minor radius (0.6 m), and outer flux surface shape as well as the density and temperature profiles. DIII-D results from this work also show reasonable agreement with neoclassical theory. Accordingly, the different aspect ratio does not explain the previously mentioned difference in poloidal rotation results.« less

Electron Cyclotron Heating system status and upgrades on DIII-D

DOE PAGES

Cengher, Mirela; Lohr, John; Gorelov, Yuri; ...

2016-06-02

The Electron Cyclotron Heating (ECH) system on the DIII-D tokamak consists of six 110 GHz gyrotrons with corrugated coaxial 31.75 mm waveguide transmission lines and steerable launching mirrors. The system has been gradually updated, leading to increased experimental flexibility and a high system reliability of 91% in the past year. Operationally, the gyrotrons can generate up to a total of 4.8 MW of rf power for pulses up to 5 seconds in length. The maximum ECH energy injected into the DIII-D has been 16.6 MJ. The HE11 mode content is over 85% for all the lines, and the transmission coefficientmore » is better than -1.1 dB for all the transmission lines, close to the theoretical value. A new depressed collector gyrotron was recently installed and was injecting up to 640 kW of power into the plasma during 2014-2015 tokamak operations. Three dual waveguide launchers, which can steer the RF beams ±20 degrees poloidally and toroidally, were used for real-time neoclassical tearing mode control and suppression. The launchers now have increased poloidal scanning speed and beam positioning accuracy of ~±2 mm at the plasma center. A new method of in-situ calibration of the mirror angle was used in conjunction with the upgrading of the encoders and motors for the launchers. Two more gyrotrons are expected to be installed and operational in 2015-2016. The first is a repaired 110 GHz, 1 MW gyrotron that had a gun failure after more than 11 years of operation at DIII-D. The second is a newly designed depressed collector tube in the 1.5 MW class, operating at 117.5 GHz, manufactured by Communications and Power Industries (CPI). It operates in the TE20,9 mode and has achieved 1.8 MW for short pulses during factory testing. Furthermore, this gyrotron is undergoing rework to address a high voltage standoff problem.« less

Electron temperature critical gradient and transport stiffness in DIII-D

DOE PAGES

Smith, Sterling P.; Petty, Clinton C.; White, Anne E.; ...

2015-07-06

The electron energy flux has been probed as a function of electron temperature gradient on the DIII-D tokamak, in a continuing effort to validate turbulent transport models. In the scan of gradient, a critical electron temperature gradient has been found in the electron heat fluxes and stiffness at various radii in L-mode plasmas. The TGLF reduced turbulent transport model [G.M. Staebler et al, Phys. Plasmas 14, 055909 (2007)] and full gyrokinetic GYRO model [J. Candy and R.E. Waltz, J. Comput. Phys. 186, 545 (2003)] recover the general trend of increasing electron energy flux with increasing electron temperature gradient scale length,more » but they do not predict the absolute level of transport at all radii and gradients. Comparing the experimental observations of incremental (heat pulse) diffusivity and stiffness to the models’ reveals that TGLF reproduces the trends in increasing diffusivity and stiffness with increasing electron temperature gradient scale length with a critical gradient behavior. Furthermore, the critical gradient of TGLF is found to have a dependence on q 95, contrary to the independence of the experimental critical gradient from q 95.« less

Statistical characterization of surface features from tungsten-coated divertor inserts in the DIII-D Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Adams, Jacob; Unterberg, Ezekial; Chrobak, Christopher; Stahl, Brian; Abrams, Tyler

2017-10-01

Continuing analysis of tungsten-coated inserts from the recent DIII-D Metal Rings Campaign utilizes a statistical approach to study carbon migration and deposition on W surfaces and to characterize the pre- versus post-exposure surface morphology. A TZM base was coated with W using both CVD and PVD and allowed for comparison between the two coating methods. The W inserts were positioned in the lower DIII-D divertor in both the upper (shelf) region and lower (floor) region and subjected to multiple plasma shots, primarily in H-mode. Currently, the post-exposure W inserts are being characterized using SEM/EDX to qualify the surface morphology and to quantify the surface chemical composition. In addition, profilometry is being used to measure the surface roughness of the inserts both before and after plasma exposure. Preliminary results suggest a correlation between the pre-exposure surface roughness and the level of carbon deposited on the surface. Furthermore, ongoing in-depth analysis may reveal insights into the formation mechanism of nanoscale bumps found in the carbon-rich regions of the W surfaces that have not yet been explained. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

Effect of lithium in the DIII-D SOL and plasma-facing surfaces

NASA Astrophysics Data System (ADS)

Jackson, G. L.; Chrobak, C. P.; McLean, A. G.; Maingi, R.; Mansfield, D. K.; Roquemore, A. L.; Diwakar, P.; Hassanein, A.; Lietz, A.; Rudakov, D. L.; Sizyuk, T.; Tripathi, J.

2015-08-01

Lithium has been introduced into the DIII-D tokamak, and migration and retention in graphite have been characterized since no lithium was present in DIII-D initially. A new regime with an enhanced edge electron pedestal and H98y2 ⩽ 2 has been obtained with lithium. Lithium deposition was not uniform, but rather preferentially deposited near the strike points, consistent with previous 13C experiments. Edge visible lithium light (LiI) remained well above the previous background during the entire DIII-D campaign, decaying with a 2600 plasma-second e-fold, but plasma performance was only affected on the discharge with lithium injection. Lithium injection demonstrated the capability of reducing hydrogenic recycling, density, and ELM frequency. Graphite and silicon samples were exposed to a lithium-injected discharge, using the DiMES system and then removed for ex-situ analysis. The deposited lithium layer remained detectable to a depth up to 1 μm.

Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Chung, H.M.; Smith, D.L.

1997-04-01

Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma.more » Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.« less

Impurity confinement and transport in high confinement regimes without edge localized modes on DIII-D [Impurity confinement and transport in high confinement regimes without ELMs on DIII-D

DOE PAGES

Grierson, Brian A.; Burrell, Keith H.; Nazikian, Raffi M.; ...

2015-04-17

Here, impurity transport in the DIII-D tokamak is investigated in stationary high confinement (H-mode) regimes without edge localized modes (ELMs). In plasmas maintained by resonant magnetic perturbation (RMP) ELM-suppression and QH-mode the confinement time of fluorine (Z=9) is equivalent to that in ELMing discharges with 40 Hz ELMs. For selected discharges with impurity injection the impurity particle confinement time compared to the energy confinement time is in the range of τ p/τ e ≈ 2 $-$ 3. In QH-mode operation the impurity confinement time is shown to be smaller for intense, coherent magnetic and density fluctuations of the edge harmonicmore » oscillation than weaker fluctuations. Transport coefficients are derived from the time evolution of the impurity density profile and compared to neoclassical and turbulent transport models NEO and TGLF. Neoclassical transport of fluorine is found to be small compared to the experimental values. In the ELMing and RMP ELM-suppressed plasma the impurity transport is affected by the presence of tearing modes. For radii larger than the mode radius the TGLF diffusion coefficient is smaller than the experimental value by a factor of 2-3, while the convective velocity is within error estimates. Low levels of diffusion are observed for radii smaller than the tearing mode radius. In the QH-mode plasma investigated, the TGLF diffusion coefficient higher inside of ρ = 0.4 and lower outside of 0.4 than the experiment, and the TGLF convective velocity is more negative by a factor of approximately 1.7.« less

OEDGE modeling for the planned tungsten ring experiment on DIII-D

DOE PAGES

Elder, J. David; Stangeby, Peter C.; Abrams, Tyler W.; ...

2017-04-19

The OEDGE code is used to model tungsten erosion and transport for DIII-D experiments with toroidal rings of high-Z metal tiles. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasmamore » impurity simulations. We developed a new model for tungsten erosion in OEDGE which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. Furthermore, these values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport), the choice of tungsten atomic physics data used in the model (in particular sviz(Te) for W-atoms), and the model of the carbon flux and energy used for 2 calculating the tungsten source due to sputtering. The core tungsten density is found to be of order 10 15 m -3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned) with density decaying into the scrape off layer.« less

Validating the energy transport modeling of the DIII-D and EAST ramp up experiments using TSC

NASA Astrophysics Data System (ADS)

Liu, Li; Guo, Yong; Chan, Vincent; Mao, Shifeng; Wang, Yifeng; Pan, Chengkang; Luo, Zhengping; Zhao, Hailin; Ye, Minyou

2017-06-01

The confidence in ramp up scenario design of the China fusion engineering test reactor (CFETR) can be significantly enhanced using validated transport models to predict the current profile and temperature profile. In the tokamak simulation code (TSC), two semi-empirical energy transport models (the Coppi-Tang (CT) and BGB model) and three theory-based models (the GLF23, MMM95 and CDBM model) are investigated on the CFETR relevant ramp up discharges, including three DIII-D ITER-like ramp up discharges and one EAST ohmic discharge. For the DIII-D discharges, all the transport models yield dynamic {{\\ell}\\text{i}} within +/- 0.15 deviations except for some time points where the experimental fluctuation is very strong. All the models agree with the experimental {β\\text{p}} except that the CT model strongly overestimates {β\\text{p}} in the first half of ramp up phase. When applying the CT, CDBM and GLF23 model to estimate the internal flux, they show maximum deviations of more than 10% because of inaccuracies in the temperature profile predictions, while the BGB model performs best on the internal flux. Although all the models fall short in reproducing the dynamic {{\\ell}\\text{i}} evolution for the EAST tokamak, the result of the BGB model is the closest to the experimental {{\\ell}\\text{i}} . Based on these comparisons, we conclude that the BGB model is the most consistent among these models for simulating CFETR ohmic ramp-up. The CT model with improvement for better simulation of the temperature profiles in the first half of ramp up phase will also be attractive. For the MMM95, GLF23 and CDBM model, better prediction of the edge temperature will improve the confidence for CFETR L-mode simulation. Conclusive validation of any transport model will require extensive future investigation covering a larger variety discharges.

Advances in understanding quiescent H-mode plasmas in DIII-Da)

NASA Astrophysics Data System (ADS)

Burrell, K. H.; West, W. P.; Doyle, E. J.; Austin, M. E.; Casper, T. A.; Gohil, P.; Greenfield, C. M.; Groebner, R. J.; Hyatt, A. W.; Jayakumar, R. J.; Kaplan, D. H.; Lao, L. L.; Leonard, A. W.; Makowski, M. A.; McKee, G. R.; Osborne, T. H.; Snyder, P. B.; Solomon, W. M.; Thomas, D. M.; Rhodes, T. L.; Strait, E. J.; Wade, M. R.; Wang, G.; Zeng, L.

2005-05-01

Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasing power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by the peeling-ballooning theory is consistent with the substantial improvement in pedestal pressure achieved by changing the plasma shape from a single null divertor to a high triangularity double null. Detailed ELITE calculations for the high triangularity plasmas have demonstrated that the plasma operating point is marginally stable to peeling-ballooning modes. Comparison of ELMing, coinjected and quiescent, counterinjected discharges with the same shape, current, toroidal field, electron density, and electron temperature indicates that the edge radial electric field or the edge toroidal rotation are also playing a role in edge stability. The EHO produces electron, main ion, and impurity particle transport at the plasma edge which is more rapid than that produced by ELMs under similar conditions. The EHO also decreases the edge rotation while producing little change in the edge electron and ion temperatures. Other edge electromagnetic modes also produce particle transport; this includes the incoherent, broadband activity seen at high triangularity. Pedestal values of ν* and βT bracketing, those required for International Experimental Thermonuclear Reactor [Nucl. Fusion 39, 2137 (1999)] have been achieved in DIII-D, demonstrating the QH-mode edge densities are sufficient for future devices.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Covele, Brent; Kotschenreuther, M.; Mahajan, S.

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Mapping and uncertainty analysis of energy and pitch angle phase space in the DIII-D fast ion loss detector

DOE PAGES

Pace, D. C.; Pipes, R.; Fisher, R. K.; ...

2014-08-05

New phase space mapping and uncertainty analysis of energetic ion loss data in the DIII-D tokamak provides experimental results that serve as valuable constraints in first-principles simulations of energetic ion transport. Beam ion losses are measured by the fast ion loss detector (FILD) diagnostic system consisting of two magnetic spectrometers placed independently along the outer wall. Monte Carlo simulations of mono-energetic and single-pitch ions reaching the FILDs are used to determine the expected uncertainty in the measurements. Modeling shows that the variation in gyrophase of 80 keV beam ions at the FILD aperture can produce an apparent measured energy signaturemore » spanning across 50-140 keV. As a result, these calculations compare favorably with experiments in which neutral beam prompt loss provides a well known energy and pitch distribution.« less

Numerical study on wave-induced beam ion prompt losses in DIII-D tokamak

DOE PAGES

Feng, Zhichen; Zhu, Jia; Fu, Guo -Yong; ...

2017-08-30

A numerical study is performed on the coherent beam ion prompt losses driven by Alfven eigenmodes (AEs) in DIII-D plasmas using realistic parameters and beam ion deposition profiles. The synthetic signal of a fast-ion loss detector (FILD) is calculated for a single AE mode. The first harmonic of the calculated FILD signal is linearly proportional to the AE amplitude with the same AE frequency in agreement with the experimental measurement. The calculated second harmonic is proportional to the square of the first harmonic for typical AE amplitudes. The coefficient of quadratic scaling is found to be sensitive to the AEmore » mode width. The second part of this work considers the AE drive due to coherent prompt loss. As a result, it is shown that the loss-induced mode drive is much smaller than the previous estimate and can be ignored for mode stability.« less

Poloidal flux profile reconstruction from pointwise measurements via extended Kalman filtering in the DIII-D Tokamak

DOE PAGES

Wang, Hexiang; Barton, Justin E.; Schuster, Eugenio

2015-09-01

The accuracy of the internal states of a tokamak, which usually cannot be measured directly, is of crucial importance for feedback control of the plasma dynamics. A first-principles-driven plasma response model could provide an estimation of the internal states given the boundary conditions on the magnetic axis and at plasma boundary. However, the estimation would highly depend on initial conditions, which may not always be known, disturbances, and non-modeled dynamics. Here in this work, a closed-loop state observer for the poloidal magnetic flux is proposed based on a very limited set of real-time measurements by following an Extended Kalman Filteringmore » (EKF) approach. Comparisons between estimated and measured magnetic flux profiles are carried out for several discharges in the DIII-D tokamak. The experimental results illustrate the capability of the proposed observer in dealing with incorrect initial conditions and measurement noise.« less

Magnetic Transport Barriers in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Kessler, J.; Volpe, F.; Evans, T. E.; Ali, H.; Punjabi, A.

2009-11-01

Large overlapping magnetic islands generate chaotic fields. However, a previous work [1] showed that second or third order perturbations of special topology and strength can also generate magnetic diffusion ``barriers" in the middle of stochastic regions. In the present study, we numerically assess their experimental feasibility at DIII-D. For this, realistic I- and C-coils perturbations are superimposed on the equilibrium field and puncture plots are generated with a field-line tracer. A criterion is defined for the automatic recognition of barriers and successfully tested on earlier symplectic maps in magnetic coordinates. The criterion is systematically applied to the new puncture plots in search for dependencies, e.g. upon the edge safety factor q95, which might be relevant to edge localized mode (ELM) stability, as well as to assess the robustness of barriers against fluctuations of the plasma parameters and coil currents. 8pt [1] H. Ali and A. Punjabi, Plasma Phys. Control. Fusion 49, 1565 (2007).

Nonlinear MHD study on the influence of E×B flow in QH-mode plasma of DIII-D

NASA Astrophysics Data System (ADS)

Liu, Feng; Huijsmans, Guido; Loarte, Alberto; Garofalo, Andrea; Solomon, Wayne; Nkonga, Boniface; Hoelzl, Matthias

2017-10-01

In QH-mode experiments with zero-net NBI torque show that there remains a finite E×B rotation in the pedestal region implying that a minimum E×B flow or flow shear is required for the plasma to develop the Edge Harmonic Oscillation (EHO), which is a saturated KPM (kink-peeling mode) characteristic of the QH-mode. To understand the roles of E×B flow and its shear in the saturation of KPMs, non-linear MHD simulations of DIII-D QH-mode plasmas including toroidal mode numbers n = 0 to 10 with different E×B rotation speed have been performed. These simulation show that ExB rotation strongly stabilizes high-n modes but destabilizes low-n modes (particularly the n =2 mode) in the linear growth phase, which is consistent experimental observations and previous linear MHD modelling. US DOE under DE-FC02-04ER54698.

Modeling of high-Z materials erosion and its suppression in DIII-D

NASA Astrophysics Data System (ADS)

Ding, Rui; Guo, H. Y.; Chan, V. S.; Snyder, P. B.; Rudakov, D. L.; Stangeby, P. C.; Elder, J. D.; Tskhakaya, D.; Wampler, W. R.; Kirschner, A.; McLean, A. G.

2015-11-01

Erosion of plasma facing components is a key issue for high-power, long pulse operation. The 3D Monte Carlo code ERO has been used to simulate the erosion/redeposition of Mo and W samples exposed to DIII-D divertor plasma using the DiMES. The net erosion rate is significantly reduced due to the high local re-deposition ratio of eroded materials, which is mainly controlled by the electric field and plasma density within the Chodura sheath as indicated by ERO modeling. Similar re-deposition ratios were obtained from the modeling using three sheath models for small inclined magnetic field angle, all being close to the measured value. ERO modeling shows that local CH4 injection can create a carbon coating on the Mo sample to mitigate Mo erosion; the local decrease of electron temperature due to gas injection also suppresses net erosion, consistent with experimental observation. Supported by the US DOE under DE-FC02-04ER54698 and PSI-SciDAC project.

Flux tube gyrokinetic simulations of the edge pedestal

NASA Astrophysics Data System (ADS)

Parker, Scott; Wan, Weigang; Chen, Yang

2011-10-01

The linear instabilities of DIII-D H-mode pedestal are studied with gyrokinetic micro-turbulence simulations. The simulation code GEM is an electromagnetic δf code with global tokamak geometry in the form of Miller equilibrium. Local flux tube simulations are carried out for multiple positions of two DIII-D profiles: shot #98889 and shot #131997. Near the top of the pedestal, the instability is clearly ITG. The dominant instability of the pedestal appears at the steep gradient region, and it is identified as a low frequency mode mostly driven by electron temperature gradient. The mode propagates along the electron diamagnetic direction for low n and may propagate along the ion direction for high n. At some positions near the steep gradient region, an ion instability is found which shows some characteristics of kinetic ballooning mode (KBM). These results will be compared to the results of E. Wang et al. and D. Fulton et al. in the same session. We thank R. Groebner and P. Snyder for providing experimental profiles and helpful discussions.

X-ray and extreme ultraviolet spectroscopy on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Victor, Brian S.; Allen, Steve L.; Beiersdorfer, P.

Two spectrometers were installed to measure tungsten emission in the core of DIII-D plasmas during a metal rings experimental campaign. The spectral range of the high-resolution (1340 spectral channels), variable-ruled grating X-ray and Extreme Ultraviolet Spectrometer (XEUS) extends from10–71more » $$\\dot{A}$$ . The spectral range of the second spectrometer, the Long-Wavelength Extreme Ultraviolet Spectrometer (LoWEUS), measures between 31–174$$\\dot{A}$$ . Three groups of tungsten lines were identified with XEUS: W 38+-W 45+ from 47–63$$\\dot{A}$$ , W 27+-W 35+ from 45–55$$\\dot{A}$$ , and W 28+-W 33+ from 16–30$$\\dot{A}$$ . Emission lines from tungsten charge states W 28+, W 43+, W 44+, and W 45+ are identified and the line amplitude is presented versus time. Peak emission of W 43+-W 45+ occurs between core Te=2.5-3 keV, and peak emission of W28+ occurs at core Te 1:3 keV. One group of tungsten lines, W 40+-W 45+, between 120–140$$\\dot{A}$$ , was identified with LoWEUS. W 43+- W 45+ lines measured with LoWEUS track the sawtooth cycle. Furthermore, sensitivity to the sawtooth cycle and the correlation of the peak emission with core electron temperature show that these spectrometers track the on-axis tungsten emission of DIII-D plasmas.« less

X-ray and extreme ultraviolet spectroscopy on DIII-D

DOE PAGES

Victor, Brian S.; Allen, Steve L.; Beiersdorfer, P.; ...

2017-06-14

Two spectrometers were installed to measure tungsten emission in the core of DIII-D plasmas during a metal rings experimental campaign. The spectral range of the high-resolution (1340 spectral channels), variable-ruled grating X-ray and Extreme Ultraviolet Spectrometer (XEUS) extends from10–71more » $$\\dot{A}$$ . The spectral range of the second spectrometer, the Long-Wavelength Extreme Ultraviolet Spectrometer (LoWEUS), measures between 31–174$$\\dot{A}$$ . Three groups of tungsten lines were identified with XEUS: W 38+-W 45+ from 47–63$$\\dot{A}$$ , W 27+-W 35+ from 45–55$$\\dot{A}$$ , and W 28+-W 33+ from 16–30$$\\dot{A}$$ . Emission lines from tungsten charge states W 28+, W 43+, W 44+, and W 45+ are identified and the line amplitude is presented versus time. Peak emission of W 43+-W 45+ occurs between core Te=2.5-3 keV, and peak emission of W28+ occurs at core Te 1:3 keV. One group of tungsten lines, W 40+-W 45+, between 120–140$$\\dot{A}$$ , was identified with LoWEUS. W 43+- W 45+ lines measured with LoWEUS track the sawtooth cycle. Furthermore, sensitivity to the sawtooth cycle and the correlation of the peak emission with core electron temperature show that these spectrometers track the on-axis tungsten emission of DIII-D plasmas.« less

Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

2000-12-01

Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

1.5D quasilinear model and its application on beams interacting with Alfvén eigenmodes in DIII-D

NASA Astrophysics Data System (ADS)

Ghantous, K.; Gorelenkov, N. N.; Berk, H. L.; Heidbrink, W. W.; Van Zeeland, M. A.

2012-09-01

We propose a model, denoted here by 1.5D, to study energetic particle (EP) interaction with toroidal Alfvenic eigenmodes (TAE) in the case where the local EP drive for TAE exceeds the stability limit. Based on quasilinear theory, the proposed 1.5D model assumes that the particles diffuse in phase space, flattening the pressure profile until its gradient reaches a critical value where the modes stabilize. Using local theories and NOVA-K simulations of TAE damping and growth rates, the 1.5D model calculates the critical gradient and reconstructs the relaxed EP pressure profile. Local theory is improved from previous study by including more sophisticated damping and drive mechanisms such as the numerical computation of the effect of the EP finite orbit width on the growth rate. The 1.5D model is applied on the well-diagnosed DIII-D discharges #142111 [M. A. Van Zeeland et al., Phys. Plasmas 18, 135001 (2011)] and #127112 [W. W. Heidbrink et al., Nucl. Fusion. 48, 084001 (2008)]. We achieved a very satisfactory agreement with the experimental results on the EP pressure profiles redistribution and measured losses. This agreement of the 1.5D model with experimental results allows the use of this code as a guide for ITER plasma operation where it is desired to have no more than 5% loss of fusion alpha particles as limited by the design.

Real Time Computer Control of Neutral Beam Energy and Current During a DIII-D Tokamak Shot

NASA Astrophysics Data System (ADS)

Pawley, C. J.; Pace, D. C.; Rauch, J. M.; Scoville, J. T.

2017-10-01

A new control system has been implemented on DIII-D neutral beams which has been used during the 2016 and 2017 experimental campaign to directly change the beam acceleration voltage (V) and beam current (I) by the Plasma Control System (PCS) during a shot. Small changes in the beam voltage of 1-2 kV can be made in 1 msec or larger changes of up to 20kV in 0.5 seconds. The beam current can be modified by as much as +/-20% at a fixed beam voltage. Since both can be independently and simultaneously changed it is possible to change beam power (IV) at fixed voltage, keep constant power while sweeping beam voltage, or to maintain minimum beam divergence during a beam voltage sweep by changing I simultaneously to keep a constant beam perveance. The limitations of the variability will be presented with required changes in equipment to extend either the speed or range of the controls. Some of the effects on fast ion plasma instabilities or other plasma mode changes made possible by this control will also be presented (see also D.C. Pace, this conference). Design and changes to the control system was performed under General Atomics Internal Research and Development support, while plasma experiments on DIII-D were supported in part by the US Department of Energy under Award No. DE-FC02-04ER54698.

Connection between plasma response and resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression in DIII-D [Connection between plasma response and RMP ELM suppression in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wingen, Andreas; Ferraro, Nathaniel M.; Shafer, Morgan W.

Calculations of the plasma response to applied non-axisymmetric fields in several DIII-D discharges show that predicted displacements depend strongly on the edge current density. This result is found using both a linear two-fluid-MHD model (M3D-C1) and a nonlinear ideal-MHD model (VMEC). Furthermore, it is observed that the probability of a discharge being edge localized mode (ELM)-suppressed is most closely related to the edge current density, as opposed to the pressure gradient. It is found that discharges with a stronger kink response are closer to the peeling–ballooning stability limit in ELITE simulations and eventually cross into the unstable region, causing ELMsmore » to reappear. Thus for effective ELM suppression, the RMP has to prevent the plasma from generating a large kink response, associated with ELM instability. Experimental observations are in agreement with the finding; discharges which have a strong kink response in the MHD simulations show ELMs or ELM mitigation during the RMP phase of the experiment, while discharges with a small kink response in the MHD simulations are fully ELM suppressed in the experiment by the applied resonant magnetic perturbation. The results are cross-checked against modeled 3D ideal MHD equilibria using the VMEC code. The procedure of constructing optimal 3D equilibria for diverted H-mode discharges using VMEC is presented. As a result, kink displacements in VMEC are found to scale with the edge current density, similar to M3D-C1, but the displacements are smaller. A direct correlation in the flux surface displacements to the bootstrap current is shown.« less

Connection between plasma response and resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression in DIII-D [Connection between plasma response and RMP ELM suppression in DIII-D

DOE PAGES

Wingen, Andreas; Ferraro, Nathaniel M.; Shafer, Morgan W.; ...

2015-09-03

Calculations of the plasma response to applied non-axisymmetric fields in several DIII-D discharges show that predicted displacements depend strongly on the edge current density. This result is found using both a linear two-fluid-MHD model (M3D-C1) and a nonlinear ideal-MHD model (VMEC). Furthermore, it is observed that the probability of a discharge being edge localized mode (ELM)-suppressed is most closely related to the edge current density, as opposed to the pressure gradient. It is found that discharges with a stronger kink response are closer to the peeling–ballooning stability limit in ELITE simulations and eventually cross into the unstable region, causing ELMsmore » to reappear. Thus for effective ELM suppression, the RMP has to prevent the plasma from generating a large kink response, associated with ELM instability. Experimental observations are in agreement with the finding; discharges which have a strong kink response in the MHD simulations show ELMs or ELM mitigation during the RMP phase of the experiment, while discharges with a small kink response in the MHD simulations are fully ELM suppressed in the experiment by the applied resonant magnetic perturbation. The results are cross-checked against modeled 3D ideal MHD equilibria using the VMEC code. The procedure of constructing optimal 3D equilibria for diverted H-mode discharges using VMEC is presented. As a result, kink displacements in VMEC are found to scale with the edge current density, similar to M3D-C1, but the displacements are smaller. A direct correlation in the flux surface displacements to the bootstrap current is shown.« less

Enabling co-simulation of tokamak plant models and plasma control systems

DOE PAGES

Walker, M. L.

2017-12-22

A system for connecting the Plasma Control System and a model of the tokamak Plant in closed loop co-simulation for plasma control development has been in routine use at DIII-D for more than 20 years and at other fusion labs that use variants of the DIII-D PCS for approximately the last decade. Here, co-simulation refers to the simultaneous execution of two independent codes with the exchange of data - Plant actuator commands and tokamak diagnostic data - between them during execution. Interest in this type of PCS-Plant simulation technology has also been growing recently at other fusion facilities. In fact,more » use of such closed loop control simulations is assumed to play an even larger role in the development of both the ITER Plasma Control System (PCS) and the experimental operation of the ITER device, where they will be used to support verification/validation of the PCS and also for ITER pulse schedule development and validation. We describe the key use cases that motivate the co-simulation capability and the features that must be provided by the Plasma Control System to support it. These features could be provided by the PCS itself or by a model of the PCS. If the PCS itself is chosen to provide them, there are requirements imposed on its architecture. If a PCS model is chosen, there are requirements imposed on the initial implementation of this simulation as well as long-term consequences for its continued development and maintenance. We describe these issues for each use case and discuss the relative merits of the two choices. Several examples are given illustrating uses of the co-simulation method to address problems of plasma control during the operation of DIII-D and of other devices that use the DIII-D PCS.« less

Enabling co-simulation of tokamak plant models and plasma control systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walker, M. L.

A system for connecting the Plasma Control System and a model of the tokamak Plant in closed loop co-simulation for plasma control development has been in routine use at DIII-D for more than 20 years and at other fusion labs that use variants of the DIII-D PCS for approximately the last decade. Here, co-simulation refers to the simultaneous execution of two independent codes with the exchange of data - Plant actuator commands and tokamak diagnostic data - between them during execution. Interest in this type of PCS-Plant simulation technology has also been growing recently at other fusion facilities. In fact,more » use of such closed loop control simulations is assumed to play an even larger role in the development of both the ITER Plasma Control System (PCS) and the experimental operation of the ITER device, where they will be used to support verification/validation of the PCS and also for ITER pulse schedule development and validation. We describe the key use cases that motivate the co-simulation capability and the features that must be provided by the Plasma Control System to support it. These features could be provided by the PCS itself or by a model of the PCS. If the PCS itself is chosen to provide them, there are requirements imposed on its architecture. If a PCS model is chosen, there are requirements imposed on the initial implementation of this simulation as well as long-term consequences for its continued development and maintenance. We describe these issues for each use case and discuss the relative merits of the two choices. Several examples are given illustrating uses of the co-simulation method to address problems of plasma control during the operation of DIII-D and of other devices that use the DIII-D PCS.« less

Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

DOE PAGES

Wang, Y. M.; Xu, X. Q.; Yan, Z.; ...

2018-01-05

A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n=30 ormore » $$k_\\theta\\rho_i\\sim0.12$$ . The ion diamagnetic drift and $$E\\times B$$ convection flow are balanced when the radial electric field (E r) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density $$n_e\\sim1.5\\times10^{19}$$ m -3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40kHz and 10 kHz respectively. The poloidal wave number $$k_\\theta$$ is about 0.2 cm -1 ($$k_\\theta\\rho_i\\sim0.05$$ ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are~3.5–6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. In conclusion, the electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.« less

Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Y. M.; Xu, X. Q.; Yan, Z.

A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n=30 ormore » $$k_\\theta\\rho_i\\sim0.12$$ . The ion diamagnetic drift and $$E\\times B$$ convection flow are balanced when the radial electric field (E r) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density $$n_e\\sim1.5\\times10^{19}$$ m -3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40kHz and 10 kHz respectively. The poloidal wave number $$k_\\theta$$ is about 0.2 cm -1 ($$k_\\theta\\rho_i\\sim0.05$$ ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are~3.5–6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. In conclusion, the electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.« less

Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-06-02

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).« less

Alpha-channeling simulation experiment in the DIII-D tokamak.

PubMed

Wong, K L; Budny, R; Nazikian, R; Petty, C C; Greenfield, C M; Heidbrink, W W; Ruskov, E

2004-08-20

Alfvén instabilities can reduce the central magnetic shear via redistribution of energetic ions. They can sustain a steady state internal transport barrier as demonstrated in this DIII-D tokamak experiment. Improvement in burning plasma performance based on this mechanism is discussed.

Using AORSA to simulate helicon waves in DIII-D

NASA Astrophysics Data System (ADS)

Lau, C.; Jaeger, E. F.; Bertelli, N.; Berry, L. A.; Blazevski, D.; Green, D. L.; Murakami, M.; Park, J. M.; Pinsker, R. I.; Prater, R.

2015-12-01

Recent efforts have shown that helicon waves (fast waves at > 20ωci) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.

Kinetic neoclassical transport in the H-mode pedestal

DOE PAGES

Battaglia, D. J.; Burrell, K. H.; Chang, C. S.; ...

2014-07-16

Multi-species kinetic neoclassical transport through the QH-mode pedestal and scrapeoff layer on DIII-D is calculated using XGC0, a 5D full-f particle-in-cell drift-kinetic solver with self-consistent neutral recycling and sheath potentials. We achieved quantitative agreement between the fluxdriven simulation and the experimental electron density, impurity density and orthogonal measurements of impurity temperature and flow profiles by adding random-walk particle diffusion to the guiding-center drift motion. Furthermore, we computed the radial electric field (Er) that maintains ambipolar transport across flux surfaces and to the wall self-consistently on closed and open magnetic field lines, and is in excellent agreement with experiment. The Ermore » inside the separatrix is the unique solution that balances the outward flux of thermal tail deuterium ions against the outward neoclassical electron flux and inward pinch of impurity and colder deuterium ions. Particle transport in the pedestal is primarily due to anomalous transport, while the ion heat and momentum transport is primarily due to the neoclassical transport. The full-f treatment quantifies the non-Maxwellian energy distributions that describe a number of experimental observations in low-collisionallity pedestals on DIII-D, including intrinsic co-Ip parallel flows in the pedestal, ion temperature anisotropy and large impurity temperatures in the scrape-off layer.« less

Drifts, currents, and power scrape-off width in SOLPS-ITER modeling of DIII-D

DOE PAGES

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; ...

2016-12-27

The effects of drifts and associated flows and currents on the width of the parallel heat flux channel (λ q) in the tokamak scrape-off layer (SOL) are analyzed using the SOLPS-ITER 2D fluid transport code. Motivation is supplied by Goldston’s heuristic drift (HD) model for λ q, which yields the same approximately inverse poloidal magnetic field dependence seen in multi-machine regression. The analysis, focusing on a DIII-D H-mode discharge, reveals HD-like features, including comparable density and temperature fall-off lengths in the SOL, and up-down ion pressure asymmetry that allows net cross-separatrix ion magnetic drift flux to exceed net anomalous ionmore » flux. In experimentally relevant high-recycling cases, scans of both toroidal and poloidal magnetic field (B tor and B pol) are conducted, showing minimal λ q dependence on either component of the field. Insensitivity to B tor is expected, and suggests that SOLPS-ITER is effectively capturing some aspects of HD physics. Absence of λ q dependence on B pol, however, is inconsistent with both the HD model and experimental results. As a result, the inconsistency is attributed to strong variation in the parallel Mach number, which violates one of the premises of the HD model.« less

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

2016-06-02

A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

Thermal management of tungsten leading edges in DIII-D

DOE PAGES

Nygren, Richard E.; Rudakov, Dmitry L.; Murphy, Christopher; ...

2017-04-29

The DiMES materials probe exposed tungsten blocks with 0.3 and 1 mm high leading edges to DIII-D He plasmas in 2015 and 2016 viewed with high resolution IRTV. The 1-mm edge may have reached >2400° C in a 3-s shot with a (parallel) heat load of ~50 MW/m 2 and ~10 MW/m 2 on the surface based on modeling. The experiments support ITER. Leading edges were also a concern in the DIII-D Metal Tile Experiment in 2016. Two toroidal rings of divertor tiles had W-coated molybdenum inserts 50 mm wide radially. This study presents data and thermal analyses.

Thermal management of tungsten leading edges in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nygren, Richard E.; Rudakov, Dmitry L.; Murphy, Christopher

The DiMES materials probe exposed tungsten blocks with 0.3 and 1 mm high leading edges to DIII-D He plasmas in 2015 and 2016 viewed with high resolution IRTV. The 1-mm edge may have reached >2400° C in a 3-s shot with a (parallel) heat load of ~50 MW/m 2 and ~10 MW/m 2 on the surface based on modeling. The experiments support ITER. Leading edges were also a concern in the DIII-D Metal Tile Experiment in 2016. Two toroidal rings of divertor tiles had W-coated molybdenum inserts 50 mm wide radially. This study presents data and thermal analyses.

In-situ Testing of the EHT High Gain and Frequency Ultra-Stable Integrators

NASA Astrophysics Data System (ADS)

Miller, Kenneth; Ziemba, Timothy; Prager, James; Slobodov, Ilia; Lotz, Dan

2014-10-01

Eagle Harbor Technologies (EHT) has developed a long-pulse integrator that exceeds the ITER specification for integration error and pulse duration. During the Phase I program, EHT improved the RPPL short-pulse integrators, added a fast digital reset, and demonstrated that the new integrators exceed the ITER integration error and pulse duration requirements. In Phase II, EHT developed Field Programmable Gate Array (FPGA) software that allows for integrator control and real-time signal digitization and processing. In the second year of Phase II, the EHT integrator will be tested at a validation platform experiment (HIT-SI) and tokamak (DIII-D). In the Phase IIB program, EHT will continue development of the EHT integrator to reduce overall cost per channel. EHT will test lower cost components, move to surface mount components, and add an onboard Field Programmable Gate Array and data acquisition to produce a stand-alone system with lower cost per channel and increased the channel density. EHT will test the Phase IIB integrator at a validation platform experiment (HIT-SI) and tokamak (DIII-D). Work supported by the DOE under Contract Number (DE-SC0006281).

Promoting Pre-college Science Education

NASA Astrophysics Data System (ADS)

Taylor, P. L.; Lee, R. L.

2000-10-01

The Fusion Education Program, with continued support from DOE, has strengthened its interactions with educators in promoting pre-college science education for students. Projects aggressively pursued this year include an on-site, college credited, laboratory-based 10-day educator workshop on plasma and fusion science; completion of `Starpower', a fusion power plant simulation on interactive CD; expansion of scientist visits to classrooms; broadened participation in an internet-based science olympiad; and enhancements to the tours of the DIII-D Facility. In the workshop, twelve teachers used bench top devices to explore basic plasma physics. Also included were radiation experiments, computer aided drafting, techniques to integrate fusion science and technology in the classroom, and visits to a University Physics lab and the San Diego Supercomputer Center. Our ``Scientist in a Classroom'' program reached more than 2200 students at 20 schools. Our `Starpower' CD allows a range of interactive learning from the effects of electric and magnetic fields on charged particles to operation of a Tokamak-based power plant. Continuing tours of the DIII-D facility were attended by more than 800 students this past year.

Prospects for Off-axis Current Drive via High Field Side Lower Hybrid Current Drive in DIII-D

NASA Astrophysics Data System (ADS)

Wukitch, S. J.; Shiraiwa, S.; Wallace, G. M.; Bonoli, P. T.; Holcomb, C.; Park, J. M.; Pinsker, R. I.

2017-10-01

An outstanding challenge for an economical, steady state tokamak is efficient off-axis current drive scalable to reactors. Previous studies have focused on high field side (HFS) launch of lower hybrid waves for current drive (LHCD) in double null configurations in reactor grade plasmas. The goal of this work is to find a HFS LHCD scenario for DIII-D that balances coupling, power penetration and damping. The higher magnetic field on the HFS improves wave accessibility, which allows for lower n||waves to be launched. These waves penetrate farther into the plasma core before damping at higher Te yielding a higher current drive efficiency. Utilizing advanced ray tracing and Fokker Planck simulation tools (GENRAY+CQL3D), wave penetration, absorption and drive current profiles in high performance DIII-D H-Mode plasmas were investigated. We found LH scenarios with single pass absorption, excellent wave penetration to r/a 0.6-0.8, FWHM r/a=0.2 and driven current up to 0.37 MA/MW coupled. These simulations indicate that HFS LHCD has potential to achieve efficient off-axis current drive in DIII-D and the latest results will be presented. Work supported by U.S. Dept. of Energy, Office of Science, Office of Fusion Energy Sciences, using User Facility DIII-D, under Award No. DE-FC02-04ER54698 and Contract No. DE-FC02-01ER54648 under Scientific Discovery through Advanced Computing Initiative.

Experimental investigation of stability, frequency and toroidal mode number of compressional Alfvén eigenmodes in DIII-D

NASA Astrophysics Data System (ADS)

Tang, S.; Thome, K.; Pace, D.; Heidbrink, W. W.; Carter, T. A.; Crocker, N. A.; NSTX-U Collaboration; DIII-D Collaboration

2017-10-01

An experimental investigation of the stability of Doppler-shifted cyclotron resonant compressional Alfvén eigenmodes (CAE) using the flexible DIII-D neutral beams has begun to validate a theoretical understanding and realize the CAE's diagnostic potential. CAEs are excited by energetic ions from neutral beams [Heidbrink, NF 2006], with frequencies and toroidal mode numbers sensitive to the fast-ion phase space distribution, making them a potentially powerful passive diagnostic. The experiment also contributes to a predictive capability for spherical tokamak temperature profiles, where CAEs may play a role in energy transport [Crocker, NF 2013]. CAE activity was observed using the recently developed Ion Cyclotron Emission diagnostic-high bandwidth edge magnetic sensors sampled at 200 MS/s. Preliminary results show CAEs become unstable in BT ramp discharges below a critical threshold in the range 1.7 - 1.9 T, with the exact value increasing as density increases. The experiment will be used to validate simulations from relevant codes such as the Hybrid MHD code [Belova, PRL 2015]. This work was supported by US DOE Grants DE-SC0011810 and DE-FC02-04ER54698.

Gyrokinetic simulations of DIII-D near-edge L-mode plasmas

NASA Astrophysics Data System (ADS)

Neiser, Tom; Jenko, Frank; Carter, Troy; Schmitz, Lothar; Merlo, Gabriele; Told, Daniel; Banon Navarro, Alejandro; McKee, George; Yan, Zheng

2017-10-01

In order to understand the L-H transition, a good understanding of the L-mode edge region is necessary. We perform nonlinear gyrokinetic simulations of a DIII-D L-mode discharge with the GENE code in the near-edge, which we define as ρtor >= 0.8 . At ρ = 0.9 , ion-scale simulations reproduce experimental heat fluxes within the uncertainty of the experiment. At ρ = 0 . 8 , electron-scale simulations reproduce the experimental electron heat flux while ion-scale simulations do not reproduce the respective ion heat flux due to a strong poloidal zonal flow. However, we reproduce both electron and ion heat fluxes by increasing the local ion temperature gradient by 80 % . Local fitting to the CER data in the domain 0.7 <= ρ <= 0.9 is compatible with such an increase in ion temperature gradient within the error bars. Ongoing multi-scale simulations are investigating whether radial electron streamers could dampen the poloidal zonal flows at ρ = 0.8 and increase the radial ion-scale flux. Supported by U.S. DOE under Contract Numbers DE-FG02-08ER54984, DE-FC02-04ER54698, and DE-AC02-05CH11231.

Turbulent particle transport as a function of toroidal rotation in DIII-D H-mode plasmas

DOE PAGES

Wang, Xin; Mordijck, Saskia; Zeng, Lei; ...

2016-03-01

In this paper we show how changes in toroidal rotation, by controlling the injected torque, affect particle transport and confinement. The toroidal rotation is altered using the co- and counter neutral beam injection (NBI) in low collisionality H-mode plasmas on DIII-D with dominant electron cyclotron heating (ECH). We find that there is no correlation between the toroidal rotation shear and the inverse density gradient, which is observed on AUG whenmore » $${{T}_{\\text{e}}}/{{T}_{\\text{i}}}$$ is varied using ECH (Angioni et al 2011 Phys. Rev. Lett. 107 215003). In DIII-D, we find that in a discharge with balanced torque injection, the $$E\\times B$$ shear is smaller than the linear gyrokinetic growth rate for small $${{k}_{\\theta}}{{\\rho}_{s}}$$ for $$\\rho =0.6$$ –0.85. This results in lower particle confinement. In the co- and counter- injected discharges the $$E\\times B$$ shear is larger or close to the linear growth rate at the plasma edge and both configurations have higher particle confinement. In order to measure particle transport, we use a small periodic perturbative gas puff. This gas puff perturbs the density profiles and allows us to extract the perturbed diffusion and inward pinch coefficients. We observe a strong increase in the inward particle pinch in the counter-torque injected plasma. Lastly, the calculated quasi-linear particle flux, nor the linear growth rates using TGLF agree with experimental observations.« less

Development and validation of a critical gradient energetic particle driven Alfven eigenmode transport model for DIII-D tilted neutral beam experiments

NASA Astrophysics Data System (ADS)

Waltz, R. E.; Bass, E. M.; Heidbrink, W. W.; VanZeeland, M. A.

2015-11-01

Recent experiments with the DIII-D tilted neutral beam injection (NBI) varying the beam energetic particle (EP) source profiles have provided strong evidence that unstable Alfven eigenmodes (AE) drive stiff EP transport at a critical EP density gradient [Heidbrink et al 2013 Nucl. Fusion 53 093006]. Here the critical gradient is identified by the local AE growth rate being equal to the local ITG/TEM growth rate at the same low toroidal mode number. The growth rates are taken from the gyrokinetic code GYRO. Simulation show that the slowing down beam-like EP distribution has a slightly lower critical gradient than the Maxwellian. The ALPHA EP density transport code [Waltz and Bass 2014 Nucl. Fusion 54 104006], used to validate the model, combines the low-n stiff EP critical density gradient AE mid-core transport with the Angioni et al (2009 Nucl. Fusion 49 055013) energy independent high-n ITG/TEM density transport model controling the central core EP density profile. For the on-axis NBI heated DIII-D shot 146102, while the net loss to the edge is small, about half the birth fast ions are transported from the central core r/a  <  0.5 and the central density is about half the slowing down density. These results are in good agreement with experimental fast ion pressure profiles inferred from MSE constrained EFIT equilibria.

Electron cyclotron heating and core intrinsic rotation reversal in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grassie, J. S. de, E-mail: degrassie@fusion.gat.com; Boedo, J. A.; Grierson, B. A.

2015-12-10

The effect of electron cyclotron heating (ECH) on the intrinsic rotation profile in DIII-D is shown experimentally. Former DIII-D experiments have shown that ECH tends to cause an interior reduction in the normally co-Ip directed intrinsic rotation profile, and this core rotation can be fully reversed to the opposite direction. This effect is due to a turbulent rearrangement of the interior rotation profile. Here, we show results that there is more than one mechanism causing this. We compare two low density L-mode discharges where the only operational difference is the location of the ECH deposition. At low ECH power, comparablemore » to the Ohmic power, the primary change is in the q-profile accompanied by a reversal of the core intrinsic rotation direction for the more off-axis deposition. The change in the shear of the q-profile fits well with a recent theoretical prediction for this rotation reversal. At higher ECH power, the primary change is in the core electron temperature, Te, accompanied by a hollowing of the rotation profile near the magnetic axis. This effect appears to be due to the change in electron collisionality, consistent with another theoretical, gyrokinetic prediction. The variety of phenomena that could allow ECH to modify the intrinsic rotation profile give some expectation that regions of large velocity shear in the interior could be generated, with the possibility of triggering internal transport barriers.« less

3D Equilibrium Effects Due to RMP Application on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Lazerson, E. Lazarus, S. Hudson, N. Pablant and D. Gates

2012-06-20

The mitigation and suppression of edge localized modes (ELMs) through application of resonant magnetic perturbations (RMPs) in Tokamak plasmas is a well documented phenomenon [1]. Vacuum calculations suggest the formation of edge islands and stochastic regions when RMPs are applied to the axisymmetric equilibria. Self-consistent calculations of the plasma equilibrium with the VMEC [2] and SPEC [3] codes have been performed for an up-down symmetric shot (142603) in DIII-D. In these codes, a self-consistent calculation of the plasma response due to the RMP coils is calculated. The VMEC code globally enforces the constraints of ideal MHD; consequently, a continuously nestedmore » family of flux surfaces is enforced throughout the plasma domain. This approach necessarily precludes the observation of islands or field-line chaos. The SPEC code relaxes the constraints of ideal MHD locally, and allows for islands and field line chaos at or near the rational surfaces. Equilibria with finite pressure gradients are approximated by a set of discrete "ideal-interfaces" at the most irrational flux surfaces and where the strongest pressure gradients are observed. Both the VMEC and SPEC calculations are initialized from EFIT reconstructions of the plasma that are consistent with the experimental pressure and current profiles. A 3D reconstruction using the STELLOPT code, which fits VMEC equilibria to experimental measurements, has also been performed. Comparisons between the equilibria generated by the 3D codes and between STELLOPT and EFIT are presented.« less

3D Equilibrium Effects Due to RMP Application on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lazerson, S.; Lazarus, E.; Hudson, S.

2012-06-20

The mitigation and suppression of edge localized modes (ELMs) through application of resonant magnetic perturbations (RMPs) in Tokamak plasmas is a well documented phenomenon. Vacuum calculations suggest the formation of edge islands and stochastic regions when RMPs are applied to the axisymmetric equilibria. Self-consistent calculations of the plasma equilibrium with the VMEC and SPEC codes have been performed for an up-down symmetric shot in DIII-D. In these codes, a self-consistent calculation of the plasma response due to the RMP coils is calculated. The VMEC code globally enforces the constraints of ideal MHD; consequently, a continuously nested family of flux surfacesmore » is enforced throughout the plasma domain. This approach necessarily precludes the observation of islands or field-line chaos. The SPEC code relaxes the constraints of ideal MHD locally, and allows for islands and field line chaos at or near the rational surfaces. Equilibria with finite pressure gradients are approximated by a set of discrete "ideal-interfaces" at the most irrational flux surfaces and where the strongest pressure gradients are observed. Both the VMEC and SPEC calculations are initialized from EFIT reconstructions of the plasma that are consistent with the experimental pressure and current profiles. A 3D reconstruction using the STELLOPT code, which fits VMEC equilibria to experimental measurements, has also been performed. Comparisons between the equilibria generated by the 3D codes and between STELLOPT and EFIT are presented.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okabayashi, M.; Zanca, P.; Strait, E. J.

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Benchmark testing of DIII-D neutral beam modeling with water flow calorimetry

DOE PAGES

Rauch, J. M.; Crowley, B. J.; Scoville, J. T.; ...

2016-06-02

Power loading on beamline components in the DIII-D neutral beam system is measured in this paper using water flow calorimetry. The results are used to benchmark beam transport models. Finally, anomalously high heat loads in the magnet region are investigated and a speculative hypothesis as to their origin is presented.

Using AORSA to simulate helicon waves in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, C., E-mail: lauch@ornl.gov; Blazevski, D.; Green, D. L.

2015-12-10

Recent efforts have shown that helicon waves (fast waves at > 20ω{sub ci}) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored,more » it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.« less

Using AORSA to simulate helicon waves in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, Cornwall H; Jaeger, E. F.; Bertelli, Nicola

2015-01-01

Recent efforts have shown that helicon waves (fast waves at >20 omega(ci)) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, itmore » will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.« less

Sawtooth mitigation in 3D MHD tokamak modelling with applied magnetic perturbations

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Veranda, M.; Cappello, S.; Chacón, L.; Escande, D. F.

2017-01-01

The effect of magnetic perturbations (MPs) on the sawtoothing dynamics of the internal kink mode in the tokamak is discussed in the framework of nonlinear 3D MHD modelling. Numerical simulations are performed with the pixie3d code (Chacón 2008 Phys. Plasmas 15 056103) based on a D-shaped configuration in toroidal geometry. MPs are applied as produced by two sets of coils distributed along the toroidal direction, one set located above and the other set below the outboard midplane, like in experimental devices such as DIII-D and ASDEX Upgrade. The capability of n  =  1 MPs to affect quasi-periodic sawteeth is shown to depend on the toroidal phase difference Δ φ between the perturbations produced by the two sets of coils. In particular, sawtooth mitigation is obtained for the Δ φ =π phasing, whereas no significant effect is observed for Δ φ =0 . Numerical findings are explained by the interplay between different poloidal harmonics in the spectrum of applied MPs, and appear to be consistent with experiments performed in the DIII-D device. Sawtooth mitigation and stimulation of self-organized helical states by applied MPs have been previously demonstrated in both circular tokamak and reversed-field pinch (RFP) experiments in the RFX-mod device, and in related 3D MHD modelling.

Interrupting an Imminent Body Current Fault and Restoring Full Power in Milliseconds on a DIII-D National Fusion Facility Gyrotron

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ponce, Dan; Brambila, Rigo E.; Cengher, Mirela

The ECH Group at DIII-D has installed in-house engineered, FPGA-based, high voltage reference waveform generators on its gyrotron control systems to enhance the capabilities of the systems and replace obsolete equipment. The new hardware, named D-Wavegen, outputs 16-bit signals every microsecond and can respond to events and anomalies in real-time. These generators have been reliably pausing gyrotron rf output during periods of DIII-D plasma density that exceed the fault density trip level and restarting the rf output if the density falls below the trip level. While tightly monitoring gyrotron body current and internal pressure, D-Wavegen has also been reliably restarting,more » in a little over 10ms, gyrotrons that spontaneously ceased rf generation.« less

Interrupting an Imminent Body Current Fault and Restoring Full Power in Milliseconds on a DIII-D National Fusion Facility Gyrotron

DOE PAGES

Ponce, Dan; Brambila, Rigo E.; Cengher, Mirela; ...

2017-10-19

The ECH Group at DIII-D has installed in-house engineered, FPGA-based, high voltage reference waveform generators on its gyrotron control systems to enhance the capabilities of the systems and replace obsolete equipment. The new hardware, named D-Wavegen, outputs 16-bit signals every microsecond and can respond to events and anomalies in real-time. These generators have been reliably pausing gyrotron rf output during periods of DIII-D plasma density that exceed the fault density trip level and restarting the rf output if the density falls below the trip level. While tightly monitoring gyrotron body current and internal pressure, D-Wavegen has also been reliably restarting,more » in a little over 10ms, gyrotrons that spontaneously ceased rf generation.« less

CAMAC throughput of a new RISC-based data acquisition computer at the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Vanderlaan, J. F.; Cummings, J. W.

1993-10-01

The amount of experimental data acquired per plasma discharge at DIII-D has continued to grow. The largest shot size in May 1991 was 49 Mbyte; in May 1992, 66 Mbyte; and in April 1993, 80 Mbyte. The increasing load has prompted the installation of a new Motorola 88100-based MODCOMP computer to supplement the existing core of three older MODCOMP data acquisition CPU's. New Kinetic Systems CAMAC serial highway driver hardware runs on the 88100 VME bus. The new operating system is MODCOMP REAL/IX version of AT&T System V UNIX with real-time extensions and networking capabilities; future plans call for installation of additional computers of this type for tokamak and neutral beam control functions. Experiences with the CAMAC hardware and software will be chronicled, including observation of data throughput. The Enhanced Serial Highway crate controller is advertised as twice as fast as the previous crate controller, and computer I/O speeds are expected to also increase data rates.

Predictive Rotation Profile Control for the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Wehner, W. P.; Schuster, E.; Boyer, M. D.; Walker, M. L.; Humphreys, D. A.

2017-10-01

Control-oriented modeling and model-based control of the rotation profile are employed to build a suitable control capability for aiding rotation-related physics studies at DIII-D. To obtain a control-oriented model, a simplified version of the momentum balance equation is combined with empirical representations of the momentum sources. The control approach is rooted in a Model Predictive Control (MPC) framework to regulate the rotation profile while satisfying constraints associated with the desired plasma stored energy and/or βN limit. Simple modifications allow for alternative control objectives, such as maximizing the plasma rotation while maintaining a specified input torque. Because the MPC approach can explicitly incorporate various types of constraints, this approach is well suited to a variety of control objectives, and therefore serves as a valuable tool for experimental physics studies. Closed-loop TRANSP simulations are presented to demonstrate the effectiveness of the control approach. Supported by the US DOE under DE-SC0010661 and DE-FC02-04ER54698.

Transient-Free Operations With Physics-Based Real-time Analysis and Control

NASA Astrophysics Data System (ADS)

Kolemen, Egemen; Burrell, Keith; Eggert, William; Eldon, David; Ferron, John; Glasser, Alex; Humphreys, David

2016-10-01

In order to understand and predict disruptions, the two most common methods currently employed in tokamak analysis are the time-consuming ``kinetic EFITs,'' which are done offline with significant human involvement, and the search for correlations with global precursors using various parameterization techniques. We are developing automated ``kinetic EFITs'' at DIII-D to enable calculation of the stability as the plasma evolves close to the disruption. This allows us to quantify the probabilistic nature of the stability calculations and provides a stability metric for all possible linear perturbations to the plasma. This study also provides insight into how the control system can avoid the unstable operating space, which is critical for high-performance operations close to stability thresholds at ITER. A novel, efficient ideal stability calculation method and new real-time CER acquisition system are being developed, and a new 77-core server has been installed on the DIII-D PCS to enable experimental use. Sponsored by US DOE under DE-SC0015878 and DE-FC02-04ER54698.

Characterization of cross-section correction to charge exchange recombination spectroscopy rotation measurements using co- and counter-neutral-beam views.

PubMed

Solomon, W M; Burrell, K H; Feder, R; Nagy, A; Gohil, P; Groebner, R J

2008-10-01

Measurements of rotation using charge exchange recombination spectroscopy can be affected by the energy dependence of the charge exchange cross section. On DIII-D, the associated correction to the rotation can exceed 100 kms at high temperatures. In reactor-relevant low rotation conditions, the correction can be several times larger than the actual plasma rotation and therefore must be carefully validated. New chords have been added to the DIII-D CER diagnostic to view the counter-neutral-beam line. The addition of these views allows determination of the toroidal rotation without depending on detailed atomic physics calculations, while also allowing experimental characterization of the atomic physics. A database of rotation comparisons from the two views shows that the calculated cross-section correction can adequately describe the measurements, although there is a tendency for "overcorrection." In cases where accuracy better than about 15% is desired, relying on calculation of the cross-section correction may be insufficient.

\\mathscr{H}_2 optimal control techniques for resistive wall mode feedback in tokamaks

NASA Astrophysics Data System (ADS)

Clement, Mitchell; Hanson, Jeremy; Bialek, Jim; Navratil, Gerald

2018-04-01

DIII-D experiments show that a new, advanced algorithm enables resistive wall mode (RWM) stability control in high performance discharges using external coils. DIII-D can excite strong, locked or nearly locked external kink modes whose rotation frequencies and growth rates are on the order of the magnetic flux diffusion time of the vacuum vessel wall. Experiments have shown that modern control techniques like linear quadratic Gaussian (LQG) control require less current than the proportional controller in use at DIII-D when using control coils external to DIII-D’s vacuum vessel. Experiments were conducted to develop control of a rotating n  =  1 perturbation using an LQG controller derived from VALEN and external coils. Feedback using this LQG algorithm outperformed a proportional gain only controller in these perturbation experiments over a range of frequencies. Results from high βN experiments also show that advanced feedback techniques using external control coils may be as effective as internal control coil feedback using classical control techniques.

Scaling of the stochastic broadening from low mn, high mn, and peeling-ballooning magnetic perturbations in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Zhao, Michael; Punjabi, Alkesh; Ali, Halima

2009-11-01

The equilibrium EFIT data for the DIII-D shot 115467 is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [A. Punjabi, and H. Ali, Phys. Plasmas 15, 122502 (2008)]. A canonical transformation is used to construct an area-preserving map for field line trajectories in the natural canonical coordinates in the DIII-D. Maps in natural canonical coordinates have the advantage that natural canonical coordinates can be inverted to calculate real space coordinates (R,Z,φ), and there is no problem in crossing the separatrix. This is not possible for magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)]. This map is applied to calculate stochastic broadening from the low mn (m,n)=(1,1)+(1,-1); high mn (m,n)=(4,1)+(3,1); and the peeling-ballooning (m,n)=(40,10)+(30,10) magnetic perturbations. In all three cases, the scaling of the widths of stochastic layer near the X-point in the principal plane of the DIII-D deviates at most by 6% from the .5ex1 -.1em/ -.15em.25ex2 power Boozer-Rechester scaling [A. Boozer, and A. Rechester, Phys. Fluids 21, 682 (1978)]. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

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2013-02-20

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Interaction of external n = 1 magnetic fields with the sawtooth instability in low- q RFX-mod and DIII-D tokamaks

DOE PAGES

Piron, C.; Martin, P.; Bonfiglio, D.; ...

2016-08-11

External n = 1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m = 1, n = 1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigationmore » in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q 95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n = 1 plasma response, that indicates an enhancement of the coupling to the marginally stable n = 1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Finally, numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.« less

Testing neoclassical and turbulent effects on poloidal rotation in the core of DIII-D

DOE PAGES

Chrystal, Colin; Burrell, Keith H.; Grierson, Brian A.; ...

2014-07-09

Experimental tests of ion poloidal rotation theories have been performed on DIII-D using a novel impurity poloidal rotation diagnostic. These tests show significant disagreements with theoretical predictions in various conditions, including L-mode plasmas with internal transport barriers (ITB), H-mode plasmas, and QH-mode plasmas. The theories tested include standard neoclassical theory, turbulence driven Reynolds stress, and fast-ion friction on the thermal ions. Poloidal rotation is observed to spin up at the formation of an ITB and makes a significant contribution to the measurement of themore » $$\\vec{E}$$ × $$\\vec{B}$$ shear that forms the ITB. In ITB cases, neoclassical theory agrees quantitatively with the experimental measurements only in the steep gradient region. Significant quantitative disagreement with neoclassical predictions is seen in the cores of ITB, QH-, and H-mode plasmas, demonstrating that neoclassical theory is an incomplete description of poloidal rotation. The addition of turbulence driven Reynolds stress does not remedy this disagreement; linear stability calculations and Doppler backscattering measurements show that disagreement increases as turbulence levels decline. Furthermore, the effect of fast-ion friction, by itself, does not lead to improved agreement; in QH-mode plasmas, neoclassical predictions are closest to experimental results in plasmas with the largest fast ion friction. Finally, predictions from a new model that combines all three effects show somewhat better agreement in the H-mode case, but discrepancies well outside the experimental error bars remain.« less

Overview of Recent DIII-D Experimental Results

NASA Astrophysics Data System (ADS)

Fenstermacher, Max; DIII-D Team

2017-10-01

Recent DIII-D experiments contributed to the ITER physics basis and to physics understanding for extrapolation to future devices. A predict-first analysis showed how shape can enhance access to RMP ELM suppression. 3D equilibrium changes from ELM control RMPs, were linked to density pumpout. Ion velocity imaging in the SOL showed 3D C2+flow perturbations near RMP induced n =1 islands. Correlation ECE reveals a 40% increase in Te turbulence during QH-mode and 70% during RMP ELM suppression vs. ELMing H-mode. A long-lived predator-prey oscillation replaces edge MHD in recent low-torque QH-mode plasmas. Spatio-temporally resolved runaway electron measurements validate the importance of synchrotron and collisional damping on RE dissipation. A new small angle slot divertor achieves strong plasma cooling and facilitates detachment access. Fast ion confinement was improved in high q_min scenarios using variable beam energy optimization. First reproducible, stable ITER baseline scenarios were established. Studies have validated a model for edge momentum transport that predicts the pedestal main-ion intrinsic velocity value and direction. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

Joint DIII-D/EAST research on the development of a high poloidal beta scenario for the steady state missions of ITER and CFETR

NASA Astrophysics Data System (ADS)

Garofalo, A. M.; Gong, X. Z.; Ding, S. Y.; Huang, J.; McClenaghan, J.; Pan, C. K.; Qian, J.; Ren, Q. L.; Staebler, G. M.; Chen, J.; Cui, L.; Grierson, B. A.; Hanson, J. M.; Holcomb, C. T.; Jian, X.; Li, G.; Li, M.; Pankin, A. Y.; Peysson, Y.; Zhai, X.; Bonoli, P.; Brower, D.; Ding, W. X.; Ferron, J. R.; Guo, W.; Lao, L. L.; Li, K.; Liu, H.; Lyv, B.; Xu, G.; Zang, Q.

2018-01-01

Experimental and modeling investigations on the DIII-D and EAST tokamaks show the attractive transport and stability properties of fully noninductive, high poloidal-beta (β P ) plasmas, and their suitability for steady-state operating scenarios in ITER and CFETR. A key feature of the high-β P regime is the large-radius (ρ > 0.6) internal transport barrier (ITB), often observed in all channels (ne, Te, Ti, rotation), and responsible for both excellent energy confinement quality and excellent stability properties. Experiments on DIII-D have shown that, with a large-radius ITB, very high β N and β P values (both ≥ 4) can be reached by taking advantage of the stabilizing effect of a nearby conducting wall. Synergistically, higher plasma pressure provides turbulence suppression by Shafranov shift, leading to ITB sustainment independent of the plasma rotation. Experiments on EAST have been used to assess the long pulse potential of the high-β P regime. Using RF-only heating and current drive, EAST achieved minute-long fully noninductive steady state H-mode operation with strike points on an ITER-like tungsten divertor. Improved confinement (relative to standard H-mode) and steady state ITB features are observed with a monotonic q-profile with q min ˜ 1.5. Separately, experiments have shown that increasing the density in plasmas driven by lower hybrid wave broadens the q-profile, a technique that could enable a large radius ITB. These experimental results have been used to validate MHD, current drive, and turbulent transport models, and to project the high-β P regime to a burning plasma. These projections suggest the Shafranov shift alone will not suffice to provide improved confinement (over standard H-mode) without rotation and rotation shear. However, increasing the negative magnetic shear (higher q on axis) provides a similar turbulence suppression mechanism to Shafranov shift, and can help devices such as ITER and CFETR achieve their steady-state fusion goals.

Modeling of 3D magnetic equilibrium effects on edge turbulence stability during RMP ELM suppression in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilcox, R. S.; Wingen, Andreas; Cianciosa, Mark R.

Some recent experimental observations have found turbulent fluctuation structures that are non-axisymmetric in a tokamak with applied 3D fields. Here, two fluid resistive effects are shown to produce changes relevant to turbulent transport in the modeled 3D magnetohydrodynamic (MHD) equilibrium of tokamak pedestals with these 3D fields applied. Ideal MHD models are insufficient to reproduce the relevant effects. By calculating the ideal 3D equilibrium using the VMEC code, the geometric shaping parameters that determine linear turbulence stability, including the normal curvature and local magnetic shear, are shown to be only weakly modified by applied 3D fields in the DIII-D tokamak.more » These ideal MHD effects are therefore not sufficient to explain the observed changes to fluctuations and transport. Using the M3D-C1 code to model the 3D equilibrium, density is shown to be redistributed on flux surfaces in the pedestal when resistive two fluid effects are included, while islands are screened by rotation in this region. Furthermore, the redistribution of density results in density and pressure gradient scale lengths that vary within pedestal flux surfaces between different helically localized flux tubes. This would produce different drive terms for trapped electron mode and kinetic ballooning mode turbulence, the latter of which is expected to be the limiting factor for pedestal pressure gradients in DIII-D.« less

Modeling of 3D magnetic equilibrium effects on edge turbulence stability during RMP ELM suppression in tokamaks

DOE PAGES

Wilcox, R. S.; Wingen, Andreas; Cianciosa, Mark R.; ...

2017-07-28

Some recent experimental observations have found turbulent fluctuation structures that are non-axisymmetric in a tokamak with applied 3D fields. Here, two fluid resistive effects are shown to produce changes relevant to turbulent transport in the modeled 3D magnetohydrodynamic (MHD) equilibrium of tokamak pedestals with these 3D fields applied. Ideal MHD models are insufficient to reproduce the relevant effects. By calculating the ideal 3D equilibrium using the VMEC code, the geometric shaping parameters that determine linear turbulence stability, including the normal curvature and local magnetic shear, are shown to be only weakly modified by applied 3D fields in the DIII-D tokamak.more » These ideal MHD effects are therefore not sufficient to explain the observed changes to fluctuations and transport. Using the M3D-C1 code to model the 3D equilibrium, density is shown to be redistributed on flux surfaces in the pedestal when resistive two fluid effects are included, while islands are screened by rotation in this region. Furthermore, the redistribution of density results in density and pressure gradient scale lengths that vary within pedestal flux surfaces between different helically localized flux tubes. This would produce different drive terms for trapped electron mode and kinetic ballooning mode turbulence, the latter of which is expected to be the limiting factor for pedestal pressure gradients in DIII-D.« less

The radial electric field as a measure for field penetration of resonant magnetic perturbations

DOE PAGES

Mordijck, Saskia; Moyer, Richard A.; Ferraro, Nathaniel M.; ...

2014-06-18

In this study, we introduce a new indirect method for identifying the radial extent of the stochastic layer due to applying resonant magnetic perturbations (RMPs) in H-mode plasmas by measuring the spin-up of the plasma near the separatrix. This spin-up is a predicted consequence of enhanced loss of electrons due to magnetic stochastization. We find that in DIII-D H-mode plasmas with n = 3 RMPs applied for edge localized mode (ELM) suppression, the stochastic layer is limited to the outer 5% region in normalized magnetic flux, Ψ N. This is in contrast to vacuum modeling predictions where this layer canmore » penetrate up to 20% in Ψ N. Theoretical predictions of a stochastic red radial electric field, E r component exceed the experimental measurements by about a factor 3 close to the separatrix, suggesting that the outer region of the plasma is weakly stochastic. Linear response calculations with M3D-C1, a resistive two-fluid model, show that in this outer 5% region, plasma response often reduces the resonant magnetic field components by 67% or more in comparison with vacuum calculations. These results for DIII-D are in reasonable agreement with results from the MAST tokamak, where the magnetic field perturbation from vacuum field calculations needed to be reduced by 75% for agreement with experimental measurements of the x-point lobe structures.« less

Study of SOL in DIII-D tokamak with SOLPS suite of codes.

NASA Astrophysics Data System (ADS)

Pankin, Alexei; Bateman, Glenn; Brennan, Dylan; Coster, David; Hogan, John; Kritz, Arnold; Kukushkin, Andrey; Schnack, Dalton; Snyder, Phil

2005-10-01

The scrape-of-layer (SOL) region in DIII-D tokamak is studied with the SOLPS integrated suite of codes. The SOLPS package includes the 3D multi-species Monte-Carlo neutral code EIRINE and 2D multi-fluid code B2. The EIRINE and B2 codes are cross-coupled through B2-EIRINE interface. The results of SOLPS simulations are used in the integrated modeling of the plasma edge in DIII-D tokamak with the ASTRA transport code. Parameterized dependences for neutral particle fluxes that are computed with the SOLPS code are implemented in a model for the H-mode pedestal and ELMs [1] in the ASTRA code. The effects of neutrals on the H-mode pedestal and ELMs are studied in this report. [1] A. Y. Pankin, I. Voitsekhovitch, G. Bateman, et al., Plasma Phys. Control. Fusion 47, 483 (2005).

Marshall N. Rosenbluth Outstanding Doctoral Thesis Award Talk: Simultaneous Measurement of Electron Temperature and Density Fluctuations in the Core of DIII-D Plasmas

NASA Astrophysics Data System (ADS)

White, A. E.

2009-11-01

Multi-field fluctuation measurements provide opportunities for rigorous comparison between experiment and nonlinear gyrokinetic turbulence simulations. A unique set of diagnostics on DIII-D allows for simultaneous study of local, long-wavelength (0 < kθρs< 0.5) electron temperature and density fluctuations in the core plasma (0.4 < ρ< 0.8). Previous experiments in L-mode indicate that normalized electron temperature fluctuation levels (40 < f < 400,kHz) increase with radius from ˜0.4% at ρ= 0.5 to ˜2% at ρ=0.8, similar to simultaneously measured density fluctuations. Electron cyclotron heating (ECH) is used to increase Te, which increases electron temperature fluctuation levels and electron heat transport in the experiments. In contrast, long wavelength density fluctuation levels change very little. The different responses are consistent with increased TEM drive relative to ITG-mode drive. A new capability at DIII-D is the measurement of phase angle between electron temperature and density fluctuations using coupled correlation electron cyclotron emission radiometer and reflectometer diagnostics. Linear and nonlinear GYRO runs have been used to design validation experiments that focus on measurements of the phase angle. GYRO shows that if Te and ∇Te increase 50% in a beam-heated L-mode plasma (ρ=0.5), then the phase angle between electron temperature and density fluctuations decreases 30%-50% and electron temperature fluctuation levels increase a factor of two more than density fluctuations. Comparisons between these predictions and experimental results will be presented.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

2016-12-01

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.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

DOE PAGES

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

2016-09-14

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, whichmore » 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). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

Interaction of external n  =  1 magnetic fields with the sawtooth instability in low-q RFX-mod and DIII-D tokamaks

NASA Astrophysics Data System (ADS)

Piron, C.; Martin, P.; Bonfiglio, D.; Hanson, J.; Logan, N. C.; Paz-Soldan, C.; Piovesan, P.; Turco, F.; Bialek, J.; Franz, P.; Jackson, G.; Lanctot, M. J.; Navratil, G. A.; Okabayashi, M.; Strait, E.; Terranova, D.; Turnbull, A.

2016-10-01

External n  =  1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m  =  1, n  =  1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigation in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q 95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n  =  1 plasma response, that indicates an enhancement of the coupling to the marginally stable n  =  1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.

AORSA full wave calculations of helicon waves in DIII-D and ITER

NASA Astrophysics Data System (ADS)

Lau, C.; Jaeger, E. F.; Bertelli, N.; Berry, L. A.; Green, D. L.; Murakami, M.; Park, J. M.; Pinsker, R. I.; Prater, R.

2018-06-01

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases. These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10%–20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.

AORSA full wave calculations of helicon waves in DIII-D and ITER

DOE PAGES

Lau, Cornwall; Jaeger, E.F.; Bertelli, Nicola; ...

2018-04-11

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases.more » These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10-20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.« less

Computational Study of Anomalous Transport in High Beta DIII-D Discharges with ITBs

NASA Astrophysics Data System (ADS)

Pankin, Alexei; Garofalo, Andrea; Grierson, Brian; Kritz, Arnold; Rafiq, Tariq

2015-11-01

The advanced tokamak scenarios require a large bootstrap current fraction and high β. These large values are often outside the range that occurs in ``conventional'' tokamak discharges. The GLF23, TGLF, and MMM transport models have been previously validated for discharges with parameters associated with ``conventional'' tokamak discharges. It has been demonstrated that the TGLF model under-predicts anomalous transport in high β DIII-D discharges [A.M. Garofalo et al. 2015 TTF Workshop]. In this research, the validity of MMM7.1 model [T. Rafiq et al. Phys. Plasmas 20 032506 (2013)] is tested for high β DIII-D discharges with low and high torque. In addition, the sensitivity of the anomalous transport to β is examined. It is shown that the MMM7.1 model over-predicts the anomalous transport in the DIII-D discharge 154406. In particular, a significant level of anomalous transport is found just outside the internal transport barrier. Differences in the anomalous transport predicted using TGLF and MMM7.1 are reviewed. Mechanisms for quenching of anomalous transport in the ITB regions of high-beta discharges are investigated. This research is supported by US Department of Energy.

ITER Baseline Scenario with ECCD Applied to Neoclassical Tearing Modes in DIII-D

NASA Astrophysics Data System (ADS)

Welander, A. G.; La Haye, R. J.; Lohr, J. M.; Humphreys, D. A.; Prater, R.; Paz-Soldan, C.; Kolemen, E.; Turco, F.; Olofsson, E.

2015-11-01

The neoclassical tearing mode (NTM) is a magnetic island that can occur on flux surfaces where the safety factor q is a rational number. Both m/n=3/2 and 2/1 NTM's degrade confinement, and the 2/1 mode often locks to the wall and disrupts the plasma. An NTM can be suppressed by depositing electron cyclotron current drive (ECCD) on the q-surface by injecting microwave beams into the plasma from gyrotrons. Recent DIII-D experiments have studied the application of ECCD/ECRH in the ITER Baseline Scenario. The power required from the gyrotrons can be significant enough to impact the fusion gain, Q in ITER. However, if gyrotron power could be minimized or turned off in ITER when not needed, this impact would be small. In fact, tearing-stable operation at low torque has been achieved previously in DIII-D without EC power. A vision for NTM control in ITER will be described together with results obtained from simulations and experiments in DIII-D under ITER like conditions. Work supported by the US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-FG02-04ER54761.

Studies of the DIII-D disruption database using Machine Learning algorithms

NASA Astrophysics Data System (ADS)

Rea, Cristina; Granetz, Robert; Meneghini, Orso

2017-10-01

A Random Forests Machine Learning algorithm, trained on a large database of both disruptive and non-disruptive DIII-D discharges, predicts disruptive behavior in DIII-D with about 90% of accuracy. Several algorithms have been tested and Random Forests was found superior in performances for this particular task. Over 40 plasma parameters are included in the database, with data for each of the parameters taken from 500k time slices. We focused on a subset of non-dimensional plasma parameters, deemed to be good predictors based on physics considerations. Both binary (disruptive/non-disruptive) and multi-label (label based on the elapsed time before disruption) classification problems are investigated. The Random Forests algorithm provides insight on the available dataset by ranking the relative importance of the input features. It is found that q95 and Greenwald density fraction (n/nG) are the most relevant parameters for discriminating between DIII-D disruptive and non-disruptive discharges. A comparison with the Gradient Boosted Trees algorithm is shown and the first results coming from the application of regression algorithms are presented. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-SC0014264 and DE-FG02-95ER54309.

AORSA full wave calculations of helicon waves in DIII-D and ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lau, Cornwall; Jaeger, E.F.; Bertelli, Nicola

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases.more » These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10-20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.« less

Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Suppression of type-I ELMs with reduced RMP coil set on DIII-D

DOE PAGES

Orlov, Dmitriy M.; Moyer, Richard A.; Evans, Todd E.; ...

2016-02-19

Recent experiments on DIII-D have demonstrated that having a toroidally-monochromatic spectral content of edge-resonant magnetic perturbations (RMPs) is not a necessary condition for suppression of Edge Localized Modes (ELMs). Robust ELM suppression has been reproducibly obtained on DIII-D during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly producing a variety of n=1, n=2, and n=3 spectral contributions. It was shown that RMP ELM suppression could be achieved with as few as 5 out of 12 internal coil loops (I-coils) on DIII-D at similar coil currents and with good plasma confinement. Linear MHD plasma response (M3DC1, IPEC, MARS)more » and vacuum (SURFMN, TRIP3D) modeling have been performed in order to understand the effects of the perturbation spectrum on the plasma response and ELM suppression. The results suggest that reduction of the dominant n=3 perturbation field is compensated by increased n=2 field in the plasma that may lead to RMP ELM suppression at lower levels of n=3 perturbative magnetic flux from the I-coils. These results provide additional confidence that ITER may be capable of RMP ELM suppression in the event of multiple internal coil failures.« less

A frequency tunable, eight-channel correlation ECE system for electron temperature turbulence measurements on the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sung, C., E-mail: csung@physics.ucla.edu; Peebles, W. A.; Wannberg, C.

2016-11-15

A new eight-channel correlation electron cyclotron emission diagnostic has recently been installed on the DIII-D tokamak to study both turbulent and coherent electron temperature fluctuations under various plasma conditions and locations. This unique system is designed to cover a broad range of operation space on DIII-D (1.6-2.1 T, detection frequency: 72-108 GHz) via four remotely selected local oscillators (80, 88, 96, and 104 GHz). Eight radial locations are measured simultaneously in a single discharge covering as much as half the minor radius. In this paper, we present design details of the quasi-optical system, the receiver, as well as representative datamore » illustrating operation of the system.« less

A study for the installation of the TEXT heavy-ion beam probe on DIII-D

NASA Astrophysics Data System (ADS)

Edmonds, P. H.; Solano, E. R.; Bravenec, R. V.; Wootton, A. J.; Schoch, P. M.; Crowley, T. P.; Hickok, R. L.; West, W. P.; Leuer, J.; Anderson, P.

1997-01-01

An assessment of the feasibility of installing the TEXT 2 MeV heavy-ion beam probe on the DIII-D tokamak has been completed. Detailed drawings of the machine cross section were imported into the CAD application AutoCAD. A set of programs written in AutoLisp were used to generate trajectories. Displays of the accessible cross section of the plasma, scan lines for the entire range of primary beam energy and injection angle ranges, and sample-volume dimensions can be rapidly generated. Because of the large deflection between the primary input beam and the emergent secondary beam, either the analyzer needs to be tracked over a ±20° angle or secondary poloidal deflector plates need to be installed at the exit port. Toroidal deflector plates will be installed at both the injection and exit ports to compensate for toroidal displacements and deflections. The sample volumes generated by this procedure are within a few centimeters of the locations derived from a full three-dimensional calculation.

Extension of high poloidal beta scenario in DIII-D to lower q95 for steady state fusion reactor

NASA Astrophysics Data System (ADS)

Huang, J.; Gong, X.; Qian, J.; Ding, S.; Ren, Q.; Guo, W.; Pan, C.; Li, G.; Xia, T.; Garofalo, A.; Lao, L.; Hyatt, A.; Ferron, J.; Collins, C.; Lin, D.; McKee, G.; Rhode, T.; McClenaghan, J.; Holcomb, C.; Cui, L.; Heidbrink, W.; Zhu, Y.; Diiid Team; East Team

2017-10-01

DIII-D/EAST joint experiments have improved the high poloidal beta scenario with sustained large-radius internal transport barrier (ITB) extended to high plasma current Ip 1MA with q95 6.0. Slight off-axis NBCD is applied to obtain broader current density profile, ITBs can now be sustained below the previously observed βp threshold with excellent confinement (H98y2 1.8). The scenario also exhibits a local negative shear appearing with q increased at rho 0.4, which helps ITB formation and sustainment. This confirms TGLF prediction that negative magnetic shear can help recover ITB and achieve high confinement with reduced q95. Detailed analysis shows that the Shafranov shift and q profile is critical in the ITB formation at high βp regime. Supported in part by National Magnetic Confinement Fusion Program of China 2015GB102000, 2015GB110005, and US Department of Energy under DE-FC02-04ER54698.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scime, Earl E.

The magnitude and spatial dependence of neutral density in magnetic confinement fusion experiments is a key physical parameter, particularly in the plasma edge. Modeling codes require precise measurements of the neutral density to calculate charge-exchange power losses and drag forces on rotating plasmas. However, direct measurements of the neutral density are problematic. In this work, we proposed to construct a laser-based diagnostic capable of providing spatially resolved measurements of the neutral density in the edge of plasma in the DIII-D tokamak. The diagnostic concept is based on two-photon absorption laser induced fluorescence (TALIF). By injecting two beams of 205 nmmore » light (co or counter propagating), ground state hydrogen (or deuterium or tritium) can be excited from the n = 1 level to the n = 3 level at the location where the two beams intersect. Individually, the beams experience no absorption, and therefore have no difficulty penetrating even dense plasmas. After excitation, a fraction of the hydrogen atoms decay from the n = 3 level to the n = 2 level and emit photons at 656 nm (the H α line). Calculations based on the results of previous TALIF experiments in magnetic fusion devices indicated that a laser pulse energy of approximately 3 mJ delivered in 5 ns would provide sufficient signal-to-noise for detection of the fluorescence. In collaboration with the DIII-D engineering staff and experts in plasma edge diagnostics for DIII-D from Oak Ridge National Laboratory (ORNL), WVU researchers designed a TALIF system capable of providing spatially resolved measurements of neutral deuterium densities in the DIII-D edge plasma. The laser systems were specified, purchased, and assembled at WVU. The TALIF system was tested on a low-power hydrogen discharge at WVU and the plan was to move the instrument to DIII-D for installation in collaboration with ORNL researchers. After budget cuts at DIII-D, the DIII-D facility declined to support installation on their tokamak. Instead, after a no-cost extension, the apparatus was moved to the University of Washington-Seattle and successfully tested on the HIT-SI3 spheromak experiment. As a result of this project, TALIF measurements of the absolutely calibrated neutral density hydrogen and deuterium were obtained in a helicon source and in a spheromak, designs were developed for installation of a TALIF system on a tokamak, and a new, xenon-based calibration scheme was proposed and demonstrated. The xenon-calibration scheme eliminates significant problems that were identified with the standard krypton calibration scheme.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.

The redistribution and potential loss of energetic particles due to MHD modes can limit the performance of fusion plasmas by reducing the plasma heating rate. In this work, we present validation studies of the 1.5D critical gradient model (CGM) for Alfvén eigenmode (AE) induced EP transport in NSTX and DIII-D neutral beam heated plasmas. In previous comparisons with a single DIII-D L-mode case, the CGM model was found to be responsible for 75% of measured AE induced neutron deficit [1]. A fully kinetic HINST is used to compute mode stability for the non-perturbative version of CGM (or nCGM). We have found that AEs show strong local instability drive up tomore » $$\\gamma /\\omega \\sim 20\\%$$ violating assumptions of perturbative approaches used in NOVA-K code. Lastly, we demonstrate that both models agree with each other and both underestimate the neutron deficit measured in DIII-D shot by approximately a factor of 2.« less

Validation of the model for ELM suppression with 3D magnetic fields using low torque ITER baseline scenario discharges in DIII-D

DOE PAGES

Moyer, Richard A.; Paz-Soldan, Carlos; Nazikian, Raffi; ...

2017-09-18

Here, experiments have been executed in the DIII-D tokamak to extend suppression of Edge Localized Modes (ELMs) with Resonant Magnetic Perturbations (RMPs) to ITER-relevant levels of beam torque. The results support the hypothesis for RMP ELM suppression based on transition from an ideal screened response to a tearing response at a resonant surface that prevents expansion of the pedestal to an unstable width.

78 FR 15995 - Self-Regulatory Organizations; the NASDAQ Stock Market LLC; Notice of Filing of Proposed Rule...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-13

...-Regulatory Organizations; the NASDAQ Stock Market LLC; Notice of Filing of Proposed Rule Change To Adopt... Proposed Rule Change The Exchange proposes to adopt a new Chapter V, Section 3(d)(iii) to provide for how... 1. Purpose The Exchange proposes to adopt Chapter V, Section 3(d)(iii) \\3\\ to provide for how the...

Deuterium velocity and temperature measurements on the DIII-D tokamak.

PubMed

Grierson, B A; Burrell, K H; Solomon, W M; Pablant, N A

2010-10-01

Newly installed diagnostic capabilities on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 46, 6114 (2002)] enable the measurement of main ion (deuterium) velocity and temperature by charge exchange recombination spectroscopy. The uncertainty in atomic physics corrections for determining the velocity is overcome by exploiting the geometrical dependence of the apparent velocity on the viewing angle with respect to the neutral beam.

Development of a New System for Transport Simulation and Analysis at General Atomics

NASA Astrophysics Data System (ADS)

St. John, H. E.; Peng, Q.; Freeman, J.; Crotinger, J.

1997-11-01

General Atomics has begun a long term program to improve all aspects of experimental data analysis related to DIII--D. The object is to make local and visiting physicists as productive as possible, with only a small investment in training, by developing intuitive, sophisticated interfaces to existing and newly created computer programs. Here we describe our initial work and results of a pilot project in this program. The pilot project is a collaboratory effort between LLNL and GA which will ultimately result in the merger of Corsica and ONETWO (and selected modules from other codes) into a new advanced transport code system. The initial goal is to produce a graphical user interface to the transport code ONETWO which will couple to a programmable (steerable) front end designed for the transport system. This will be an object oriented scheme written primarily in python. The programmable application will integrate existing C, C^++, and Fortran methods in a single computational paradigm. Its most important feature is the use of plug in physics modules which will allow a high degree of customization.

Fusion programs in applied plasma physics

NASA Astrophysics Data System (ADS)

1992-07-01

The Applied Plasma Physics (APP) program at General Atomics (GA) described here includes four major elements: (1) Applied Plasma Physics Theory Program, (2) Alpha Particle Diagnostic, (3) Edge and Current Density Diagnostic, and (4) Fusion User Service Center (USC). The objective of the APP theoretical plasma physics research at GA is to support the DIII-D and other tokamak experiments and to significantly advance our ability to design a commercially-attractive fusion reactor. We categorize our efforts in three areas: magnetohydrodynamic (MHD) equilibria and stability; plasma transport with emphasis on H-mode, divertor, and boundary physics; and radio frequency (RF). The objective of the APP alpha particle diagnostic is to develop diagnostics of fast confined alpha particles using the interactions with the ablation cloud surrounding injected pellets and to develop diagnostic systems for reacting and ignited plasmas. The objective of the APP edge and current density diagnostic is to first develop a lithium beam diagnostic system for edge fluctuation studies on the Texas Experimental Tokamak (TEXT). The objective of the Fusion USC is to continue to provide maintenance and programming support to computer users in the GA fusion community. The detailed progress of each separate program covered in this report period is described.

The Multiple Gyrotron System on the DIII-D Tokamak

DOE PAGES

Lohr, J.; Anderson, J.; Brambila, R.; ...

2015-08-28

A major component of the versatile heating systems on the DIII-D tokamak is the gyrotron complex. This system routinely operates at 110 GHz with 4.7 MW generated rf power for electron cyclotron heating and current drive. The complex is being upgraded with the addition of new depressed collector potential gyrotrons operating at 117.5 GHz and generating rf power in excess of 1.0 MW each. The long term upgrade plan calls for 10 gyrotrons at the higher frequency being phased in as resources permit, for an injected power near 10 MW. This article presents a summary of the current status ofmore » the DIII-D gyrotron complex, its performance, individual components, testing procedures, operational parameters, plans, and a brief summary of the experiments for which the system is currently being used.« less

Divertor sheath power studies in DIII-D using fixed Langmuir probes and three-dimensional modeling of tile heat flows

NASA Astrophysics Data System (ADS)

Donovan, D.; Nygren, R.; Buchenauer, D.; Watkins, J.; Rudakov, D.; Leonard, A.; Wong, C. P. C.; Makowski, M.

2014-04-01

Experimental results are presented from the three-Langmuir probe (LP) diagnostic head of the divertor material evaluation system (DiMES) on DIII-D that confirm the size of the projected current collection area of the LPs, which is essential for properly measuring ion saturation current density (Jsat) and the sheath power transmission factor (SPTF). Also using the 3-LP DiMES head, the hypothesis that collisional effects on plasma density occurring in the magnetic sheath of the tile are responsible for a lower than expected SPTF is tested and deemed not to have a significant impact on the SPTF. Three-dimensional thermal modeling of wall tiles is presented that accounts for lateral heat conduction, temperature dependence of tile material properties and radiative heat loss from the tile surface. This modeling was developed to be used in the analysis of temperature profiles of the divertor embedded thermocouple (TC) array to obtain more accurate interpretations of TC temperature profiles to infer divertor surface heat flux than have previously been accomplished using more basic one-dimensional methods.

ELM Behavior in High- βp EAST-Demonstration Plasmas on DIII-D

NASA Astrophysics Data System (ADS)

Li, G. Q.; Gong, X. Z.; Garofalo, A. M.; Lao, L. L.; Meneghini, O.; Snyder, P. B.; Ren, Q. L.; Ding, S. Y.; Guo, W. F.; Qian, J. P.; Wan, B. N.; Xu, G. S.; Holcomb, C. T.; Solomon, W. M.

2015-11-01

In the DIII-D high- βp EAST-demonstration experiment, for several similar discharges when the experimental parameters such as the toroidal magnetic field or ECH power are varied slightly, the changes in ELM frequency response are observed to be much larger. Kinetic EFIT equilibrium reconstructions for these discharges have been performed, which suggest that the ELM frequency changes are likely due to the variations of pedestal width, height, and edge current density. Kinetic profile analyses further indicate that the strong ITB that are located at large minor radii (rho=0.6 ~0.7) in these discharges are affecting the pedestal structure. The ITB could broaden the pedestal width and decrease the pedestal height, thus changing the ELM frequency and size. With the GATO and ELITE MHD codes, the linear growth rates and mode structures of these ELMs are analyzed. The impact of ITB on the ELMs behavior will be discussed. Work supported by China MOST under 2014GB106001 and 2015GB102001 and US DOE under DE-FC02-04ER54698 and DE-FG03-95ER54309.

Detailed study of spontaneous rotation generation in diverted H-mode plasma using the full-f gyrokinetic code XGC1

NASA Astrophysics Data System (ADS)

Seo, Janghoon; Chang, C. S.; Ku, S.; Kwon, J. M.; Yoon, E. S.

2013-10-01

The Full-f gyrokinetic code XGC1 is used to study the details of toroidal momentum generation in H-mode plasma. Diverted DIII-D geometry is used, with Monte Carlo neutral particles that are recycled at the limiter wall. Nonlinear Coulomb collisions conserve particle, momentum, and energy. Gyrokinetic ions and adiabatic electrons are used in the present simulation to include the effects from ion gyrokinetic turbulence and neoclassical physics, under self-consistent radial electric field generation. Ion orbit loss physics is automatically included. Simulations show a strong co-Ip flow in the H-mode layer at outside midplane, similarly to the experimental observation from DIII-D and ASDEX-U. The co-Ip flow in the edge propagates inward into core. It is found that the strong co-Ip flow generation is mostly from neoclassical physics. On the other hand, the inward momentum transport is from turbulence physics, consistently with the theory of residual stress from symmetry breaking. Therefore, interaction between the neoclassical and turbulence physics is a key factor in the spontaneous momentum generation.

Dependency of Tearing Mode Stability on Current and Pressure Profiles in DIII-D Hybrid Discharges

NASA Astrophysics Data System (ADS)

Kim, K.; Park, J. M.; Murakami, M.; La Haye, R. J.; Na, Y.-S.; SNU/ORAU; ORNL; Atomics, General; SNU; DIII-D Team

2016-10-01

Understanding the physics of the onset and evolution of tearing modes (TMs) in tokamak plasmas is important for high- β steady-state operation. Based on DIII-D steady-state hybrid experiments with accurate equilibrium reconstruction and well-measured plasma profiles, the 2/1 tearing mode can be more stable with increasing local current and pressure gradient at rational surface and with lower pressure peaking and plasma inductance. The tearing stability index Δ', estimated by the Rutherford equation with experimental mode growth rate was validated against Δ' calculated by linear eigenvalue solver (PEST3); preliminary comprehensive MHD modeling by NIMROD reproduced the TM onset reasonably well. We present a novel integrated modeling for the purpose of predicting TM onset in experiment by combining a model equilibrium reconstruction using IPS/FASTRAN, linear stability Δ' calculation using PEST3, and fitting formula for critical Δ' from NIMROD. Work supported in part by the US DoE under DE-AC05-06OR23100, DE-AC05-00OR22725, and DEFC02-04ER54698.

Study of High Temperature Failure Mechanisms in Ceramics

DTIC Science & Technology

1988-06-01

The major experimental 4 techniques employed in the program are the use of small- angle neutron scattering to characterize cavity nucleation and growth...creep crackgrowth. Of particular interest are the development of a stochastic model of grainboundary sliding and a micromechanical model that relates...Accession For NTIS GF.A&I DTIC T,’ IDi st ribut Ion’ ;i Avillii~diii l l= (~~ I. RESEARCH OBJECTIVES I. Utilize small- angle neutron scattering to

Rotational shear effects on edge harmonic oscillations in DIII-D quiescent H-mode discharges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Xi; Burrell, Keith H.; Ferraro, Nathaniel M.

In the quiescent H-mode (QH-mode) regime, edge harmonic oscillations (EHO) play an important role in avoiding transient edge localized mode (ELM) power fluxes by providing benign and continuous edge particle transport. A detailed theoretical, experimental and modeling comparison has been made of low-n (n ≤ 5) EHO in DIII-D QH-mode plasmas. The calculated linear eigenmode structure from the extended MHD code M3D-C1 matches closely the coherent EHO properties from external magnetics data and internal measurements using the ECE, BES, ECE-Imaging and microwave imaging reflectometer (MIR) diagnostics, as well as the kink/peeling mode properties found by the ideal MHD code ELITE.more » Numerical investigations indicate that the low-n EHO-like solutions from M3D-C1 are destabilized by the rotational shear while high-n modes are stabilized. This effect is independent of the rotation direction, suggesting that EHO can be destabilized in principle with rotation in either direction. Furthermore, the modeling results are consistent with observations of the EHO, support the proposed theory of the EHO as a rotational shear driven kink/peeling mode, and improve our understanding and confidence in creating and sustaining QH-mode in present and future devices.« less

Rotational shear effects on edge harmonic oscillations in DIII-D quiescent H-mode discharges

DOE PAGES

Chen, Xi; Burrell, Keith H.; Ferraro, Nathaniel M.; ...

2016-06-21

In the quiescent H-mode (QH-mode) regime, edge harmonic oscillations (EHO) play an important role in avoiding transient edge localized mode (ELM) power fluxes by providing benign and continuous edge particle transport. A detailed theoretical, experimental and modeling comparison has been made of low-n (n ≤ 5) EHO in DIII-D QH-mode plasmas. The calculated linear eigenmode structure from the extended MHD code M3D-C1 matches closely the coherent EHO properties from external magnetics data and internal measurements using the ECE, BES, ECE-Imaging and microwave imaging reflectometer (MIR) diagnostics, as well as the kink/peeling mode properties found by the ideal MHD code ELITE.more » Numerical investigations indicate that the low-n EHO-like solutions from M3D-C1 are destabilized by the rotational shear while high-n modes are stabilized. This effect is independent of the rotation direction, suggesting that EHO can be destabilized in principle with rotation in either direction. Furthermore, the modeling results are consistent with observations of the EHO, support the proposed theory of the EHO as a rotational shear driven kink/peeling mode, and improve our understanding and confidence in creating and sustaining QH-mode in present and future devices.« less

Rotational shear effects on edge harmonic oscillations in DIII-D quiescent H-mode discharges

NASA Astrophysics Data System (ADS)

Chen, Xi; Burrell, K. H.; Ferraro, N. M.; Osborne, T. H.; Austin, M. E.; Garofalo, A. M.; Groebner, R. J.; Kramer, G. J.; Luhmann, N. C., Jr.; McKee, G. R.; Muscatello, C. M.; Nazikian, R.; Ren, X.; Snyder, P. B.; Solomon, W. M.; Tobias, B. J.; Yan, Z.

2016-07-01

In the quiescent H-mode (QH-mode) regime, edge harmonic oscillations (EHOs) play an important role in avoiding transient edge localized mode (ELM) power fluxes by providing benign and continuous edge particle transport. A detailed theoretical, experimental and modeling comparison has been made of low-n (n  ⩽  5) EHO in DIII-D QH-mode plasmas. The calculated linear eigenmode structure from the extended magentoohydrodynamics (MHD) code M3D-C1 matches closely the coherent EHO properties from external magnetics data and internal measurements using the ECE, BES, ECE-Imaging and microwave imaging reflectometer (MIR) diagnostics, as well as the kink/peeling mode properties found by the ideal MHD code ELITE. Numerical investigations indicate that the low-n EHO-like solutions from M3D-C1 are destabilized by rotation and/or rotational shear while high-n modes are stabilized. This effect is independent of the rotation direction, suggesting that EHOs can be destabilized in principle with rotation in either direction. The modeling results are consistent with observations of EHO, support the proposed theory of the EHO as a low-n kink/peeling mode destabilized by edge E  ×  B rotational shear, and improve our understanding and confidence in creating and sustaining QH-mode in present and future devices.

Use of reconstructed 3D equilibria to determine onset conditions of helical cores in tokamaks for extrapolation to ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wingen, A.; Wilcox, R. S.; Seal, S. K.

In this paper, large, spontaneous m/n = 1/1 helical cores are shown to be expected in tokamaks such as ITER with extended regions of low- or reversed- magnetic shear profiles and q near 1 in the core. The threshold for this spontaneous symmetry breaking is determined using VMEC scans, beginning with reconstructed 3D equilibria from DIII-D and Alcator C-Mod based on observed internal 3D deformations. The helical core is a saturated internal kink mode (Wesson 1986 Plasma Phys. Control. Fusion 28 243); its onset threshold is shown to be proportional tomore » $$({\\rm d}p/{\\rm d}\\rho)/B_t^2$$ around q = 1. Below the threshold, applied 3D fields can drive a helical core to finite size, as in DIII-D. The helical core size thereby depends on the magnitude of the applied perturbation. Above it, a small, random 3D kick causes a bifurcation from axisymmetry and excites a spontaneous helical core, which is independent of the kick size. Systematic scans of the q-profile show that the onset threshold is very sensitive to the q-shear in the core. Helical cores occur frequently in Alcator C-Mod during ramp-up when slow current penetration results in a reversed shear q-profile, which is favorable for helical core formation. In conclusion, a comparison of the helical core onset threshold for discharges from DIII-D, Alcator C-Mod and ITER confirms that while DIII-D is marginally stable, Alcator C-Mod and especially ITER are highly susceptible to helical core formation without being driven by an externally applied 3D magnetic field.« less

Use of reconstructed 3D equilibria to determine onset conditions of helical cores in tokamaks for extrapolation to ITER

DOE PAGES

Wingen, A.; Wilcox, R. S.; Seal, S. K.; ...

2018-01-15

In this paper, large, spontaneous m/n = 1/1 helical cores are shown to be expected in tokamaks such as ITER with extended regions of low- or reversed- magnetic shear profiles and q near 1 in the core. The threshold for this spontaneous symmetry breaking is determined using VMEC scans, beginning with reconstructed 3D equilibria from DIII-D and Alcator C-Mod based on observed internal 3D deformations. The helical core is a saturated internal kink mode (Wesson 1986 Plasma Phys. Control. Fusion 28 243); its onset threshold is shown to be proportional tomore » $$({\\rm d}p/{\\rm d}\\rho)/B_t^2$$ around q = 1. Below the threshold, applied 3D fields can drive a helical core to finite size, as in DIII-D. The helical core size thereby depends on the magnitude of the applied perturbation. Above it, a small, random 3D kick causes a bifurcation from axisymmetry and excites a spontaneous helical core, which is independent of the kick size. Systematic scans of the q-profile show that the onset threshold is very sensitive to the q-shear in the core. Helical cores occur frequently in Alcator C-Mod during ramp-up when slow current penetration results in a reversed shear q-profile, which is favorable for helical core formation. In conclusion, a comparison of the helical core onset threshold for discharges from DIII-D, Alcator C-Mod and ITER confirms that while DIII-D is marginally stable, Alcator C-Mod and especially ITER are highly susceptible to helical core formation without being driven by an externally applied 3D magnetic field.« less

Perturbative transport modeling and comparison to cold-pulse and heat-pulse propagation experiments in Alcator C-Mod and DIII-D

NASA Astrophysics Data System (ADS)

Rodriguez Fernandez, P.; White, A. E.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Howard, N. T.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Petty, C. C.; Rice, J. E.; Alcator C-Mod Team

2016-10-01

Possible ``non-local'' transport phenomena are often observed in tokamak plasmas. Different models have been proposed to explain fast responses during perturbative transport experiments, including non-diffusive effects. Specific tools to characterize the dynamic behavior and power balance analysis using TRANSP and the quasi-linear trapped gyro-landau fluid code TGLF have been developed to analyze Alcator C-Mod experiments. Recent results from cold pulse experiments show that fast core temperature increases following edge cold-pulse injections (peak within 10ms , while τE 25ms) are not correlated with the direction of intrinsic rotation, and instead the amplitude of the core response depends on density, plasma current and RF input power. The propagation of the cold pulse can be compared with propagation of heat pulses from sawteeth, and both may be used to probe changes in temperature profile stiffness. A Laser Blow Off (LBO) system is being developed for DIII-D that will allow further validation and cross-machine comparison of cold pulse experiments. LBO at DIII-D will also allow for direct comparisons with ECH perturbative heat pulse experiments. Work supported by US DOE under Grants DE-FC02-99ER54512 (C-Mod) and DE-FC02-04ER54698 (DIII-D) and La Caixa Fellowship.

Numerical Solution of the Electron Heat Transport Equation and Physics-Constrained Modeling of the Thermal Conductivity via Sequential Quadratic Programming Optimization in Nuclear Fusion Plasmas

NASA Astrophysics Data System (ADS)

Paloma, Cynthia S.

The plasma electron temperature (Te) plays a critical role in a tokamak nu- clear fusion reactor since temperatures on the order of 108K are required to achieve fusion conditions. Many plasma properties in a tokamak nuclear fusion reactor are modeled by partial differential equations (PDE's) because they depend not only on time but also on space. In particular, the dynamics of the electron temperature is governed by a PDE referred to as the Electron Heat Transport Equation (EHTE). In this work, a numerical method is developed to solve the EHTE based on a custom finite-difference technique. The solution of the EHTE is compared to temperature profiles obtained by using TRANSP, a sophisticated plasma transport code, for specific discharges from the DIII-D tokamak, located at the DIII-D National Fusion Facility in San Diego, CA. The thermal conductivity (also called thermal diffusivity) of the electrons (Xe) is a plasma parameter that plays a critical role in the EHTE since it indicates how the electron temperature diffusion varies across the minor effective radius of the tokamak. TRANSP approximates Xe through a curve-fitting technique to match experimentally measured electron temperature profiles. While complex physics-based model have been proposed for Xe, there is a lack of a simple mathematical model for the thermal diffusivity that could be used for control design. In this work, a model for Xe is proposed based on a scaling law involving key plasma variables such as the electron temperature (Te), the electron density (ne), and the safety factor (q). An optimization algorithm is developed based on the Sequential Quadratic Programming (SQP) technique to optimize the scaling factors appearing in the proposed model so that the predicted electron temperature and magnetic flux profiles match predefined target profiles in the best possible way. A simulation study summarizing the outcomes of the optimization procedure is presented to illustrate the potential of the proposed modeling method.

Effects of resistivity and rotation on the linear plasma response to non-axisymmetric magnetic perturbations on DIII-D

DOE PAGES

Haskey, Shaun R.; Lanctot, Matthew J.; Liu, Y. Q.; ...

2015-01-05

Parameter scans show the strong dependence of the plasma response on the poloidal structure of the applied field highlighting the importance of being able to control this parameter using non-axisymmetric coil sets. An extensive examination of the linear single fluid plasma response to n = 3 magnetic perturbations in L-mode DIII-D lower single null plasmas is presented. The effects of plasma resistivity, toroidal rotation and applied field structure are calculated using the linear single fluid MHD code, MARS-F. Measures which separate the response into a pitch-resonant and resonant field amplification (RFA) component are used to demonstrate the extent to whichmore » resonant screening and RFA occurs. The ability to control the ratio of pitch-resonant fields to RFA by varying the phasing between upper and lower resonant magnetic perturbations coils sets is shown. The predicted magnetic probe outputs and displacement at the x-point are also calculated for comparison with experiments. Additionally, modelling of the linear plasma response using experimental toroidal rotation profiles and Spitzer like resistivity profiles are compared with results which provide experimental evidence of a direct link between the decay of the resonant screening response and the formation of a 3D boundary. As a result, good agreement is found during the initial application of the MP, however, later in the shot a sudden drop in the poloidal magnetic probe output occurs which is not captured in the linear single fluid modelling.« less

Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-11-16

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment ofmore » the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (see standard divertor).« less

A High Power Helicon Antenna Design for DIII-D

DOE PAGES

Nagy, A.; deGrassie, J.; Moeller, C.; ...

2017-08-02

A new antenna design for driving current in high beta tokamaks using electromagnetic waves, called Helicons, will be experimentally tested for the first time at power approaching 1 megawatt (MW) in the DIII-D Tokamak. This method is expected to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and may be well suited to reactor-like configurations. A low power (100 watt (W)) 476 megahertz (MHz) “comb-line” antenna, consisting of 12 inductively coupled electrostatically shielded, modular resonators, was tested in DIII-D and showed strong coupling to the plasma without disturbing its characteristics or introducing metal impurities.more » The high power antenna consists of 30 modules affixed to back-plates and mounted on the outer wall of the vacuum vessel above the mid-plane. The antenna design follows a similar low power antenna design modified to minimize RF loss. Heat removal is provided by water cooling and a novel heat conducting path using pyrolytic graphite sheet. The CuCrZr antenna modules are designed to handle high eddy current forces. The modules use molybdenum Faraday shields that have the plasma side coated with boron carbide to enhance thermal resistance and minimize high Z impurities. A RF strip-line feed routes the RF power from coaxial vacuum feed-throughs to the antenna. Multipactor analysis of the antenna, strip line, and feedthrough will be performed. A 1.2 MW, 476 MHz klystron system, provided by the Stanford Linear Accelerator (SLAC) will provide RF power to the new antenna. Lastly, a description of the design of the high power antenna, the RF strip-line feeds, and the vessel installation will be presented.« less

A High Power Helicon Antenna Design for DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nagy, A.; deGrassie, J.; Moeller, C.

A new antenna design for driving current in high beta tokamaks using electromagnetic waves, called Helicons, will be experimentally tested for the first time at power approaching 1 megawatt (MW) in the DIII-D Tokamak. This method is expected to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and may be well suited to reactor-like configurations. A low power (100 watt (W)) 476 megahertz (MHz) “comb-line” antenna, consisting of 12 inductively coupled electrostatically shielded, modular resonators, was tested in DIII-D and showed strong coupling to the plasma without disturbing its characteristics or introducing metal impurities.more » The high power antenna consists of 30 modules affixed to back-plates and mounted on the outer wall of the vacuum vessel above the mid-plane. The antenna design follows a similar low power antenna design modified to minimize RF loss. Heat removal is provided by water cooling and a novel heat conducting path using pyrolytic graphite sheet. The CuCrZr antenna modules are designed to handle high eddy current forces. The modules use molybdenum Faraday shields that have the plasma side coated with boron carbide to enhance thermal resistance and minimize high Z impurities. A RF strip-line feed routes the RF power from coaxial vacuum feed-throughs to the antenna. Multipactor analysis of the antenna, strip line, and feedthrough will be performed. A 1.2 MW, 476 MHz klystron system, provided by the Stanford Linear Accelerator (SLAC) will provide RF power to the new antenna. Lastly, a description of the design of the high power antenna, the RF strip-line feeds, and the vessel installation will be presented.« less

Kinetic simulations of scrape-off layer physics in the DIII-D tokamak

DOE PAGES

Churchill, Randy M.; Canik, John M.; Chang, C. S.; ...

2016-12-27

Simulations using the fully kinetic code XGCa were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total- f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Fluid simulations are normally used to simulate the SOL, due to its high collisionality. However, depending on plasma conditions, a number of discrepancies have been observed between experiment and leading SOL fluid codes (e.g. SOLPS), including underestimating outer target temperatures, radial electric field in the SOL, parallel ion SOL flowsmore » at the low field side, and impurity radiation. Many of these discrepancies may be linked to the fluid treatment, and might be resolved by including kinetic effects in SOL simulations. The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels ( Mi ~0.5), with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/T e ~ 3–4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. Finally, these two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.« less

Extracting 3D Information from 1D and 2D Diagnostic Systems on the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Brookman, Michael

2017-10-01

The interpretation of tokamak data often hinges on assumptions of axisymetry and flux surface equilibria, neglecting 3D effects. This work discusses examples on the DIII-D tokamak where this assumption is an insufficient approximation, and explores the diagnostic information available to resolve 3D effects while preserving 1D profiles. Methods for extracting 3D data from the electron cyclotron emission radiometers, density profile reflectometer, and Thomson scattering system are discussed. Coordinating diagnostics around the tokamak shows the significance of 3D features, such as sawteeth[1] and resonant magnetic perturbations. A consequence of imposed 3D perturbations is a shift in major radius of measured profiles between diagnostics at different toroidal locations. Integrating different diagnostics requires a database containing information about their toroidal, poloidal, and radial locations. Through the data analysis framework OMFIT, it is possible to measure the magnitude of the apparent shifts from 3D effects and enforce consistency between diagnostics. Using the existing 1D and 2D diagnostic systems on DIII-D, this process allows the effects of the 3D perturbations on 1D profiles to be addressed. Supported by US DOE contracts DE-FC02-04ER54698, DE-FG03-97ER54415.

Use of reconstructed 3D equilibria to determine onset conditions of helical cores in tokamaks for extrapolation to ITER

NASA Astrophysics Data System (ADS)

Wingen, A.; Wilcox, R. S.; Seal, S. K.; Unterberg, E. A.; Cianciosa, M. R.; Delgado-Aparicio, L. F.; Hirshman, S. P.; Lao, L. L.

2018-03-01

Large, spontaneous m/n  =  1/1 helical cores are shown to be expected in tokamaks such as ITER with extended regions of low- or reversed- magnetic shear profiles and q near 1 in the core. The threshold for this spontaneous symmetry breaking is determined using VMEC scans, beginning with reconstructed 3D equilibria from DIII-D and Alcator C-Mod based on observed internal 3D deformations. The helical core is a saturated internal kink mode (Wesson 1986 Plasma Phys. Control. Fusion 28 243); its onset threshold is shown to be proportional to (dp/dρ)/B_t2 around q  =  1. Below the threshold, applied 3D fields can drive a helical core to finite size, as in DIII-D. The helical core size thereby depends on the magnitude of the applied perturbation. Above it, a small, random 3D kick causes a bifurcation from axisymmetry and excites a spontaneous helical core, which is independent of the kick size. Systematic scans of the q-profile show that the onset threshold is very sensitive to the q-shear in the core. Helical cores occur frequently in Alcator C-Mod during ramp-up when slow current penetration results in a reversed shear q-profile, which is favorable for helical core formation. Finally, a comparison of the helical core onset threshold for discharges from DIII-D, Alcator C-Mod and ITER confirms that while DIII-D is marginally stable, Alcator C-Mod and especially ITER are highly susceptible to helical core formation without being driven by an externally applied 3D magnetic field.

Chaotic density fluctuations in L-mode plasmas of the DIII-D tokamak

DOE PAGES

Maggs, J. E.; Rhodes, Terry L.; Morales, G. J.

2015-03-05

Analysis of the time series obtained with the Doppler backscattering system (DBS) in the DIII-D tokamak shows that intermediate wave number plasma density fluctuations in low confinement (L-mode) tokamak plasmas are chaotic. Here, the supporting evidence is based on the shape of the power spectrum; the location of the signal in the complexity-entropy plane (C-H plane); and the population of the corresponding Bandt-Pompe probability distributions.

Arcing and its role in PFC erosion and dust production in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Chrobak, C. P.; Doerner, R. P.; Krasheninnikov, S. I.; Moyer, R. A.; Umstadter, K. R.; Wampler, W. R.; Wong, C. P. C.

2013-07-01

Two types of arc tracks are observed on the plasma-facing components (PFCs) in DIII-D. "Unmagnetized" random walk tracks are produced during glow discharges; they are rare and have no importance for PFC erosion but may degrade diagnostic mirrors. "Magnetized" scratch-like type II tracks are produced by unipolar arcs during plasma operations; they are formed by "retrograde BxJ" motion of the cathode spot and are roughly perpendicular to the local magnetic field. Type II arcs cause measurable erosion of graphite, but based on the evidence available they are relatively small contributors to the total erosion of carbon in DIII-D compared to other mechanisms such as physical and chemical sputtering and ablation from leading edges. Erosion by arcing of tungsten films deposited on graphite samples was observed in Divertor Material Evaluation System (DiMES) experiments. New DiMES experiments aimed at time-resolved arc measurements are proposed.

Feedback Control of Resistive Wall Modes in Slowly Rotating DIII-D Plasmas

NASA Astrophysics Data System (ADS)

Okabayashi, M.; Chance, M. S.; Takahashi, H.; Garofalo, A. M.; Reimerdes, H.; in, Y.; Chu, M. S.; Jackson, G. L.; La Haye, R. J.; Strait, E. J.

2006-10-01

In slowly rotating plasmas on DIII-D, the requirement of RWM control feedback have been identified, using a MHD code along with measured power supply characteristics. It was found that a small time delay is essential for achieving high beta if no rotation stabilization exists. The overall system delay or the band pass time constant should be in the range of 0.4 of the RWM growth time. Recently the control system was upgraded using twelve linear audio amplifiers and a faster digital control system, reducing the time-delay from 600 to 100 μs. The advantage has been clearly observed when the RWMs excited by ELMs were effectively controlled by feedback even if the rotation transiently slowed nearly to zero. This study provides insight on stability in the low- rotation plasmasw with balanced NBI in DIII-D and also in ITER.

Exposures of tungsten nanostructures to divertor plasmas in DIII-D

DOE PAGES

Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; ...

2016-01-22

Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with littlemore » obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.« less

Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henline, P.A.

1995-12-31

The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DIII-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape controlmore » due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.« less

Currents in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Azari, A.; Eidietis, N. W.

2012-10-01

Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

Nonlinear Simulation of DIII-D Plasma and Poloidal Systems Using DINA and Simulink

NASA Astrophysics Data System (ADS)

Walker, M. L.; Leuer, J. A.; Deranian, R. D.; Humphreys, D. A.; Khayrutdinov, R. R.

2002-11-01

Hardware-in-the-loop simulation capability was developed previously for poloidal shape control testing using Matlab Simulink [1]. This has been upgraded by replacing a linearized plasma model with the DINA nonlinear plasma evolution code [2]. In addition to its use for shape control studies, this new capability will allow study of current profile control using the DINA model of electron cyclotron current drive (ECCD) and current profile information soon to be available from the Plasma Control System (PCS) real time EFIT [3] calculation. We describe the incorporation of DINA into the Simulink DIII-D tokamak systems model and results of validating this combined model against DIII-D data. \\vspace0.1em [1] J.A. Leuer, et al., 18th IEEE/NPSS SOFE (1999), p. 531. [2] R.R. Khayrutdinov, V.E. Lukash, J. Comput. Phys. 109, 193 (1993). [3] J.R. Ferron, et al., Nucl. Fusion 38, 1055 (1988).

Control of high-Z PFC erosion by local gas injection in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Stangeby, P. C.; Wong, C. P. C.; McLean, A. G.; Wampler, W. R.; Watkins, J. G.; Boedo, J. A.; Briesemeister, A.; Buchenauer, D. A.; Chrobak, C. P.; Elder, J. D.; Fenstermacher, M. E.; Guo, H. Y.; Lasnier, C. J.; Leonard, A. W.; Maingi, R.; Moyer, R. A.

2015-08-01

Reduced erosion of a high-Z PFC divertor surface was observed in DIII-D with local injection of methane and deuterium gases. Molybdenum-coated silicon samples were exposed in the lower divertor of DIII-D using DiMES under plasma conditions previously shown to cause significant net erosion of Mo. Three exposures with 13CH4 and one exposure with D2 gas injection about 12 cm upstream of the samples located within 1-2 cm of the attached strike point were performed. Reduction of Mo erosion was evidenced in-situ by the suppression of MoI line radiation at 386.4 nm once the gas injection started. Post-mortem ion beam analysis demonstrated that the net erosion of molybdenum near the center of the samples exposed with 13CH4 injection was below the measurement resolution of 0.5 nm, corresponding to a rate of ⩽0.04 nm/s. Compared to the previously measured erosion rates, this constitutes a reduction by a factor of >10.

Investigation of He-W interactions using DiMES on DIII-D

NASA Astrophysics Data System (ADS)

Doerner, R. P.; Rudakov, D. L.; Chrobak, C. P.; Briesemeister, A. R.; Corr, C.; De Temmerman, G.; Kluth, P.; Lasnier, C. J.; McLean, A. G.; Pace, D. C.; Pitts, R. A.; Schmitz, O.; Thompson, M.; Winters, V.

2016-02-01

Tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225-850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in the surface of W buttons after only a single DIII-D (3 s, ˜150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04-0.1 MJ m-2.

Validating predictive models for fast ion profile relaxation in burning plasmas

DOE PAGES

Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; ...

2016-07-22

The redistribution and potential loss of energetic particles due to MHD modes can limit the performance of fusion plasmas by reducing the plasma heating rate. In this work, we present validation studies of the 1.5D critical gradient model (CGM) for Alfvén eigenmode (AE) induced EP transport in NSTX and DIII-D neutral beam heated plasmas. In previous comparisons with a single DIII-D L-mode case, the CGM model was found to be responsible for 75% of measured AE induced neutron deficit [1]. A fully kinetic HINST is used to compute mode stability for the non-perturbative version of CGM (or nCGM). We have found that AEs show strong local instability drive up tomore » $$\\gamma /\\omega \\sim 20\\%$$ violating assumptions of perturbative approaches used in NOVA-K code. Lastly, we demonstrate that both models agree with each other and both underestimate the neutron deficit measured in DIII-D shot by approximately a factor of 2.« less

Bifurcation of quiescent H-mode to a wide pedestal regime in DIII-D and advances in the understanding of edge harmonic oscillations

DOE PAGES

Chen, Xi; Burrell, K. H.; Osborne, T. H.; ...

2017-06-14

New experimental studies and modelling of the coherent edge harmonic oscillation (EHO), which regulates the conventional Quiescent H-mode (QH-mode) edge, validate the proposed hypothesis of edge rotational shear in destabilizing the low-n kink-peeling mode as the additional drive mechanism for the EHO. The observed minimum edge E×B shear required for the EHO decreases linearly with pedestal collisionalitymore » $$\

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shafer, Morgan W; Battaglia, D. J.; Unterberg, Ezekial A

A new tangential 2D Soft X-Ray Imaging System (SXRIS) is being designed to examine the edge magnetic island structure in the lower X-point region of DIII-D. A synthetic diagnostic calculation coupled to 3D emissivity estimates is used to generate phantom images. Phillips-Tikhonov regularization is used to invert the phantom images for comparison to the original emissivity model. Noise level, island size, and equilibrium accuracy are scanned to assess the feasibility of detecting edge island structures. Models of typical DIII-D discharges indicate integration times > 1 ms with accurate equilibrium reconstruction are needed for small island (< 3 cm) detection.

Gyrokinetic simulations and experiment

NASA Astrophysics Data System (ADS)

Ross, David W.; Bravenec, R. V.; Dorland, W.

2002-11-01

Nonlinear gyrokinetic simulations with the code GS2 have been carried out in an effort to predict transport fluxes and fluctuation levels in the tokamaks DIII-D and Alcator C-Mod.(W. Dorland et al. in Fusion Energy 2000 (International Atomic Energy Agency, Vienna, 2000).)^,( W. Ross and W. Dorland, submitted to Phys. Plasmas (2002).) These simulations account for full electron dynamics and, in some instances, electromagnetic waves.( D. W. Ross, W. Dorland, and B. N. Rogers, Bull. Am. Phys. Soc. 46, 115 (2001).) Here, some issues of the necessary resolution, precision and wave number range are examined in connection with the experimental comparisons and parameter scans.

Improved confinement in highly powered high performance scenarios on DIII-D

NASA Astrophysics Data System (ADS)

Petrie, T. W.; Osborne, T.; Fenstermacher, M. E.; Ferron, J.; Groebner, R.; Grierson, B.; Holcomb, C.; Lasnier, C.; Leonard, A.; Luce, T.; Makowski, M.; Turco, F.; Solomon, W.; Victor, B.; Watkins, J.

2017-08-01

DIII-D has recently demonstrated improved energy confinement by injecting neutral deuterium gas into high performance near-double null divertor (DND) plasmas during high power operation. Representative parameters for these plasmas are: q 95  =  6, P IN up to 15 MW, H 98  =  1.4-1.8, and β N  =  2.5-4.0. The ion B   ×  \

Experiments on Helicon Excitation and Off-Axis Current Drive on DIII-D: Status and Plans

NASA Astrophysics Data System (ADS)

Pinsker, R. I.; Prater, R.; Moeller, C. P.; Degrassie, J. S.; Tooker, J. F.; Anderson, J. P.; Torreblanca, H.; Hansink, M.; Nagy, A.; Porkolab, M.

2015-11-01

Fast waves in the LHRF, also called ``whistlers'' or ``helicons,'' will be studied in experiments on the DIII-D tokamak beginning in autumn 2015. In the first stage, a 12-element traveling wave antenna (``comb-line'') is installed in the DIII-D vessel for operation at very low power (~ 0.1 kW) at 476 MHz, with a well-defined launched n| | spectrum peaked at 3.0. The goals of the low-power experiment include: (1) determining the efficiency with which the desired fast waves can be excited under a variety of plasma conditions in discharges relevant to the subsequent high-power current drive experiments and (2) proving that the radial and poloidal location at which the antenna will be mounted does not cause deleterious effects in the DIII-D discharges with high neutral beam power, and that the antenna is not damaged by fast ion losses, etc. Plans for 1 MW-level experiments with a single klystron beginning in FY17 are discussed. In addition to demonstrating off-axis current drive at an efficiency of ~ 60 kA/MW in high-performance plasmas, these experiments will explore non-linear aspects of wave excitation, propagation and absorption such as ponderomotive effects and parametric decay instabilities. Supported by US DOE DE-FC02-04ER54698, DE-AC02-09CH11466 and DE-FG02-94ER54084.

Nonlinear MHD simulations of QH-mode DIII-D plasmas and implications for ITER high Q scenarios

NASA Astrophysics Data System (ADS)

Liu, F.; Huijsmans, G. T. A.; Loarte, A.; Garofalo, A. M.; Solomon, W. M.; Hoelzl, M.; Nkonga, B.; Pamela, S.; Becoulet, M.; Orain, F.; Van Vugt, D.

2018-01-01

In nonlinear MHD simulations of DIII-D QH-mode plasmas it has been found that low n kink/peeling modes (KPMs) are unstable and grow to a saturated kink-peeling mode. The features of the dominant saturated KPMs, which are localised toroidally by nonlinear coupling of harmonics, such as mode frequencies, density fluctuations and their effect on pedestal particle and energy transport, are in good agreement with the observations of the edge harmonic oscillation typically present in DIII-D QH-mode experiments. The nonlinear evolution of MHD modes including both kink-peeling modes and ballooning modes, is investigated through MHD simulations by varying the pedestal current and pressure relative to the initial conditions of DIII-D QH-mode plasma. The edge current and pressure at the pedestal are key parameters for the plasma either saturating to a QH-mode regime or a ballooning mode dominant regime. The influence of E × B flow and its shear on the QH-mode plasma has been investigated. E × B flow shear has a strong stabilisation effect on the medium to high-n modes but is destabilising for the n = 2 mode. The QH-mode extrapolation results of an ITER Q = 10 plasma show that the pedestal currents are large enough to destabilise n = 1-5 KPMs, leading to a stationary saturated kink-peeling mode.

Validation of TGLF in C-Mod and DIII-D using machine learning and integrated modeling tools

NASA Astrophysics Data System (ADS)

Rodriguez-Fernandez, P.; White, Ae; Cao, Nm; Creely, Aj; Greenwald, Mj; Grierson, Ba; Howard, Nt; Meneghini, O.; Petty, Cc; Rice, Je; Sciortino, F.; Yuan, X.

2017-10-01

Predictive models for steady-state and perturbative transport are necessary to support burning plasma operations. A combination of machine learning algorithms and integrated modeling tools is used to validate TGLF in C-Mod and DIII-D. First, a new code suite, VITALS, is used to compare SAT1 and SAT0 models in C-Mod. VITALS exploits machine learning and optimization algorithms for the validation of transport codes. Unlike SAT0, the SAT1 saturation rule contains a model to capture cross-scale turbulence coupling. Results show that SAT1 agrees better with experiments, further confirming that multi-scale effects are needed to model heat transport in C-Mod L-modes. VITALS will next be used to analyze past data from DIII-D: L-mode ``Shortfall'' plasma and ECH swing experiments. A second code suite, PRIMA, allows for integrated modeling of the plasma response to Laser Blow-Off cold pulses. Preliminary results show that SAT1 qualitatively reproduces the propagation of cold pulses after LBO injections and SAT0 does not, indicating that cross-scale coupling effects play a role in the plasma response. PRIMA will be used to ``predict-first'' cold pulse experiments using the new LBO system at DIII-D, and analyze existing ECH heat pulse data. Work supported by DE-FC02-99ER54512, DE-FC02-04ER54698.

Building a database for statistical characterization of ELMs on DIII-D

NASA Astrophysics Data System (ADS)

Fritch, B. J.; Marinoni, A.; Bortolon, A.

2017-10-01

Edge localized modes (ELMs) are bursty instabilities which occur in the edge region of H-mode plasmas and have the potential to damage in-vessel components of future fusion machines by exposing the divertor region to large energy and particle fluxes during each ELM event. While most ELM studies focus on average quantities (e.g. energy loss per ELM), this work investigates the statistical distributions of ELM characteristics, as a function of plasma parameters. A semi-automatic algorithm is being used to create a database documenting trigger times of the tens of thousands of ELMs for DIII-D discharges in scenarios relevant to ITER, thus allowing statistically significant analysis. Probability distributions of inter-ELM periods and energy losses will be determined and related to relevant plasma parameters such as density, stored energy, and current in order to constrain models and improve estimates of the expected inter-ELM periods and sizes, both of which must be controlled in future reactors. Work supported in part by US DoE under the Science Undergraduate Laboratory Internships (SULI) program, DE-FC02-04ER54698 and DE-FG02- 94ER54235.

A study for the installation of the TEXT heavy-ion beam probe on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edmonds, P.H.; Solano, E.R.; Bravenec, R.V.

1997-01-01

An assessment of the feasibility of installing the TEXT 2 MeV heavy-ion beam probe on the DIII-D tokamak has been completed. Detailed drawings of the machine cross section were imported into the CAD application AutoCAD. A set of programs written in AutoLisp were used to generate trajectories. Displays of the accessible cross section of the plasma, scan lines for the entire range of primary beam energy and injection angle ranges, and sample{endash}volume dimensions can be rapidly generated. Because of the large deflection between the primary input beam and the emergent secondary beam, either the analyzer needs to be tracked overmore » a {plus_minus}20{degree} angle or secondary poloidal deflector plates need to be installed at the exit port. Toroidal deflector plates will be installed at both the injection and exit ports to compensate for toroidal displacements and deflections. The sample volumes generated by this procedure are within a few centimeters of the locations derived from a full three-dimensional calculation.{copyright} {ital 1997 American Institute of Physics.}« less

Analysis of Alfven eigenmode destabilization in DIII-D high poloidal β discharges using a Landau closure model

NASA Astrophysics Data System (ADS)

Varela, J.; Spong, D. A.; Garcia, L.; Huang, J.; Murakami, M.; Garofalo, A. M.; Qian, J. P.; Holcomb, C. T.; Hyatt, A. W.; Ferron, J. R.; Collins, C. S.; Ren, Q. L.; McClenaghan, J.; Guo, W.

2018-07-01

Alfvén eigenmodes are destabilized at the DIII-D pedestal during transient beta drops in high poloidal β discharges with internal transport barriers (ITBs), driven by n  =  1 external kink modes, leading to energetic particle losses. There are two different scenarios in the thermal β recovery phase: with bifurcation (two instability branches with different frequencies) or without bifurcation (single instability branch). We use the reduced MHD equations in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles as well as the geodesic acoustic wave dynamics, to study the properties of the instabilities observed in the DIII-D high poloidal β discharges and identify the conditions to trigger the bifurcation. The simulations suggest that instabilities with lower frequency in the bifurcation case are ballooning modes driven at the plasma pedestal, while the instability branch with higher frequencies are low n (n  <  4) toroidal Alfvén eigenmodes nearby the pedestal. The reverse shear region between the middle and plasma periphery in the non-bifurcated case avoids the excitation of ballooning modes at the pedestal, although toroidal Alfvén eigenmodes and reverse shear Alfvén eigenmodes are unstable in the reverse shear region. The n  =  1 and n  =  2 Alfvén eigenmode activity can be suppressed or minimized if the neutral beam injector (NBI) intensity is lower than the experimental value (). In addition, if the beam energy or neutral beam injector voltage is lower than in the experiment (), the resonance between beam and thermal plasma is weaker. The and 6 AE activity can not be fully suppressed, although the growth rate and frequency is smaller for an optimized neutral beam injector operation regime. In conclusion, AE activity in high poloidal β discharges can be minimized for optimized NBI operation regimes.

Optimization of DIII-D discharges to avoid AE destabilization

NASA Astrophysics Data System (ADS)

Varela, Jacobo; Spong, Donald; Garcia, Luis; Huang, Juan; Murakami, Masanori

2017-10-01

The aim of the study is to analyze the stability of Alfven Eigenmodes (AE) perturbed by energetic particles (EP) during DIII-D operation. We identify the optimal NBI operational regimes that avoid or minimize the negative effects of AE on the device performance. We use the reduced MHD equations to describe the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles, including the effect of the acoustic modes. We add the Landau damping and resonant destabilization effects using a closure relation. We perform parametric studies of the MHD and AE stability, taking into account the experimental profiles of the thermal plasma and EP, also using a range of values of the energetic particles β, density and velocity as well the effect of the toroidal couplings. We reproduce the AE activity observed in high poloidal β discharge at the pedestal and reverse shear discharges. This material based on work is supported both by the U.S. Department of Energy, Office of Science, under Contract DE-AC05-00OR22725 with UT-Battelle, LLC. Research sponsored in part by the Ministerio de Economia y Competitividad of Spain under the project.

Dust Studies in DIII-D and TEXTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudakov, D L; Litnovsky, A; West, W P

2009-02-17

Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Direct heating of the dust particles by the neutralmore » beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust. Large flakes or debris falling into the plasma may result in a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micron-size dust in plasma discharges. In DIII-D, a sample holder filled with {approx}30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure ({approx}0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. At the given configuration of the launch, the dust did not penetrate the core plasma and only moderately perturbed the edge plasma, as evidenced by an increase of the edge carbon content.« less

Dust studies in DIII-D and TEXTOR

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Litnovsky, A.; West, W. P.; Yu, J. H.; Boedo, J. A.; Bray, B. D.; Brezinsek, S.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Hollmann, E. M.; Huber, A.; Hyatt, A. W.; Krasheninnikov, S. I.; Lasnier, C. J.; McLean, A. G.; Moyer, R. A.; Pigarov, A. Yu.; Philipps, V.; Pospieszczyk, A.; Smirnov, R. D.; Sharpe, J. P.; Solomon, W. M.; Watkins, J. G.; Wong, C. P. C.

2009-08-01

Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicrometre sized dust is routinely observed using Mie scattering from a Nd : Yag laser. The source is strongly correlated with the presence of type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust; on the other hand, large flakes or debris falling into the plasma may induce a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micrometre-size particles into plasma discharges. In DIII-D, a sample holder filled with 30-40 mg of dust is inserted in the lower divertor and exposed, via sweeping of the strike points, to the diverted plasma flux of high-power ELMing H-mode discharges. After a brief dwell (~0.1 s) of the outer strike point on the sample holder, part of the dust penetrates into the core plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase in the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off-layer 0-2 cm radially outside of the last closed flux surface in discharges heated with 1.4 MW of NBI. Launched in this configuration, the dust perturbed the edge plasma, as evidenced by a moderate increase in the edge carbon content, but did not penetrate into the core plasma.

Energetic-particle-induced geodesic acoustic mode.

PubMed

Fu, G Y

2008-10-31

A new energetic particle-induced geodesic acoustic mode (EGAM) is shown to exist. The mode frequency and mode structure are determined nonperturbatively by energetic particle kinetic effects. In particular the EGAM frequency is found to be substantially lower than the standard GAM frequency. The radial mode width is determined by the energetic particle drift orbit width and can be fairly large for high energetic particle pressure and large safety factor. These results are consistent with the recent experimental observation of the beam-driven n=0 mode in DIII-D.

ELM Mitigation in Low-rotation ITER Baseline Scenario Plasmas on DIII-D with Deuterium Pellet Injection

NASA Astrophysics Data System (ADS)

Baylor, L. R.

2016-10-01

ELM mitigation using high frequency D2 pellet ELM pacing has been demonstrated in ITER baseline scenario plasmas on DIII D with low rotation obtained with low NBI input torque. The ITER burning plasmas will have relatively low input torque and are expected to have low rotation. ELM mitigation by on-demand pellet ELM triggering has not been observed before in these conditions. New experiments on DIII-D in these conditions with 90 Hz D2 pellets have shown that significant mitigation of the divertor ELM peak heat flux by a factor of 8 is possible without detrimental effects to the plasma confinement. High-Z impurity accumulation is dramatically reduced at all input torques from 0.1 to 2.5 N-m. Fueling with high field side injection of D2 pellets has been employed to demonstrate that density buildup can be obtained simultaneously with ELM mitigation. The implications are that rapid pellet injection remains a promising technique to trigger on-demand ELMs in low rotating plasmas with greatly reduced peak flux while preventing impurity accumulation in ITER. Supported by the US DOE under DE-AC05-00OR22725, DE-FC02-04ER54698.

DIII-D Neutron Measurement: Status and Plan for Simplification and Upgrade

NASA Astrophysics Data System (ADS)

Zhu, Y. B.; Heidbrink, W. W.; Taylor, P. L.; Finkenthal, D.

2017-10-01

Neutron diagnostics play key essential roles on DIII-D. Historically an 18-channel 2.45MeV D-D neutron measurement system based on 3He and BF3 proportional counters was inherited from Doublet-III including associated electronics and CAMAC data acquisition. Three fission chambers and two neutron scintillators were added in the 1980s and middle 1990s respectively. For Tritium burn-up studies, two 14MeV D-T neutron measurement systems were installed in 2009 and 2010. Operation and maintenance experience have led to a plan to simplify and upgrade these aging systems to provide a more economical and reliable solution for future DIII-D experiments. On simplification, most conventional expensive NIM and CAMAC modules will be removed. Advanced technologies like ultra-fast data acquisition and software-based pulse identification have been successfully tested. Significant data reduction and efficiency improvement will be achieved by real-time digital pulse identification with a field-programmable gate array. The partly renewed system will consist of 4 neutron counters for absolute calibration and 4 relatively calibrated neutron scintillators covering a wide measurement range. Work supported by US DOE under DE-FC02-04ER54698.

Modeling non-stationary, non-axisymmetric heat patterns in DIII-D tokamak

DOE PAGES

Ciro, D.; Evans, T. E.; Caldas, I. L.

2016-10-27

Non-axisymmetric stationary magnetic perturbations lead to the formation of homoclinic tangles near the divertor magnetic saddle in tokamak discharges. These tangles intersect the divertor plates in static helical structures that delimit the regions reached by open magnetic field lines reaching the plasma column and leading the charged particles to the strike surfaces by parallel transport. In this article we introduce a non-axisymmetric rotating magnetic perturbation to model the time evolution of the three-dimensional magnetic field of a singlenull DIII-D tokamak discharge developing a rotating tearing mode. The non-axiymmetric field is modeled using the magnetic signals to adjust the phases andmore » currents of a set of internal filamentary currents that approximate the magnetic field in the plasma edge region. The stable and unstable manifolds of the asymmetric magnetic saddle are obtained through an adaptive calculation providing the cuts at a given poloidal plane and the strike surfaces. Lastly, for the modeled shot, the experimental heat pattern and its time development are well described by the rotating unstable manifold, indicating the emergence of homoclinic lobes in a rotating frame due to the plasma instabilities.« less

Overview of Recent DIII-D Experimental Results

NASA Astrophysics Data System (ADS)

Fenstermacher, Max

2015-11-01

Recent DIII-D experiments have added to the ITER physics basis and to physics understanding for extrapolation to future devices. ELMs were suppressed by RMPs in He plasmas consistent with ITER non-nuclear phase conditions, and in steady state hybrid plasmas. Characteristics of the EHO during both standard high torque, and low torque enhanced pedestal QH-mode with edge broadband fluctuations were measured, including edge localized density fluctuations with a microwave imaging reflectometer. The path to Super H-mode was verified at high beta with a QH-mode edge, and in plasmas with ELMs triggered by Li granules. ITER acceptable TQ mitigation was obtained with low Ne fraction Shattered Pellet Injection. Divertor ne and Te data from Thomson Scattering confirm predicted drift-driven asymmetries in electron pressure, and X-divertor heat flux reduction and detachment were characterized. The crucial mechanisms for ExB shear control of turbulence were clarified. In collaboration with EAST, high beta-p scenarios were obtained with 80 % bootstrap fraction, high H-factor and stability limits, and large radius ITBs leading to low AE activity. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

Central Safety Factor and Normalized Beta Control Under Near-Zero Input Torque Constraints in DIII-D

NASA Astrophysics Data System (ADS)

Pajares, Andres; Wehner, William; Schuster, Eugenio; Burrell, Keith; Ferron, John; Walker, Michael; Humphreys, David; Lehigh University Team; Atomics Team, General

2017-10-01

DIII-D experiments have assessed the capability of combined central safety factor (q0) and normalized beta (βN) control under near-zero net torque to facilitate access to QH-mode with reverse Ip and normal Bt. Regulation of q0 and βN can prevent magneto-hydrodynamic instabilities that deteriorate plasma performance in discharges with a monotonically increasing safety-factor profile. Zero-input-torque scenarios are of special interest because future burning plasma tokamaks such as ITER will most likely operate with very low input torque, which makes these scenarios more susceptible to locked modes. To support studies of such scenarios, a controller for simultaneous regulation of q0 and βN has been developed using near-zero net input torque actuators including balanced neutral beam injection (NBI) and electron-cyclotron heating & current drive (ECH/ECCD). Experimental results show that in spite of the presence of locked modes the use of feedback control resulted in good tracking of the commanded q0 and βN when both ECCD/ECH and NBI were available. Supported by the US DOE under DE-SC0010661 and DE-FC02-04ER54698.

Modeling non-stationary, non-axisymmetric heat patterns in DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ciro, D.; Evans, T. E.; Caldas, I. L.

Non-axisymmetric stationary magnetic perturbations lead to the formation of homoclinic tangles near the divertor magnetic saddle in tokamak discharges. These tangles intersect the divertor plates in static helical structures that delimit the regions reached by open magnetic field lines reaching the plasma column and leading the charged particles to the strike surfaces by parallel transport. In this article we introduce a non-axisymmetric rotating magnetic perturbation to model the time evolution of the three-dimensional magnetic field of a singlenull DIII-D tokamak discharge developing a rotating tearing mode. The non-axiymmetric field is modeled using the magnetic signals to adjust the phases andmore » currents of a set of internal filamentary currents that approximate the magnetic field in the plasma edge region. The stable and unstable manifolds of the asymmetric magnetic saddle are obtained through an adaptive calculation providing the cuts at a given poloidal plane and the strike surfaces. Lastly, for the modeled shot, the experimental heat pattern and its time development are well described by the rotating unstable manifold, indicating the emergence of homoclinic lobes in a rotating frame due to the plasma instabilities.« less

Measurements of tungsten migration in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Wampler, W. R.; Rudakov, D. L.; Watkins, J. G.; McLean, A. G.; Unterberg, E. A.; Stangeby, P. C.

2017-12-01

An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.

Three-dimensional drift kinetic response of high- β plasmas in the DIII-D tokamak

DOE PAGES

Wang, Zhirui R.; Lanctot, Matthew J.; Liu, Y. Q.; ...

2015-04-07

A quantitative interpretation of the experimentally measured high pressure plasma response to externally applied three-dimensional (3D) magnetic field perturbations, across the no-wall Troyon limit, is achieved. The key to success is the self-consistent inclusion of the drift kinetic resonance effects in numerical modeling using the MARS-K code. This resolves an outstanding issue of ideal magneto-hydrodynamic model, which signi cantly over-predicts the plasma induced field ampli fication near the no-wall limit, as compared to experiments. The self-consistent drift kinetic model leads to quantitative agreement not only for the measured 3D field amplitude and toroidal phase, but also for the measured internalmore » 3D displacement of the plasma.« less

Measurement of deuterium density profiles in the H-mode steep gradient region using charge exchange recombination spectroscopy on DIII-D

DOE PAGES

Haskey, S. R.; Grierson, B. A.; Burrell, K. H.; ...

2016-09-26

Recent completion of a thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on the DIII-D tokamak enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. In a H-mode DIII-D discharge, these new measurement capabilities are used to provide the deuterium density profile, demonstrate the importance of profile alignment between Thomson scattering and CER diagnostics, and aid in determining the electron temperature at the separatrix. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the plasma edge, providing high resolution measurements across the pedestal and steep gradient region inmore » H-mode plasmas. Extracting useful physical quantities such as deuterium density is challenging due to multiple photoemission processes. Finally, these challenges are overcome using a detailed fitting model and by forward modeling the photoemission using the FIDASIM code, which implements a comprehensive collisional radiative model. Published by AIP Publishing.« less

Reduction of toroidal rotation by fast wave power in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grassie, J.S. de; Baker, D.R.; Burrell, K.H.

1997-04-01

The application of fast wave power in DIII-D has proven effective for both electron heating and current drive. Since the last RIF Conference FW power has been applied to advanced confinement regimes in DIII-D; negative central shear (NCS), VH- and H-modes, high {beta}{sub p}, and high-{ell}i. Typically these regimes show enhanced confinement of toroidal momentum exhibited by increased toroidal rotation velocity. Indeed, layers of large shear in toroidal velocity are associated with transport barriers. A rather common occurrence in these experiments is that the toroidal rotation velocity is decreased when the FW power is turned on, to lowest order independentmore » of whether the antennas are phased for co or counter current drive. At present all the data is for co-injected beams. The central toroidal rotation can be reduced to 1/2 of the non-FW level. Here the authors describe the effect in NCS discharges with co-beam injection.« less

A combined phase contrast imaging-interferometer system for the detection of multiscale density fluctuations on DIII-D

NASA Astrophysics Data System (ADS)

Davis, E. M.; Rost, J. C.; Porkolab, M.; Marinoni, A.; van Zeeland, M. A.

2016-10-01

A heterodyne interferometer channel has been added to the DIII-D phase contrast imaging (PCI) system. Both measurements share a single 10.6 μm probe beam. Whereas the PCI excels at detecting medium- to high- k fluctuations (1.5 cm-1 <= k <= 20 cm-1), the interferometer extends the system sensitivity to low- k fluctuations (k <= 5 cm-1), allowing simultaneous measurement of electron- and ion-scale instabilities with sub-microsecond resolution. Further, correlating measurements from the interferometer channel with those from DIII-D's pre-existing, toroidally separated interferometer (Δ∅ = 45°) allows identification of low- n modes. This new capability has been corroborated against magnetic measurements and may allow novel investigations of core - localized MHD that is otherwise inaccessible via external magnetic measurements, with potential applications to fast particle transport and disruptions. Work supported by USDOE under DE-FG02-94ER54235, DE-FC02-04ER54698, and DE-FC02-99ER54512.

Comparison and prediction of chirping in NSTX and DIII-D

NASA Astrophysics Data System (ADS)

Duarte, Vinicius; Berk, Herbert; Gorelenkov, Nikolai; Heidbrink, William; Kramer, Gerrit; Nazikian, Raffi; Pace, David; Podesta, Mario; van Zeeland, Michael

2016-10-01

We present an explanation of why frequency chirping of Alfven waves is ubiquitous in NSTX and rarely observed in DIII-D. A time-delayed cubic nonlinear equation is employed for the study of the onset of nonlinear phase-space structures. Its explosive solutions are chirping precursors. We employ NOVA and NOVA-K codes to provide consistent Alfvenic eigenmodes and weighted physical contributions from all regions of phase space. In addition, TRANSP is employed to determine the diffusivity needed to fulfill power balance. Though background micro-turbulence is usually unimportant in determining the energetic particle spatial profile, it may still be important with regard to whether chirping structures likely form. We show that the energetic particle micro-turbulent induced scattering often competes with collisional pitch-angle scattering. This competition explains the tendency for NSTX, where micro-turbulence is weak, to exhibit Alfvénic chirping, whereas in DIII-D turbulent diffusion usually dominates and chirping is not observed except when micro-turbulence markedly reduces.

Technical overview of the millimeter-wave imaging reflectometer on the DIII-D tokamak (invited)

DOE PAGES

Muscatello, Christopher M.; Domier, Calvin W.; Hu, Xing; ...

2014-07-22

The two-dimensional mm-wave imaging reflectometer (MIR) on DIII-D is a multi-faceted device for diagnosing electron density fluctuations in fusion plasmas. Its multi-channel, multi-frequency capabilities and high sensitivity permit visualization and quantitative diagnosis of density perturbations, including correlation length, wavenumber, mode propagation velocity, and dispersion. The two-dimensional capabilities of MIR are made possible with twelve vertically separated sightlines and four-frequency operation (corresponding to four radial channels). The 48-channel DIII-D MIR system has a tunable source that can be stepped in 500 µs increments over a range of 56 to 74 GHz. An innovative optical design keeps both on-axis and off-axis channelsmore » focused at the cutoff surface, permitting imaging over an extended poloidal region. As a result, the integrity of the MIR optical design is confirmed by comparing Gaussian beam calculations to laboratory measurements of the transmitter beam pattern and receiver antenna patterns.« less

Measurement of deuterium density profiles in the H-mode steep gradient region using charge exchange recombination spectroscopy on DIII-D.

PubMed

Haskey, S R; Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Kaplan, D H; Pablant, N A; Stagner, L

2016-11-01

Recent completion of a thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. In a H-mode DIII-D discharge, these new measurement capabilities are used to provide the deuterium density profile, demonstrate the importance of profile alignment between Thomson scattering and CER diagnostics, and aid in determining the electron temperature at the separatrix. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the plasma edge, providing high resolution measurements across the pedestal and steep gradient region in H-mode plasmas. Extracting useful physical quantities such as deuterium density is challenging due to multiple photoemission processes. These challenges are overcome using a detailed fitting model and by forward modeling the photoemission using the FIDASIM code, which implements a comprehensive collisional radiative model.

Measurement of deuterium density profiles in the H-mode steep gradient region using charge exchange recombination spectroscopy on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haskey, S. R.; Grierson, B. A.; Burrell, K. H.

Recent completion of a thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on the DIII-D tokamak enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. In a H-mode DIII-D discharge, these new measurement capabilities are used to provide the deuterium density profile, demonstrate the importance of profile alignment between Thomson scattering and CER diagnostics, and aid in determining the electron temperature at the separatrix. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the plasma edge, providing high resolution measurements across the pedestal and steep gradient region inmore » H-mode plasmas. Extracting useful physical quantities such as deuterium density is challenging due to multiple photoemission processes. Finally, these challenges are overcome using a detailed fitting model and by forward modeling the photoemission using the FIDASIM code, which implements a comprehensive collisional radiative model. Published by AIP Publishing.« less

Striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D [On the striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D

DOE PAGES

Frerichs, H.; Schmitz, Oliver; Reiter, D.; ...

2014-02-04

The application of resonant magnetic perturbations (RMPs) results in a non-axisymmetric striation pattern of magnetic field lines from the plasma interior which intersect the divertor targets. The impact on related particle and heat fluxes is investigated by three dimensional computer simulations for two different recycling conditions (controlled via neutral gas pumping). It is demonstrated that a mismatch between the particle and heat flux striation pattern, as is repeatedly observed in ITER similar shape H-mode plasmas at DIII-D, can be reproduced by the simulations for high recycling conditions at the onset of partial detachment. Finally, these results indicate that a detailedmore » knowledge of the particle and energy balance is at least as important for realistic simulations as the consideration of a change in the magnetic field structure by plasma response effects.« less

Measurement of deuterium density profiles in the H-mode steep gradient region using charge exchange recombination spectroscopy on DIII-D

NASA Astrophysics Data System (ADS)

Haskey, S. R.; Grierson, B. A.; Burrell, K. H.; Chrystal, C.; Groebner, R. J.; Kaplan, D. H.; Pablant, N. A.; Stagner, L.

2016-11-01

Recent completion of a thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. In a H-mode DIII-D discharge, these new measurement capabilities are used to provide the deuterium density profile, demonstrate the importance of profile alignment between Thomson scattering and CER diagnostics, and aid in determining the electron temperature at the separatrix. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the plasma edge, providing high resolution measurements across the pedestal and steep gradient region in H-mode plasmas. Extracting useful physical quantities such as deuterium density is challenging due to multiple photoemission processes. These challenges are overcome using a detailed fitting model and by forward modeling the photoemission using the FIDASIM code, which implements a comprehensive collisional radiative model.

Neutron Detection With Ultra-Fast Digitizer and Pulse Identification Techniques on DIII-D

NASA Astrophysics Data System (ADS)

Zhu, Y. B.; Heidbrink, W. W.; Piglowski, D. A.

2013-10-01

A prototype system for neutron detection with an ultra-fast digitizer and pulse identification techniques has been implemented on the DIII-D tokamak. The system consists of a cylindrical neutron fission chamber, a charge sensitive amplifier, and a GaGe Octopus 12-bit CompuScope digitizer card installed in a Linux computer. Digital pulse identification techniques have been successfully performed at maximum data acquisition rate of 50 MSPS with on-board memory of 2 GS. Compared to the traditional approach with fast nuclear electronics for pulse counting, this straightforward digital solution has many advantages, including reduced expense, improved accuracy, higher counting rate, and easier maintenance. The system also provides the capability of neutron-gamma pulse shape discrimination and pulse height analysis. Plans for the upgrade of the old DIII-D neutron counting system with these techniques will be presented. Work supported by the US Department of Energy under SC-G903402, and DE-FC02-04ER54698.

Extrapolation of the DIII-D high poloidal beta scenario to ITER steady-state using transport modeling

NASA Astrophysics Data System (ADS)

McClenaghan, J.; Garofalo, A. M.; Meneghini, O.; Smith, S. P.

2016-10-01

Transport modeling of a proposed ITER steady-state scenario based on DIII-D high βP discharges finds that the core confinement may be improved with either sufficient rotation or a negative central shear q-profile. The high poloidal beta scenario is characterized by a large bootstrap current fraction( 80%) which reduces the demands on the external current drive, and a large radius internal transport barrier which is associated with improved normalized confinement. Typical temperature and density profiles from the non-inductive high poloidal beta scenario on DIII-D are scaled according to 0D modeling predictions of the requirements for achieving Q=5 steady state performance in ITER with ``day one'' H&CD capabilities. Then, TGLF turbulence modeling is carried out under systematic variations of the toroidal rotation and the core q-profile. Either strong negative central magnetic shear or rotation are found to successfully provide the turbulence suppression required to maintain the temperature and density profiles. This work supported by the US Department of Energy under DE-FC02-04ER54698.

Modeling carbon production and transport during ELMs in DIII-D

NASA Astrophysics Data System (ADS)

Hogan, J.; Wade, M.; Coster, D.; Lasnier, C.

2004-11-01

Large-scale Type I ELM events could provide a significant C source in ITER, and C production rates depend on incident D flux density and surface temperature, quantities which can vary significantly during an ELM event. Recent progress on DIII-D has improved opportunities for code comparison. Fast time-scale measurements of divertor CIII evolution [1] and fast edge CER measurements of C profile evolution during low-density DIII-D LSN ELMy H-modes (type I) [2] have been modeled using the solps5.0/Eirene99 coupled edge code and time dependent thermal analysis codes. An ELM model based on characteristics of MHD peeling-ballooning modes reproduces the pedestal evolution. Qualitative agreement for the CIII evolution during an ELM event is found using the Roth et al annealing model for chemical sputtering and the sensitivity to other models is described. Significant ELM-to-ELM variations in observed maximum divertor target IR temperature during nominally identical ELMs are investigated with models for C emission from micron-scale dust particles. [1] M Groth, M Fenstermacher et al J Nucl Mater 2003, [2] M Wade, K Burrell et al PSI-16

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

NASA Astrophysics Data System (ADS)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

Developing DIII-D To Prepare For ITER And The Path To Fusion Energy

NASA Astrophysics Data System (ADS)

Buttery, Richard; Hill, David; Solomon, Wayne; Guo, Houyang; DIII-D Team

2017-10-01

DIII-D pursues the advancement of fusion energy through scientific understanding and discovery of solutions. Research targets two key goals. First, to prepare for ITER we must resolve how to use its flexible control tools to rapidly reach Q =10, and develop the scientific basis to interpret results from ITER for fusion projection. Second, we must determine how to sustain a high performance fusion core in steady state conditions, with minimal actuators and a plasma exhaust solution. DIII-D will target these missions with: (i) increased electron heating and balanced torque neutral beams to simulate burning plasma conditions (ii) new 3D coil arrays to resolve control of transients (iii) off axis current drive to study physics in steady state regimes (iv) divertors configurations to promote detachment with low upstream density (v) a reactor relevant wall to qualify materials and resolve physics in reactor-like conditions. With new diagnostics and leading edge simulation, this will position the US for success in ITER and a unique knowledge to accelerate the approach to fusion energy. Supported by the US DOE under DE-FC02-04ER54698.

Investigation of He–W interactions using DiMES on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doerner, R. P.; Rudakov, D. L.; Chrobak, C. P.

Here, tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in themore » surface of W buttons after only a single DIII-D (3 s, ~150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m –2.« less

Investigation of He–W interactions using DiMES on DIII-D

DOE PAGES

Doerner, R. P.; Rudakov, D. L.; Chrobak, C. P.; ...

2016-01-22

Here, tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in themore » surface of W buttons after only a single DIII-D (3 s, ~150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m –2.« less

The inter-ELM tungsten erosion profile in DIII-D H-mode discharges and benchmarking with ERO+OEDGE modeling [The inter-ELM W erosion profile in DIII-D H-mode discharges and benchmarking with OEDGE+ERO modeling

DOE PAGES

Abrams, Tyler; Ding, Rui; Guo, Houyang Y.; ...

2017-04-03

It is important to develop a predictive capability for the tungsten source rate near the strike points during H-mode operation in ITER and beyond. H-mode deuterium plasma exposures were performed on W-coated graphite and TZM molybdenum substrates in the DIII-D divertor using DiMES. The W-I 400.9 nm spectral line was monitored by fast filtered diagnostics cross calibrated via a high-resolution spectrometer to resolve inter-ELM W erosion. The effective ionization/photon (S/XB) was calibrated using a unique method developed on DIII-D based on surface analysis. Inferred S/XB values agree with an existing empirical scaling at low electron density (n e) but divergemore » at higher densities, consistent with recent ADAS atomic physics modeling results. Edge modeling of the inter-ELM phase is conducted via OEDGE utilizing the new capability for charge-state resolved carbon impurity fluxes. ERO modeling is performed with the calculated main ion and impurity plasma background from OEDGE. ERO results demonstrate the importance a mixed-material surface model in the interpretation of W sourcing measurements. As a result, it is demonstrated that measured inter-ELM W erosion rates can be well explained by C→W sputtering only if a realistic mixed material model is incorporated.« less

The inter-ELM tungsten erosion profile in DIII-D H-mode discharges and benchmarking with ERO+OEDGE modeling [The inter-ELM W erosion profile in DIII-D H-mode discharges and benchmarking with OEDGE+ERO modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abrams, Tyler; Ding, Rui; Guo, Houyang Y.

It is important to develop a predictive capability for the tungsten source rate near the strike points during H-mode operation in ITER and beyond. H-mode deuterium plasma exposures were performed on W-coated graphite and TZM molybdenum substrates in the DIII-D divertor using DiMES. The W-I 400.9 nm spectral line was monitored by fast filtered diagnostics cross calibrated via a high-resolution spectrometer to resolve inter-ELM W erosion. The effective ionization/photon (S/XB) was calibrated using a unique method developed on DIII-D based on surface analysis. Inferred S/XB values agree with an existing empirical scaling at low electron density (n e) but divergemore » at higher densities, consistent with recent ADAS atomic physics modeling results. Edge modeling of the inter-ELM phase is conducted via OEDGE utilizing the new capability for charge-state resolved carbon impurity fluxes. ERO modeling is performed with the calculated main ion and impurity plasma background from OEDGE. ERO results demonstrate the importance a mixed-material surface model in the interpretation of W sourcing measurements. As a result, it is demonstrated that measured inter-ELM W erosion rates can be well explained by C→W sputtering only if a realistic mixed material model is incorporated.« less

Issues in Turbulence Simulation for Experimental Comparison

NASA Astrophysics Data System (ADS)

Ross, D. W.; Bravenec, R. V.; Dorland, W.; Beer, M. A.; Hammett, G. W.

1999-11-01

Studies of the sensitivity of fluctuation spectra and transport fluxes to local plasma parameters and gradients(D. W. Ross et al.), Bull. Am. Phys. Soc. 43, 1760 (1998); D. W. Ross et al., Transport Task Force Workshop, Portland, Oregon, (1999). are continued using nonlinear gyrofluid simulation(M. A. Beer et al.), Phys. Plasmas 2, 2687 (1995). on the T3E at NERSC. Parameters that are characteristic of discharges in DIII-D and Alcator C-Mod are employed. In the previous work, the gradients of Z_eff, n_e, and Te were varied within the experimental uncertainty. Amplitudes and fluxes are quite sensitive to dZ_eff/dr. Here, these studies are continued and extended to variation of other parameters, including T_e/T_i, and dT_i/dr, which are important for ion temperature gradient modes. The role of electric field shear is discussed. Implications for comparison with experiment, including transient perturbations, are discussed, with the goal of quantifying the accuracy of profile data needed to verify the turbulence theory.

High-Z material erosion and its control in DIII-D carbon divertor

DOE PAGES

Ding, Rui; Rudakov, Dimitry L.; Stangeby, Peter C.; ...

2017-03-16

It is expected that high-Z materials will be used as plasma-facing components (PFCs) in future fusion devices, making the erosion of high-Z material a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion ofmore » Mo and W are found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH 4 injection also provides information on radial transport due to E×B drifts and cross field diffusion. Finally, D 2 gas puffing is found to cause 2 local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.« less

Effect of resonant magnetic perturbations on microturbulence in DIII-D pedestal

DOE PAGES

Holod, I.; Lin, Z.; Taimourzadeh, S.; ...

2016-10-03

Vacuum resonant magnetic perturbations (RMP) applied to otherwise axisymmetric tokamak plasmas produce in general a combination of non-resonant effects that preserve closed flux surfaces (kink response) and resonant effects that introduce magnetic islands and/or stochasticity (tearing response). The effect of the plasma kink response on the linear stability and nonlinear transport of edge turbulence is studied using the gyrokinetic toroidal code GTC for a DIII-D plasma with applied n = 2 vacuum RMP. GTC simulations use the 3D equilibrium of DIII-D discharge 158103 (Nazikian et al 2015 Phys. Rev. Lett. 114 105002), which is provided by nonlinear ideal MHD VMECmore » equilibrium solver in order to include the effect of the plasma kink response to the external field but to exclude island formation at rational surfaces. Analysis using the GTC simulation results reveal no increase of growth rates for the electrostatic drift wave instability and for the electromagnetic kinetic-ballooning mode in the presence of the plasma kink response to the RMP. Moreover, nonlinear electrostatic simulations show that the effect of the 3D equilibrium on zonal flow damping is very weak and found to be insufficient to modify turbulent transport in the electrostatic turbulence.« less

Simulation of gross and net erosion of high-Z materials in the DIII-D divertor

DOE PAGES

Wampler, William R.; Ding, R.; Stangeby, P. C.; ...

2015-12-17

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 themore » 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.« less

Embedded calibration system for the DIII-D Langmuir probe analog fiber optic links

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watkins, J. G.; Rajpal, R.; Mandaliya, H.

2012-10-15

This paper describes a generally applicable technique for simultaneously measuring offset and gain of 64 analog fiber optic data links used for the DIII-D fixed Langmuir probes by embedding a reference voltage waveform in the optical transmitted signal before every tokamak shot. The calibrated data channels allow calibration of the power supply control fiber optic links as well. The array of fiber optic links and the embedded calibration system described here makes possible the use of superior modern data acquisition electronics in the control room.

Quiescent double barrier regime in the DIII-D tokamak.

PubMed

Greenfield, C M; Burrell, K H; DeBoo, J C; Doyle, E J; Stallard, B W; Synakowski, E J; Fenzi, C; Gohil, P; Groebner, R J; Lao, L L; Makowski, M A; McKee, G R; Moyer, R A; Rettig, C L; Rhodes, T L; Pinsker, R I; Staebler, G M; West, W P

2001-05-14

A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.

Reduction of Net Erosion of High-Z Divertor Surface by Local Redeposition in DIII-D

NASA Astrophysics Data System (ADS)

Stangeby, P. C.

2012-10-01

Utilizing the unique capability to expose material samples to well characterized diverted plasmas, recent DIII-D measurements have confirmed theoretical expectations of the relative net and gross erosion rates of molybdenum in the divertor region. Knowledge of these erosion rates is important for predicting first wall lifetime in future fusion devices. Theory suggests that the net erosion rate will be much less than gross erosion due to prompt local deposition of eroded ions by gyro-orbit motion, the strong E-field toward the target and friction with the fast plasma flow toward the target. However, experimental evidence to date has been contradictory. The results here, which are the most definitive to date, are consistent with the basic theoretical predictions. The net and gross erosion rates were measured utilizing 1-cm and 1-mm diameter Mo samples that are mounted on the DIII-D Divertor Material Evaluation System (DiMES) system and simultaneously exposed near the attached outer strike point of an L-mode plasma for 4 s. Due to the spatial extent of the re-deposition, the larger sample gives the net erosion while the smaller sample is indicative of the gross erosion. Post-mortem ion beam analysis (RBS) of the larger sample, indicates a 2.9 nm film thickness reduction (or 0.72 nm/s net erosion rate). Similar analysis of the smaller sample yields a 1.3 nm/s gross erosion rate, consistent with spectroscopic measurements of Mo I emission. The net to gross erosion ratio of 0.56 is consistent with calculations using a modeling package including REDEP/WBS and OEDGE codes. Using as input the measured plasma density and temperature profiles from divertor Langmuir probes, these codes estimate a net to gross erosion ratio of 0.46. Details of the modeling and implications for future devices will be discussed.

Validation of Kinetic-Turbulent-Neoclassical Theory for Edge Intrinsic Rotation in DIII-D Plasmas

NASA Astrophysics Data System (ADS)

Ashourvan, Arash

2017-10-01

Recent experiments on DIII-D with low-torque neutral beam injection (NBI) have provided a validation of a new model of momentum generation in a wide range of conditions spanning L- and H-mode with direct ion and electron heating. A challenge in predicting the bulk rotation profile for ITER has been to capture the physics of momentum transport near the separatrix and steep gradient region. A recent theory has presented a model for edge momentum transport which predicts the value and direction of the main-ion intrinsic velocity at the pedestal-top, generated by the passing orbits in the inhomogeneous turbulent field. In this study, this model-predicted velocity is tested on DIII-D for a database of 44 low-torque NBI discharges comprised of bothL- and H-mode plasmas. For moderate NBI powers (PNBI<4 MW), model prediction agrees well with the experiments for both L- and H-mode. At higher NBI power the experimental rotation is observed to saturate and even degrade compared to theory. TRANSP-NUBEAM simulations performed for the database show that for discharges with nominally balanced - but high powered - NBI, the net injected torque through the edge can exceed 1 N.m in the counter-current direction. The theory model has been extended to compute the rotation degradation from this counter-current NBI torque by solving a reduced momentum evolution equation for the edge and found the revised velocity prediction to be in agreement with experiment. Projecting to the ITER baseline scenario, this model predicts a value for the pedestal-top rotation (ρ 0.9) comparable to 4 kRad/s. Using the theory modeled - and now tested - velocity to predict the bulk plasma rotation opens up a path to more confidently projecting the confinement and stability in ITER. Supported by the US DOE under DE-AC02-09CH11466 and DE-FC02-04ER54698.

Modeling of fast neutral-beam-generated ion effects on MHD-spectroscopic observations of resistive wall mode stability in DIII-D plasmas [Modeling of fast neutral-beam-generated ion effects on MHD spectroscopic observations of RWM stability in DIII-D plasmas

DOE PAGES

Turco, Francesca; Turnbull, Alan D.; Hanson, Jeremy M.; ...

2015-02-03

Experiments conducted at DIII-D investigate the role of drift kinetic damping and fast neutral beam injection (NBI)-ions in the approach to the no-wall β N limit. Modelling results show that the drift kinetic effects are significant and necessary to reproduce the measured plasma response at the ideal no-wall limit. Fast neutral-beam ions and rotation play important roles and are crucial to quantitatively match the experiment. In this paper, we report on the model validation of a series of plasmas with increasing β N, where the plasma stability is probed by active magnetohydrodynamic (MHD) spectroscopy. The response of the plasma tomore » an externally applied field is used to probe the stable side of the resistive wall mode and obtain an indication of the proximity of the equilibrium to an instability limit. We describe the comparison between the measured plasma response and that calculated by means of the drift kinetic MARS-K code, which includes the toroidal rotation, the electron and ion drift-kinetic resonances, and the presence of fast particles for the modelled plasmas. The inclusion of kinetic effects allows the code to reproduce the experimental results within ~13% for both the amplitude and phase of the plasma response, which is a significant improvement with respect to the undamped MHD-only model. The presence of fast NBI-generated ions is necessary to obtain the low response at the highest β N levels (~90% of the ideal no-wall limit). Finally, the toroidal rotation has an impact on the results, and a sensitivity study shows that a large variation in the predicted response is caused by the details of the rotation profiles at high β N.« less

Transition From High Harmonic Fast Wave to Whistler/Helicon Regime in Tokamaks

NASA Astrophysics Data System (ADS)

Harris, S. P.; Pinsker, R. I.; Porkolab, M.

2014-10-01

Experiments are being prepared1 on DIII-D in which fast waves (FWs) at 0.5 GHz will be used to drive current noninductively in the mid-radius region. Previous DIII-D experiments used FWs at ~0.1 GHz to drive central current; in this work we examine the frequency dependence of wave propagation and damping in the 0.1-1.0 GHz range with the goal of identifying the optimum frequency range for a particular application. Strongly enhanced electron damping and reduced ion damping at higher frequencies must be weighed against increasing coupling difficulties at higher frequencies and more restrictive wave accessibility at low toroidal field. Wave propagation and accessibility is studied with ray tracing models in slab, cylindrical, and fully toroidal geometries. Analytic expressions for electron and ion damping will be derived with an emphasis on understanding the transition from the moderate-to-high ion cyclotron harmonic regime to the very high harmonic or ``whistler''/``helicon''/lower hybrid FW regime. Work supported in part by the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US Department of Energy under DE-FC02-04ER54698.

Origin of Non-Gaussian Spectra Observed via the Charge Exchange Recombination Spectroscopy Diagnostic in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Sulyman, Alex; Chrystal, Colin; Haskey, Shaun; Burrell, Keith; Grierson, Brian

2017-10-01

The possible observation of non-Maxwellian ion distribution functions in the pedestal of DIII-D will be investigated with a synthetic diagnostic that accounts for the effect of finite neutral beam size. Ion distribution functions in tokamak plasmas are typically assumed to be Maxwellian, however non-Gaussian features observed in impurity charge exchange spectra have challenged this concept. Two possible explanations for these observations are spatial averaging over a finite beam size and a local ion distribution that is non-Maxwellian. Non-Maxwellian ion distribution functions could be driven by orbit loss effects in the edge of the plasma, and this has implications for momentum transport and intrinsic rotation. To investigate the potential effect of finite beam size on the observed spectra, a synthetic diagnostic has been created that uses FIDAsim to model beam and halo neutral density. Finite beam size effects are investigated for vertical and tangential views in the core and pedestal region with varying gradient scale lengths. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program, DE-FC02-04ER54698, and DE-AC02-09CH11466.

Disruption prediction investigations using Machine Learning tools on DIII-D and Alcator C-Mod

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rea, C.; Granetz, R. S.; Montes, K.

Using data-driven methodology, we exploit the time series of relevant plasma parameters for a large set of disrupted and non-disrupted discharges to develop a classification algorithm for detecting disruptive phases in shots that eventually disrupt. Comparing the same methodology on different devices is crucial in order to have information on the portability of the developed algorithm and the possible extrapolation to ITER. Therefore, we use data from two very different tokamaks, DIII-D and Alcator C-Mod. We then focus on a subset of disruption predictors, most of which are dimensionless and/or machine-independent parameters, coming from both plasma diagnostics and equilibrium reconstructions,more » such as the normalized plasma internal inductance ℓ and the n = 1 mode amplitude normalized to the toroidal magnetic field. Using such dimensionless indicators facilitates a more direct comparison between DIII-D and C-Mod. We then choose a shallow Machine Learning technique, called Random Forests, to explore the databases available for the two devices. We show results from the classification task, where we introduce a time dependency through the definition of class labels on the basis of the elapsed time before the disruption (i.e. ‘far from a disruption’ and ‘close to a disruption’). The performances of the different Random Forest classifiers are discussed in terms of several metrics, by showing the number of successfully detected samples, as well as the misclassifications. The overall model accuracies are above 97% when identifying a ‘far from disruption’ and a ‘disruptive’ phase for disrupted discharges. Nevertheless, the Forests are intrinsically different in their capability of predicting a disruptive behavior, with C-Mod predictions comparable to random guesses. Indeed, we show that C-Mod recall index, i.e. the sensitivity to a disruptive behavior, is as low as 0.47, while DIII-D recall is ~0.72. The portability of the developed algorithm is also tested across the two devices, by using DIII-D data for training the forests and C-Mod for testing and vice versa.« less

Disruption prediction investigations using Machine Learning tools on DIII-D and Alcator C-Mod

DOE PAGES

Rea, C.; Granetz, R. S.; Montes, K.; ...

2018-06-18

Using data-driven methodology, we exploit the time series of relevant plasma parameters for a large set of disrupted and non-disrupted discharges to develop a classification algorithm for detecting disruptive phases in shots that eventually disrupt. Comparing the same methodology on different devices is crucial in order to have information on the portability of the developed algorithm and the possible extrapolation to ITER. Therefore, we use data from two very different tokamaks, DIII-D and Alcator C-Mod. We then focus on a subset of disruption predictors, most of which are dimensionless and/or machine-independent parameters, coming from both plasma diagnostics and equilibrium reconstructions,more » such as the normalized plasma internal inductance ℓ and the n = 1 mode amplitude normalized to the toroidal magnetic field. Using such dimensionless indicators facilitates a more direct comparison between DIII-D and C-Mod. We then choose a shallow Machine Learning technique, called Random Forests, to explore the databases available for the two devices. We show results from the classification task, where we introduce a time dependency through the definition of class labels on the basis of the elapsed time before the disruption (i.e. ‘far from a disruption’ and ‘close to a disruption’). The performances of the different Random Forest classifiers are discussed in terms of several metrics, by showing the number of successfully detected samples, as well as the misclassifications. The overall model accuracies are above 97% when identifying a ‘far from disruption’ and a ‘disruptive’ phase for disrupted discharges. Nevertheless, the Forests are intrinsically different in their capability of predicting a disruptive behavior, with C-Mod predictions comparable to random guesses. Indeed, we show that C-Mod recall index, i.e. the sensitivity to a disruptive behavior, is as low as 0.47, while DIII-D recall is ~0.72. The portability of the developed algorithm is also tested across the two devices, by using DIII-D data for training the forests and C-Mod for testing and vice versa.« less

Relationship Between Locked Modes and Disruptions in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Sweeney, Ryan

This thesis is organized into three body chapters: (1) the first use of naturally rotating tearing modes to diagnose intrinsic error fields is presented with experimental results from the EXTRAP T2R reversed field pinch, (2) a large scale study of locked modes (LMs) with rotating precursors in the DIII-D tokamak is reported, and (3) an in depth study of LM induced thermal collapses on a few DIII-D discharges is presented. The amplitude of naturally rotating tearing modes (TMs) in EXTRAP T2R is modulated in the presence of a resonant field (given by the superposition of the resonant intrinsic error field, and, possibly, an applied, resonant magnetic perturbation (RMP)). By scanning the amplitude and phase of the RMP and observing the phase-dependent amplitude modulation of the resonant, naturally rotating TM, the corresponding resonant error field is diagnosed. A rotating TM can decelerate and lock in the laboratory frame, under the effect of an electromagnetic torque due to eddy currents induced in the wall. These locked modes often lead to a disruption, where energy and particles are lost from the equilibrium configuration on a timescale of a few to tens of milliseconds in the DIII-D tokamak. In fusion reactors, disruptions pose a problem for the longevity of the reactor. Thus, learning to predict and avoid them is important. A database was developed consisting of ˜ 2000 DIII-D discharges exhibiting TMs that lock. The database was used to study the evolution, the nonlinear effects on equilibria, and the disruptivity of locked and quasi-stationary modes with poloidal and toroidal mode numbers m = 2 and n = 1 at DIII-D. The analysis of 22,500 discharges shows that more than 18% of disruptions present signs of locked or quasi-stationary modes with rotating precursors. A parameter formulated by the plasma internal inductance li divided by the safety factor at 95% of the toroidal flux, q95, is found to exhibit predictive capability over whether a locked mode will cause a disruption or not, and does so up to hundreds of milliseconds before the disruption. Within 20 ms of the disruption, the shortest distance between the island separatrix and the unperturbed last closed flux surface, referred to as dedge, performs comparably to l i/q95 in its ability to discriminate disruptive locked modes, and it also correlates well with the duration of the locked mode. On average, and within errors, the n=1 perturbed field grows exponentially in the final 50 ms before a disruption, however, the island width cannot discern whether a LM will disrupt or not up to 20 ms before the disruption. A few discharges are selected to analyze the evolution of the electron temperature profile in the presence of multiple coexisting locked modes during partial and full thermal quenches. Partial thermal quenches are often an initial, distinct stage in the full thermal quench caused by radiation, conduction, or convection losses. Here we explore the fundamental mechanism that causes the partial quench. Near the onset of partial thermal quenches, locked islands are observed to align in a unique way, or island widths are observed to grow above a threshold. Energy analysis on one discharge suggests that about half of the energy is lost in the divertor region. In discharges with minimum values of the safety factor above ˜1.2, and with current profiles expected to be classically stable, locked modes are observed to self-stabilize by inducing a full thermal quench, possibly by double tearing modes that remove the pressure gradient across the island, thus removing the neoclassical drive.

Predictions of the near edge transport shortfall in DIII-D L-mode plasmas using the trapped gyro-Landau-fluid model [Predictions of the near edge transport shortfall in DIII-D L-mode plasmas using the TGLF model

DOE PAGES

Kinsey, Jon E.; Staebler, Gary M.; Candy, Jefferey M.; ...

2015-01-14

Previous studies of DIII-D L-mode plasmas have shown that a transport shortfall exists in that our current models of turbulent transport can significantly underestimate the energy transport in the near edge region. In this paper, the Trapped Gyro-Landau-Fluid (TGLF) drift wave transport model is used to simulate the near edge transport in a DIII-D L-mode experiment designed to explore the impact of varying the safety factor on the shortfall. We find that the shortfall systematically increases with increasing safety factor and is more pronounced for the electrons than for the ions. Within the shortfall dataset, a single high current casemore » has been found where no transport shortfall is predicted. Reduced neutral beam injection power has been identified as the key parameter separating this discharge from other discharges exhibiting a shortfall. Further analysis shows that the energy transport in the L-mode near edge region is not stiff according to TGLF. Unlike the H-mode core region, the predicted temperature profiles are relatively more responsive to changes in auxiliary heating power. In testing the fidelity of TGLF for the near edge region, we find that a recalibration of the collision model is warranted. A recalibration improves agreement between TGLF and nonlinear gyrokinetic simulations performed using the GYRO code with electron-ion collisions. As a result, the recalibration only slightly impacts the predicted shortfall.« less

Compressional Alfven Eigenmode Similarity Study

NASA Astrophysics Data System (ADS)

Heidbrink, W. W.; Fredrickson, E. D.; Gorelenkov, N. N.; Rhodes, T. L.

2004-11-01

NSTX and DIII-D are nearly ideal for Alfven eigenmode (AE) similarity experiments, having similar neutral beams, fast-ion to Alfven speed v_f/v_A, fast-ion pressure, and shape of the plasma, but with a factor of 2 difference in the major radius. Toroidicity-induced AE with ˜100 kHz frequencies were compared in an earlier study [1]; this paper focuses on higher frequency AE with f ˜ 1 MHz. Compressional AE (CAE) on NSTX have a polarization, dependence on the fast-ion distribution function, frequency scaling, and low-frequency limit that are qualitatively consistent with CAE theory [2]. Global AE (GAE) are also observed. On DIII-D, coherent modes in this frequency range are observed during low-field (0.6 T) similarity experiments. Experiments will compare the CAE stability limits on DIII-D with the NSTX stability limits, with the aim of determining if CAE will be excited by alphas in a reactor. Predicted differences in the frequency splitting Δ f between excited modes will also be used. \\vspace0.25em [1] W.W. Heidbrink, et al., Plasmas Phys. Control. Fusion 45, 983 (2003). [2] E.D. Fredrickson, et al., Princeton Plasma Physics Laboratory Report PPPL-3955 (2004).

Design upgrades to the DIII-D gamma ray imager

NASA Astrophysics Data System (ADS)

Lvovskiy, A.; Cooper, C. M.; Eidietis, N. W.; Pace, D.; Paz-Soldan, C.

2016-10-01

Generation of runaway electrons (RE) in tokamak disruptions can cause damage of plasma facing components. RE studies are necessary in order to provide a reliable mechanism of RE mitigation. For that task a gamma ray imager (GRI) has been developed for DIII-D. It measures the bremsstrahlung emission by RE providing information on RE energy spectrum and RE distribution across a poloidal cross-section. The GRI consists of a lead pinhole camera illuminating a 2D array of 30 BGO detectors placed in the DIII-D mid-plane. First results showed the successful measurements of RE energy spectra in the range 1 - 60 MeV with time resolution 100 μs. They have been obtained in the low-flux quiescent RE regime via pulse-high analysis. The measurements in the high gamma flux post-disruption RE regime showed strong signal saturation. Here we present GRI design upgrades towards signal attenuation and better detector shielding including Monte-Carlo Neutral Particle modeling of GRI irradiation, as well as improved calibration techniques and options to improve electronic noise rejection. Work supported by US DOE under DE-AC05-06OR23100 and DE-FC02-04ER54698.

Stability analysis of the high poloidal bet scenario on DIII-Dtowards operation athigher plasma current

NASA Astrophysics Data System (ADS)

Guo, W. F.; Gong, X. Z.; Huang, J.; Ren, Q. L.; Qian, J. P.; Ding, S. Y.; Pan, C. K.; Li, G. Q.; Xia, T. Y.; Garofalo, A. M.; Lao, L.; Hyatt, A.; Ferron, J.; Meneghini, O.; Liu, Y. Q.; McClenaghan, J.; Holcomb, C. T.

2017-10-01

The high poloidal beta scenario with plasma current IP 600 kA and large-radius internal transport barrier (ITB) on DIII-D is subject to n =1 MHD kink modes when the current profile becomes very broad at internal inductance values li 0.5-0.6. It is desirable to extend this scenario to higer plasma current ( 1 MA) for highernormalized fusionperformance. However, higher current at constant normalized beta, ?N 3, would reducethe poloidal bet, ?P, below the threshold for ITB sustainment, observed at ?P 1.9. Thus, to avoid loss of the IT, ?N?? must be increased together with IP while avoiding the kink instability. MHD analysis is presented that explains possible paths to high ?N stability limit for the kink mode in tis scenario. Work supported by National Magnetic Confinement Fusion Program of Chin under 2015GB110001 and 2015GB102000 - National Natural Science Foundation of China under Grant No. 1147521 and by US DOE under DE-FC02-04ER54698.

Comparison of Sheath Power Transmission Factor for Neutral Beam Injection and Electron Cyclotron Heated Discharges in DIII-D

NASA Astrophysics Data System (ADS)

Donovan, D. C.; Buchenauer, D. A.; Watkins, J. G.; Leonard, A. W.; Lasnier, C. J.; Stangeby, P. C.

2011-10-01

The sheath power transmission factor (SPTF) is examined in DIII-D with a new IR camera, a more thermally robust Langmuir probe array, fast thermocouples, and a unique probe configuration on the Divertor Materials Evaluation System (DiMES). Past data collected from the fixed Langmuir Probes and Infrared Camera on DIII-D have indicated a SPTF near 1 at the strike point. Theory indicates that the SPTF should be approximately 7 and cannot be less than 5. SPTF values are calculated using independent measurements from the IR camera and fast thermocouples. Experiments have been performed with varying levels of electron cyclotron heating and neutral beam power. The ECH power does not involve fast ions, so the SPTF can be calculated and compared to previous experiments to determine the extent to which fast ions may be influencing the SPTF measurements, and potentially offer insight into the disagreement with the theory. Work supported in part by US DOE under DE-AC04-94AL85000, DE-FC02-04ER54698, and DE-AC52-07NA27344.

$$\\mathscr{H}_2$$ optimal control techniques for resistive wall mode feedback in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clement, Mitchell; Hanson, Jeremy; Bialek, Jim

DIII-D experiments show that a new, advanced algorithm improves resistive wall mode (RWM) stability control in high performance discharges using external coils. DIII-D can excite strong, locked or nearly locked external kink modes whose rotation frequencies and growth rates are on the order of the magnetic ux di usion time of the vacuum vessel wall. The VALEN RWM model has been used to gauge the e ectiveness of RWM control algorithms in tokamaks. Simulations and experiments have shown that modern control techniques like Linear Quadratic Gaussian (LQG) control will perform better, using 77% less current, than classical techniques when usingmore » control coils external to DIII-D's vacuum vessel. Experiments were conducted to develop control of a rotating n = 1 perturbation using an LQG controller derived from VALEN and external coils. Feedback using this LQG algorithm outperformed a proportional gain only controller in these perturbation experiments over a range of frequencies. Results from high N experiments also show that advanced feedback techniques using external control coils may be as e ective as internal control coil feedback using classical control techniques.« less

Tokamak Operation with Safety Factor q 95 < 2 via Control of MHD Stability

DOE PAGES

Piovesan, Paolo; Hanson, Jeremy M.; Martin, Piero; ...

2014-07-24

Magnetic feedback control of the resistive-wall mode has enabled DIII-D to access stable operation at safety factor q95 = 1:9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at given toroidal magnetic eld. In tokamaks with a divertor, the limit occurs at q95 = 2, as con rmed in DIII-D. Since the energy con cement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a wholemore » new high-current regime not accessible before. This result brings signi cant possible bene ts in terms of fusion performance, but it also extends resistive wall mode physics and its control to conditions never explored before. In present experiments, q95 < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.« less

Exploration of the Super H-mode regime on DIII-D and potential advantages for burning plasma devices

DOE PAGES

Solomon, W. M.; Snyder, P. B.; Bortolon, A.; ...

2016-03-25

In a new high pedestal regime ("Super H-mode") we predicted and accessed DIII-D. Super H-mode was first achieved on DIII-D using a quiescent H-mode edge, enabling a smooth trajectory through pedestal parameter space. By exploiting Super H-mode, it has been possible to access high pedestal pressures at high normalized densities. And while elimination of Edge localized modes (ELMs) is beneficial for Super H-mode, it may not be a requirement, as recent experiments have maintained high pedestals with ELMs triggered by lithium granule injection. Simulations using TGLF for core transport and the EPED model for the pedestal find that ITER canmore » benefit from the improved performance associated with Super H-mode, with increased values of fusion power and gain possible. In similar studies demonstrate that the Super H-mode pedestal can be advantageous for a steady-state power plant, by providing a path to increasing the bootstrap current while simultaneously reducing the demands on the core physics performance.« less

Ion beam analysis of C-13 and deuterium deposition in DIII-D and their removal by in-situ oxygen baking

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wampler, W. R.; Allen, S. L.; Brooks, N. H.

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. C-13 methane was injected toroidally symmetrically near the secondary separatrix into ELMy H-mode deuterium plasmas. The resulting deposition of C-13 was determined by nuclear reaction analysis. These results show that very little of the injected C-13 was deposited at the primary separatrix, whereas a large fraction of injected C-13 was deposited close to the point of injection near the secondary separatrix. Six of the tiles weremore » put back into DIII-D, where they were baked at 350-360 degrees C for 2 h at similar to 1 kPa in a 20% O-2/80% He gas mixture. Subsequent ion beam analysis of these tiles showed that about 21% of the C-13 and 54% of the deuterium were removed by the bake.« less

Inductive flux usage and its optimization in tokamak operation

DOE PAGES

Luce, Timothy C.; Humphreys, David A.; Jackson, Gary L.; ...

2014-07-30

The energy flow from the poloidal field coils of a tokamak to the electromagnetic and kinetic stored energy of the plasma are considered in the context of optimizing the operation of ITER. The goal is to optimize the flux usage in order to allow the longest possible burn in ITER at the desired conditions to meet the physics objectives (500 MW fusion power with energy gain of 10). A mathematical formulation of the energy flow is derived and applied to experiments in the DIII-D tokamak that simulate the ITER design shape and relevant normalized current and pressure. The rate ofmore » rise of the plasma current was varied, and the fastest stable current rise is found to be the optimum for flux usage in DIII-D. A method to project the results to ITER is formulated. The constraints of the ITER poloidal field coil set yield an optimum at ramp rates slower than the maximum stable rate for plasmas similar to the DIII-D plasmas. Finally, experiments in present-day tokamaks for further optimization of the current rise and validation of the projections are suggested.« less

$$\\mathscr{H}_2$$ optimal control techniques for resistive wall mode feedback in tokamaks

DOE PAGES

Clement, Mitchell; Hanson, Jeremy; Bialek, Jim; ...

2018-02-28

DIII-D experiments show that a new, advanced algorithm improves resistive wall mode (RWM) stability control in high performance discharges using external coils. DIII-D can excite strong, locked or nearly locked external kink modes whose rotation frequencies and growth rates are on the order of the magnetic ux di usion time of the vacuum vessel wall. The VALEN RWM model has been used to gauge the e ectiveness of RWM control algorithms in tokamaks. Simulations and experiments have shown that modern control techniques like Linear Quadratic Gaussian (LQG) control will perform better, using 77% less current, than classical techniques when usingmore » control coils external to DIII-D's vacuum vessel. Experiments were conducted to develop control of a rotating n = 1 perturbation using an LQG controller derived from VALEN and external coils. Feedback using this LQG algorithm outperformed a proportional gain only controller in these perturbation experiments over a range of frequencies. Results from high N experiments also show that advanced feedback techniques using external control coils may be as e ective as internal control coil feedback using classical control techniques.« less

Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

DOE PAGES

Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; ...

2016-06-20

Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of β p and β N despite strong ITBs. Good confinement has been achieved with reduced toroidal rotation. These high β p plasmas challenge the energy transport understanding, especiallymore » in the electron energy channel. A new turbulent transport model, named 2 TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. Finally, more investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.« less

Investigation of runaway electron dissipation in DIII-D using a gamma ray imager

NASA Astrophysics Data System (ADS)

Lvovskiy, A.; Paz-Soldan, C.; Eidietis, N.; Pace, D.; Taussig, D.

2017-10-01

We report the findings of a novel gamma ray imager (GRI) to study runaway electron (RE) dissipation in the quiescent regime on the DIII-D tokamak. The GRI measures the bremsstrahlung emission by RE providing information on RE energy spectrum and distribution across a poloidal cross-section. It consists of a lead pinhole camera illuminating a matrix of BGO detectors placed in the DIII-D mid-plane. The number of detectors was recently doubled to provide better spatial resolution and additional detector shielding was implemented to reduce un-collimated gamma flux and increase single-to-noise ratio. Under varying loop voltage, toroidal magnetic field and plasma density, a non-monotonic RE distribution function has been revealed as a result of the interplay between electric field, synchrotron radiation and collisional damping. A fraction of the high-energy RE population grows forming a bump at the RE distribution function while synchrotron radiation decreases. A possible destabilizing effect of Parail-Pogutse instability on the RE population will be also discussed. Work supported by the US DOE under DE-FC02-04ER54698.

Measurements of the internal magnetic field on DIII-D using intensity and spacing of the motional Stark multiplet.

PubMed

Pablant, N A; Burrell, K H; Groebner, R J; Kaplan, D H; Holcomb, C T

2008-10-01

We describe a version of a motional Stark effect (MSE) diagnostic based on the relative line intensities and spacing of Stark split D(alpha) emission from the neutral beams. This system, named B-Stark, has been recently installed on the DIII-D tokamak. To find the magnetic pitch angle, we use the ratio of the intensities of the pi(3) and sigma(1) lines. These lines originate from the same upper level and so are not dependent on the level populations. In future devices, such as ITER, this technique may have advantages over diagnostics based on MSE polarimetry. We have done an optimization of the viewing direction for the available ports on DIII-D to choose the installation location. With this placement, we have a near optimal viewing angle of 59.6 degrees from the vertical direction. All hardware has been installed for one chord, and we have been routinely taking data since January 2007. We fit the spectra using a simple Stark model in which the upper level populations of the D(alpha) transition are treated as free variables. The magnitude and direction of the magnetic field obtained using this diagnostic technique compare well with measurements from MSE polarimetry and EFIT.

Parametric Instabilities During High Power Helicon Wave Injection on DIII-D

NASA Astrophysics Data System (ADS)

Porkolab, M.; Pinsker, R. I.

2017-10-01

High power helicon (whistler) waves at a frequency of 0.47 GHz are being considered for efficient off-axis current generation in high performance DIII-D plasmas and in K-Star [3]. The need for deploying helicon waves for current profile control has been noted in previous publications since penetration to the core of reactor grade plasmas is easier than with lower hybrid slow waves (LHCD) which suffer from accessibility limitations and strong electron Landau absorption in fusion grade high temperature plasmas. In this work we show that under typical experimental conditions in present day tokamaks with 1 MW of RF power coupled per antenna, the associated perpendicular electric fields of the order of 40 kV/m can drive strong parametric decay instabilities near the lower hybrid layer. The EXB and polarization drift velocities which are the dominant driver of the PDI can be comparable to the speed of sound in the outer plasma layers, a key measure of driving PDI instabilities. Here we calculate growth rates and convective thresholds for PDIs, and we find that decay waves into hot ion lower hybrid waves and ion cyclotron quasi modes dominate in the vicinity of the lower hybrid layer, possibly leading to pump depletion. Such instabilities in future reactor grade high temperature plasmas are less likely.

Multi-scale transport in the DIII-D ITER baseline scenario with direct electron heating and projection to ITER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grierson, B. A.; Staebler, G. M.; Solomon, W. M.

Multi-scale fluctuations measured by turbulence diagnostics spanning long and short wavelength spatial scales impact energy confinement and the scale-lengths of plasma kinetic profiles in the DIII-D ITER baseline scenario with direct electron heating. Contrasting discharge phases with ECH + neutral beam injection (NBI) and NBI only at similar rotation reveal higher energy confinement and lower fluctuations when only NBI heating is used. Modeling of the core transport with TGYRO using the TGLF turbulent transport model and NEO neoclassical transport reproduces the experimental profile changes upon application of direct electron heating and indicates that multi-scale transport mechanisms are responsible for changesmore » in the temperature and density profiles. Intermediate and high-k fluctuations appear responsible for the enhanced electron thermal flux, and intermediate-k electron modes produce an inward particle pinch that increases the inverse density scale length. Projection to ITER is performed with TGLF and indicates a density profile that has a finite scale length due to intermediate-k electron modes at low collisionality and increases the fusion gain. Finally, for a range of E×B shear, the dominant mechanism that increases fusion performance is suppression of outward low-k particle flux and increased density peaking.« less

Multi-scale transport in the DIII-D ITER baseline scenario with direct electron heating and projection to ITER

DOE PAGES

Grierson, B. A.; Staebler, G. M.; Solomon, W. M.; ...

2018-02-01

Multi-scale fluctuations measured by turbulence diagnostics spanning long and short wavelength spatial scales impact energy confinement and the scale-lengths of plasma kinetic profiles in the DIII-D ITER baseline scenario with direct electron heating. Contrasting discharge phases with ECH + neutral beam injection (NBI) and NBI only at similar rotation reveal higher energy confinement and lower fluctuations when only NBI heating is used. Modeling of the core transport with TGYRO using the TGLF turbulent transport model and NEO neoclassical transport reproduces the experimental profile changes upon application of direct electron heating and indicates that multi-scale transport mechanisms are responsible for changesmore » in the temperature and density profiles. Intermediate and high-k fluctuations appear responsible for the enhanced electron thermal flux, and intermediate-k electron modes produce an inward particle pinch that increases the inverse density scale length. Projection to ITER is performed with TGLF and indicates a density profile that has a finite scale length due to intermediate-k electron modes at low collisionality and increases the fusion gain. Finally, for a range of E×B shear, the dominant mechanism that increases fusion performance is suppression of outward low-k particle flux and increased density peaking.« less

Computational Study of Poloidal Angular Momentum Transport in DIII-D

NASA Astrophysics Data System (ADS)

Pankin, Alexei; Kruger, Scott; Kritz, Arnold; Rafiq, Tariq; Weiland, Jan

2013-10-01

The new Multi-Mode Model, MMM8.1, includes the capability to predict the anomalous poloidal momentum diffusivity [T. Rafiq et al., Phys. Plasmas 20, 032506 (2013)]. It is important to consider the effect of this diffusivity on the poloidal rotation of tokamak plasmas since some experimental observations suggest that neoclassical effects are not always sufficient to explain the observed poloidal rotation [B.A. Grierson et al., Phys. Plasmas 19, 056107 (2012)]. One of the objectives of this research is to determine if the anomalous contribution to the poloidal rotation can be significant in the regions of internal transport barriers (ITBs). In this study, the MMM8.1 model is used to compute the poloidal momentum diffusivity for a range of plasma parameters that correspond to the parameters that occur in DIII-D discharges. The parameters that are considered include the temperature and density gradients, and magnetic shear. The role of anomalous poloidal transport in the possible poloidal spin up in the ITB regions is discussed. Progress in the implementation of poloidal transport equations in the ASTRA transport code is reported and initial predictive simulation results for the poloidal rotation profiles are presented. This research is partially support by the DOE Grants DE-SC0006629 and DE-FG02-92ER54141.

Multi-scale transport in the DIII-D ITER baseline scenario with direct electron heating and projection to ITER

NASA Astrophysics Data System (ADS)

Grierson, B. A.; Staebler, G. M.; Solomon, W. M.; McKee, G. R.; Holland, C.; Austin, M.; Marinoni, A.; Schmitz, L.; Pinsker, R. I.; DIII-D Team

2018-02-01

Multi-scale fluctuations measured by turbulence diagnostics spanning long and short wavelength spatial scales impact energy confinement and the scale-lengths of plasma kinetic profiles in the DIII-D ITER baseline scenario with direct electron heating. Contrasting discharge phases with ECH + neutral beam injection (NBI) and NBI only at similar rotation reveal higher energy confinement and lower fluctuations when only NBI heating is used. Modeling of the core transport with TGYRO using the TGLF turbulent transport model and NEO neoclassical transport reproduces the experimental profile changes upon application of direct electron heating and indicates that multi-scale transport mechanisms are responsible for changes in the temperature and density profiles. Intermediate and high-k fluctuations appear responsible for the enhanced electron thermal flux, and intermediate-k electron modes produce an inward particle pinch that increases the inverse density scale length. Projection to ITER is performed with TGLF and indicates a density profile that has a finite scale length due to intermediate-k electron modes at low collisionality and increases the fusion gain. For a range of E × B shear, the dominant mechanism that increases fusion performance is suppression of outward low-k particle flux and increased density peaking.

Modeling Giant Sawtooth Modes in DIII-D using the NIMROD code

NASA Astrophysics Data System (ADS)

Kruger, Scott; Jenkins, Thomas; Held, Eric; King, Jacob; NIMROD Team

2014-10-01

Ongoing efforts to model giant sawtooth cycles in DIII-D shot 96043 using NIMROD are summarized. In this discharge, an energetic ion population induced by RF heating modifies the sawtooth stability boundary, supplanting the conventional sawtooth cycle with longer-period giant sawtooth oscillations of much larger amplitude. NIMROD has the unique capability of being able to use both continuum kinetic and particle-in-cell numerical schemes to model the RF-induced hot-particle distribution effects on the sawtooth stability. This capability is used to numerically investigate the role played by the form of the energetic particle distribution, including a possible high-energy tail drawn out by the RF, to study the sawtooth threshold and subsequent nonlinear evolution. Equilibrium reconstructions from the experimental data are used to enable these detailed validation studies. Effects of other parameters on the sawtooth behavior (such as the plasma Lundquist number and hot-particle β-fraction) are also considered. Ultimately, we hope to assess the degree to which NIMROD's extended MHD model correctly simulates the observed linear onset and nonlinear behavior of the giant sawtooth, and to establish its reliability as a predictive modeling tool for these modes. This work was initiated by the late Dr. Dalton Schnack. Equilibria were provided by Dr. A. Turnbull of General Atomics.

An evaluation of the susceptibility of V-4Cr-4Ti to stress corrosion cracking in room temperature DIII-D water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kurtz, R.J.; Jones, R.H.; Johnson, W.R.

1997-04-01

Two fatigue precracked compact tension (CT) specimens of V-4Cr-4Ti were statically loaded to a stress intensity factor of about 30 MPa{radical}m in room temperature DIII-D water. The first specimen was tested for a period of about 30 days and the second specimen for about 54 days. At the conclusion of each test the specimens were fractured, and the fracture surfaces examined with a scanning electron microscope (SEM) to determine if SCC had occurred. No SCC was found in either test specimen.

Toroidal rotation in neutral beam heated discharges in DIII-D

NASA Astrophysics Data System (ADS)

de Grassie, J. S.; Baker, D. R.; Burrell, K. H.; Gohil, P.; Greenfield, C. M.; Groebner, R. J.; Thomas, D. M.

2003-02-01

It is known that the toroidal angular momentum and the ion thermal energy are correlated in tokamak discharges heated by neutral beam injection. Here, data from ten years of measurements on DIII-D are considered, for representative discharges from all types and all conditions. The ratio of simple replacement times for momentum and energy is found to order this correlation indicating that these times are approximately equal, across the minor radius. Representative discharges of several types are discussed in more detail, as well as transport analysis results for the momentum and thermal ion diffusivities.

Modulation of Core Turbulent Density Fluctuations by Large-Scale Neoclassical Tearing Mode Islands in the DIII-D Tokamak

DOE PAGES

Bardóczi, L.; Rhodes, T. L.; Carter, T. A.; ...

2016-05-26

We report the first observation of localized modulation of turbulent density uctuations en (via Beam Emission Spectroscopy) by neoclassical tearing modes (NTMs) in the core of the DIII-D tokamak. NTMs are important as they often lead to severe degradation of plasma confinement and disruptions in high-confinement fusion experiments. Magnetic islands associated with NTMs significantly modify the profiles and turbulence drives. In this experiment n was found to be modulated by 14% across the island. Gyrokinetic simulations suggest that en could be dominantly driven by the ion temperature gradient (ITG) instability.

Changes in particle transport as a result of resonant magnetic perturbations in DIII-D (vol 19, 056503, 2012)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mordijck, S.; Doyle, E. J.; McKee, G. R.

2012-01-01

Publisher s Note: Changes in particle transport as a result of resonant magnetic perturbations in DIII-D [Phys. Plasmas 19, 056503 (2012)]a) In the Invited Papers from the 53rd Annual Meeting of the APS Division of Plasma Physics of the May 2012 issue of the journal, this article was originally published online and in print in the incorrect section; it was published within Ionospheric, Solar-System and Astrophysical Plasmas (Sec. 65) instead of Magnetically Confined Plasmas, Heating, Confinement (Sec. 61). AIP apologizes for this error. a)

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Christopher; Nygren, R. E.; Chrobak, C P.

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels ofmore » isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.« less

3-D plasma boundary and plasma wall interaction research at UW-Madison

NASA Astrophysics Data System (ADS)

Schmitz, Oliver; Akerson, Adrian; Bader, Aaron; Barbui, Tullio; Effenberg, Florian; Flesch, Kurt; Frerichs, Heinke; Green, Jonathan; Hinson, Edward; Kremeyer, Thierry; Norval, Ryan; Stephey, Laurie; Waters, Ian; Winters, Victoria

2016-10-01

The necessity of considering 3-D effects on the plasma boundary and plasma wall interaction (PWI) in tokamaks, stellarators and reversed field pinches has been highlighted by abundant experimental and numerical results in the recent past. Prominent examples with 3-D boundary situations are numerous: ELM controlled H-modes by RMP fields in tokamaks, research on boundary plasmas and PWI in stellarators in general, quasi-helical states in RFPs, asymmetric fueling situations, and structural and wall elements which are not aligned with the magnetic guiding fields. A systematic approach is being taken at UW-Madison to establish a targeted experimental basis for identifying the most significant effects for plasma edge transport and resulting PWI in such 3-D plasma boundary situations. We deploy advanced 3-D modeling using the EMC3-EIRENE, ERO and MCI codes in combination with laboratory experiments at UW-Madison to investigate the relevance of 3-D effects in large scale devices with a concerted approach on DIII-D, NSTX-U, and Wendelstein 7-X. Highlights of experimental results from the on-site laboratory activities at UW-Madison and the large scale facilities are presented and interlinks will be discussed. This work was supported by US DOE DE-SC0013911, DE-SC00012315 and DE-SC00014210.

Control of Post-disruption Runaway Electron Beams in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Eidietis, N. W.

2011-10-01

Recent experiments on DIII-D have demonstrated real-time control of post-disruption runaway electron (RE) beams, presenting the possibility for slow, controlled dissipation of the beam energy. RE beams will present a greater challenge to ITER than present tokamaks due to ITER's high RE avalanche gain constant [Nucl.Fusion 37, 1355-62 (1997)] and the difficulty repairing potential damage to its first wall. In the rare event that disruption control and mitigation schemes fail to suppress RE generation, active control of the RE beam may be an important line of defense to prevent rapid, localized deposition of RE beam energy on the first wall. Initially, sustaining a RE beam plateau requires avoiding radial collapse of the beam into the inner wall during the first 1-2 wall penetration times following the current quench (CQ). This collapse is caused by attractive induced currents in the wall and a lack of radial equilibrium with slow vertical field coils. The collapse is avoided by slewing the inner PF coils to push the RE beam off the wall while reducing the outer PF coil currents. Beam survival through this phase requires sufficient RE plateau current (IRE) and power supply slew rates to re-establish equilibrium. Following that transient period, RE beam vertical position was dynamically controlled, and stabilization was maintained in an elongated (κ <= 1 . 8) DND configuration for up 250ms. Most controlled RE beams end in a rapid vertical displacement event (VDE), indicating that the profiles evolve even as the position is controlled. Experimental radial evolution and VDE onset are shown to be consistent with theoretical calculations of controllability boundaries. However, ohmic regulation of IRE has been shown to delay VDEs to the pre-programmed ramp-down time, indicating that steady-state control may be achievable. Supported by the US DOE under DE-FC02-04ER54698.

High performance advanced tokamak regimes in DIII-D for next-step experiments

NASA Astrophysics Data System (ADS)

Greenfield, C. M.; Murakami, M.; Ferron, J. R.; Wade, M. R.; Luce, T. C.; Petty, C. C.; Menard, J. E.; Petrie, T. W.; Allen, S. L.; Burrell, K. H.; Casper, T. A.; DeBoo, J. C.; Doyle, E. J.; Garofalo, A. M.; Gorelov, I. A.; Groebner, R. J.; Hobirk, J.; Hyatt, A. W.; Jayakumar, R. J.; Kessel, C. E.; La Haye, R. J.; Jackson, G. L.; Lohr, J.; Makowski, M. A.; Pinsker, R. I.; Politzer, P. A.; Prater, R.; Strait, E. J.; Taylor, T. S.; West, W. P.; DIII-D Team

2004-05-01

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts.

Fast-ion transport in qmin>2, high- β steady-state scenarios on DIII-D

DOE PAGES

Holcomb, C. T.; Heidbrink, W. W.; Ferron, J. R.; ...

2015-05-22

The results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-qminqmin confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing β N and the noninductive current drive. However, in scenarios with q min>2 that target the typical range of q 95= 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. Thismore » enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable β N. Conversely similar plasmas except with q min just above 1 have approximately classical fast-ion transport. Experiments that take q min>3 plasmas to higher β P with q 95= 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-q min scenario, the high β P cases have shorter slowing-down time and lower ∇β fast, and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, β N, and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q 95, high-q min plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes.« less

Fast-ion transport in q{sub min}>2, high-β steady-state scenarios on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holcomb, C. T.; Heidbrink, W. W.; Collins, C.

2015-05-15

Results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-q{sub min} confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing β{sub N} and the noninductive current drive. However, in scenarios with q{sub min}>2 that target the typical range of q{sub 95}= 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. Thismore » enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable β{sub N}. In contrast, similar plasmas except with q{sub min} just above 1 have approximately classical fast-ion transport. Experiments that take q{sub min}>3 plasmas to higher β{sub P} with q{sub 95}= 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-q{sub min} scenario, the high β{sub P} cases have shorter slowing-down time and lower ∇β{sub fast}, and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, β{sub N}, and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q{sub 95}, high-q{sub min} plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes.« less

3D-DIVIMP-HC modeling analysis of methane injection into DIII-D using the DiMES porous plug injector

NASA Astrophysics Data System (ADS)

Mu, Y.; McLean, A. G.; Elder, J. D.; Stangeby, P. C.; Bray, B. D.; Brooks, N. H.; Davis, J. W.; Fenstermacher, M. E.; Groth, M.; Lasnier, C. J.; Rudakov, D. L.; Watkins, J. G.; West, W. P.; Wong, C. P. C.

2009-06-01

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D, the porous plug injector (PPI), has been employed for in situ study of chemical erosion in the tokamak divertor environment by injection of CH 4 [A.G. McLean et al., these Proceedings]. A new interpretive code, 3D-DIVIMP-HC, has been developed and applied to the interpretation of the CH, CI, and CII emissions. Particular emphasis is placed on the interpretation of 2D filtered-camera (TV) pictures in CH, CI and CII light taken from a view essentially straight down on the PPI. The code replicates sufficient measurements to conclude that most of the basic elements of the controlling physics and chemistry have been identified and incorporated in the code-model.

Numerical simulation of runaway electrons: 3-D effects on synchrotron radiation and impurity-based runaway current dissipation

NASA Astrophysics Data System (ADS)

del-Castillo-Negrete, D.; Carbajal, L.; Spong, D.; Izzo, V.

2018-05-01

Numerical simulations of runaway electrons (REs) with a particular emphasis on orbit dependent effects in 3-D magnetic fields are presented. The simulations were performed using the recently developed Kinetic Orbit Runaway electron Code (KORC) that computes the full-orbit relativistic dynamics in prescribed electric and magnetic fields including radiation damping and collisions. The two main problems of interest are synchrotron radiation and impurity-based RE dissipation. Synchrotron radiation is studied in axisymmetric fields and in 3-D magnetic configurations exhibiting magnetic islands and stochasticity. For passing particles in axisymmetric fields, neglecting orbit effects might underestimate or overestimate the total radiation power depending on the direction of the radial shift of the drift orbits. For trapped particles, the spatial distribution of synchrotron radiation exhibits localized "hot" spots at the tips of the banana orbits. In general, the radiation power per particle for trapped particles is higher than the power emitted by passing particles. The spatial distribution of synchrotron radiation in stochastic magnetic fields, obtained using the MHD code NIMROD, is strongly influenced by the presence of magnetic islands. 3-D magnetic fields also introduce a toroidal dependence on the synchrotron spectra, and neglecting orbit effects underestimates the total radiation power. In the presence of magnetic islands, the radiation damping of trapped particles is larger than the radiation damping of passing particles. Results modeling synchrotron emission by RE in DIII-D quiescent plasmas are also presented. The computation uses EFIT reconstructed magnetic fields and RE energy distributions fitted to the experimental measurements. Qualitative agreement is observed between the numerical simulations and the experiments for simplified RE pitch angle distributions. However, it is noted that to achieve quantitative agreement, it is necessary to use pitch angle distributions that depart from simplified 2-D Fokker-Planck equilibria. Finally, using the guiding center orbit model (KORC-GC), a preliminary study of pellet mitigated discharges in DIII-D is presented. The dependence of RE energy decay and current dissipation on initial energy and ionization levels of neon impurities is studied. The computed decay rates are within the range of experimental observations.

Analysis of Island Formation Due to RMPs in D3D Plasmas Using SIESTA

NASA Astrophysics Data System (ADS)

Hirshman, Steven; Shafer, Morgan; Seal, Sudip; Canik, John

2015-11-01

By varying the initial helical perturbation amplitude of Resonant Magnetic Perturbations (RMPs) applied to a Doublet III-D (DIII-D) plasma, a variety of meta-stable equilibrium are scanned using the SIESTA MHD equilibrium code. It is found that increasing the perturbation strength at the dominant m =2 resonant surface leads to lower MHD energies and significant increases in the equilibrium island widths at the m =2 (and sidebands) surfaces. Island overlap eventually leads to stochastic magnetic fields which correlate well with the experimentally inferred field line structure. The magnitude and spatial phase (around associated rational surfaces) of resonant (shielding) components of the parallel current is shown to be correlated with the magnetic island topology. Work supported by U.S. DOE under Contract DE-AC05-00OR22725 with UT-Battelle, LLC.

Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cooper, C. M., E-mail: coopercm@fusion.gat.com; Pace, D. C.; Paz-Soldan, C.

2016-11-15

A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5–100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead “pinhole camera” mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses permore » second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.« less

Modeling impurity production and transport during ELM transients in a DIII-D lower single null configuration.

NASA Astrophysics Data System (ADS)

Hogan, J.; Fenstermacher, M.; Groth, M.; West, P.; Coster, D.; Thomas, P.

2003-10-01

Better understanding of carbon production and eventual pathways is an important need for ITER. ELM events can provide a significant carbon source, and small scale experiments predict significant dependence of C production rates on incident deuterium flux and surface temperature, quantities which change significantly during an ELM event. Thus, development of better quantitative models has been hampered by lack of sufficient time resolution during ELMs. Recent progress on DIII-D has significantly improved the spectroscopic resolution [1, 2]. Measured CIII evolution during low- and high-density DIII-D LSN ELMy H-modes (type I and type I-III, respectively) has been compared with modeling using the solps5.0/Eirene99 coupled edge code, and the CASTEM- 2000 3-D, time dependent thermal analysis code. The latter provides time-resolved absolute surface temperature distributions for the cases described in [2]. Comparison with observations using the Roth et al annealing model for chemical sputtering finds qualitative agreement. However, the transition in ELM type/frequency as density increases is the most important factor, and this is an input to the calculation given the present state of first principles ELM models. [1] M Fenstermacher et al EPS2003 [2] M Groth et al J Nucl Mater 2003

Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

NASA Astrophysics Data System (ADS)

Cooper, C. M.; Pace, D. C.; Paz-Soldan, C.; Commaux, N.; Eidietis, N. W.; Hollmann, E. M.; Shiraki, D.

2016-11-01

A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5-100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead "pinhole camera" mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited).

PubMed

Cooper, C M; Pace, D C; Paz-Soldan, C; Commaux, N; Eidietis, N W; Hollmann, E M; Shiraki, D

2016-11-01

A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5-100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead "pinhole camera" mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

DOE PAGES

Cooper, C. M.; Pace, D. C.; Paz-Soldan, C.; ...

2016-08-30

A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5-100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead "pinhole camera" mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20,000 pulses permore » second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Furthermore, magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.« less

Measurements of the internal magnetic field using the B-Stark motional Stark effect diagnostic on DIII-D (inivited).

PubMed

Pablant, N A; Burrell, K H; Groebner, R J; Holcomb, C T; Kaplan, D H

2010-10-01

Results are presented from the B-Stark diagnostic installed on the DIII-D tokamak. This diagnostic provides measurements of the magnitude and direction of the internal magnetic field. The B-Stark system is a version of a motional Stark effect (MSE) diagnostic based on the relative line intensities and spacing of the Stark split D(α) emission from injected neutral beams. This technique may have advantages over MSE polarimetry based diagnostics in future devices, such as the ITER. The B-Stark diagnostic technique and calibration procedures are discussed. The system is shown to provide accurate measurements of B(θ)/B(T) and ∣B∣ over a range of plasma conditions. Measurements have been made with toroidal fields in the range of 1.2-2.1 T, plasma currents in the range 0.5-2.0 MA, densities between 1.7 and 9.0×10(19) m(-3), and neutral beam voltages between 50 and 81 keV. The viewing direction and polarization dependent transmission properties of the collection optics are found using an in situ beam into gas calibration. These results are compared to values found from plasma equilibrium reconstructions and the MSE polarimetry system on DIII-D.

Linear ideal MHD predictions for n = 2 non-axisymmetric magnetic perturbations on DIII-D

DOE PAGES

Haskey, Shaun R.; Lanctot, Matthew J.; Liu, Y. Q.; ...

2014-02-05

Here, an extensive examination of the plasma response to dominantly n = 2 non-axisymmetric magnetic perturbations (MPs) on the DIII-D tokamak shows the potential to control 3D field interactions by varying the poloidal spectrum of the radial magnetic field. The plasma response is calculated as a function of the applied magnetic field structure and plasma parameters, using the linear magnetohydrodynamic code MARS-F. The ideal, single fluid plasma response is decomposed into two main components: a local pitch-resonant response occurring at rational magnetic flux surfaces, and a global kink response. The efficiency with which the field couples to the total plasmamore » response is determined by the safety factor and the structure of the applied field. In many cases, control of the applied field has a more significant effect than control of plasma parameters, which is of particular interest since it can be modified at will throughout a shot to achieve a desired effect. The presence of toroidal harmonics, other than the dominant n = 2 component, is examined revealing a significant n = 4 component in the perturbations applied by the DIII-D MP coils; however, modeling shows the plasma responses to n = 4 perturbations are substantially smaller than the dominant n = 2 responses in most situations.« less

Transport Simulations of DIII-D Discharges with Impurity Injection

NASA Astrophysics Data System (ADS)

Mandrekas, J.; Stacey, W. M.; Murakami, M.

2001-10-01

Several recent DIII-D discharges with external impurity injection into L-mode plasmas are analyzed with a coupled main plasma and multi-charge state 1frac 12-D impurity transport code. These discharges exhibit various degrees of confinement improvement, which has been attributed to the synergistic effects of impurity induced enhancement of the E×B shearing rate and reduction of the drift wave turbulence growth rate (M. Murakami, et. al., Nucl. Fusion 41) (2001) 317.. Impurity transport is described by empirical and neoclassical transport models. Both the standard neoclassical theory as well as an enhanced theory which takes into account the effects of external momentum input and radial momentum transport (W.M. Stacey, Phys. Plasmas 8) (2001) 158. have been considered.

Lack of dependence on resonant error field of locked mode island size in ohmic plasmas in DIII-D

DOE PAGES

Haye, R. J. La; Paz-Soldan, C.; Strait, E. J.

2015-01-23

DIII-D experiments show that fully penetrated resonant n=1 error field locked modes in Ohmic plasmas with safety factor q 95≳3 grow to similar large disruptive size, independent of resonant error field correction. Relatively small resonant (m/n=2/1) static error fields are shielded in Ohmic plasmas by the natural rotation at the electron diamagnetic drift frequency. However, the drag from error fields can lower rotation such that a bifurcation results, from nearly complete shielding to full penetration, i.e., to a driven locked mode island that can induce disruption.

Core tungsten radiation diagnostic calibration by small shell pellet injection in the DIII-D tokamak

DOE PAGES

Hollmann, Eric M.; Commaux, Nicolas; Shiraki, Daisuke; ...

2017-10-04

Injection of small (OD = 0.8 mm) plastic pellets carrying embedded smaller (10 μg) tungsten grains is used to check calibrations of core tungsten line radiation diagnostics in support of the 2016 tungsten rings campaign in the DIII-D tokamak. The total (1 eV – 10 keV) and soft x-ray (1 keV – 10 keV) brightnesses we observed were found to be reasonably well (< factor 2) predicted using existing calibration factors and rate calculations. Individual core (EUV/SXR) tungsten line brightnesses appear to be somewhat less reliable (factor 2-4) for prediction of core tungsten concentration.

Evaluation of Particle Pinch and Diffusion Coefficients in the Edge Pedestal of DIII-D H-mode Discharges

NASA Astrophysics Data System (ADS)

Stacey, W. M.; Groebner, R. J.

2009-11-01

Momentum balance requires that the radial particle flux satisfy a pinch-diffusion relationship. The pinch can be evaluated in terms of measurable quantities (rotation velocities, Er, etc.) by the use of momentum and particle balance [1,2], the radial particle flux can be determined by momentum balance, and then the diffusion coefficient can be evaluated from the pinch diffusion relation using the measured density gradient. Applications to several DIII-D H-mode plasmas are presented. 6pt [1] W.M. Stacey, Contr. Plasma Phys. 48, 94 (2008). [2] W.M. Stacey and R.J. Groebner, Phys. Plasmas 15, 012503 (2008).

Development of a cross-polarization scattering system for the measurement of internal magnetic fluctuations in the DIII-D tokamak

DOE PAGES

Rhodes, Terry L.; Peebles, William A.; Crocker, Neal A.; ...

2014-08-05

The design and performance of a new cross-polarization scattering (CPS) system for the localized measurement of internal magnetic fluctuations is presented. CPS is a process whereby magnetic fluctuations scatter incident electromagnetic radiation into a perpendicular polarization which is subsequently detected. A new CPS design that incorporates a unique scattering geometry was laboratory tested, optimized, and installed on the DIII-D tokamak. Plasma tests of signal-to-noise, polarization purity, and frequency response indicate proper functioning of the system. Lastly, CPS data show interesting features related to internal MHD perturbations known as sawteeth that are not observed on density fluctuations.

Core tungsten radiation diagnostic calibration by small shell pellet injection in the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hollmann, Eric M.; Commaux, Nicolas; Shiraki, Daisuke

Injection of small (OD = 0.8 mm) plastic pellets carrying embedded smaller (10 μg) tungsten grains is used to check calibrations of core tungsten line radiation diagnostics in support of the 2016 tungsten rings campaign in the DIII-D tokamak. The total (1 eV – 10 keV) and soft x-ray (1 keV – 10 keV) brightnesses we observed were found to be reasonably well (< factor 2) predicted using existing calibration factors and rate calculations. Individual core (EUV/SXR) tungsten line brightnesses appear to be somewhat less reliable (factor 2-4) for prediction of core tungsten concentration.

Spectroscopic measurements and modeling of tungsten erosion in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Abrams, T. D.; Ding, R.; Guo, H. Y.; Leonard, A. W.; Thomas, D. M.; Allen, S. L.; McLean, A. G.; Briesemeister, A. R.; Unterberg, E. A.; Chrobak, C.; Doerner, R. P.; Rudakov, D. L.; Elder, J. D.; Stangeby, P. C.; Wampler, W. R.; Watkins, J. G.

2015-11-01

In situ time-resolved measurements of the gross W erosion rate have been performed in DIII-D by monitoring W/I (400.9 nm) emission in the divertor via a filtered camera and high-resolution spectrometer. The erosion rate of a thin W coating on DiMES, inferred via the S/XB method, was found to be ~ 0.7 nm/s during deuterim L-mode exposure, in fair agreement with post-mortem IBA analysis but lower than REDEP/WBC modeling. During H-mode He bombardment of W disks, average erosion rates of ~ 2.9 nm/s and ~ 9.0 nm/s were estimated during the inter-ELM and intra-ELM phases, using ne and Te from divertor Thomson scattering and Langmuir probes. Results will also be presented from additional W erosion experiments in preparation for the DIII-D mini-campaign to measure high-Z transport in the edge plasma. Comparisons will be made with ERO modeling Supported by US DOE DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-SC0001961, DE-AC04-94AL85000.

Test of bootstrap current models using high- β p EAST-demonstration plasmas on DIII-D

DOE PAGES

Ren, Qilong; Lao, Lang L.; Garofalo, Andrea M.; ...

2015-01-12

Magnetic measurements together with kinetic profile and motional Stark effect measurements are used in full kinetic equilibrium reconstructions to test the Sauter and NEO bootstrap current models in a DIII-D high-more » $${{\\beta}_{\\text{p}}}$$ EAST-demonstration experiment. This aims at developing on DIII-D a high bootstrap current scenario to be extended on EAST for a demonstration of true steady-state at high performance and uses EAST-similar operational conditions: plasma shape, plasma current, toroidal magnetic field, total heating power and current ramp-up rate. It is found that the large edge bootstrap current in these high-$${{\\beta}_{\\text{p}}}$$ plasmas allows the use of magnetic measurements to clearly distinguish the two bootstrap current models. In these high collisionality and high-$${{\\beta}_{\\text{p}}}$$ plasmas, the Sauter model overpredicts the peak of the edge current density by about 30%, while the first-principle kinetic NEO model is in close agreement with the edge current density of the reconstructed equilibrium. Furthermore, these results are consistent with recent work showing that the Sauter model largely overestimates the edge bootstrap current at high collisionality.« less

Detachment experiments in new DIII-D upper divertor

NASA Astrophysics Data System (ADS)

Moser, A. L.; Leonard, A. W.; Groebner, R. J.; Guo, H.; Wang, H.; Watkins, J. G.; McLean, A. G.; Fenstermacher, M. E.; Shafer, M. W.; Briesemeister, A. R.; Hinson, E. T.

2017-10-01

Installation of the Small Angle Slot (SAS) in the upper divertor of DIII-D enables new studies of the effect of target and baffle geometry on divertor detachment. This structure provides a more-closed upper divertor as well as the SAS divertor itself. Initial SAS experiment results indicate that divertor detachment occurs at a lower line-averaged density than in the more-open, lower single null divertor configurations on DIII-D. In contrast, the increased divertor closure of the new installation did not reduce the upstream density required for detachment beyond that achieved with the previous upper divertor structure. Particle pumping in the upper divertor structure is found to produce a 10 % reduction in the pedestal density required for detachment compared to the case with no pumping. Comparisons focus on both the onset of detachment (measured by in-target Langmuir probes) as a function of upstream density, as well as the effect of the new divertor configurations on pedestal density profiles. Work supported by US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-SC00013911.

Compatibility of internal transport barrier with steady-state operation in the high bootstrap fraction regime on DIII-D

DOE PAGES

Garofalo, Andrea M.; Gong, Xianzu; Grierson, Brian A.; ...

2015-11-16

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by ≥30% relative to earlier work. The advancement was enabled by improved understanding of the “relaxation oscillations”, previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the “relaxation oscillations” are coupled core-edge modes 2 amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced tomore » classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ≥ 80%, β N ≤ 4 , β P ≥ 3 , and β T ≥ 2%. Finally, these results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.« less

Anomalous Transport in High Beta Poloidal DIII-D Discharges

NASA Astrophysics Data System (ADS)

Pankin, A.; Garofalo, A.; Kritz, A.; Rafiq, T.; Weiland, J.

2016-10-01

Dominant instabilities that drive anomalous transport in high beta poloidal DIII-D discharges are investigated using the MMM7.1, and TGLF models in the predictive integrated modeling TRANSP code. The ion thermal transport is found to be strongly reduced in these discharges, but turbulence driven by the ITG modes along with the neoclassical transport still play a role in determining the ion temperature profiles. The electron thermal transport driven by the ETG modes impact the electron temperature profiles. The E × B flow shear is found to have a small effect in reducing the electron thermal transport. The Shafranov shift is found to strongly reduce the anomalous transport in the high beta poloidal DIII-D discharges. The reduction of Shafranov shift can destroy the ion internal transport barrier and can result in significantly lower core temperatures. The MMM7.1 model predicts electron and ion temperature profiles reasonably well, but it fails to accurately predict the properties of electron internal transport barrier, which indicates that the ETG model in MMM7.1 needs to be improved in the high beta poloidal operational regime. Research supported by the Office of Science, US DOE.

Inferring Core Tungsten Behavior Using SPRED During the DIII-D Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Thomas, D. M.; Kaplan, D.; Groebner, R.; Grierson, B.; Unterberg, Z.; Victor, B.

2016-10-01

The GA SPRED EUV spectrometer was used to study core emission of highly charged tungsten ions (W40+-W45+) in the 120-135Å region during the recent Metal Rings Campaign. These experiments used two 5-cm wide toroidal rings of W-coated metal inserts exposed to a variety of DIII-D discharges to study effects of high-Z divertor erosion, migration, core uptake, and effects on advanced tokamak performance. For the proper core temperature range (2-4 keV), the measured multistate W emission forms a well defined spectral pattern that can be used to study the relative importance of strike point location, flux expansion, injected power, ELM characteristics and magnetic drift direction for high-Z core contamination in DIII-D. The spectra are fit using simple Gaussians to estimate concentrations using the historical SPRED intensity calibration. Calibration shots using known core dosages of pellet injected W are used to help infer the relative response of the instrument. Supported by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-AC05-00OR22725, DE-AC52-07NA27344.

Advances in Low-Temperature Tungsten Spectroscopy Capability to Quantify DIII-D Divertor Erosion

DOE PAGES

Abrams, Tyler; Thomas, Daniel M.; Unterberg, Ezekial A.; ...

2018-01-05

Recent emphasis of tungsten (W) plasma-materialsinteractions (PMI) experiments on DIII-D has made it essential to enhance the W I and W II measurement capabilities of its spectroscopy diagnostic suite to acquire W sourcing measurements with high temporal, spatial, and wavelength resolution. To this end, four new viewing chords for the Multichordal Divertor Spectrometer (MDS) and diverter filterscope systems were installed, leading to a 7x increase in blue light sensitivity. W I and low-Z impurity line identifications were performed in the 3995-4030 Å region, placing wavelengths within 0.1 Å of the NIST values. A novel method was also developed for themore » DIII-D high temporal resolution filterscopes to distinguish between W I light and background contamination, important due to the relatively weak intensity of this line, using two different bandpass filters width different widths but the same center wavelength. Lastly, fast imaging of the W I 4008.75 Å spectral line with a PCO Pixelfly VGA 200/205 camera allowed for discrimination between ELMy and intra-ELM W I emission profiles with very high (~1 mm) spatial resolution.« less

Particle transport in DIII-D plasmas

NASA Astrophysics Data System (ADS)

Kress, Peter; Mordijck, Saskia

2017-10-01

By analyzing the plasma opacity and density evolution during the ELM cycle in DIII-D H-mode plasmas in which the amount of gas fueling was altered, we find evidence for an inward particle pinch at the plasma edge which seems to become more pronounced at higher density. Furthermore, at the plasma edge we find a correlation between the pedestal density and opacity, which measures neutral penetration depth. The changes in edge opacity during an ELM cycle were calculated by using a detailed time history of measured plasma profiles. At the same time, the density evolution during an ELM cycle was investigated. We find that if the edge density increases through an increase in gas fueling, then opacity increases and neutral fueling penetration depth decreases. We also find that density at the top of the pedestal recovers faster following an ELM when the overall density level is higher, leading to a hollow profile inside of the pedestal top. All these results indicate that there must be an inward particle pinch in the pedestal which will be crucial in the fueling of future burning plasma devices. Supported by US DOE DE-SC0007880, DIII-D Grant Number DE-FC02-04ER54698.

Bayesian Inference of Physics Parameters in the DIII-D Charge-Exchange Recombination Spectroscopy System

NASA Astrophysics Data System (ADS)

Bowman, C.; Gibson, K. J.; La Haye, R. J.; Groebner, R. J.; Taylor, N. Z.; Grierson, B. A.

2014-10-01

A Bayesian inference framework has been developed for the DIII-D charge-exchange recombination (CER) system, capable of computing probability distribution functions (PDFs) for desired parameters. CER is a key diagnostic system at DIII-D, measuring important physics parameters such as plasma rotation and impurity ion temperature. This work is motivated by a case in which the CER system was used to probe the plasma rotation radial profile around an m/n = 2/1 tearing mode island rotating at ~ 1 kHz. Due to limited resolution in the tearing mode phase and short integration time, it has proven challenging to observe the structure of the rotation profile across the island. We seek to solve this problem by using the Bayesian framework to improve the estimation accuracy of the plasma rotation, helping to reveal details of how it is perturbed in the magnetic island vicinity. Examples of the PDFs obtained through the Bayesian framework will be presented, and compared with results from a conventional least-squares analysis of the CER data. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC02-09CH11466.

Advances in Low-Temperature Tungsten Spectroscopy Capability to Quantify DIII-D Divertor Erosion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abrams, Tyler; Thomas, Daniel M.; Unterberg, Ezekial A.

Recent emphasis of tungsten (W) plasma-materialsinteractions (PMI) experiments on DIII-D has made it essential to enhance the W I and W II measurement capabilities of its spectroscopy diagnostic suite to acquire W sourcing measurements with high temporal, spatial, and wavelength resolution. To this end, four new viewing chords for the Multichordal Divertor Spectrometer (MDS) and diverter filterscope systems were installed, leading to a 7x increase in blue light sensitivity. W I and low-Z impurity line identifications were performed in the 3995-4030 Å region, placing wavelengths within 0.1 Å of the NIST values. A novel method was also developed for themore » DIII-D high temporal resolution filterscopes to distinguish between W I light and background contamination, important due to the relatively weak intensity of this line, using two different bandpass filters width different widths but the same center wavelength. Lastly, fast imaging of the W I 4008.75 Å spectral line with a PCO Pixelfly VGA 200/205 camera allowed for discrimination between ELMy and intra-ELM W I emission profiles with very high (~1 mm) spatial resolution.« less

Observation of EHO in NSTX and Theoretical Study of its Active Control Using HHFW Antenna

DOE Office of Scientific and Technical Information (OSTI.GOV)

J.-K. Park, et. al.

2013-01-14

Two important topics in the tokamak ELM control, using the non-axisymmetric (3D) magnetic perturbations, are studied in NSTX and combined envisioning ELM control in the future NSTX-U operation: Experimental observations of the edge harmonic oscillation in NSTX (not necessarily the same as EHOs in DIII-D), and theoretical study of its external drive using the high harmonic fast wave (HHFW) antenna as a 3D field coil. Edge harmonic oscillations were observed particularly well in NSTX ELM-free operation with low n core modes, with various diagnostics confirming n = 4 ~ 6 edge-localized and coherent oscillations in 2 ~ 8kHz frequency range.more » These oscillations seem to have a favored operational window in rotational shear, similarly to EHOs in DIII-D QH modes . However, in NSTX, they are not observed to provide particle or impurity control, possibly due to their weak amplitudes, of a few mm displacements, as measured by reflectometry. The external drive of these modes has been proposed in NSTX, by utilizing audio-frequency currents in the HHFW antenna straps. Analysis shows that the HHFW straps can be optimized to maximize n = 4 ~ 6 while minimizing n = 1 ~ 3. Also, IPEC calculations show that the optimized configuration with only 1kAt current can produce comparable or larger displacements than the observed internal modes. If this optimized external drive can be constructively combined, or further resonated with the internal modes, the edge harmonic oscillations in NSTX may be able to produce sufficient particle control to modify ELMs.« less

Update on the DIII-D ECH system: experiments, gyrotrons, advanced diagnostics, and controls

NASA Astrophysics Data System (ADS)

Lohr, John; Brambila, Rigoberto; Cengher, Mirela; Gorelov, Yuri; Grosnickle, William; Moeller, Charles; Ponce, Dan; Torrezan, Antonio; Ives, Lawrence; Reed, Michael; Blank, Monica; Felch, Kevin; Parisuaña, Claudia; LeViness, Alexandra

2017-08-01

The ECH system on DIII-D is continuing to be upgraded, while simultaneously being operated nearly daily for plasma experiments. The latest major hardware addition is a new 117.5 GHz gyrotron, which generated 1.7 MW for short pulses during factory testing. A new gyrotron control system based on Field Programmable Gate Array (FPGA) technology with very high speed system data acquisition has significantly increased the flexibility and reliability of individual gyrotron operation. We have improved the performance of the fast mirror scanning, both by increasing the scan speeds and by adding new algorithms for controlling the aiming using commands generated by the Plasma Control System (PCS). The system is used for transport studies, ELM control, current profile control, non-inductive current generation, suppression of MHD modes, startup assist, plasma density control, and other applications. A program of protective measures, which has been in place for more than two years, has eliminated damage to hardware and diagnostics caused by overdense operation. Other activities not directly related to fusion research have used the ECH system to test components, study methods for improving production of semiconductor junctions and materials, and test the feasibility of using ground based microwave systems to power satellites into orbit.

Dynamic diagnostics of the error fields in tokamaks

NASA Astrophysics Data System (ADS)

Pustovitov, V. D.

2007-07-01

The error field diagnostics based on magnetic measurements outside the plasma is discussed. The analysed methods rely on measuring the plasma dynamic response to the finite-amplitude external magnetic perturbations, which are the error fields and the pre-programmed probing pulses. Such pulses can be created by the coils designed for static error field correction and for stabilization of the resistive wall modes, the technique developed and applied in several tokamaks, including DIII-D and JET. Here analysis is based on the theory predictions for the resonant field amplification (RFA). To achieve the desired level of the error field correction in tokamaks, the diagnostics must be sensitive to signals of several Gauss. Therefore, part of the measurements should be performed near the plasma stability boundary, where the RFA effect is stronger. While the proximity to the marginal stability is important, the absolute values of plasma parameters are not. This means that the necessary measurements can be done in the diagnostic discharges with parameters below the nominal operating regimes, with the stability boundary intentionally lowered. The estimates for ITER are presented. The discussed diagnostics can be tested in dedicated experiments in existing tokamaks. The diagnostics can be considered as an extension of the 'active MHD spectroscopy' used recently in the DIII-D tokamak and the EXTRAP T2R reversed field pinch.

Control of power, torque, and instability drive using in-shot variable neutral beam energy in tokamaks

DOE PAGES

Pace, D. C.; Collins, C. S.; Crowley, B.; ...

2016-09-28

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak. Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significantmore » changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. As a result, developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.« less

Enabling High Fidelity Measurements of Energy and Pitch Angle for Escaping Energetic Ions with a Fast Ion Loss Detector

NASA Astrophysics Data System (ADS)

Chaban, R.; Pace, D. C.; Marcy, G. R.; Taussig, D.

2016-10-01

Energetic ion losses must be minimized in burning plasmas to maintain fusion power, and existing tokamaks provide access to energetic ion parameter regimes that are relevant to burning machines. A new Fast Ion Loss Detector (FILD) probe on the DIII-D tokamak has been optimized to resolve beam ion losses across a range of 30 - 90 keV in energy and 40° to 80° in pitch angle, thereby providing valuable measurements during many different experiments. The FILD is a magnetic spectrometer; once inserted into the tokamak, the magnetic field allows energetic ions to pass through a collimating aperture and strike a scintillator plate that is imaged by a wide view camera and narrow view photomultiplier tubes (PMTs). The design involves calculating scintillator strike patterns while varying probe geometry. Calculated scintillator patterns are then used to design an optical system that allows adjustment of the focus regions for the 1 MS/s resolved PMTs. A synthetic diagnostic will be used to determine the energy and pitch angle resolution that can be attained in DIII-D experiments. Work supported in part by US DOE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

Control of power, torque, and instability drive using in-shot variable neutral beam energy in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pace, D. C.; Collins, C. S.; Crowley, B.

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak. Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significantmore » changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. As a result, developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.« less

Control of power, torque, and instability drive using in-shot variable neutral beam energy in tokamaks

NASA Astrophysics Data System (ADS)

Pace, D. C.; Collins, C. S.; Crowley, B.; Grierson, B. A.; Heidbrink, W. W.; Pawley, C.; Rauch, J.; Scoville, J. T.; Van Zeeland, M. A.; Zhu, Y. B.; The DIII-D Team

2017-01-01

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak (Luxon 2002 Nucl. Fusion 42 614). Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significant changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. Developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.

Development and applications of 3D-DIVIMP(HC) Monte Carlo impurity modeling code

NASA Astrophysics Data System (ADS)

Mu, Yarong

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D, the Porous Plug Injector (PPI), has been employed by A. McLean for in-situ study of chemical erosion in the tokamak divertor environment by injection of CH4. The principal contribution of the present thesis is a new interpretive code, 3D-DIVIMP(HC), which has been developed and successfully applied to the interpretation of the CH, C I, and C II emissions measured during the PPI experiments. The two principal types of experimental data which are compared here with 3D-DIVIMP(HC) code modeling are (a) absolute emissivities measured with a high resolution spectrometer, and (b) 2D filtered camera (TV) pictures taken from a view essentially straight down on the PPI. Incorporating the Janev-Reiter database for the breakup reactions of methane molecules in a plasma, 3D-DIVIMP(HC) is able to replicate these measurements to within the combined experimental and database uncertainties. It is therefore concluded that the basic elements of the physics and chemistry controlling the breakup of methane entering an attached divertor plasma have been identified and are incorporated in 3D-DIVIMP(HC).

Projecting High Beta Steady-State Scenarios from DIII-D Advanced Tokamk Discharges

NASA Astrophysics Data System (ADS)

Park, J. M.

2013-10-01

Fusion power plant studies based on steady-state tokamak operation suggest that normalized beta in the range of 4-6 is needed for economic viability. DIII-D is exploring a range of candidate high beta scenarios guided by FASTRAN modeling in a repeated cycle of experiment and modeling validation. FASTRAN is a new iterative numerical procedure coupled to the Integrated Plasma Simulator (IPS) that integrates models of core transport, heating and current drive, equilibrium and stability self-consistently to find steady state (d / dt = 0) solutions, and reproduces most features of DIII-D high beta discharges with a stationary current profile. Separately, modeling components such as core transport (TGLF) and off-axis neutral beam current drive (NUBEAM) show reasonable agreement with experiment. Projecting forward to scenarios possible on DIII-D with future upgrades, two self-consistent noninductive scenarios at βN > 4 are found: high qmin and high internal inductance li. Both have bootstrap current fraction fBS > 0 . 5 and rely on the planned addition of a second off-axis neutral beamline and increased electron cyclotron heating. The high qmin > 2 scenario achieves stable operation at βN as high as 5 by a very broad current density profile to improve the ideal-wall stabilization of low-n instabilities along with confinement enhancement from low magnetic shear. The li near 1 scenario does not depend on ideal-wall stabilization. Improved confinement from strong magnetic shear makes up for the lower pedestal needed to maintain li high. The tradeoff between increasing li and reduced edge pedestal determines the achievable βN (near 4) and fBS (near 0.5). This modeling identifies the necessary upgrades to achieve target scenarios and clarifies the pros and cons of particular scenarios to better inform the development of steady-state fusion. Supported by the US Department of Energy under DE-AC05-00OR22725 & DE-FC02-04ER54698.

The pattern of parallel edge plasma flows due to pressure gradients, recycling, and resonant magnetic perturbations in DIII-D

DOE PAGES

Frerichs, H.; Schmitz, Oliver; Evans, Todd; ...

2015-07-13

High resolution plasma transport simulations with the EMC3-EIRENE code have been performed to address the parallel plasma flow structure in the boundary of a poloidal divertor configuration with non-axisymmetric perturbations at DIII-D. Simulation results show that a checkerboard pattern of flows with alternating direction is generated inside the separatrix. This pattern is aligned with the position of the main resonances (i.e. where the safety factor is equal to rational values q = m/n for a perturbation field with base mode number n): m pairs of alternating forward and backward flow channel exist for each resonance. The poloidal oscillations are alignedmore » with the subharmonic Melnikov function, which indicates that the plasma flow is generated by parallel pressure gradients along perturbed field lines. Lastly, an additional scrape-off layer-like domain is introduced by the perturbed separatrix which guides field lines from the interior to the divertor targets, resulting in an enhanced outward flow that is consistent with the experimentally observed particle pump-out effect. However, while the lobe structure of the perturbed separatrix is very well reflected in the temperature profile, the same lobes can appear to be smaller in the flow profile due to a competition between high upstream pressure and downstream particle sources driving flows in opposite directions.« less

Extended-MHD Studies of Flow-Profile Effects on Edge Harmonic Oscillations in QH-mode Discharges

NASA Astrophysics Data System (ADS)

King, J. R.; Burrell, K. H.; Garofalo, A. M.; Jenkins, T. G.; Kruger, S. E.; Snyder, P. B.

2012-10-01

It is desirable to have an ITER H-mode regime that is quiescent to edge-localized modes (ELMs). ELMs deposit large, localized, impulsive, surface heat loads that can damage the divertor. One such quiescent regime with edge harmonic oscillations (EHO) is observed on DIII-D, JET, JT-60U, and ASDEX-U [1]. The physical mechanisms of EHO are not fully understood, but linear MHD calculations suggest EHO may be a saturated kink-peeling mode partially driven by flow-profile shear [2]. We present preliminary EHO computations using the extended-MHD NIMROD code. The model incorporates first-order FLR effects and parallel heat flows. Using reconstructed DIII-D profiles from discharges with EHO, we scan the ExB and polodial flow profiles and compute linear stability. The aim is to ascertain the role of the ExB flow shear, as motivated by experimental results [3], and to compare with theoretical predictions where the growth rate is enhanced at intermediate wavenumbers and cut-off at large wavenumbers by diamagnetic effects [4]. Initial nonlinear computations exploring the EHO saturation mechanism are presented.[4pt] [1] Phys. Plasmas, v19, p056117, 2012 (and refs. within).[0pt] [2] Nucl. Fusion, v47, p961, 2007.[0pt] [3] Nucl. Fusion, v51, p083018, 2011.[0pt] [4] Phys. Plasmas v10, p4405, 2003.

Impurity mixing and radiation asymmetry in massive gas injection simulations of DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Izzo, V. A.

Simulations of neon massive gas injection into DIII-D are performed with the 3D MHD code NIMROD. The poloidal and toroidal distribution of the impurity source is varied. This report will focus on the effects of the source variation on impurity mixing and radiated power asymmetry. Even toroidally symmetric impurity injection is found to produce asymmetric radiated power due to asymmetric convective heat flux produced by the 1/1 mode. When the gas source is toroidally localized, the phase relationship between the mode and the source location is important, affecting both radiation peaking and impurity mixing. Under certain circumstances, a single, localizedmore » gas jet could produce better radiation symmetry during the disruption thermal quench than evenly distributed impurities.« less

Integrated modeling of plasma ramp-up in DIII-D ITER-like and high bootstrap current scenario discharges

NASA Astrophysics Data System (ADS)

Wu, M. Q.; Pan, C. K.; Chan, V. S.; Li, G. Q.; Garofalo, A. M.; Jian, X.; Liu, L.; Ren, Q. L.; Chen, J. L.; Gao, X.; Gong, X. Z.; Ding, S. Y.; Qian, J. P.; Cfetr Physics Team

2018-04-01

Time-dependent integrated modeling of DIII-D ITER-like and high bootstrap current plasma ramp-up discharges has been performed with the equilibrium code EFIT, and the transport codes TGYRO and ONETWO. Electron and ion temperature profiles are simulated by TGYRO with the TGLF (SAT0 or VX model) turbulent and NEO neoclassical transport models. The VX model is a new empirical extension of the TGLF turbulent model [Jian et al., Nucl. Fusion 58, 016011 (2018)], which captures the physics of multi-scale interaction between low-k and high-k turbulence from nonlinear gyro-kinetic simulation. This model is demonstrated to accurately model low Ip discharges from the EAST tokamak. Time evolution of the plasma current density profile is simulated by ONETWO with the experimental current ramp-up rate. The general trend of the predicted evolution of the current density profile is consistent with that obtained from the equilibrium reconstruction with Motional Stark effect constraints. The predicted evolution of βN , li , and βP also agrees well with the experiments. For the ITER-like cases, the predicted electron and ion temperature profiles using TGLF_Sat0 agree closely with the experimental measured profiles, and are demonstrably better than other proposed transport models. For the high bootstrap current case, the predicted electron and ion temperature profiles perform better in the VX model. It is found that the SAT0 model works well at high IP (>0.76 MA) while the VX model covers a wider range of plasma current ( IP > 0.6 MA). The results reported in this paper suggest that the developed integrated modeling could be a candidate for ITER and CFETR ramp-up engineering design modeling.

Spatially resolved measurements of two-dimensional turbulent structures in DIII-D plasmas

DOE PAGES

Zemedkun, Samuel E.; Che, S.; Chen, Y.; ...

2015-12-21

Here, two-dimensional observations of spatially-coherent electron temperature fluctuations at drift wave scales (k ~1 cm -1) have been made using the electron cyclotron emission imaging (ECEI) diagnostic on the DIII-D tokamak. These measurements enable the extraction of spectral properties, including poloidal dispersion relations. Temperature fluctuation levels are found to be ˜ T e/< T e > = 1.2%, and the phase velocity of the fluctuations is found to be constant across frequencies, consistent with modes having real frequencies low compared to the rotation-induced Doppler shifts. Comparisons with radially global linear gyrokinetic simulations suggest that the observed modes may be trappedmore » electron modes (TEM).« less

Modeling of Synergy Between 4th and 6th Harmonic Absorptions of Fast Waves on Injected Beams in DIII-D Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, M.; Pinsker, R. I.; Chan, V. S.

2011-12-23

In recent moderate to high harmonic fast wave heating and current drive experiments in DIII-D, a synergy effect was observed when the 6{sup th} harmonic 90 MHz fast wave power is applied to the plasma preheated by neutral beams and the 4{sup th} harmonic 60 MHz fast wave. In this paper, we investigate how the synergy can occur using ORBIT-RF coupled with AORSA. Preliminary simulations suggest that damping of 4{sup th} harmonic FW on beam ions accelerates them above the injection energy, which may allow significant damping of 6{sup th} harmonic FW on beam ion tails to produce synergy.

Measurements and modeling of intra-ELM tungsten sourcing and transport in DIII-D

NASA Astrophysics Data System (ADS)

Abrams, T.; Leonard, A. W.; Thomas, D. M.; McLean, A. G.; Makowski, M. A.; Wang, H. Q.; Unterberg, E. A.; Briesemeister, A. R.; Rudakov, D. L.; Bykov, I.; Donovan, D.

2017-10-01

Intra-ELM tungsten erosion profiles in the DIII-D divertor, acquired via W I spectroscopy with high temporal and spatial resolution, are consistent with SDTrim.SP sputtering modeling using measured ion saturation currents and impact energies during ELMs as input and an ad-hoc 2% C2+ impurity flux. The W sputtering profile peaks close to the OSP both during and between ELMs in the favorable BT direction. In reverse BT the W source peaks close to the OSP between ELMs but strongly broadens and shifts outboard during ELMs, heuristically consistent with radially outward ion transport via ExB drifts. Ion impact energies during ELMs (inferred taking the ratio of divertor heat flux to the ion saturation current) are found to be approximately equal to Te,ped, lower than the 4*Te,ped value predicted by the Fundamenski/Moulton free streaming model. These impact energies imply both D main ions and C impurities contribute strongly to W sputtering during ELMs on DIII-D. This work represents progress towards a predictive model to link upstream conditions (i.e., pedestal height) and SOL impurity levels to the ELM-induced W impurity source at both the strike-point and far-target regions in the ITER divertor. Correlations between ELM size/frequency and SOL W fluxes measured via a midplane deposition probe will also be presented. Work supported by US DOE under DE-FC02-04ER54698.

Transport modeling of the DIII-D high $${{\\beta}_{p}}$$ scenario and extrapolations to ITER steady-state operation

DOE PAGES

McClenaghan, Joseph; Garofalo, Andrea M.; Meneghini, Orso; ...

2017-08-03

In this study, transport modeling of a proposed ITER steady-state scenario based on DIII-D high poloidal-beta (more » $${{\\beta}_{p}}$$ ) discharges finds that ITB formation can occur with either sufficient rotation or a negative central shear q-profile. The high $${{\\beta}_{p}}$$ scenario is characterized by a large bootstrap current fraction (80%) which reduces the demands on the external current drive, and a large radius internal transport barrier which is associated with excellent normalized confinement. Modeling predictions of the electron transport in the high $${{\\beta}_{p}}$$ scenario improve as $${{q}_{95}}$$ approaches levels similar to typical existing models of ITER steady-state and the ion transport is turbulence dominated. Typical temperature and density profiles from the non-inductive high $${{\\beta}_{p}}$$ scenario on DIII-D are scaled according to 0D modeling predictions of the requirements for achieving a $Q=5$ steady-state fusion gain in ITER with 'day one' heating and current drive capabilities. Then, TGLF turbulence modeling is carried out under systematic variations of the toroidal rotation and the core q-profile. A high bootstrap fraction, high $${{\\beta}_{p}}$$ scenario is found to be near an ITB formation threshold, and either strong negative central magnetic shear or rotation in a high bootstrap fraction are found to successfully provide the turbulence suppression required to achieve $Q=5$.« less

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

NASA Astrophysics Data System (ADS)

Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

2018-05-01

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

Validation of the kinetic-turbulent-neoclassical theory for edge intrinsic rotation in DIII-D

NASA Astrophysics Data System (ADS)

Ashourvan, Arash; Grierson, B. A.; Battaglia, D. J.; Haskey, S. R.; Stoltzfus-Dueck, T.

2018-05-01

In a recent kinetic model of edge main-ion (deuterium) toroidal velocity, intrinsic rotation results from neoclassical orbits in an inhomogeneous turbulent field [T. Stoltzfus-Dueck, Phys. Rev. Lett. 108, 065002 (2012)]. This model predicts a value for the toroidal velocity that is co-current for a typical inboard X-point plasma at the core-edge boundary (ρ ˜ 0.9). Using this model, the velocity prediction is tested on the DIII-D tokamak for a database of L-mode and H-mode plasmas with nominally low neutral beam torque, including both signs of plasma current. Values for the flux-surface-averaged main-ion rotation velocity in the database are obtained from the impurity carbon rotation by analytically calculating the main-ion—impurity neoclassical offset. The deuterium rotation obtained in this manner has been validated by direct main-ion measurements for a limited number of cases. Key theoretical parameters of ion temperature and turbulent scale length are varied across a wide range in an experimental database of discharges. Using a characteristic electron temperature scale length as a proxy for a turbulent scale length, the predicted main-ion rotation velocity has a general agreement with the experimental measurements for neutral beam injection (NBI) powers in the range PNBI < 4 MW. At higher NBI power, the experimental rotation is observed to saturate and even degrade compared to theory. TRANSP-NUBEAM simulations performed for the database show that for discharges with nominally balanced—but high powered—NBI, the net injected torque through the edge can exceed 1 Nm in the counter-current direction. The theory model has been extended to compute the rotation degradation from this counter-current NBI torque by solving a reduced momentum evolution equation for the edge and found the revised velocity prediction to be in agreement with experiment. Using the theory modeled—and now tested—velocity to predict the bulk plasma rotation opens up a path to more confidently projecting the confinement and stability in ITER.

Orbit Tomography: A Method for Determining the Population of Individual Fast-ion Orbits from Experimental Measurements

NASA Astrophysics Data System (ADS)

Stagner, L.; Heidbrink, W. W.

2017-10-01

Due to the complicated nature of the fast-ion distribution function, diagnostic velocity-space weight functions are used to analyze experimental data. In a technique known as Velocity-space Tomography (VST), velocity-space weight functions are combined with experimental measurements to create a system of linear equations that can be solved. However, VST (which by definition ignores spatial dependencies) is restricted, both by the accuracy of its forward model and also by the availability of spatially overlapping diagnostics. In this work we extend velocity-space weight functions to a full 6D generalized coordinate system and then show how to reduce them to a 3D orbit-space without loss of generality using an action-angle formulation. Furthermore, we show how diagnostic orbit-weight functions can be used to infer the full fast-ion distribution function, i.e. Orbit Tomography. Examples of orbit weights functions for different diagnostics and reconstructions of fast-ion distributions are shown for DIII-D experiments. This work was supported by the U.S. Department of Energy under DE-AC02-09CH11466 and DE-FC02-04ER54698.

The inter-ELM tungsten erosion profile in DIII-D H-mode discharges and benchmarking with ERO+OEDGE modeling

NASA Astrophysics Data System (ADS)

Abrams, T.; Ding, R.; Guo, H. Y.; Thomas, D. M.; Chrobak, C. P.; Rudakov, D. L.; McLean, A. G.; Unterberg, E. A.; Briesemeister, A. R.; Stangeby, P. C.; Elder, J. D.; Wampler, W. R.; Watkins, J. G.

2017-05-01

It is important to develop a predictive capability for the tungsten source rate near the strike points during H-mode operation in ITER and beyond. H-mode deuterium plasma exposures were performed on W-coated graphite and molybdenum substrates in the DIII-D divertor using DiMES. The W-I 400.9 nm spectral line was monitored by fast filtered diagnostics cross calibrated via a high-resolution spectrometer to resolve inter-ELM W erosion. The effective ionization/photon (S/XB) was calibrated using a unique method developed on DIII-D based on surface analysis. Inferred S/XB values agree with an existing empirical scaling at low electron density (n e) but diverge at higher densities, consistent with recent ADAS atomic physics modeling results. Edge modeling of the inter-ELM phase is conducted via OEDGE utilizing the new capability for charge-state resolved carbon impurity fluxes. ERO modeling is performed with the calculated main ion and impurity plasma background from OEDGE. ERO results demonstrate the importance a mixed-material surface model in the interpretation of W sourcing measurements. It is demonstrated that measured inter-ELM W erosion rates can be well explained by C→W sputtering only if a realistic mixed material model is incorporated.

Quasistationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges [A New, Quasi-stationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barada, Kshitish; Rhodes, Terry L.; Burrell, Keith H.

A new, long-lived limit cycle oscillation (LCO) regime has been observed in the edge of near zero torque high-performance DIII-D tokamak plasma discharges. These LCOs are localized and comprised of density turbulence, gradient drives, and E X B velocity shear damping ( E and B are the local radial electric and total magnetic fields). Density turbulence sequentially acts as a predator (via turbulence transport) of profile gradients and a prey (via shear suppression) to the E X B velocity shear. Reported here for the first time, a unique spatiotemporal variation of the local E X B velocity which is foundmore » to be essential for the existence of this system. The LCO system is quasi-stationary, existing from 3 to 12 plasma energy confinement times (~30 to 900 LCO cycles) limited by hardware constraints. In conclusion, this plasma system appears to contribute strongly to the edge transport in these high-performance and transient-free plasmas as evident from oscillations in transport relevant edge parameters at LCO timescale.« less

High heat flux Langmuir probe array for the DIII-D divertor platesa)

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Taussig, D.; Boivin, R. L.; Mahdavi, M. A.; Nygren, R. E.

2008-10-01

Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m2 for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5° surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of Jsat, Te, and Vf with 4 mm spatial resolution are shown.

Quasistationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges [A New, Quasi-stationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E × B Flow During High Performance DIII-D Tokamak Discharges

DOE PAGES

Barada, Kshitish; Rhodes, Terry L.; Burrell, Keith H.; ...

2018-03-27

A new, long-lived limit cycle oscillation (LCO) regime has been observed in the edge of near zero torque high-performance DIII-D tokamak plasma discharges. These LCOs are localized and comprised of density turbulence, gradient drives, and E X B velocity shear damping ( E and B are the local radial electric and total magnetic fields). Density turbulence sequentially acts as a predator (via turbulence transport) of profile gradients and a prey (via shear suppression) to the E X B velocity shear. Reported here for the first time, a unique spatiotemporal variation of the local E X B velocity which is foundmore » to be essential for the existence of this system. The LCO system is quasi-stationary, existing from 3 to 12 plasma energy confinement times (~30 to 900 LCO cycles) limited by hardware constraints. In conclusion, this plasma system appears to contribute strongly to the edge transport in these high-performance and transient-free plasmas as evident from oscillations in transport relevant edge parameters at LCO timescale.« less

Recycling Reduction and Density Control with Lithium Injection in DIII-D

NASA Astrophysics Data System (ADS)

Jackson, G. L.; Chrobak, C. P.; Maingi, R.; Mansfield, D.; Roquemore, A.; McLean, A. G.

2013-10-01

Lithium conditioning has been effective in tokamaks for reducing recycling and providing density control, particularly in NSTX and EAST. Since DIII-D has not injected lithium in more than a decade (and then in only very small amounts, 0.4 g total), a unique opportunity exists to extend this experience and examine the physical effects of lithium in a well-conditioned lithium-free machine. A lithium dropper, developed by PPPL, has recently been installed on DIII-D. By injecting 0.09 g of lithium we have observed reductions in recycling, density, and ELM frequency from the first discharge with significant lithium injection. Although modeling of individual 40 μ m diam. Li granules predicts virtually no penetration beyond the separatrix in auxiliary heated H-mode pulses, LiIII emission was detected in the core plasma, albeit with no increase in radiated power. On subsequent discharges without injection no core Li was detected, and only LiI emission was observed in the SOL and divertor regions. We will present the effects of Li on recycling, ELM frequency, and the edge pedestal, and discuss the long-term observations of lithium on plasma facing components. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-AC02-09CH11466, and DE-AC52-07NA27344.

Global simulation of edge pedestal micro-instabilities

NASA Astrophysics Data System (ADS)

Wan, Weigang; Parker, Scott; Chen, Yang

2011-10-01

We study micro turbulence of the tokamak edge pedestal with global gyrokinetic particle simulations. The simulation code GEM is an electromagnetic δf code. Two sets of DIII-D experimental profiles, shot #131997 and shot #136051 are used. The dominant instabilities appear to be two kinds of modes both propagating in the electron diamagnetic direction, with comparable linear growth rates. The low n mode is at the Alfven frequency range and driven by density and ion temperature gradients. The high n mode is driven by electron temperature gradient and has a low real frequency. A β scan shows that the low n mode is electromagnetic. Frequency analysis shows that the high n mode is sometimes mixed with an ion instability. Experimental radial electric field is applied and its effects studied. We will also show some preliminary nonlinear results. We thank R. Groebner, P. Snyder and Y. Zheng for providing experimental profiles and helpful discussions.

Transport and Performance in DIII--D Discharges with Weak or Negative Central Magnetic Shear

NASA Astrophysics Data System (ADS)

Greenfield, C. M.

1996-11-01

The previously reported [B.W. Rice et al., Phys. Plasmas 3, 1983 (1996)] improved performance in DIII-D plasmas with weak or negative central magnetic shear has been additionally enhanced in recent experiments where controlled L-H transitions were used to further broaden the pressure profile and delay detrimental MHD activity [E.A. Lazarus et al., submitted to Phys. Rev. Letters]. These discharges exhibit the highest plasma energy (>=4 MJ) and fusion reactivity (R_DD <= 4.8 × 10^16 s-1, Q_DD <= 0.00146, equivalent Q_DT <= 0.32) yet realized in DIII-D. In such discharges, the core magnetic shear is reversed by tailoring the current profile through application of early, low power, neutral beam injection. These plasmas often undergo a transition to a high performance state, usually following an increase in the applied heating power. At the transition time, we observe the formation of an internal transport barrier near the location of the minimum safety factor, q_min. Formation of this barrier, which can result in central peaking of the temperature and density profiles, is consistent with suppression of turbulence by locally enhanced E×B flow shear. Beam emission spectroscopy and far infrared scattering measurements made in the vicinity of the barrier show that at the time of transition to high performance, fluctuation levels are reduced to below the threshold of detection (tilden/n <= 0.1%). Analysis with the ONETWO and TRANSP transport codes indicates concomitant reductions in the core ion thermal diffusivity to levels at or below Chang-Hinton neoclassical. Smaller reductions are indicated for the electrons. An L-H transition is programmed shortly before the plasma would become MHD unstable in order to broaden the profiles and delay the onset of instabilities. In the resulting state, the region exhibiting ion diffusivities at or below neoclassical is extended to nearly the entire plasma. Analysis to date suggests that the effect of strongly negative vs. weak magnetic shear on transport is negligible, although there is a significant effect on stability. A comparison of transport in strong and weakly sheared discharges will be shown, both in L- (peaked profiles) and H-mode (broadened profiles).

The effect of pre-existing islands on disruption mitigation in MHD simulations of DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Izzo, V. A.

Locked-modes are the most likely cause of disruptions in ITER, so large islands are expected to be common when the ITER disruption mitigation system is deployed. MHD modeling of disruption mitigation by massive gas injection is carried out for DIII-D plasmas with stationary, pre-existing islands. Results show that the magnetic topology at the q=2 surface can affect the parallel spreading of injected impurities, and that, in particular, the break-up of large 2/1 islands into smaller 4/2 islands chains can favorably affect mitigation metrics. The direct imposition of a 4/2 mode is found to have similar results to the case inmore » which the 4/2 harmonic grows spontaneously.« less

Improving fast-ion confinement in high-performance discharges by suppressing Alfvén eigenmodes

DOE PAGES

Kramer, Geritt J.; Podestà, Mario; Holcomb, Christopher; ...

2017-03-28

Here, we show that the degradation of fast-ion confinement in steady-state DIII-D discharges is quantitatively consistent with predictions based on the effects of multiple unstable Alfven eigenmodes on beam-ion transport. Simulation and experiment show that increasing the radius where the magnetic safety factor has its minimum is effective in minimizing beam-ion transport. This is favorable for achieving high performance steady-state operation in DIII-D and future reactors. A comparison between the experiments and a critical gradient model, in which only equilibrium profiles were used to predict the most unstable modes, show that in a number of cases this model reproduces themore » measured neutron rate well.« less

The effect of pre-existing islands on disruption mitigation in MHD simulations of DIII-D

DOE PAGES

Izzo, V. A.

2017-02-27

Locked-modes are the most likely cause of disruptions in ITER, so large islands are expected to be common when the ITER disruption mitigation system is deployed. MHD modeling of disruption mitigation by massive gas injection is carried out for DIII-D plasmas with stationary, pre-existing islands. Results show that the magnetic topology at the q=2 surface can affect the parallel spreading of injected impurities, and that, in particular, the break-up of large 2/1 islands into smaller 4/2 islands chains can favorably affect mitigation metrics. The direct imposition of a 4/2 mode is found to have similar results to the case inmore » which the 4/2 harmonic grows spontaneously.« less

Three-dimensional drift kinetic response of high-β plasmas in the DIII-D tokamak.

PubMed

Wang, Z R; Lanctot, M J; Liu, Y Q; Park, J-K; Menard, J E

2015-04-10

A quantitative interpretation of the experimentally measured high-pressure plasma response to externally applied three-dimensional (3D) magnetic field perturbations, across the no-wall Troyon β limit, is achieved. The self-consistent inclusion of the drift kinetic effects in magnetohydrodynamic (MHD) modeling [Y. Q. Liu et al., Phys. Plasmas 15, 112503 (2008)] successfully resolves an outstanding issue of the ideal MHD model, which significantly overpredicts the plasma-induced field amplification near the no-wall limit, as compared to experiments. The model leads to quantitative agreement not only for the measured field amplitude and toroidal phase but also for the measured internal 3D displacement of the plasma. The results can be important to the prediction of the reliable plasma behavior in advanced fusion devices, such as ITER [K. Ikeda, Nucl. Fusion 47, S1 (2007)].

Results from core-edge experiments in high Power, high performance plasmas on DIII-D

DOE PAGES

Petrie, T. W.; Fenstermacher, M. E.; Holcomb, C. T.; ...

2016-12-24

Here, significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q ⊥ P) ∝ [P SOL x I P] 0.92 for P SOL = 8-19 MW and I P = 1.0–1.4 MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-Dmore » plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q ⊥ P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core.« less

Understanding rotation profile structures in ECH-heated plasmas using nonlinear gyrokinetic simulations

NASA Astrophysics Data System (ADS)

Wang, Weixing; Brian, B.; Ethier, S.; Chen, J.; Startsev, E.; Diamond, P. H.; Lu, Z.

2015-11-01

A non-diffusive momentum flux connecting edge momentum sources/sinks and core plasma flow is required to establish the off-axis peaked ion rotation profile typically observed in ECH-heated DIII-D plasmas without explicit external momentum input. The understanding of the formation of such profile structures provides an outstanding opportunity to test the physics of turbulence driving intrinsic rotation, and validate first-principles-based gyrokinetic simulation models. Nonlinear, global gyrokinetic simulations of DIII-D ECH plasmas indicate a substantial ITG fluctuation-induced residual stress generated around the region of peaked toroidal rotation, along with a diffusive momentum flux. The residual stress profile shows an anti-gradient, dipole structure, which is critical for accounting for the formation of the peaked rotation profile. It is showed that both turbulence intensity gradient and zonal flow ExB shear contribute to the generation of k// asymmetry needed for residual stress generation. By balancing the simulated residual stress and the momentum diffusion, a rotation profile is calculated. In general, the radial structure of core rotation profile is largely determined by the residual stress profile, while the amplitude of core rotation depends on the edge toroidal rotation velocity, which is determined by edge physics and used as a boundary condition in our model. The calculated core rotation profile is consistent with the experimental measurements. Also discussed is the modification of turbulence-generated Reynolds stress on poloidal rotation in those plasmas. Work supported by U.S. DOE Contract DE-AC02-09-CH11466.

Enhanced reproducibility of L-mode plasma discharges via physics-model-based q-profile feedback control in DIII-D

NASA Astrophysics Data System (ADS)

Schuster, E.; Wehner, W. P.; Barton, J. E.; Boyer, M. D.; Luce, T. C.; Ferron, J. R.; Holcomb, C. T.; Walker, M. L.; Humphreys, D. A.; Solomon, W. M.; Penaflor, B. G.; Johnson, R. D.

2017-11-01

Recent experiments on DIII-D demonstrate the potential of physics-model-based q-profile control to improve reproducibility of plasma discharges. A combined feedforward  +  feedback control scheme is employed to optimize the current ramp-up phase by consistently achieving target q profiles (Target 1: q_min=1.3, q95=4.4 ; Target 2: q_min=1.65, q95=5.0 ; Target 3: q_min=2.1, q95=6.2 ) at prescribed times during the plasma formation phase (Target 1: t=1.5 s; Target 2: t=1.3 s; Target 3: t=1.0 s). At the core of the control scheme is a nonlinear, first-principles-driven, physics-based, control-oriented model of the plasma dynamics valid for low confinement (L-mode) scenarios. To prevent undesired L-H transitions, a constraint on the maximum allowable total auxiliary power is imposed in addition to the maximum powers for the individual heating and current-drive sources. Experimental results are presented to demonstrate the effectiveness of the combined feedforward  +  feedback control scheme to consistently achieve the desired target profiles at the predefined times. These results also show how the addition of feedback control significantly improves upon the feedforward-only control solution by reducing the matching error and also how the feedback controller is able to reduce the matching error as the constraint on the maximum allowable total auxiliary power is relaxed while keeping the plasma in L-mode.

Scientist in the Classroom: Highlights of a Plasma Outreach Program

NASA Astrophysics Data System (ADS)

Nagy, A.; Lee, R. L.

2000-10-01

The General Atomics education program ``Scientist in the Classroom'' now in its third year, uses scientists and engineers to present ``Plasma the fourth state of matter,'' to students in the classroom. A program goal is to make science an enjoyable experience while showing students how plasma plays an important role in their world. A fusion overview is presented, including topics on energy and environment. Using hands-on equipment, students manipulate a plasma discharge using magnets, observe its spectral properties and observe the plasma in a fluorescent tube. In addition, they observe physical properties of liquid nitrogen, and use an infrared camera to observe radiant heat energy. Several program benefits are; it costs less than facility tours, is more flexible in scheduling, and is adaptable for grades 2--adult. The program has doubled in coverage since last year, with over 2200 students at 20 schools visited by 8 scientists. Increased participation by the DIII-D staff in this program has been achieved by enlisting them to bring the program to their children's school.

Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henline, P.A.

1995-10-01

The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DRI-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape controlmore » due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.« less

Boundary perturbations coupled to core 3/2 tearing modes on the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Tobias, B.; Yu, L.; Domier, C. W.; Luhmann, N. C., Jr.; Austin, M. E.; Paz-Soldan, C.; Turnbull, A. D.; Classen, I. G. J.; the DIII-D Team

2013-09-01

High confinement (H-mode) discharges on the DIII-D tokamak are routinely subject to the formation of long-lived, non-disruptive magnetic islands that degrade confinement and limit fusion performance. Simultaneous, 2D measurement of electron temperature fluctuations in the core and edge regions allows for reconstruction of the radially resolved poloidal mode number spectrum and phase of the global plasma response associated with these modes. Coherent, n = 2 excursions of the plasma boundary are found to be the result of coupling to an n = 2, kink-like mode which arises locked in phase to the 3/2 island chain. This coupling dictates the relative phase of the displacement at the boundary with respect to the tearing mode. This unambiguous phase relationship, for which no counter-examples are observed, is presented as a test for modeling of the perturbed fields to be expected outside the confined plasma.

Applications of Collisional Radiative Modeling of Helium and Deuterium for Image Tomography Diagnostic of Te, Ne, and ND in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Munoz Burgos, J. M.; Brooks, N. H.; Fenstermacher, M. E.; Meyer, W. H.; Unterberg, E. A.; Schmitz, O.; Loch, S. D.; Balance, C. P.

2011-10-01

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. Supported in part by the US DOE under DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725.

Modeling of Plasma Pressure Effects on ELM Suppression With RMP in DIII-D

NASA Astrophysics Data System (ADS)

Orlov, D. M.; Moyer, R. A.; Mordijck, S.; Evans, T. E.; Osborne, T. H.; Snyder, P. B.; Unterberg, E. A.; Fenstermacher, M. E.

2009-11-01

Resonant magnetic perturbations (RMPs) are used to control the pedestal pressure gradient in both low and high (ν3^*) DIII-D plasmas. In this work we have analyzed several discharges with different levels of triangularity, different neutral beam injection power levels, and with, βN ranging from 1.5 to 2.3. The field line integration code TRIP3D was used to model the magnetic perturbation in ELMing and ELM suppressed phases during the RMP pulse. The results of this modeling showed very little effect of βN on the structure of the vacuum magnetic field during ELM suppression using n=3 RMPs. Kinetic equilibrium reconstructions showed a decrease in bootstrap current during RMP. Linear peeling-ballooning stability analysis performed with the ELITE code suggested that the ELMs, which persist during RMP, i.e. ELMing still is observed, are not Type I ELMs. Identification of these Dα spikes is an ongoing work.

Near midplane scintillator-based fast ion loss detector on DIII-D.

PubMed

Chen, X; Fisher, R K; Pace, D C; García-Muñoz, M; Chavez, J A; Heidbrink, W W; Van Zeeland, M A

2012-10-01

A new scintillator-based fast-ion loss detector (FILD) installed near the outer midplane of the plasma has been commissioned on DIII-D. This detector successfully measures coherent fast ion losses produced by fast-ion driven instabilities (≤500 kHz). Combined with the first FILD at ∼45° below the outer midplane [R. K. Fisher, et al., Rev. Sci. Instrum. 81, 10D307 (2010)], the two-detector system measures poloidal variation of losses. The phase space sensitivity of the new detector (gyroradius r(L) ∼ [1.5-8] cm and pitch angle α ∼ [35°-85°]) is calibrated using neutral beam first orbit loss measurements. Since fast ion losses are localized poloidally, having two FILDs at different poloidal locations allows for the study of losses over a wider range of plasma shapes and types of loss orbits.

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

DOE PAGES

Frerichs, H.; Schmitz, O.; Covele, B.; ...

2018-02-28

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the diverter slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frerichs, H.; Schmitz, O.; Covele, B.

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the diverter slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Numerical studies of nonlocal effects of the low mode numbers tearing modes and their mitigation in the DIII-D

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima; Evans, Todd

2006-10-01

In this work, the method of maps [1-4] is used to study the trajectories of magnetic field lines in the DIII-D tokamak. Data from the DIII-D shot 115467 is used to determine the parameters in the maps. Effects of the m=1, n=±1 tearing modes and the dipole perturbation from the C-coils on the motion of field lines are calculated. Internal tearing modes produce non-local effects on the magnetic footprints, and destroy their symmetry. Dipole perturbations mitigate the effects of the tearing modes, spread the heat-flux on the plates over a wider area, reduce the peak heat-flux, and reorganize the phase space structure in a new pattern that has the same symmetry as that of the external perturbation. The low dimensionality of the system and its symplecticity impose severe restrictions on the motion of the system in phase space forcing it to take on the symmetry properties of the perturbations. This work is done under the DOE grant number DE-FG02-01ER54624. 1. A. Punjabi, A. Boozer, and A. Verma, Phys. Rev. lett., 69, 3322 (1992). 2. H. Ali, A. Punjabi, and A. Boozer, Phys. Plasmas 11, 4527 (2004). 3. A. Punjabi, H. Ali, and A. Boozer, Phys. Plasmas 10, 3992 (2003). 4. A. Punjabi, H. Ali, and A. Boozer, Phys. Plasmas 4, 337 (1997).

Low to high confinement transition theory of finite-beta drift-wave driven shear flow and its comparison with data from DIII-D

NASA Astrophysics Data System (ADS)

Guzdar, P. N.; Kleva, R. G.; Groebner, R. J.; Gohil, P.

2004-03-01

Shear flow stabilization of edge turbulence in tokamaks has been the accepted paradigm for the improvement in confinement observed in high (H) confinement mode plasmas. Results on the generation of zonal flow and fields in finite β plasmas are presented. This theory yields a criterion for bifurcation from low to high (L-H) confinement mode, proportional to Te/√Ln , where Te is the electron temperature and Ln is the density scale-length at the steepest part of the density gradient. When this parameter exceeds a critical value (mostly determined by the strength of the toroidal magnetic field), the transition occurs. The predicted threshold based on this parameter shows good agreement with edge measurements on discharges undergoing L-H transitions in DIII-D [J. L. Luxon, R. Anderson, F. Batty et al., in Proceedings of the 11th Conference on Plasma Physics and Controlled Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. The observed differences in the transitions with the reversal of the toroidal magnetic field are reconciled in terms of this critical parameter due to the differences in the density gradient scale-lengths in the edge. The theory also provides a possible explanation for lowered threshold power, pellet injection H modes in DIII-D, thereby providing a unified picture of the varied observations on the L-H transition.

Comparisons of Measurements and Simulations of Turbulence and Transport for DIII-D Discharges with Off-Axis Modulated ECH

NASA Astrophysics Data System (ADS)

Bravenec, R. V.; Ross, D. W.; Austin, M. E.; Gentle, K. W.; Deboo, J. C.; DIII-D Team; McKee, G. R.; Dorland, W.; Rhodes, T. L.; Zeng, L.

2002-11-01

Experiments to elucidate the nature of electron thermal transport have been conducted in DIII-D plasmas using modulated off-axis electron-cyclotron heating (ECH). Density fluctuations were measured using beam-emission spectroscopy, microwave reflectometry, and far-infrared scattering. Simulations of the experiment are performed with the gyrokinetic and gyrofluid flux-tube codes GS2(F. Jenko, W. Dorland, M. Kotschenreuther, and B.N. Rogers, Phys. Plasmas 7), 1904 (2000) and refs. therein. and GRYFFIN,(W. Dorland and G.W. Hammett, Phys. Fluids B 5), 812 (1993); M.A. Beer and G.W. Hammett, Phys. Plasmas 3, 4046 (1996). respectively. Comparisons of experiment and simulation results for the fluctuations and transport fluxes (ion and electron) will be presented for both time-averaged and modulated quantities.

Design and first plasma measurements of the ITER-ECE prototype radiometer.

PubMed

Austin, M E; Brookman, M W; Rowan, W L; Danani, S; Bryerton, E W; Dougherty, P

2016-11-01

On ITER, second harmonic optically thick electron cyclotron emission (ECE) in the range of 220-340 GHz will supply the electron temperature (T e ). To investigate the requirements and capabilities prescribed for the ITER system, a prototype radiometer covering this frequency range has been developed by Virginia Diodes, Inc. The first plasma measurements with this instrument have been carried out on the DIII-D tokamak, with lab bench tests and measurements of third through fifth harmonic ECE from high T e plasmas. At DIII-D the instrument shares the transmission line of the Michelson interferometer and can simultaneously acquire data. Comparison of the ECE radiation temperature from the absolutely calibrated Michelson and the prototype receiver shows that the ITER radiometer provides accurate measurements of the millimeter radiation across the instrument band.

Synergy between fast-ion transport by core MHD and test blanket module fields in DIII-D experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidbrink, W. W.; Austin, M. E.; Collins, C. S.

2015-07-21

We measured fast-ion transport caused by the combination of MHD and a mock-up test-blanket module (TBM) coil in the DIII-D tokamak. The primary diagnostic is an infrared camera that measures the heat flux on the tiles surrounding the coil. The combined effects of the TBM and four other potential sources of transport are studied: neoclassical tearing modes, Alfvén eigenmodes, sawteeth, and applied resonant magnetic perturbation fields for the control of edge localized modes. A definitive synergistic effect is observed at sawtooth crashes where, in the presence of the TBM, the localized heat flux at a burst increases from 0.36 ±0.27more » to 2.6 ±0.5 MW/m -2.« less

Interpretation of rotation and momentum transport in the DIII-D edge plasma and comparison with neoclassical theory

DOE PAGES

Stacey, Weston M.; Grierson, Brian A.

2014-05-08

Here, a low-confinement mode discharge which optimizes the capability of the new main-ion chargeexchange-recombination spectroscopy system on DIII-D to measure deuterium toroidal velocity is interpretted in comparison with the predictions of neoclassical theory, with an emphasis on the plasma edge region. The observed peaking in the deuterium toroidal velocity near the separatrix is shown to be consistent with intrinsic co-rotation due to ion orbit loss. In general, the standard neoclassical toroidal and poloidal momentum transport rates are found to be smaller than those inferred from experiment, but a comparison has not yet been made with the more recent extended neoclassicalmore » theory that calculates the effects of poloidal asymmetries using an elongated flux surface representation.« less

Application of ECH to the study of transport in ITER baseline scenario-like discharges in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinsker, R. I.; Austin, M. E.; Ernst, D. R.

Recent DIII-D experiments in the ITER Baseline Scenario (IBS) have shown strong increases in fluctuations and correlated reduction of confinement associated with entering the electron-heating-dominated regime with strong electron cyclotron heating (ECH). The addition of 3.2 MW of 110 GHz EC power deposited at ρ~0.42 to IBS discharges with ~3 MW of neutral beam injection causes large increases in low-k and medium-k turbulent density fluctuations observed with Doppler backscatter (DBS), beam emission spectroscopy (BES) and phase-contrast imaging (PCI) diagnostics, correlated with decreases in the energy, particle, and momentum confinement times. Power balance calculations show the electron heat diffusivity χ emore » increases significantly in the mid-radius region 0.4« less

Application of ECH to the study of transport in ITER baseline scenario-like discharges in DIII-D

DOE PAGES

Pinsker, R. I.; Austin, M. E.; Ernst, D. R.; ...

2015-03-12

Recent DIII-D experiments in the ITER Baseline Scenario (IBS) have shown strong increases in fluctuations and correlated reduction of confinement associated with entering the electron-heating-dominated regime with strong electron cyclotron heating (ECH). The addition of 3.2 MW of 110 GHz EC power deposited at ρ~0.42 to IBS discharges with ~3 MW of neutral beam injection causes large increases in low-k and medium-k turbulent density fluctuations observed with Doppler backscatter (DBS), beam emission spectroscopy (BES) and phase-contrast imaging (PCI) diagnostics, correlated with decreases in the energy, particle, and momentum confinement times. Power balance calculations show the electron heat diffusivity χ emore » increases significantly in the mid-radius region 0.4« less

3D-DIVIMP(HC) code modeling of DIII-D DiMES porous plug injector experiments

NASA Astrophysics Data System (ADS)

Mu, Y.; Elder, J. D.; Stangeby, P. C.; McLean, A. G.

2011-08-01

A Porous Plug Injector (PPI) system for the Divertor Material Evaluation System (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. The 3D-DIVIMP(HC) code has been applied to the interpretation of the CI, CII and other spectroscopic measurements made at the PPI location, for (a) the synthetic source due to injection of CH4 through the PPI, and (b) the natural emission from the PPI head itself, which was inserted above surrounding graphite tiles by ˜0.3 mm.The code successfully replicated the MDS (spectrometer)-measured absolute emissions of CH, CI, CII 427 nm, 514 nm, and 658 nm [1] and the DiMES TV-measured spatial shapes of the CH, CI, and CII 514 nm [1] emission "clouds" to within the combined uncertainties. It is thus concluded that the most important physics and chemistry of chemical sputtering have most likely been included in the model.

Study of the impact of resonant magnetic perturbation fields on gross tungsten erosion using DiMES samples in DIII-D

NASA Astrophysics Data System (ADS)

Hinson, E. T.; Schmitz, O.; Frerichs, H.; Abrams, T.; Briesemeister, A.; Rudakov, D. L.; Unterberg, E. A.; Wampler, W. R.; Watkins, J. G.; Wang, H. Q.

2017-12-01

An experiment was conducted in DIII-D to compare gross tungsten (W) erosion on samples exposed to outer strike point (OSP) sweeps in L-mode plasmas for three conditions. These included two phases of resonant magnetic perturbations (RMPs), and a set with no perturbations. Upon RMP application, lobe structures indicative of strike point splitting of the OSP were evident in divertor camera data and on Langmuir probes. Gross W erosion flux, {{{Γ }}}{{W}}, inferred spectroscopically using the S/XB method applied to the 400.9 nm W-I line, was generally in the range {{{Γ }}}{{W}}/{{{Γ }}}{{D}+,\\perp }=2× {10}-4 referenced to incident deuterium ion flux {{{Γ }}}{{D}+,\\perp }, and was increased in the RMP cases by no more than 30% of the level observed in unperturbed discharges. A large reduction in gross erosion (50%) was observed in the private flux region at the W sample for one specific toroidal phase of the RMP field.

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

DOE PAGES

Pinsker, R. I.; Austin, M. E.; Diem, S. J.; ...

2014-02-12

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ~2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedlymore » strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. As a result, the AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.« less

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinsker, R. I.; Jackson, G. L.; Luce, T. C.

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ∼2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedlymore » strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. The AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.« less

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinsker, R. I.; Austin, M. E.; Diem, S. J.

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ~2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedlymore » strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. As a result, the AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.« less

Feedforward and feedback control of locked mode phase and rotation in DIII-D with application to modulated ECCD experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, W.; La Haye, R. J.; Lanctot, M. J.

The toroidal phase and rotation of otherwise locked magnetic islands of toroidal mode number n=1 are controlled in the DIII-D tokamak by means of applied magnetic perturbations of n=1. Pre-emptive perturbations were applied in feedforward to "catch" the mode as it slowed down and entrain it to the rotating field before complete locking, thus avoiding the associated major confinement degradation. Additionally, for the first time, the phase of the perturbation was optimized in real-time, in feedback with magnetic measurements, in order for the mode’s phase to closely match a prescribed phase, as a function of time. Experimental results confirm themore » capability to hold the mode in a given fixed-phase or to rotate it at up to 20 Hz with good uniformity. The controlcoil currents utilized in the experiments agree with the requirements estimated by an electromechanical model. Moreover, controlled rotation at 20 Hz was combined with Electron Cyclotron Current Drive (ECCD) modulated at the same frequency. This is simpler than regulating the ECCD modulation in feedback with spontaneous mode rotation, and enables repetitive, reproducible ECCD deposition at or near the island O-point, X-point and locations in between, for careful studies of how this affects the island stability. Current drive was found to be radially misaligned relative to the island, and resulting growth and shrinkage of islands matched expectations of the Modified Rutherford Equation for some discharges presented here. Finally, simulations predict the as designed ITER 3D coils can entrain a small island at sub-10 Hz frequencies.« less

Feedforward and feedback control of locked mode phase and rotation in DIII-D with application to modulated ECCD experiments

NASA Astrophysics Data System (ADS)

Choi, W.; La Haye, R. J.; Lanctot, M. J.; Olofsson, K. E. J.; Strait, E. J.; Sweeney, R.; Volpe, F. A.; The DIII-D Team

2018-03-01

The toroidal phase and rotation of otherwise locked magnetic islands of toroidal mode number n  =  1 are controlled in the DIII-D tokamak by means of applied magnetic perturbations of n  =  1. Pre-emptive perturbations were applied in feedforward to ‘catch’ the mode as it slowed down and entrain it to the rotating field before complete locking, thus avoiding the associated major confinement degradation. Additionally, for the first time, the phase of the perturbation was optimized in real-time, in feedback with magnetic measurements, in order for the mode’s phase to closely match a prescribed phase, as a function of time. Experimental results confirm the capability to hold the mode in a given fixed-phase or to rotate it at up to 20 Hz with good uniformity. The control-coil currents utilized in the experiments agree with the requirements estimated by an electromechanical model. Moreover, controlled rotation at 20 Hz was combined with electron cyclotron current drive (ECCD) modulated at the same frequency. This is simpler than regulating the ECCD modulation in feedback with spontaneous mode rotation, and enables repetitive, reproducible ECCD deposition at or near the island O-point, X-point and locations in between, for careful studies of how this affects the island stability. Current drive was found to be radially misaligned relative to the island, and resulting growth and shrinkage of islands matched expectations of the modified Rutherford equation for some discharges presented here. Finally, simulations predict the as designed ITER 3D coils can entrain a small island at sub-10 Hz frequencies.

Feedforward and feedback control of locked mode phase and rotation in DIII-D with application to modulated ECCD experiments

DOE PAGES

Choi, W.; La Haye, R. J.; Lanctot, M. J.; ...

2018-02-05

The toroidal phase and rotation of otherwise locked magnetic islands of toroidal mode number n=1 are controlled in the DIII-D tokamak by means of applied magnetic perturbations of n=1. Pre-emptive perturbations were applied in feedforward to "catch" the mode as it slowed down and entrain it to the rotating field before complete locking, thus avoiding the associated major confinement degradation. Additionally, for the first time, the phase of the perturbation was optimized in real-time, in feedback with magnetic measurements, in order for the mode’s phase to closely match a prescribed phase, as a function of time. Experimental results confirm themore » capability to hold the mode in a given fixed-phase or to rotate it at up to 20 Hz with good uniformity. The controlcoil currents utilized in the experiments agree with the requirements estimated by an electromechanical model. Moreover, controlled rotation at 20 Hz was combined with Electron Cyclotron Current Drive (ECCD) modulated at the same frequency. This is simpler than regulating the ECCD modulation in feedback with spontaneous mode rotation, and enables repetitive, reproducible ECCD deposition at or near the island O-point, X-point and locations in between, for careful studies of how this affects the island stability. Current drive was found to be radially misaligned relative to the island, and resulting growth and shrinkage of islands matched expectations of the Modified Rutherford Equation for some discharges presented here. Finally, simulations predict the as designed ITER 3D coils can entrain a small island at sub-10 Hz frequencies.« less

Bifurcation of quiescent H-mode to a wide pedestal regime in DIII-D and advances in the understanding of edge harmonic oscillations

NASA Astrophysics Data System (ADS)

Chen, Xi; Burrell, K. H.; Osborne, T. H.; Barada, K.; Ferraro, N. M.; Garofalo, A. M.; Groebner, R. J.; McKee, G. R.; Petty, C. C.; Porkolab, M.; Rhodes, T. L.; Rost, J. C.; Snyder, P. B.; Solomon, W. M.; Yan, Z.; The DIII-D Team

2017-08-01

New experimental studies and modelling of the coherent edge harmonic oscillation (EHO), which regulates the conventional Quiescent H-mode (QH-mode) edge, validate the proposed hypothesis of edge rotational shear in destabilizing the low-n kink-peeling mode as the additional drive mechanism for the EHO. The observed minimum edge E  ×  B shear required for the EHO decreases linearly with pedestal collisionality ν \\text{e}\\ast , which is favorable for operating QH-mode in machines with low collisionality and low rotation such as ITER. In addition, the QH-mode regime in DIII-D has recently been found to bifurcate into a new ‘wide-pedestal’ state at low torque in double-null shaped plasmas, characterized by increased pedestal height, width and thermal energy confinement (Burrell 2016 Phys. Plasmas 23 056103, Chen 2017 Nucl. Fusion 57 022007). This potentially provides an alternate path for achieving high performance ELM-stable operation at low torque, in addition to the low-torque QH-mode sustained with applied 3D fields. Multi-branch low-k and intermediate-k turbulences are observed in the ‘wide-pedestal’. New experiments support the hypothesis that the decreased edge E  ×  B shear enables destabilization of broadband turbulence, which relaxes edge pressure gradients, improves peeling-ballooning stability and allows a wider and thus higher pedestal. The ability to accurately predict the critical E  ×  B shear for EHO and maintain high performance QH-mode at low torque is an essential requirement for projecting QH-mode operation to ITER and future machines.

Parallel Energy Transport in Detached DIII-D Divertor Plasmas

NASA Astrophysics Data System (ADS)

Leonard, A. W.; Lore, J. D.; Canik, J. M.; McLean, A. G.; Makowski, M. A.

2017-10-01

A comparison of experiment and modeling of detached divertor plasmas is examined in the context of parallel energy transport. Experimental estimates of power carried by electron thermal conduction versus plasma convection are experimentally inferred from power balance measurements of radiated power and target plate heat flux combined with Thomson scattering measurements of the Te profile along the divertor leg. Experimental profiles of Te exhibit relatively low gradients with Te < 15 eV from the X-point to the target implying transport dominated by convection. In contrast, fluid modeling with SOLPS produces sharp Te gradients for Te > 3 eV, characteristic of transport dominated by electron conduction through the bulk of the divertor. This discrepancy with experimental transport dominated by convection and modeling by conduction has significant implications for the radiative capacity of divertor plasmas and may explain at least part of the difficulty for fluid modeling to obtain the experimentally observed radiative losses. Comparisons are also made for helium plasmas where the match between experiment and modeling is much better. Work supported by the US DOE under DE-FC02-04ER54698.

Scenario development for high β p low torque plasma with q min above 2 and large-radius internal transport barrier in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ding, S.; Xu, G. S.; Wang, Q.

A recent experiment on DIII-D, which was conducted by the joint research team from DIII-D and EAST, has extended the previous high β p, high q min regime, which has been tested in the 2013 DIII-D/EAST joint experiment, to inductive operation at higher plasma current (I p=0.8 MA) and significantly higher normalized fusion performance (G = H 89β N/=qmore » $$2\\atop{95}$$ = 0.16). The experiment aims at exploring high performance scenario with q min > 2 and reduced torque for long pulse operation, which can be potentially extrapolated to EAST. The effort was largely motivated by the interest in developing a feasible scenario for long-pulse high performance operation with low torque on EAST. Very high confinement, H 89 = 3.5 or H 98,y2 = 2.1 with β N ~ 3.0, has been achieved transiently in this experiment together with q min > 2 and reduced NBI torque (3~5 N-m). The excellent confinement is associated with the spontaneous formation of an internal transport barrier (ITB) in plasmas with I p=0.8 MA at large minor radius (normalized p ~ 0.7) in all channels (n e, T e, T i, V Φ, especially strong in the T e channel). Fluctuation measurements show a significant reduction in the fluctuation levels, including AE modes and broadband turbulence, at the location where an ITB forms. Linear gyrokinetic simulations also support the decrease of the growth rate of the most unstable mode during strong ITB formation. The simulation implies that strong suppression of turbulence and a positive feedback loop may be active in this process and is responsible for the spontaneous formation of large-radius ITB. Finally, in an unstable ITB phase, an ELM crash is observed to have a positive effect on transient formation of large-radius ITB. The formation of this kind of ITB is found to have a shielding (protecting) effect on the core plasma while isolating the perturbation due to ELM crash.« less

SOL effects on the pedestal structure in DIII-D discharges

DOE PAGES

Sontag, Aaron C.; Chen, Xi; Canik, John; ...

2017-05-24

SOLPS analysis explains the differences in pedestal structure associated with different ion ∇B drift directions in DIII-D. Core transport models predict that fusion power scales roughly as the square of the pressure at the top of the pedestal, so understanding the effects that determine pedestal structure in steady-state operational scenarios is important to projecting scenarios developed in DIII-D to ITER and other devices. Both experiments and modeling indicate that scrape off layer (SOL) conditions are important in optimizing the pedestal structure for high-beta steady-state scenarios. The SOLPS code is used to provide interpretive analysis of the pedestal and SOL tomore » examine the nature of flows and fueling on the pedestal structure including the effects of drifts in the fluid model. This analysis shows that flows driven by the ion ∇B drift are outward when this drift is toward the x-point in a single-null divertor configuration (favorable ∇B direction for reduced H-mode power threshold), and inward when the drift is away from the x-point (unfavorable ∇B direction). It is hypothesized that these flows decrease the density gradient in the pedestal in the favorable direction, thereby stabilizing the kinetic ballooning mode (KBM) and increasing the pedestal width. Comparisons of pedestal structures in similarly shaped DIII-D steady-state plasmas confirm this change, showing increased density pedestal width and lower peak density and lower separatrix density with the favorable drift direction. The pedestal temperature is higher in the lower density case, resulting in an increased pedestal pressure, which indicates that the increased particle flux does not significantly degrade energy confinement. Modeling of cases with constant ∇B drift direction but changing between the more open lower divertor and more closed upper divertor show that there is increased fueling inside the pedestal with the more open geometry. As a result, the pedestal fueling rate for both attached and detached cases is always lower with more closed divertor geometry than in any cases with more open geometry.« less

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2018-02-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.« less

Fast Ion Transport Studies in DIII-D High βN Steady-State Scenarios

NASA Astrophysics Data System (ADS)

Holcomb, C. T.

2014-10-01

DIII-D research is identifying paths to optimize energetic particle (EP) transport in high βN steady-state tokamak scenarios. Operation with qmin > 2 is predicted to achieve high βN, confinement, and bootstrap fraction. However DIII-D experiments have shown that Alfvén eigenmodes (AE) and correlated EP transport can limit the performance of some qmin > 2 plasmas. Enhanced EP transport occurs in plasmas with qmin = 2-2.5, q95 = 5-7, and relatively long slowing down time. Strong AEs are present, the confinement factor H89 = 1.6-1.8 and βN is limited to ~3 by the available power. These observations are consistent with EP transport models having a critical gradient in βf. However, adjusting the parameters can recover classical EP confinement or improve thermal confinement so that H89 > 2 . One example is a scenario with βP and βN ~ 3 . 2 , qmin > 3 and q95 ~ 11 developed to test control of long pulse, high heat flux operation on devices like EAST. This has an internal transport barrier at ρ ~ 0 . 7 , bootstrap fraction >75%, density limit fraction ~1, and H89 >= 2 . In these cases AE activity and EP transport is very dynamic - it varies between classical and anomalous from shot to shot and within shots. Thus these plasmas are close to a threshold for enhanced EP transport. This may be governed by a combination of a relatively low ∇βfast due to good thermal confinement and lower beam power, short slowing down time, and possibly changes to the q-profile. Another example is scenarios with qmin ~ 1.1. These typically have classical EP confinement and good thermal confinement. Thus by using its flexible parameters and profile control tools DIII-D is comparing a wide range of steady-state scenarios to identify the key physics setting EP transport. Work supported by the US Department of Energy under DE-AC52-07NA27344, SC-G903402, DE-FC02-04ER54698, and DE-AC02-09CH11466.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.« less

SOL effects on the pedestal structure in DIII-D discharges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sontag, Aaron C.; Chen, Xi; Canik, John

SOLPS analysis explains the differences in pedestal structure associated with different ion ∇B drift directions in DIII-D. Core transport models predict that fusion power scales roughly as the square of the pressure at the top of the pedestal, so understanding the effects that determine pedestal structure in steady-state operational scenarios is important to projecting scenarios developed in DIII-D to ITER and other devices. Both experiments and modeling indicate that scrape off layer (SOL) conditions are important in optimizing the pedestal structure for high-beta steady-state scenarios. The SOLPS code is used to provide interpretive analysis of the pedestal and SOL tomore » examine the nature of flows and fueling on the pedestal structure including the effects of drifts in the fluid model. This analysis shows that flows driven by the ion ∇B drift are outward when this drift is toward the x-point in a single-null divertor configuration (favorable ∇B direction for reduced H-mode power threshold), and inward when the drift is away from the x-point (unfavorable ∇B direction). It is hypothesized that these flows decrease the density gradient in the pedestal in the favorable direction, thereby stabilizing the kinetic ballooning mode (KBM) and increasing the pedestal width. Comparisons of pedestal structures in similarly shaped DIII-D steady-state plasmas confirm this change, showing increased density pedestal width and lower peak density and lower separatrix density with the favorable drift direction. The pedestal temperature is higher in the lower density case, resulting in an increased pedestal pressure, which indicates that the increased particle flux does not significantly degrade energy confinement. Modeling of cases with constant ∇B drift direction but changing between the more open lower divertor and more closed upper divertor show that there is increased fueling inside the pedestal with the more open geometry. As a result, the pedestal fueling rate for both attached and detached cases is always lower with more closed divertor geometry than in any cases with more open geometry.« less

Scenario development for high β p low torque plasma with q min above 2 and large-radius internal transport barrier in DIII-D

DOE PAGES

Ding, S.; Xu, G. S.; Wang, Q.; ...

2016-09-30

A recent experiment on DIII-D, which was conducted by the joint research team from DIII-D and EAST, has extended the previous high β p, high q min regime, which has been tested in the 2013 DIII-D/EAST joint experiment, to inductive operation at higher plasma current (I p=0.8 MA) and significantly higher normalized fusion performance (G = H 89β N/=qmore » $$2\\atop{95}$$ = 0.16). The experiment aims at exploring high performance scenario with q min > 2 and reduced torque for long pulse operation, which can be potentially extrapolated to EAST. The effort was largely motivated by the interest in developing a feasible scenario for long-pulse high performance operation with low torque on EAST. Very high confinement, H 89 = 3.5 or H 98,y2 = 2.1 with β N ~ 3.0, has been achieved transiently in this experiment together with q min > 2 and reduced NBI torque (3~5 N-m). The excellent confinement is associated with the spontaneous formation of an internal transport barrier (ITB) in plasmas with I p=0.8 MA at large minor radius (normalized p ~ 0.7) in all channels (n e, T e, T i, V Φ, especially strong in the T e channel). Fluctuation measurements show a significant reduction in the fluctuation levels, including AE modes and broadband turbulence, at the location where an ITB forms. Linear gyrokinetic simulations also support the decrease of the growth rate of the most unstable mode during strong ITB formation. The simulation implies that strong suppression of turbulence and a positive feedback loop may be active in this process and is responsible for the spontaneous formation of large-radius ITB. Finally, in an unstable ITB phase, an ELM crash is observed to have a positive effect on transient formation of large-radius ITB. The formation of this kind of ITB is found to have a shielding (protecting) effect on the core plasma while isolating the perturbation due to ELM crash.« less

Immunogenic Subviral Particles Displaying Domain III of Dengue 2 Envelope Protein Vectored by Measles Virus

PubMed Central

Harahap-Carrillo, Indira S.; Ceballos-Olvera, Ivonne; Reyes-del Valle, Jorge

2015-01-01

Vaccines against dengue virus (DV) are commercially nonexistent. A subunit vaccination strategy may be of value, especially if a safe viral vector acts as biologically active adjuvant. In this paper, we focus on an immunoglobulin-like, independently folded domain III (DIII) from DV 2 envelope protein (E), which contains epitopes that elicits highly specific neutralizing antibodies. We modified the hepatitis B small surface antigen (HBsAg, S) in order to display DV 2 DIII on a virus-like particle (VLP), thus generating the hybrid antigen DIII-S. Two varieties of measles virus (MV) vectors were developed to express DIII-S. The first expresses the hybrid antigen from an additional transcription unit (ATU) and the second additionally expresses HBsAg from a separate ATU. We found that this second MV vectoring the hybrid VLPs displaying DIII-S on an unmodified HBsAg scaffold were immunogenic in MV-susceptible mice (HuCD46Ge-IFNarko), eliciting robust neutralizing responses (averages) against MV (1:1280 NT90), hepatitis B virus (787 mIU/mL), and DV2 (1:160 NT50) in all of the tested animals. Conversely, the MV vector expressing only DIII-S induced immunity against MV alone. In summary, DV2 neutralizing responses can be generated by displaying E DIII on a scaffold of HBsAg-based VLPs, vectored by MV. PMID:26350592

Validating predictive models for fast ion profile relaxation in burning plasmas

NASA Astrophysics Data System (ADS)

Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; Lestz, J. B.; Podesta, M.; Van Zeeland, M. A.; White, R. B.

2016-11-01

The redistribution and potential loss of energetic particles due to MHD modes can limit the performance of fusion plasmas by reducing the plasma heating rate. In this work, we present validation studies of the 1.5D critical gradient model (CGM) for Alfvén eigenmode (AE) induced EP transport in NSTX and DIII-D neutral beam heated plasmas. In previous comparisons with a single DIII-D L-mode case, the CGM model was found to be responsible for 75% of measured AE induced neutron deficit [1]. A fully kinetic HINST is used to compute mode stability for the non-perturbative version of CGM (or nCGM). We have found that AEs show strong local instability drive up to γ /ω ∼ 20% violating assumptions of perturbative approaches used in NOVA-K code. We demonstrate that both models agree with each other and both underestimate the neutron deficit measured in DIII-D shot by approximately a factor of 2. On the other hand in NSTX the application of CGM shows good agreement for the measured flux deficit predictions. We attempt to understand these results with the help of the so-called kick model which is based on the guiding center code ORBIT. The kick model comparison gives important insight into the underlying velocity space dependence of the AE induced EP transport as well as it allows the estimate of the neutron deficit in the presence of the low frequency Alfvénic modes. Within the limitations of used models we infer that there are missing modes in the analysis which could improve the agreement with the experiments.

Investigation of Physical Processes Limiting Plasma Density in DIII--D

NASA Astrophysics Data System (ADS)

Maingi, R.

1996-11-01

Understanding the physical processes which limit operating density is crucial in achieving peak performance in confined plasmas. Studies from many of the world's tokamaks have indicated the existence(M. Greenwald, et al., Nucl. Fusion 28) (1988) 2199 of an operational density limit (Greenwald limit, n^GW_max) which is proportional to the plasma current and independent of heating power. Several theories have reproduced the current dependence, but the lack of a heating power dependence in the data has presented an enigma. This limit impacts the International Thermonuclear Experimental Reactor (ITER) because the nominal operating density for ITER is 1.5 × n^GW_max. In DIII-D, experiments are being conducted to understand the physical processes which limit operating density in H-mode discharges; these processes include X-point MARFE formation, high core recycling and neutral pressure, resistive MHD stability, and core radiative collapse. These processes affect plasma properties, i.e. edge/scrape-off layer conduction and radiation, edge pressure gradient and plasma current density profile, and core radiation, which in turn restrict the accessible density regime. With divertor pumping and D2 pellet fueling, core neutral pressure is reduced and X-point MARFE formation is effectively eliminated. Injection of the largest-sized pellets does cause transient formation of divertor MARFEs which occasionally migrate to the X-point, but these are rapidly extinguished in pumped discharges in the time between pellets. In contrast to Greenwald et al., it is found that the density relaxation time after pellets is largely independent of the density relative to the Greenwald limit. Fourier analysis of Mirnov oscillations indicates the de-stabilization and growth of rotating, tearing-type modes (m/n= 2/1) when the injected pellets cause large density perturbations, and these modes often reduce energy confinement back to L-mode levels. We are examining the mechanisms for de-stabilization of the mode, the primary ones being neo-classical pressure gradient drivers. Discharges with a gradual density increase are often free of large amplitude tearing modes, allowing access to the highest density regimes in which off-axis beam deposition can lead to core radiative collapse, i.e. a central power balance limit. The highest achieved barne was 1.5 × n^GW_max with τ_E/τ_E^JET-DIII-D >= 0.9. The highest density obtained in L-mode discharges was 3 × n^GW_max. Implications of these results for ITER will be discussed.

Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

DOE PAGES

Rognlien, Thomas D.; McLean, Adam G.; Fenstermacher, Max E.; ...

2017-01-27

A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult H-mode regime. The data set, which spans a range of plasmas densities for both forward and reverse toroidal magnetic field (B t) over a range of plasma densities, is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (T e) and density (n e) across both divertor legs for individual discharges. The calculations show the same features of in/out plasma asymmetries as measured inmore » the experiment, with the normal B t direction (ion ∇B drift toward the X-point) having higher n e and lower T e in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. Furthermore, these 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.« less

Recent sheath physics studies on DIII-D

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Labombard, B.; Stangeby, P. C.; Lasnier, C. J.; McLean, A. G.; Nygren, R. E.; Boedo, J. A.; Leonard, A. W.; Rudakov, D. L.

2015-08-01

A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

Design and first plasma measurements of the ITER-ECE prototype radiometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Austin, M. E.; Brookman, M. W.; Rowan, W. L.

2016-11-15

On ITER, second harmonic optically thick electron cyclotron emission (ECE) in the range of 220-340 GHz will supply the electron temperature (T{sub e}). To investigate the requirements and capabilities prescribed for the ITER system, a prototype radiometer covering this frequency range has been developed by Virginia Diodes, Inc. The first plasma measurements with this instrument have been carried out on the DIII-D tokamak, with lab bench tests and measurements of third through fifth harmonic ECE from high T{sub e} plasmas. At DIII-D the instrument shares the transmission line of the Michelson interferometer and can simultaneously acquire data. Comparison of themore » ECE radiation temperature from the absolutely calibrated Michelson and the prototype receiver shows that the ITER radiometer provides accurate measurements of the millimeter radiation across the instrument band.« less

Improved edge charge exchange recombination spectroscopy in DIII-D

NASA Astrophysics Data System (ADS)

Chrystal, C.; Burrell, K. H.; Grierson, B. A.; Haskey, S. R.; Groebner, R. J.; Kaplan, D. H.; Briesemeister, A.

2016-11-01

The charge exchange recombination spectroscopy diagnostic on the DIII-D tokamak has been upgraded with the addition of more high radial resolution view chords near the edge of the plasma (r/a > 0.8). The additional views are diagnosed with the same number of spectrometers by placing fiber optics side-by-side at the spectrometer entrance with a precise separation that avoids wavelength shifted crosstalk without the use of bandpass filters. The new views improve measurement of edge impurity parameters in steep gradient, H-mode plasmas with many different shapes. The number of edge view chords with 8 mm radial separation has increased from 16 to 38. New fused silica fibers have improved light throughput and clarify the observation of non-Gaussian spectra that suggest the ion distribution function can be non-Maxwellian in low collisionality plasmas.

Improved edge charge exchange recombination spectroscopy in DIII-D.

PubMed

Chrystal, C; Burrell, K H; Grierson, B A; Haskey, S R; Groebner, R J; Kaplan, D H; Briesemeister, A

2016-11-01

The charge exchange recombination spectroscopy diagnostic on the DIII-D tokamak has been upgraded with the addition of more high radial resolution view chords near the edge of the plasma (r/a > 0.8). The additional views are diagnosed with the same number of spectrometers by placing fiber optics side-by-side at the spectrometer entrance with a precise separation that avoids wavelength shifted crosstalk without the use of bandpass filters. The new views improve measurement of edge impurity parameters in steep gradient, H-mode plasmas with many different shapes. The number of edge view chords with 8 mm radial separation has increased from 16 to 38. New fused silica fibers have improved light throughput and clarify the observation of non-Gaussian spectra that suggest the ion distribution function can be non-Maxwellian in low collisionality plasmas.

Evolution of Edge Pedestal Profiles Between ELMs

NASA Astrophysics Data System (ADS)

Floyd, J. P.; Stacey, W. M.; Groebner, R. J.

2012-10-01

The measured edge profile evolution in DIII-D discharges is analyzed in terms of the implied thermal diffusivities, ion diffusion coefficients and pinch velocities, using the momentum-balance methodology of Ref. [1], extended to take into account ion orbit loss and X-point loss. The evolution of the density, temperature, rotation and radial electric field profiles in the edge pedestal between edge localized modes (ELMs) provides information of these diffusive and non-diffusive transport processes in the pedestal of H-mode plasmas. This methodology is incorporated in the GTEDGE code developed for DIII-D data interpretation. Using a smaller integration time for the charge exchange recombination measurements than in Ref. [1] allows a more detailed examination of the time evolution of the ion temperature and rotation profiles. 6pt [1] W.M. Stacey and R.J. Groebner, Nucl. Fusion 51, 063024 (2011).

Initial results of the high resolution edge Thomson scattering upgrade at DIII-D.

PubMed

Eldon, D; Bray, B D; Deterly, T M; Liu, C; Watkins, M; Groebner, R J; Leonard, A W; Osborne, T H; Snyder, P B; Boivin, R L; Tynan, G R

2012-10-01

Validation of models of pedestal structure is an important part of predicting pedestal height and performance in future tokamaks. The Thomson scattering diagnostic at DIII-D has been upgraded in support of validating these models. Spatial and temporal resolution, as well as signal to noise ratio, have all been specifically enhanced in the pedestal region. This region is now diagnosed by 20 view-chords with a spacing of 6 mm and a scattering length of just under 5 mm sampled at a nominal rate of 250 Hz. When mapped to the outboard midplane, this corresponds to ~3 mm spacing. These measurements are being used to test critical gradient models, in which pedestal gradients increase in time until a threshold is reached. This paper will describe the specifications of the upgrade and present initial results of the system.

A method for determining poloidal rotation from poloidal asymmetry in toroidal rotation (invited)

DOE PAGES

Chrystal, Chrystal; Burrell, Keith H.; Grierson, Brian A.; ...

2014-08-08

A new diagnostic has been developed on DIII-D that determines the impurity poloidal rotation from the poloidal asymmetry in the toroidal angular rotation velocity. This asymmetry is measured with recently added tangential charge exchange viewchords on the high-field side of the tokamak midplane. Measurements are made on co- and counter-current neutral beams, allowing the charge exchange cross section effect to be measured and eliminating the need for atomic physics calculations. The diagnostic implementation on DIII-D restricts the measurement range to the core (r/a < 0.6) where, relative to measurements made with the vertical charge exchange system, the spatial resolution ismore » improved. Furthermore, significant physics results have been obtained with this new diagnostic; for example, poloidal rotation measurements that significantly exceed neoclassical predictions.« less

Advances in the steady-state hybrid regime in DIII-D – a fully non-inductive, ELM-suppressed scenario for ITER

DOE PAGES

Petty, Craig C.; Nazikian, Raffi; Park, Jin Myung; ...

2017-07-19

Here, the hybrid regime with beta, collisionality, safety factor and plasma shape relevant to the ITER steady-state mission has been successfully integrated with ELM suppression by applying an odd parity n=3 resonant magnetic perturbation (RMP). Fully non-inductive hybrids in the DIII-D tokamak with high beta (β ≤ 2.8%) and high confinement (98y2 ≤ 1.4) in the ITER similar shape have achieved zero surface loop voltage for up to two current relaxation times using efficient central current drive from ECCD and NBCD. The n=3 RMP causes surprisingly little increase in thermal transport during ELM suppression. Poloidal magnetic flux pumping in hybridmore » plasmas maintains q above 1 without loss of current drive efficiency, except that experiments show that extremely peaked ECCD profiles can create sawteeth. During ECCD, Alfvén eigenmode (AE) activity is replaced by a more benign fishbone-like mode, reducing anomalous beam ion diffusion by a factor of 2. While the electron and ion thermal diffusivities substantially increase with higher ECCD power, the loss of confinement can be offset by the decreased fast ion transport resulting from AE suppression. Extrapolations from DIII-D along a dimensionless parameter scaling path as well as those using self-consistent theory-based modeling show that these ELM-suppressed, fully non-inductive hybrids can achieve the Q = 5 ITER steady-state mission.« less

Advances in understanding of high- Z material erosion and re-deposition in low- Z wall environment in DIII-D

DOE PAGES

Ding, R.; Rudakov, D. L.; Stangeby, P. C.; ...

2017-03-24

Dedicated DIII-D experiments coupled with modeling reveal that the net erosion rate of high-Z materials, i.e. Mo and W, is strongly affected by carbon concentration in the plasma and the magnetic pre-sheath properties. We have investigated different methods such as electrical biasing and local gas injection to control high-Z material erosion. The net erosion rate of high-Z materials is significantly reduced due to the high local re-deposition ratio. The ERO modeling shows that the local re-deposition ratio is mainly controlled by the electric field and plasma density within the magnetic pre-sheath. The net erosion can be significantly suppressed by reducingmore » the sheath potential drop. A high carbon impurity concentration in the background plasma is also found to reduce the net erosion rate of high-Z materials. Both DIII-D experiments and modeling show that local 13CH 4 injection can create a carbon coating on the metal surface. The profile of 13C deposition provides quantitative information on radial transport due to E × B drift and the cross-field diffusion. The deuterium gas injection upstream of the W sample can reduce W net erosion rate by plasma perturbation. The inter-ELM W erosion we measured in H-mode plasmas, rates at different radial locations are well reproduced by ERO modeling taking into account charge-state-resolved carbon ion flux in the background plasma calculated using the OEDGE code.« less

High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak.

PubMed

Truong, D D; Austin, M E

2014-11-01

The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of Te(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83-130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1-3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6-0.8 cm) resolution Te measurements. The high resolution subsystem branches off from the regular channels' IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2-4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters' center frequencies (250 MHz). This configuration allows for full coverage of the 83-130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a "zoomed-in" analysis of a ∼2-4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial Te measurements, which demonstrate that the desired resolution is achieved, are presented.

High Field Side Lower Hybrid Current Drive Launcher Design for DIII-D

NASA Astrophysics Data System (ADS)

Wallace, G. M.; Leccacori, R.; Doody, J.; Vieira, R.; Shiraiwa, S.; Wukitch, S. J.; Holcomb, C.; Pinsker, R. I.

2017-10-01

Efficient off-axis current drive scalable to reactors is a key enabling technology for a steady-state tokamak. Simulations of DIII-D discharges have identified high performance scenarios with excellent lower hybrid (LH) wave penetration, single pass absorption and high current drive efficiency. The strategy was to adapt known launching technology utilized in previous experiments on C-Mod (poloidal splitter) and Tore Supra (bi-junction) and remain within power density limits established in JET and Tore Supra. For a 2 MW source power antenna, the launcher consists of 32 toroidal apertures and 4 poloidal rows. The aperture is 60 mm x 5 mm with 1 mm septa and the peak n| | is 2.7+/-0.2 for 90□ phasing. Eight WR187 waveguides are routed from the R-1 port down under the lower cryopump, under the existing divertor, and up the central column with the long waveguide dimension along the vacuum vessel. Above the inner strike point region, each waveguide is twisted to orient the long dimension perpendicular to the vacuum vessel and splits into 4 toroidal apertures via bi-junctions. To protect the waveguide, the inner wall radius will need to increase by 2.5 cm. RF, disruption, and thermal analysis of the latest design will be presented. Work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using User Facility DIII-D, under Award Number DE-FC02-04ER54698 and by MIT PSFC cooperative agreement DE-SC0014264.

Modeling of fast neutral-beam-generated ions and rotation effects on RWM stability in DIII-D plasmas

DOE PAGES

Turco, Francesca; Turnbull, Alan D.; Hanson, Jeremy M.; ...

2015-10-15

Here, validation results for the MARS-K code for DIII-D equilibria, predict that the absence of fast Neutral Beam (NB) generated ions leads to a plasma response ~40–60% higher than in NB-sustained H-mode plasmas when the no-wall β N limit is reached. In a β N scan, the MARS-K model with thermal and fast-ions, reproduces the experimental measurements above the no-wall limit, except at the highest β N where the phase of the plasma response is overestimated. The dependencies extrapolate unfavorably to machines such as ITER with smaller fast ion fractions since elevated responses in the absence of fast ions indicatemore » the potential onset of a resistive wall mode (RWM). The model was also tested for the effects of rotation at high β N, and recovers the measured response even when fast-ions are neglected, reversing the effect found in lower β N cases, but consistent with the higher β N results above the no-wall limit. The agreement in the response amplitude and phase for the rotation scan is not as good, and additional work will be needed to reproduce the experimental trends. In the case of current-driven instabilities, the magnetohydrodynamic spectroscopy system used to measure the plasma response reacts differently from that for pressure driven instabilities: the response amplitude remains low up to ~93% of the current limit, showing an abrupt increase only in the last ~5% of the current ramp. This makes it much less effective as a diagnostic for the approach to an ideal limit. However, the mode structure of the current driven RWM extends radially inwards, consistent with that in the pressure driven case for plasmas with q edge~2. This suggests that previously developed RWM feedback techniques together with the additional optimizations that enabled q edge~2 operation, can be applied to control of both current-driven and pressure-driven modes at high β N.« less

Understanding the stability of the low torque ITER Baseline Scenario in DIII-D

NASA Astrophysics Data System (ADS)

Turco, Francesca

2017-10-01

Analysis of the evolving current density (J), pedestal and rotation profiles in a database of 200 ITER Baseline Scenario discharges in the DIII-D tokamak sheds light on the cause of the disruptive instability limiting both high and low torque operation of these plasmas. The m =2/n =1 tearing modes, occurring after several pressure-relaxation times, are related to the shape of the current profile in the outer region of the plasma. The q =2 surface is located just inside the current pedestal, near a minimum in J. This well in J deepens at constant betaN and at lower rotation, causing the equilibrium to evolve towards a classically unstable state. Lack of core-edge differential rotation likely biases the marginal point towards instability during the secular trend in J. New results from the 2017 experimental campaign establish the first reproducible, stable operation at T =0 Nm for this scenario. A new ramp-up recipe with delayed heating keeps the discharges stable without the need for ECCD stabilization. The J profile shape in the new shots is consistent with an expansion of the previous ``shallow well'' stable operational space. Realtime Active MHD Spectroscopy (AMS) has been applied to IBS plasmas for the first time, and the plasma response measurements show that the AMS can help sense the approach to instability during the discharges. The AMS data shows the trend towards instability at low rotation, and MARS-K modelling partially reproduces the experimental trend if collisionality and resistivity are included. The modelling results are sensitive to the edge resistivity, and this can indicate that the AMS is measuring the changes in ideal (kink) stability, to which the tearing stability index delta' is correlated. Together these results constitute a crucial step to acquire physical understanding and sensing capability for the MHD stability in the Q =10 ITER scenario. Work supported by US DOE under DE-FC02-04ER54698 and DE-FG02-04ER54761.

ITER L-mode confinement database

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaye, S.M.

This paper describes the content of an L-mode database that has been compiled with data from Alcator C-Mod, ASDEX, DIII, DIII-D, FTU, JET, JFT-2M, JT-60, PBX-M, PDX, T-10, TEXTOR, TFTR, and Tore-Supra. The database consists of a total of 2938 entries, 1881 of which are in the L-phase while 922 are ohmically heated only (OH). Each entry contains up to 95 descriptive parameters, including global and kinetic information, machine conditioning, and configuration. The paper presents a description of the database and the variables contained therein, and it also presents global and thermal scalings along with predictions for ITER.

Enhanced reproducibility of L-mode plasma discharges via physics-model-based q-profile feedback control in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schuster, Eugenio J.; Wehner, William P.; Barton, Joseph E.

Recent experiments on DIII-D demonstrate the potential of physics-model-based q-profile control to improve reproducibility of plasma discharges. A combined feed forward + feedback control scheme is employed to optimize the current ramp-up phase by consistently achieving target q profiles (Target 1: q min = 1.3,q 95 = 4:4; Target 2: q min = 1.65,q 95 = 5.0; Target 3: q min = 2.1,q 95 = 6:2) at prescribed times during the plasma formation phase (Target 1: t = 1.5 s; Target 2: t = 1:3 s; Target 3: t = 1.0 s). At the core of the control scheme ismore » a nonlinear, first-principles-driven, physics-based, control-oriented model of the plasma dynamics valid for low confinement (L-mode) scenarios. To prevent undesired L-H transitions, a constraint on the maximum allowable total auxiliary power is imposed in addition to the maximum powers for the individual heating and current-drive sources. Experimental results are presented to demonstrate the effectiveness of the combined feed forward + feedback control scheme to consistently achieve the desired target profiles at the predefined times. Here, these results also show how the addition of feedback control significantly improves upon the feed forward only control solution by reducing the matching error and also how the feedback controller is able to reduce the matching error as the constraint on the maximum allowable total auxiliary power is relaxed while keeping the plasma in L-mode.« less

Enhanced reproducibility of L-mode plasma discharges via physics-model-based q-profile feedback control in DIII-D

DOE PAGES

Schuster, Eugenio J.; Wehner, William P.; Barton, Joseph E.; ...

2017-08-09

Recent experiments on DIII-D demonstrate the potential of physics-model-based q-profile control to improve reproducibility of plasma discharges. A combined feed forward + feedback control scheme is employed to optimize the current ramp-up phase by consistently achieving target q profiles (Target 1: q min = 1.3,q 95 = 4:4; Target 2: q min = 1.65,q 95 = 5.0; Target 3: q min = 2.1,q 95 = 6:2) at prescribed times during the plasma formation phase (Target 1: t = 1.5 s; Target 2: t = 1:3 s; Target 3: t = 1.0 s). At the core of the control scheme ismore » a nonlinear, first-principles-driven, physics-based, control-oriented model of the plasma dynamics valid for low confinement (L-mode) scenarios. To prevent undesired L-H transitions, a constraint on the maximum allowable total auxiliary power is imposed in addition to the maximum powers for the individual heating and current-drive sources. Experimental results are presented to demonstrate the effectiveness of the combined feed forward + feedback control scheme to consistently achieve the desired target profiles at the predefined times. Here, these results also show how the addition of feedback control significantly improves upon the feed forward only control solution by reducing the matching error and also how the feedback controller is able to reduce the matching error as the constraint on the maximum allowable total auxiliary power is relaxed while keeping the plasma in L-mode.« less

Fast camera imaging of dust in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Yu, J. H.; Rudakov, D. L.; Pigarov, A. Yu.; Smirnov, R. D.; Brooks, N. H.; Muller, S. H.; West, W. P.

2009-06-01

Naturally occurring and injected dust particles are observed in the DIII-D tokamak in the outer midplane scrape-off-layer (SOL) using a visible fast-framing camera, and the size of dust particles is estimated using the observed particle lifetime and theoretical ablation rate of a carbon sphere. Using this method, the lower limit of detected dust radius is ˜3 μm and particles with inferred radius as large as ˜1 mm are observed. Dust particle 2D velocities range from approximately 10 to 300 m/s with velocities inversely correlated with dust size. Pre-characterized 2-4 μm diameter diamond dust particles are introduced at the lower divertor in an ELMing H-mode discharge using the divertor materials evaluation system (DiMES), and these particles are found to be at the lower size limit of detection using the camera with resolution of ˜0.2 cm 2 per pixel and exposure time of 330 μs.

Integrated Modeling of Time Evolving 3D Kinetic MHD Equilibria and NTV Torque

NASA Astrophysics Data System (ADS)

Logan, N. C.; Park, J.-K.; Grierson, B. A.; Haskey, S. R.; Nazikian, R.; Cui, L.; Smith, S. P.; Meneghini, O.

2016-10-01

New analysis tools and integrated modeling of plasma dynamics developed in the OMFIT framework are used to study kinetic MHD equilibria evolution on the transport time scale. The experimentally observed profile dynamics following the application of 3D error fields are described using a new OMFITprofiles workflow that directly addresses the need for rapid and comprehensive analysis of dynamic equilibria for next-step theory validation. The workflow treats all diagnostic data as fundamentally time dependent, provides physics-based manipulations such as ELM phase data selection, and is consistent across multiple machines - including DIII-D and NSTX-U. The seamless integration of tokamak data and simulation is demonstrated by using the self-consistent kinetic EFIT equilibria and profiles as input into 2D particle, momentum and energy transport calculations using TRANSP as well as 3D kinetic MHD equilibrium stability and neoclassical transport modeling using General Perturbed Equilibrium Code (GPEC). The result is a smooth kinetic stability and NTV torque evolution over transport time scales. Work supported by DE-AC02-09CH11466.

Impact of ideal MHD stability limits on high-beta hybrid operation

NASA Astrophysics Data System (ADS)

Piovesan, P.; Igochine, V.; Turco, F.; Ryan, D. A.; Cianciosa, M. R.; Liu, Y. Q.; Marrelli, L.; Terranova, D.; Wilcox, R. S.; Wingen, A.; Angioni, C.; Bock, A.; Chrystal, C.; Classen, I.; Dunne, M.; Ferraro, N. M.; Fischer, R.; Gude, A.; Holcomb, C. T.; Lebschy, A.; Luce, T. C.; Maraschek, M.; McDermott, R.; Odstrčil, T.; Paz-Soldan, C.; Reich, M.; Sertoli, M.; Suttrop, W.; Taylor, N. Z.; Weiland, M.; Willensdorfer, M.; The ASDEX Upgrade Team; The DIII-D Team; The EUROfusion MST1 Team

2017-01-01

The hybrid scenario is a candidate for stationary high-fusion gain tokamak operation in ITER and DEMO. To obtain such performance, the energy confinement and the normalized pressure {βN} must be maximized, which requires operating near or above ideal MHD no-wall limits. New experimental findings show how these limits can affect hybrid operation. Even if hybrids are mainly limited by tearing modes, proximity to the no-wall limit leads to 3D field amplification that affects plasma profiles, e.g. rotation braking is observed in ASDEX Upgrade throughout the plasma and peaks in the core. As a result, even the small ASDEX Upgrade error fields are amplified and their effects become visible. To quantify such effects, ASDEX Upgrade measured the response to 3D fields applied by 8× 2 non-axisymmetric coils as {βN} approaches the no-wall limit. The full n  =  1 response profile and poloidal structure were measured by a suite of diagnostics and compared with linear MHD simulations, revealing a characteristic feature of hybrids: the n  =  1 response is due to a global, marginally-stable n  =  1 kink characterized by a large m  =  1, n  =  1 core harmonic due to q min being just above 1. A helical core distortion of a few cm forms and affects various core quantities, including plasma rotation, electron and ion temperature, and intrinsic W density. In similar experiments, DIII-D also measured the effect of this helical core on the internal current profile, providing information useful to understanding of the physics of magnetic flux pumping, i.e. anomalous current redistribution by MHD modes that keeps {{q}\\text{min}}>1 . Thanks to flux pumping, a broad current profile is maintained in DIII-D even with large on-axis current drive, enabling fully non-inductive operation at high {βN} up to 3.5-4.

Impact of ideal MHD stability limits on high-beta hybrid operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piovesan, Paolo; Igochine, V.; Turco, F.

Here, the hybrid scenario is a candidate for stationary high-fusion gain tokamak operation in ITER and DEMO. To obtain such performance, the energy confinement and the normalized pressuremore » $${{\\beta}_{N}}$$ must be maximized, which requires operating near or above ideal MHD no-wall limits. New experimental findings show how these limits can affect hybrid operation. Even if hybrids are mainly limited by tearing modes, proximity to the no-wall limit leads to 3D field amplification that affects plasma profiles, e.g. rotation braking is observed in ASDEX Upgrade throughout the plasma and peaks in the core. As a result, even the small ASDEX Upgrade error fields are amplified and their effects become visible. To quantify such effects, ASDEX Upgrade measured the response to 3D fields applied by $$8\\times 2$$ non-axisymmetric coils as $${{\\beta}_{N}}$$ approaches the no-wall limit. The full n = 1 response profile and poloidal structure were measured by a suite of diagnostics and compared with linear MHD simulations, revealing a characteristic feature of hybrids: the n = 1 response is due to a global, marginally-stable n = 1 kink characterized by a large m = 1, n = 1 core harmonic due to q min being just above 1. A helical core distortion of a few cm forms and affects various core quantities, including plasma rotation, electron and ion temperature, and intrinsic W density. In similar experiments, DIII-D also measured the effect of this helical core on the internal current profile, providing information useful to understanding of the physics of magnetic flux pumping, i.e. anomalous current redistribution by MHD modes that keeps $${{q}_{\\text{min}}}>1$$ . Thanks to flux pumping, a broad current profile is maintained in DIII-D even with large on-axis current drive, enabling fully non-inductive operation at high $${{\\beta}_{N}}$$ up to 3.5–4.« less

Impact of ideal MHD stability limits on high-beta hybrid operation

DOE PAGES

Piovesan, Paolo; Igochine, V.; Turco, F.; ...

2016-10-27

Here, the hybrid scenario is a candidate for stationary high-fusion gain tokamak operation in ITER and DEMO. To obtain such performance, the energy confinement and the normalized pressuremore » $${{\\beta}_{N}}$$ must be maximized, which requires operating near or above ideal MHD no-wall limits. New experimental findings show how these limits can affect hybrid operation. Even if hybrids are mainly limited by tearing modes, proximity to the no-wall limit leads to 3D field amplification that affects plasma profiles, e.g. rotation braking is observed in ASDEX Upgrade throughout the plasma and peaks in the core. As a result, even the small ASDEX Upgrade error fields are amplified and their effects become visible. To quantify such effects, ASDEX Upgrade measured the response to 3D fields applied by $$8\\times 2$$ non-axisymmetric coils as $${{\\beta}_{N}}$$ approaches the no-wall limit. The full n = 1 response profile and poloidal structure were measured by a suite of diagnostics and compared with linear MHD simulations, revealing a characteristic feature of hybrids: the n = 1 response is due to a global, marginally-stable n = 1 kink characterized by a large m = 1, n = 1 core harmonic due to q min being just above 1. A helical core distortion of a few cm forms and affects various core quantities, including plasma rotation, electron and ion temperature, and intrinsic W density. In similar experiments, DIII-D also measured the effect of this helical core on the internal current profile, providing information useful to understanding of the physics of magnetic flux pumping, i.e. anomalous current redistribution by MHD modes that keeps $${{q}_{\\text{min}}}>1$$ . Thanks to flux pumping, a broad current profile is maintained in DIII-D even with large on-axis current drive, enabling fully non-inductive operation at high $${{\\beta}_{N}}$$ up to 3.5–4.« less

Validation of the model for ELM suppression with 3D magnetic fields using low torque ITER baseline scenario discharges in DIII-D

NASA Astrophysics Data System (ADS)

Moyer, R. A.; Paz-Soldan, C.; Nazikian, R.; Orlov, D. M.; Ferraro, N. M.; Grierson, B. A.; Knölker, M.; Lyons, B. C.; McKee, G. R.; Osborne, T. H.; Rhodes, T. L.; Meneghini, O.; Smith, S.; Evans, T. E.; Fenstermacher, M. E.; Groebner, R. J.; Hanson, J. M.; La Haye, R. J.; Luce, T. C.; Mordijck, S.; Solomon, W. M.; Turco, F.; Yan, Z.; Zeng, L.; DIII-D Team

2017-10-01

Experiments have been executed in the DIII-D tokamak to extend suppression of Edge Localized Modes (ELMs) with Resonant Magnetic Perturbations (RMPs) to ITER-relevant levels of beam torque. The results support the hypothesis for RMP ELM suppression based on transition from an ideal screened response to a tearing response at a resonant surface that prevents expansion of the pedestal to an unstable width [Snyder et al., Nucl. Fusion 51, 103016 (2011) and Wade et al., Nucl. Fusion 55, 023002 (2015)]. In ITER baseline plasmas with I/aB = 1.4 and pedestal ν * ˜ 0.15, ELMs are readily suppressed with co- I p neutral beam injection. However, reducing the beam torque from 5 Nm to ≤ 3.5 Nm results in loss of ELM suppression and a shift in the zero-crossing of the electron perpendicular rotation ω ⊥ e ˜ 0 deeper into the plasma. The change in radius of ω ⊥ e ˜ 0 is due primarily to changes to the electron diamagnetic rotation frequency ωe * . Linear plasma response modeling with the resistive MHD code m3d-c1 indicates that the tearing response location tracks the inward shift in ω ⊥ e ˜ 0. At pedestal ν * ˜ 1, ELM suppression is also lost when the beam torque is reduced, but the ω ⊥ e change is dominated by collapse of the toroidal rotation v T . The hypothesis predicts that it should be possible to obtain ELM suppression at reduced beam torque by also reducing the height and width of the ωe * profile. This prediction has been confirmed experimentally with RMP ELM suppression at 0 Nm of beam torque and plasma normalized pressure β N ˜ 0.7. This opens the possibility of accessing ELM suppression in low torque ITER baseline plasmas by establishing suppression at low beta and then increasing beta while relying on the strong RMP-island coupling to maintain suppression.

Promoting Pre-college Science Education

NASA Astrophysics Data System (ADS)

Lee, R. L.

1999-11-01

The Fusion Education Program, with support from DOE, continues to promote pre-college science education for students and teachers using multiple approaches. An important part of our program is direct scientist-student interaction. Our ``Scientist in a Classroom'' program allows students to interact with scientists and engage in plasma science activities in the students' classroom. More than 1000 students from 11 schools have participated in this exciting program. Also, this year more than 800 students and teachers have visited the DIII--D facility and interacted with scientists to cover a broad range of technical and educational issues. Teacher-scientist interaction is imperative in professional development and each year more than 100 teachers attend workshops produced by the fusion education team. We also participate in unique learning opportunities. Members of the team, in collaboration with the San Diego County Office of Education, held a pioneering Internet-based Physics Olympiad for American and Siberian students. Our teamwork with educators helps shape material that is grade appropriate, relevant, and stimulates thinking in educators and students.

Design and evaluation of noise suppression sheet for GHz band utilizing magneto-elastic effect

NASA Astrophysics Data System (ADS)

Igarashi, Toshiyuki; Kondo, Koichi; Yoshida, Shigeyoshi

2017-12-01

Feasibility of realizing a noise suppression sheet (NSS) coping with the low SHF band such as the 5 GHz band was investigated, which was composed of soft magnetic metal flakes dispersed in a polymer. For suppressing noises, the higher frequency one of the bimodal frequency dispersion (lower frequency one: Dispersion DII, higher frequency one: Dispersion DIII) seen in the imaginary permeability (μ″; magnetic loss component) spectrum of the NSS was aimed to utilize. Referring to the previous finding that Dispersion DIII is originated from a magneto-elastic effect, several magnetic composite sheets were prepared using various alloy flakes with different saturation magnetostriction (λs), and their frequency (fr(DIII)) and magnitude (μ″(DIII)) of Dispersion DIII were investigated. It was found that the NSS containing flakes with higher λs exhibited higher fr(DIII) and higher μ″(DIII)/μ″(DII), which was ratio of μ″(DIII) to the magnitude of Dispersion DII (μ″(DII)). The fr(DIII) for the NSS having the highest λs containing Fe-Co alloy flake reached 7.45 GHz and μ″ in the 5 GHz band was approximately twice as high as the conventional NSS containing Fe-Si-Al alloy flake. For transmission attenuation power ratio (Rtp) when an NSS was placed on a microstrip line with characteristic impedance of 50 Ω, NSS with larger fr(DIII)2 · μ″(DIII) ∝ Ms2 (Ms: saturation magnetization), which theoretically gave the frequency limit of imaginary permeability for a thin film, exhibited larger Rtp in the low SHF band. These results suggested that an NSS containing a magnetic flake material with both large λs and Ms was suitable for suppressing low SHF band noises.

A phase contrast imaging–interferometer system for detection of multiscale electron density fluctuations on DIII-D

DOE PAGES

Davis, E. M.; Rost, J. C.; Porkolab, M.; ...

2016-08-15

Heterodyne interferometry and phase contrast imaging (PCI) are robust, mature techniques for measuring low-k and high-k electron density fluctuations, respectively. Here, we describe the first-ever implementation of a combined PCI-interferometer. The combined system uses a single 10:6 μm probe beam, two interference schemes, and two detectors to measure electron density uctuations at large spatiotemporal bandwidth (10 kHz < f < 5MHz and 0 cm -1 ≤ k ≤ 20 cm -1), allowing simultaneous measurement of ion- and electron-scale instabilities. Further, correlating our interferometer's measurements with those from DIII-D's pre-existing, toroidally separated interferometer allows core-localized, low-n MHD studies that may otherwisemore » be inaccessible via external magnetic measurements. In the combined diagnostic's small port requirements and minimal access restrictions make it well-suited to the harsh neutron environments and limited port space expected in next-step devices.« less

Design and first plasma measurements of the ITER-ECE prototype radiometer

DOE PAGES

Austin, M. E.; Brookman, M. W.; Rowan, W. L.; ...

2016-08-09

On ITER, second harmonic optically thick electron cyclotron emission (ECE) in the range of 220-340 GHz will supply the electron temperature (T e). In order to investigate the requirements and capabilities prescribed for the ITER system, a prototype radiometer covering this frequency range has been developed by Virginia Diodes, Inc. The first plasma measurements with this instrument have been carried out on the DIII-D tokamak, with lab bench tests and measurements of third through fifth harmonic ECE from high T e plasmas. At DIII-D the instrument shares the transmission line of the Michelson interferometer and can simultaneously acquire data. Inmore » our comparison of the ECE radiation temperature from the absolutely calibrated Michelson and the prototype receiver we show that the ITER radiometer provides accurate measurements of the millimeter radiation across the instrument band.« less

A phase contrast imaging-interferometer system for detection of multiscale electron density fluctuations on DIII-D

NASA Astrophysics Data System (ADS)

Davis, E. M.; Rost, J. C.; Porkolab, M.; Marinoni, A.; Van Zeeland, M. A.

2016-11-01

Heterodyne interferometry and phase contrast imaging (PCI) are robust, mature techniques for measuring low-k and high-k electron density fluctuations, respectively. This work describes the first-ever implementation of a combined PCI-interferometer. The combined system uses a single 10.6 μm probe beam, two interference schemes, and two detectors to measure electron density fluctuations at large spatiotemporal bandwidth (10 kHz

Improved edge charge exchange recombination spectroscopy in DIII-D

DOE PAGES

Chrystal, Colin; Burrell, K. H.; Grierson, Brian A.; ...

2016-08-02

The charge exchange recombination spectroscopy diagnostic on the DIII-D tokamak has been upgraded with the addition of more high radial resolution view chords near the edge of the plasma ( r/a > 0.8). The additional views are diagnosed with the same number of spectrometers by placing fiber optics side-by-side at the spectrometer entrance with a precise separation that avoids wavelength shifted crosstalk without the use of bandpass filters. The new views improve measurement of edge impurity parameters in steep gradient, H-mode plasmas with many different shapes. The number of edge view chords with 8 mm radial separation has increased frommore » 16 to 38.As a result, new fused silica fibers have improved light throughput and clarify the observation of non-Gaussian spectra that suggest the ion distribution function can be non-Maxwellian in low collisionality plasmas.« less

Optimized Baking of the DIII-D Vessel

DOE Office of Scientific and Technical Information (OSTI.GOV)

P.M. Anderson; A.G. Kellman

The DIII-D tokamak vacuum vessel baking system is used to heat the vessel walls and internal hardware to an average temperature of 350 C to allow rapid conditioning of the vacuum surfaces. The system combines inductive heating and a circulating hot air system to provide rapid heating with temperature uniformity required by stress considerations. In recent years, the time to reach 350 C had increased from 9 hrs to 14 hrs. To understand and remedy this sluggish heating rate, an evaluation of the baking system was recently performed. The evaluation indicated that the mass of additional in-vessel hardware (50% increasemore » in mass) was primarily responsible. This paper reports on this analysis and the results of the addition of an electric air heater and procedural changes that have been implemented. Preliminary results indicate that the time to 350 C has been decreased to 4.5 hours and the temperature uniformity has improved.« less

Intershot Analysis of Flows in DIII-D

NASA Astrophysics Data System (ADS)

Meyer, W. H.; Allen, S. L.; Samuell, C. M.; Howard, J.

2016-10-01

Analysis of the DIII-D flow diagnostic data require demodulation of interference images, and inversion of the resultant line integrated emissivity and flow (phase) images. Four response matrices are pre-calculated: the emissivity line integral and the line integral of the scalar product of the lines-of-site with the orthogonal unit vectors of parallel flow. Equilibrium data determines the relative weight of the component matrices used in the final flow inversion matrix. Serial processing has been used for the lower divertor viewing flow camera 800x600 pixel image. The full cross section viewing camera will require parallel processing of the 2160x2560 pixel image. We will discuss using a Posix thread pool and a Tesla K40c GPU in the processing of this data. Prepared by LLNL under Contract DE-AC52-07NA27344. This material is based upon work supported by the U.S. DOE, Office of Science, Fusion Energy Sciences.

Main-ion intrinsic toroidal rotation profile driven by residual stress torque from ion temperature gradient turbulence in the DIII-D tokamak

DOE PAGES

Grierson, B. A.; Wang, W. X.; Ethier, S.; ...

2017-01-06

Intrinsic toroidal rotation of the deuterium main ions in the core of the DIII-D tokamak is observed to transition from flat to hollow, forming an off-axis peak, above a threshold level of direct electron heating. Nonlinear gyrokinetic simulations show that the residual stress associated with electrostatic ion temperature gradient turbulence possesses the correct radial location and stress structure to cause the observed hollow rotation profile. Residual stress momentum flux in the gyrokinetic simulations is balanced by turbulent momentum diffusion, with negligible contributions from turbulent pinch. Finally, the prediction of the velocity profile by integrating the momentum balance equation produces amore » rotation profile that qualitatively and quantitatively agrees with the measured main-ion profile, demonstrating that fluctuation-induced residual stress can drive the observed intrinsic velocity profile.« less

Models of SOL transport and their relation to scaling of the divertor heat flux width in DIII-D

DOE PAGES

Makowski, M. A.; Lasnier, C. J.; Leonard, A. W.; ...

2014-10-06

Strong support for the critical pressure gradient model for the heat flux width has been obtained, in that the measured separatrix pressure gradient lies below and scales similarly to the pressure gradient limit obtained from the ideal, infinite-n stability codes, BALOO and 2DX, in all cases that have been examined. Predictions of a heuristic drift model for the heat flux width are also in qualitative agreement with the measurements. We obtained these results by using an improved high rep-rate and higher edge spatial resolution Thomson scattering system on DIII-D to measure the upstream electron temperature and density profiles. In ordermore » to compare theory and experiment, profiles of density, temperature, and pressure for both electrons and ions are needed as well values of these quantitities at the separatrix. We also developed a simple method to identify a proxy for the separatrix.« less

DIII-D accomplishments and plans in support of fusion next steps

DOE PAGES

Buttery, R. J; Eidietis, N.; Holcomb, C.; ...

2013-06-01

DIII-D is using its flexibility and diagnostics to address the critical science required to enable next step fusion devices. We have adapted operating scenarios for ITER to low torque and are now being optimized for transport. Three ELM mitigation scenarios have been developed to near-ITER parameters. New control techniques are managing the most challenging plasma instabilities. Disruption mitigation tools show promising dissipation strategies for runaway electrons and heat load. An off axis neutral beam upgrade has enabled sustainment of high βN capable steady state regimes. Divertor research is identifying the challenge, physics and candidate solutions for handling the hot plasmamore » exhaust with notable progress in heat flux reduction using the snowflake configuration. Our work is helping optimize design choices and prepare the scientific tools for operation in ITER, and resolve key elements of the plasma configuration and divertor solution for an FNSF.« less

Dynamics of energetic particle driven modes and MHD modes in wall-stabilized high-β plasmas on JT-60U and DIII-D

NASA Astrophysics Data System (ADS)

Matsunaga, G.; Okabayashi, M.; Aiba, N.; Boedo, J. A.; Ferron, J. R.; Hanson, J. M.; Hao, G. Z.; Heidbrink, W. W.; Holcomb, C. T.; In, Y.; Jackson, G. L.; Liu, Y. Q.; Luce, T. C.; McKee, G. R.; Osborne, T. H.; Pace, D. C.; Shinohara, K.; Snyder, P. B.; Solomon, W. M.; Strait, E. J.; Turnbull, A. D.; Van Zeeland, M. A.; Watkins, J. G.; Zeng, L.; the DIII-D Team; the JT-60 Team

2013-12-01

In the wall-stabilized high-β plasmas in JT-60U and DIII-D, interactions between energetic particle (EP) driven modes (EPdMs) and edge localized modes (ELMs) have been observed. The interaction between the EPdM and ELM are reproducibly observed. Many EP diagnostics indicate a strong correlation between the distorted waveform of the EPdM and the EP transport to the edge. The waveform distortion is composed of higher harmonics (n ⩾ 2) and looks like a density snake near the plasma edge. According to statistical analyses, ELM triggering by the EPdMs requires a finite level of waveform distortion and pedestal recovery. ELM pacing by the EPdMs occurs when the repetition frequency of the EPdMs is higher than the natural ELM frequency. EPs transported by EPdMs are thought to contribute to change the edge stability.

Increased electron temperature turbulence during suppression of edge localized mode by resonant magnetic perturbations in the DIII-D tokamak [Increased electron temperature turbulence during edge localized mode (ELM) suppression by resonant magnetic perturbations (RMPs) in the DIII-D tokamak

DOE PAGES

Sung, Choongki; Wang, G.; Rhodes, Terry L.; ...

2017-11-16

We report the first observation of increased edge electron temperature turbulence correlated with changes in gradients and the ELM suppression time which occurs after the application of resonant magnetic perturbations (RMP) on DIII-D H-mode plasmas. This increase (T ~ e/T e approximately doubles) occurs in the region extending from the top of the pedestal outward to the upper part of the edge steep gradient region. This is significant as it is consistent with increased turbulence driven transport potentially replacing some part of the edge localized mode (ELM) driven transport. However, temperature turbulence does not change with the initial RMP applicationmore » while ELMs are still present, indicating the turbulence changes are not causative in the development of ELM suppression or initial profile evolution with RMP – but rather a response to these effects. This temperature turbulence is broadband and long wavelength, k θρ s < 0.5, where k θ = poloidal wavenumber, ρ s = ion sound gyroradius. As has been reported previously, long wavelength density turbulence (k θρ s < 1.0) in the same location also increases after ELMs were suppressed by the RMP. Since the decrease of the density starts nearly immediately with RMP application, these results suggest that the so-called RMP “density pump-out” is not linked to these long wavelength turbulent transport changes. Comparison with linear stability analysis finds both consistencies and inconsistencies in this important region.« less

Dependence of Helicon Antenna Loading on the Antenna/Plasma Gap and n|| in DIII-D Experiments

NASA Astrophysics Data System (ADS)

Pinsker, R. I.; Moeller, C. P.

2017-10-01

A comprehensive set of measurements of the plasma loading of a 12-element antenna array, designed to launch helicon waves (i.e., very-high-harmonic fast waves), were performed on DIII-D in 2016. The antenna, operated in the 466 - 486 MHz band, is prototypical of a wider array for a 1-MW-level experiment planned for 2018-9. The dependence of the antenna loading on antenna/plasma gap is of great practical significance, as the gap must be kept greater than a minimum distance to suppress deleterious plasma-material interactions, while the loading must be high enough to retain good efficiency of power transfer to the plasma. While the loading in all examined plasma regimes, including both limited and diverted L-mode discharges and H-mode discharges, decayed exponentially with increasing gap in agreement with simple theory, the characteristic decay length was in all cases larger than expected, motivating the development of a more realistic model. Furthermore, the characteristic decay length did not depend on the launched n||, though the absolute level of loading at a given gap increased as |n||| was decreased from 4 to 2. After the antenna was removed from DIII-D, measurements of the loading produced by a 100 Ω/sq resistive film were carried out on the bench. Both the antenna/film gap and n|| were scanned varied and the results compared with calculations done with the QuickWave FDTD electromagnetics solver. Very good agreement was found in this case. Work supported by the US DOE under DE-FC02-04ER54698.

Lack of dependence on resonant error field of locked mode island size in ohmic plasmas in DIII-D

NASA Astrophysics Data System (ADS)

La Haye, R. J.; Paz-Soldan, C.; Strait, E. J.

2015-02-01

DIII-D experiments show that fully penetrated resonant n = 1 error field locked modes in ohmic plasmas with safety factor q95 ≳ 3 grow to similar large disruptive size, independent of resonant error field correction. Relatively small resonant (m/n = 2/1) static error fields are shielded in ohmic plasmas by the natural rotation at the electron diamagnetic drift frequency. However, the drag from error fields can lower rotation such that a bifurcation results, from nearly complete shielding to full penetration, i.e., to a driven locked mode island that can induce disruption. Error field correction (EFC) is performed on DIII-D (in ITER relevant shape and safety factor q95 ≳ 3) with either the n = 1 C-coil (no handedness) or the n = 1 I-coil (with ‘dominantly’ resonant field pitch). Despite EFC, which allows significantly lower plasma density (a ‘figure of merit’) before penetration occurs, the resulting saturated islands have similar large size; they differ only in the phase of the locked mode after typically being pulled (by up to 30° toroidally) in the electron diamagnetic drift direction as they grow to saturation. Island amplification and phase shift are explained by a second change-of-state in which the classical tearing index changes from stable to marginal by the presence of the island, which changes the current density profile. The eventual island size is thus governed by the inherent stability and saturation mechanism rather than the driving error field.

Increased electron temperature turbulence during suppression of edge localized mode by resonant magnetic perturbations in the DIII-D tokamak [Increased electron temperature turbulence during edge localized mode (ELM) suppression by resonant magnetic perturbations (RMPs) in the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sung, Choongki; Wang, G.; Rhodes, Terry L.

We report the first observation of increased edge electron temperature turbulence correlated with changes in gradients and the ELM suppression time which occurs after the application of resonant magnetic perturbations (RMP) on DIII-D H-mode plasmas. This increase (T ~ e/T e approximately doubles) occurs in the region extending from the top of the pedestal outward to the upper part of the edge steep gradient region. This is significant as it is consistent with increased turbulence driven transport potentially replacing some part of the edge localized mode (ELM) driven transport. However, temperature turbulence does not change with the initial RMP applicationmore » while ELMs are still present, indicating the turbulence changes are not causative in the development of ELM suppression or initial profile evolution with RMP – but rather a response to these effects. This temperature turbulence is broadband and long wavelength, k θρ s < 0.5, where k θ = poloidal wavenumber, ρ s = ion sound gyroradius. As has been reported previously, long wavelength density turbulence (k θρ s < 1.0) in the same location also increases after ELMs were suppressed by the RMP. Since the decrease of the density starts nearly immediately with RMP application, these results suggest that the so-called RMP “density pump-out” is not linked to these long wavelength turbulent transport changes. Comparison with linear stability analysis finds both consistencies and inconsistencies in this important region.« less

Electron cyclotron heating can drastically alter reversed shear Alfvén eigenmode activity in DIII-D through finite pressure effects

NASA Astrophysics Data System (ADS)

Van Zeeland, M. A.; Heidbrink, W. W.; Sharapov, S. E.; Spong, D.; Cappa, A.; Chen, Xi; Collins, C.; García-Muñoz, M.; Gorelenkov, N. N.; Kramer, G. J.; Lauber, P.; Lin, Z.; Petty, C.

2016-11-01

A recent DIII-D experiment investigating the impact of electron cyclotron heating (ECH) on neutral beam driven reversed shear Alfvén eigenmode (RSAE) activity is presented. The experiment includes variations of ECH injection location and timing, current ramp rate, beam injection geometry (on/off-axis), and neutral beam power. Essentially all variations carried out in this experiment were observed to change the impact of ECH on AE activity significantly. In some cases, RSAEs were observed to be enhanced with ECH near the off-axis minimum in magnetic safety factor ({{q}\\min} ), in contrast to the original DIII-D experiments where the modes were absent when ECH was deposited near {{q}\\min} . It is found that during intervals when the geodesic acoustic mode (GAM) frequency at {{q}\\min} is elevated and the calculated RSAE minimum frequency, including contributions from thermal plasma gradients, is very near or above the nominal TAE frequency (f TAE), RSAE activity is not observed or RSAEs with a much reduced frequency sweep range are found. This condition is primarily brought about by ECH modification of the local electron temperature (T e) which can raise both the local T e at {{q}\\min} as well as its gradient. A q-evolution model that incorporates this reduction in RSAE frequency sweep range is in agreement with the observed spectra and appears to capture the relative balance of TAE or RSAE-like modes throughout the current ramp phase of over 38 DIII-D discharges. Detailed ideal MHD calculations using the NOVA code show both modification of plasma pressure and pressure gradient at {{q}\\min} play an important role in modifying the RSAE activity. Analysis of the ECH injection near the {{q}\\min} case where no frequency sweeping RSAEs are observed shows the typical RSAE is no longer an eigenmode of the system. What remains is an eigenmode with poloidal harmonic content reminiscent of the standard RSAE, but absent of the typical frequency sweeping behavior. The remaining eigenmode is also often strongly coupled to gap TAEs. Analysis with the non-perturbative gyro fluid code TAEFL confirms this change in RSAE activity and also shows a large drop in the resultant mode growth rates.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks.

Optimal Control Techniques for ResistiveWall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Clement, Mitchell Dobbs Pearson

Tokamaks can excite kink modes that can lock or nearly lock to the vacuum vessel wall, and whose rotation frequencies and growth rates vary in time but are generally inversely proportional to the magnetic flux diffusion time of the vacuum vessel wall. This magnetohydrodynamic (MHD) instability is pressure limiting in tokamaks and is called the Resistive Wall Mode (RWM). Future tokamaks that are expected to operate as fusion reactors will be required to maximize plasma pressure in order to maximize fusion performance. The DIII-D tokamak is equipped with electromagnetic control coils, both inside and outside of its vacuum vessel, which create magnetic fields that are small by comparison to the machine's equilibrium field but are able to dynamically counteract the RWM. Presently for RWM feedback, DIII-D uses its interior control coils using a classical proportional gain only controller to achieve high plasma pressure. Future advanced tokamak designs will not likely have the luxury of interior control coils and a proportional gain algorithm is not expected to be effective with external control coils. The computer code VALEN was designed to calculate the performance of an MHD feedback control system in an arbitrary geometry. VALEN models the perturbed magnetic field from a single MHD instability and its interaction with surrounding conducting structures using a finite element approach. A linear quadratic gaussian (LQG) control, or H 2 optimal control, algorithm based on the VALEN model for RWM feedback was developed for use with DIII-D's external control coil set. The algorithm is implemented on a platform that combines a graphics processing unit (GPU) for real-time control computation with low latency digital input/output control hardware and operates in parallel with the DIII-D Plasma Control System (PCS). Simulations and experiments showed that modern control techniques performed better, using 77% less current, than classical techniques when using coils external to the vacuum vessel for RWM feedback. RWM feedback based on VALEN outperformed a classical control algorithm using external coils to suppress the normalized plasma response to a rotating n=1 perturbation applied by internal coils over a range of frequencies. This study describes the design, development and testing of the GPU based control hardware and algorithm along with its performance during experiment and simulation.

Electron cyclotron heating can drastically alter reversed shear Alfvén eigenmode activity in DIII-D through finite pressure effects

DOE PAGES

Van Zeeland, M. A.; Heidbrink, W. W.; Sharapov, S. E.; ...

2016-07-22

Here, a recent DIII-D experiment investigating the impact of electron cyclotron heating (ECH) on neutral beam driven reversed shear Alfvén eigenmode (RSAE) activity is presented. The experiment includes variations of ECH injection location and timing, current ramp rate, beam injection geometry (on/off-axis), and neutral beam power. Essentially all variations carried out in this experiment were observed to change the impact of ECH on AE activity significantly. In some cases, RSAEs were observed to be enhanced with ECH near the off-axis minimum in magnetic safety factor (more » $${{q}_{\\min}}$$ ), in contrast to the original DIII-D experiments where the modes were absent when ECH was deposited near $${{q}_{\\min}}$$ . It is found that during intervals when the geodesic acoustic mode (GAM) frequency at $${{q}_{\\min}}$$ is elevated and the calculated RSAE minimum frequency, including contributions from thermal plasma gradients, is very near or above the nominal TAE frequency (f TAE), RSAE activity is not observed or RSAEs with a much reduced frequency sweep range are found. This condition is primarily brought about by ECH modification of the local electron temperature (T e) which can raise both the local T e at $${{q}_{\\min}}$$ as well as its gradient. A q-evolution model that incorporates this reduction in RSAE frequency sweep range is in agreement with the observed spectra and appears to capture the relative balance of TAE or RSAE-like modes throughout the current ramp phase of over 38 DIII-D discharges. Detailed ideal MHD calculations using the NOVA code show both modification of plasma pressure and pressure gradient at $${{q}_{\\min}}$$ play an important role in modifying the RSAE activity. Analysis of the ECH injection near the $${{q}_{\\min}}$$ case where no frequency sweeping RSAEs are observed shows the typical RSAE is no longer an eigenmode of the system. What remains is an eigenmode with poloidal harmonic content reminiscent of the standard RSAE, but absent of the typical frequency sweeping behavior. The remaining eigenmode is also often strongly coupled to gap TAEs. Analysis with the non-perturbative gyro fluid code TAEFL confirms this change in RSAE activity and also shows a large drop in the resultant mode growth rates.« less

Regulation of Na+ channel inactivation by the DIII and DIV voltage-sensing domains.

PubMed

Hsu, Eric J; Zhu, Wandi; Schubert, Angela R; Voelker, Taylor; Varga, Zoltan; Silva, Jonathan R

2017-03-06

Functional eukaryotic voltage-gated Na + (Na V ) channels comprise four domains (DI-DIV), each containing six membrane-spanning segments (S1-S6). Voltage sensing is accomplished by the first four membrane-spanning segments (S1-S4), which together form a voltage-sensing domain (VSD). A critical Na V channel gating process, inactivation, has previously been linked to activation of the VSDs in DIII and DIV. Here, we probe this interaction by using voltage-clamp fluorometry to observe VSD kinetics in the presence of mutations at locations that have been shown to impair Na V channel inactivation. These locations include the DIII-DIV linker, the DIII S4-S5 linker, and the DIV S4-S5 linker. Our results show that, within the 10-ms timeframe of fast inactivation, the DIV-VSD is the primary regulator of inactivation. However, after longer 100-ms pulses, the DIII-DIV linker slows DIII-VSD deactivation, and the rate of DIII deactivation correlates strongly with the rate of recovery from inactivation. Our results imply that, over the course of an action potential, DIV-VSDs regulate the onset of fast inactivation while DIII-VSDs determine its recovery. © 2017 Hsu et al.

Integrated modeling applications for tokamak experiments with OMFIT

NASA Astrophysics Data System (ADS)

Meneghini, O.; Smith, S. P.; Lao, L. L.; Izacard, O.; Ren, Q.; Park, J. M.; Candy, J.; Wang, Z.; Luna, C. J.; Izzo, V. A.; Grierson, B. A.; Snyder, P. B.; Holland, C.; Penna, J.; Lu, G.; Raum, P.; McCubbin, A.; Orlov, D. M.; Belli, E. A.; Ferraro, N. M.; Prater, R.; Osborne, T. H.; Turnbull, A. D.; Staebler, G. M.

2015-08-01

One modeling framework for integrated tasks (OMFIT) is a comprehensive integrated modeling framework which has been developed to enable physics codes to interact in complicated workflows, and support scientists at all stages of the modeling cycle. The OMFIT development follows a unique bottom-up approach, where the framework design and capabilities organically evolve to support progressive integration of the components that are required to accomplish physics goals of increasing complexity. OMFIT provides a workflow for easily generating full kinetic equilibrium reconstructions that are constrained by magnetic and motional Stark effect measurements, and kinetic profile information that includes fast-ion pressure modeled by a transport code. It was found that magnetic measurements can be used to quantify the amount of anomalous fast-ion diffusion that is present in DIII-D discharges, and provide an estimate that is consistent with what would be needed for transport simulations to match the measured neutron rates. OMFIT was used to streamline edge-stability analyses, and evaluate the effect of resonant magnetic perturbation (RMP) on the pedestal stability, which have been found to be consistent with the experimental observations. The development of a five-dimensional numerical fluid model for estimating the effects of the interaction between magnetohydrodynamic (MHD) and microturbulence, and its systematic verification against analytic models was also supported by the framework. OMFIT was used for optimizing an innovative high-harmonic fast wave system proposed for DIII-D. For a parallel refractive index {{n}\\parallel}>3 , the conditions for strong electron-Landau damping were found to be independent of launched {{n}\\parallel} and poloidal angle. OMFIT has been the platform of choice for developing a neural-network based approach to efficiently perform a non-linear multivariate regression of local transport fluxes as a function of local dimensionless parameters. Transport predictions for thousands of DIII-D discharges showed excellent agreement with the power balance calculations across the whole plasma radius and over a broad range of operating regimes. Concerning predictive transport simulations, the framework made possible the design and automation of a workflow that enables self-consistent predictions of kinetic profiles and the plasma equilibrium. It is found that the feedback between the transport fluxes and plasma equilibrium can significantly affect the kinetic profiles predictions. Such a rich set of results provide tangible evidence of how bottom-up approaches can potentially provide a fast track to integrated modeling solutions that are functional, cost-effective, and in sync with the research effort of the community.

Dynamic ELM and divertor control using resonant toroidal multi-mode magnetic fields in DIII-D and EAST

NASA Astrophysics Data System (ADS)

Sun, Youwen

2017-10-01

A rotating n = 2 Resonant Magnetic Perturbation (RMP) field combined with a stationary n = 3 RMP field has validated predictions that access to ELM suppression can be improved, while divertor heat and particle flux can also be dynamically controlled in DIII-D. Recent observations in the EAST tokamak indicate that edge magnetic topology changes, due to nonlinear plasma response to magnetic perturbations, play a critical role in accessing ELM suppression. MARS-F code MHD simulations, which include the plasma response to the RMP, indicate the nonlinear transition to ELM suppression is optimized by configuring the RMP coils to drive maximal edge stochasticity. Consequently, mixed toroidal multi-mode RMP fields, which produce more densely packed islands over a range of additional rational surfaces, improve access to ELM suppression, and further spread heat loading on the divertor. Beneficial effects of this multi-harmonic spectrum on ELM suppression have been validated in DIII-D. Here, the threshold current required for ELM suppression with a mixed n spectrum, where part of the n = 3 RMP field is replaced by an n = 2 field, is smaller than the case with pure n = 3 field. An important further benefit of this multi-mode approach is that significant changes of 3D particle flux footprint profiles on the divertor are found in the experiment during the application of a rotating n = 2 RMP field superimposed on a static n = 3 RMP field. This result was predicted by modeling studies of the edge magnetic field structure using the TOP2D code which takes into account plasma response from MARS-F code. These results expand physics understanding and potential effectiveness of the technique for reliably controlling ELMs and divertor power/particle loading distributions in future burning plasma devices such as ITER. Work supported by USDOE under DE-FC02-04ER54698 and NNSF of China under 11475224.

Helical core reconstruction of a DIII-D hybrid scenario tokamak discharge

DOE PAGES

Cianciosa, Mark; Wingen, Andreas; Hirshman, Steven P.; ...

2017-05-18

Our paper presents the first fully 3-dimensional (3D) equilibrium reconstruction of a helical core in a tokamak device. Using a new parallel implementation of the Variational Moments Equilibrium Code (PARVMEC) coupled to V3FIT, 3D reconstructions can be performed at resolutions necessary to produce helical states in nominally axisymmetric tokamak equilibria. In a flux pumping experiment performed on DIII-D, an external n=1 field was applied while a 3/2 neoclassical tearing mode was suppressed using ECCD. The externally applied field was rotated past a set of fixed diagnostics at a 20 Hz frequency. Furthermore, the modulation, were found to be strongest in the core SXR and MSE channels, indicates a localized rotating 3D structure locked in phase with the applied field. Signals from multiple time slices are converted to a virtual rotation of modeled diagnostics adding 3D signal information. In starting from an axisymmetric equilibrium reconstruction solution, the reconstructed broader current profile flattens the q-profile, resulting in an m=1, n=1 perturbation of the magnetic axis that ismore » $$\\sim 50\\times $$ larger than the applied n=1 deformation of the edge. Error propagation confirms that the displacement of the axis is much larger than the uncertainty in the axis position validating the helical equilibrium.« less

Helical core reconstruction of a DIII-D hybrid scenario tokamak discharge

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cianciosa, Mark; Wingen, Andreas; Hirshman, Steven P.

Our paper presents the first fully 3-dimensional (3D) equilibrium reconstruction of a helical core in a tokamak device. Using a new parallel implementation of the Variational Moments Equilibrium Code (PARVMEC) coupled to V3FIT, 3D reconstructions can be performed at resolutions necessary to produce helical states in nominally axisymmetric tokamak equilibria. In a flux pumping experiment performed on DIII-D, an external n=1 field was applied while a 3/2 neoclassical tearing mode was suppressed using ECCD. The externally applied field was rotated past a set of fixed diagnostics at a 20 Hz frequency. Furthermore, the modulation, were found to be strongest in the core SXR and MSE channels, indicates a localized rotating 3D structure locked in phase with the applied field. Signals from multiple time slices are converted to a virtual rotation of modeled diagnostics adding 3D signal information. In starting from an axisymmetric equilibrium reconstruction solution, the reconstructed broader current profile flattens the q-profile, resulting in an m=1, n=1 perturbation of the magnetic axis that ismore » $$\\sim 50\\times $$ larger than the applied n=1 deformation of the edge. Error propagation confirms that the displacement of the axis is much larger than the uncertainty in the axis position validating the helical equilibrium.« less

Fast ion transport during applied 3D magnetic perturbations on DIII-D

DOE PAGES

Van Zeeland, Michael A.; Ferraro, Nathaniel M.; Grierson, Brian A.; ...

2015-06-26

In this paper, measurements show fast ion losses correlated with applied three-dimensional (3D) fields in a variety of plasmas ranging from L-mode to resonant magnetic perturbation (RMP) edge localized mode (ELM) suppressed H-mode discharges. In DIII-D L-mode discharges with a slowly rotatingmore » $n=2$ magnetic perturbation, scintillator detector loss signals synchronized with the applied fields are observed to decay within one poloidal transit time after beam turn-off indicating they arise predominantly from prompt loss orbits. Full orbit following using M3D-C1 calculations of the perturbed fields and kinetic profiles reproduce many features of the measured losses and points to the importance of the applied 3D field phase with respect to the beam injection location in determining the overall impact on prompt beam ion loss. Modeling of these results includes a self-consistent calculation of the 3D perturbed beam ion birth profiles and scrape-off-layer ionization, a factor found to be essential to reproducing the experimental measurements. Extension of the simulations to full slowing down timescales, including fueling and the effects of drag and pitch angle scattering, show the applied $n=3$ RMPs in ELM suppressed H-mode plasmas can induce a significant loss of energetic particles from the core. With the applied $n=3$ fields, up to 8.4% of the injected beam power is predicted to be lost, compared to 2.7% with axisymmetric fields only. These fast ions, originating from minor radii $$\\rho >0.7$$ , are predicted to be primarily passing particles lost to the divertor region, consistent with wide field-of-view infrared periscope measurements of wall heating in $n=3$ RMP ELM suppressed plasmas. Edge fast ion $${{\\text{D}}_{\\alpha}}$$ (FIDA) measurements also confirm a large change in edge fast ion profile due to the $n=3$ fields, where the effect was isolated by using short 50 ms RMP-off periods during which ELM suppression was maintained yet the fast ion profile was allowed to recover. Finally, the role of resonances between fast ion drift motion and the applied 3D fields in the context of selectively targeting regions of fast ion phase space is also discussed.« less

Validation of conducting wall models using magnetic measurements

DOE PAGES

Hanson, Jeremy M.; Bialek, James M.; Turco, Francesca; ...

2016-08-16

The impact of conducting wall eddy currents on perturbed magnetic field measurements is a key issue for understanding the measurement and control of long-wavelength MHD stability in tokamak devices. As plasma response models have growth in sophistication, the need to understand and resolve small changes in these measurements has become more important, motivating increased fidelity in simulations of externally applied fields and the wall eddy current response. In this manuscript, we describe thorough validation studies of the wall models in the MARS-F and VALEN stability codes, using coil–sensor vacuum coupling measurements from the DIII-D tokamak. The valen formulation treats conductingmore » structures with arbitrary threedimensional geometries, while mars-f uses an axisymmetric wall model and a spectral decomposition of the problem geometry with a fixed toroidal harmonic n. The vacuum coupling measurements have a strong sensitivity to wall eddy currents induced by timechanging coil currents, owing to the close proximities of both the sensors and coils to the wall. Measurements from individual coil and sensor channels are directly compared with valen predictions. It is found that straightforward improvements to the valen model, such as refining the wall mesh and simulating the vertical extent of the DIII-D poloidal field sensors, lead to good agreement with the experimental measurements. In addition, couplings to multi-coil, n = 1 toroidal mode perturbations are calculated from the measurements and compared with predictions from both codes. Lastly, the toroidal mode comparisons favor the fully three-dimensional simulation approach, likely because this approach naturally treats n > 1 sidebands generated by the coils and wall eddy currents, as well as the n = 1 fundamental.« less

Validation of conducting wall models using magnetic measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanson, Jeremy M.; Bialek, James M.; Turco, Francesca

The impact of conducting wall eddy currents on perturbed magnetic field measurements is a key issue for understanding the measurement and control of long-wavelength MHD stability in tokamak devices. As plasma response models have growth in sophistication, the need to understand and resolve small changes in these measurements has become more important, motivating increased fidelity in simulations of externally applied fields and the wall eddy current response. In this manuscript, we describe thorough validation studies of the wall models in the MARS-F and VALEN stability codes, using coil–sensor vacuum coupling measurements from the DIII-D tokamak. The valen formulation treats conductingmore » structures with arbitrary threedimensional geometries, while mars-f uses an axisymmetric wall model and a spectral decomposition of the problem geometry with a fixed toroidal harmonic n. The vacuum coupling measurements have a strong sensitivity to wall eddy currents induced by timechanging coil currents, owing to the close proximities of both the sensors and coils to the wall. Measurements from individual coil and sensor channels are directly compared with valen predictions. It is found that straightforward improvements to the valen model, such as refining the wall mesh and simulating the vertical extent of the DIII-D poloidal field sensors, lead to good agreement with the experimental measurements. In addition, couplings to multi-coil, n = 1 toroidal mode perturbations are calculated from the measurements and compared with predictions from both codes. Lastly, the toroidal mode comparisons favor the fully three-dimensional simulation approach, likely because this approach naturally treats n > 1 sidebands generated by the coils and wall eddy currents, as well as the n = 1 fundamental.« less

Symplectic approach to calculation of magnetic field line trajectories in physical space with realistic magnetic geometry in divertor tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Punjabi, Alkesh; Ali, Halima

A new approach to integration of magnetic field lines in divertor tokamaks is proposed. In this approach, an analytic equilibrium generating function (EGF) is constructed in natural canonical coordinates ({psi},{theta}) from experimental data from a Grad-Shafranov equilibrium solver for a tokamak. {psi} is the toroidal magnetic flux and {theta} is the poloidal angle. Natural canonical coordinates ({psi},{theta},{phi}) can be transformed to physical position (R,Z,{phi}) using a canonical transformation. (R,Z,{phi}) are cylindrical coordinates. Another canonical transformation is used to construct a symplectic map for integration of magnetic field lines. Trajectories of field lines calculated from this symplectic map in natural canonicalmore » coordinates can be transformed to trajectories in real physical space. Unlike in magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)], the symplectic map in natural canonical coordinates can integrate trajectories across the separatrix surface, and at the same time, give trajectories in physical space. Unlike symplectic maps in physical coordinates (x,y) or (R,Z), the continuous analog of a symplectic map in natural canonical coordinates does not distort trajectories in toroidal planes intervening the discrete map. This approach is applied to the DIII-D tokamak [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The EGF for the DIII-D gives quite an accurate representation of equilibrium magnetic surfaces close to the separatrix surface. This new approach is applied to demonstrate the sensitivity of stochastic broadening using a set of perturbations that generically approximate the size of the field errors and statistical topological noise expected in a poloidally diverted tokamak. Plans for future application of this approach are discussed.« less

Symplectic approach to calculation of magnetic field line trajectories in physical space with realistic magnetic geometry in divertor tokamaks

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima

2008-12-01

A new approach to integration of magnetic field lines in divertor tokamaks is proposed. In this approach, an analytic equilibrium generating function (EGF) is constructed in natural canonical coordinates (ψ,θ) from experimental data from a Grad-Shafranov equilibrium solver for a tokamak. ψ is the toroidal magnetic flux and θ is the poloidal angle. Natural canonical coordinates (ψ,θ,φ) can be transformed to physical position (R,Z,φ) using a canonical transformation. (R,Z,φ) are cylindrical coordinates. Another canonical transformation is used to construct a symplectic map for integration of magnetic field lines. Trajectories of field lines calculated from this symplectic map in natural canonical coordinates can be transformed to trajectories in real physical space. Unlike in magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)], the symplectic map in natural canonical coordinates can integrate trajectories across the separatrix surface, and at the same time, give trajectories in physical space. Unlike symplectic maps in physical coordinates (x,y) or (R,Z), the continuous analog of a symplectic map in natural canonical coordinates does not distort trajectories in toroidal planes intervening the discrete map. This approach is applied to the DIII-D tokamak [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The EGF for the DIII-D gives quite an accurate representation of equilibrium magnetic surfaces close to the separatrix surface. This new approach is applied to demonstrate the sensitivity of stochastic broadening using a set of perturbations that generically approximate the size of the field errors and statistical topological noise expected in a poloidally diverted tokamak. Plans for future application of this approach are discussed.

Flow Shear Effects in the Onset Physics of Resistive MHD Instabilities in Tokamaks. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brennan, Dylan P.

The progress in this research centers around the computational analysis of flow shear effects in the onset of a 3/2 mode driven by a 1/1 mode in DIII-D equilibria. The initial idea was to try and calculate, via nonlinear simulations with NIMROD, the effects of rotation shear on driven 3/2 and 2/1 seed island physics, in experimentally relevant DIIID equilibria. The simulations indicated that very small seed islands were directly driven, as shielding between the sawtooth and the surfaces is significant at the high Lundquist numbers of the experiment. Instead, long after the initial crash the difference in linear stabilitymore » of the 3/2, which remained prevalent despite the flattening of the core profiles from the sawtooth, contributed to a difference in the eventual seed island evolution. Essentially the seed islands grew or decayed long after the sawtooth crash, and not directly from it. Effectively the dominant 1/1 mode was found to be dragging the coupled modes surrounding it at a high rate through the plasma at their surfaces. The 1/1 mode is locked to the local frame of the plasma in the core, where the flow rate is greatest. The resonant perturbations at the surrounding surfaces propagate in the 'high slip regime' in the language of Fitzpatrick. Peaked flux averaged jxb forces (see Figs. 1 and 2) agree with localized flow modifications at the surfaces in analogy with Ebrahimi, PRL 2007. We track the mode into nonlinear saturation and have found oscillatory states in the evolution. During a visit (11/09) to Tulsa by R.J. LaHaye (GA), it became clear that similar oscillatory states are observed in DIII-D for these types of discharges.« less

A new gamma ray imaging diagnostic for runaway electron studies at DIII-D

NASA Astrophysics Data System (ADS)

Cooper, C. M.; Pace, D. C.; Eidietis, N. W.; Paz-Soldan, C.; Commaux, N.; Shiraki, D.; Hollmann, E. M.; Moyer, R. A.; Risov, V.

2015-11-01

A new Gamma Ray Imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at DIII-D. The diagnostic is sensitive to 0.5 - 50 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE dissipation from pellet injection. The GRI consists of a lead ``pinhole camera'' mounted on the midplane with 11x11 counter-current tangential chords 20 cm wide that span the vessel. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE Bremsstrahlung radiation. Detectors operate in current saturation mode at 10 MHz, or the flux is attenuated for Pulse Height Analysis (PHA) capable of discriminating up to ~10k pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. Work supported by the US DOE under DE-AC05-00OR22725, DE-FG02-07ER54917 & DE-FC02-04ER54698.