National Fusion Facility Research Articles

GaN-based W-band receiver chip development for fusion plasma diagnostics

Millimeter-wave diagnostics have proven effective on various magnetic fusion devices worldwide, yet the formidable challenges posed by the harsh environments of future burning plasma devices, characterized by extreme temperatures, pressures, and radiation levels, remain a significant hurdle. To address these challenges, the utilization of wide bandgap Gallium Nitride (GaN)-based millimeter-wave diagnostics is a most promising solution for fusion reactor safety monitoring and control. A noteworthy W-band GaN-based system-on-chip receiver has been the demonstrated by employing HRL T3 40 nm GaN technology. This receiver chip, compactly designed with dimensions of 3 × 5 mm2, incorporates essential components such as the 75–110 GHz RF Low-Noise Amplifier (LNA), mixer, Intermediate Frequency (IF) amplifier, and Local Oscillator (LO) chain. This receiver chip will be packaged as a millimeter-wave receiver module and applied on the DIII-D National Fusion Facility, for fusion plasma edge shape monitoring for operational safety and dangerous disruption prediction. The laboratory measurement results have demonstrated suitable performance. This advancement is pivotal for accurate analysis of plasma behavior in the extreme conditions of burning plasma devices, driving progress in fusion research and technology.

Remote operation of the DIII-D National Fusion Facility

Full remote scientific operation of the DIII-D National Fusion Facility is now possible through significant advances in the computer science hardware and software infrastructure made over the last decade. Capabilities around information visualization, data movement, and communication have all been enhanced. The level of capability deployed to remotely operate DIII-D required an infrastructure advancement over what had previously been achieved in the fusion community. The large quantity of real-time data that is automatically displayed on DIII-D’s control room screens can now be visualized by remote participants via web-based applications. New audio/video solutions using the VoIP and instant messaging application Discord have been implemented to mimic the dynamic and ad-hoc scientific conversations that are critical in successfully operating an experimental campaign. Discord’s ability for a user to rapidly move between audio channels, text with images, and share screens is a significant enhancement over traditional videoconferencing tools. In addition, multiple combinations of broadcast audio are made available via a web-based application to allow remote participants to simultaneously listen to general announcements/sounds while conducting their own specific conversations. Secure methodologies have been put into place to allow remote control of hardware including DIII-D’s plasma control system application. Secure methods also included the ability of the on-site team to closely coordinate their work with remote team members which has been enhanced through extensions to the wireless network and the use of tablet computers for audio/video/screen sharing. However, no amount of software can fully replace the need for ‘hands on hardware.’ This infrastructure was severely stress tested during the COVID-19 pandemic where occupancy of the DIII-D control room was restricted. Operational efficiency during the pandemic, measured in discharges per hour, remained high (3.8 ± 0.8) compared to values obtained pre-pandemic (3.7 ± 0.8).

DIII-D Electron Cyclotron Heating and Current Drive System Status and Plans

The DIII-D 110-GHz electron cyclotron heating and current drive (ECH/ECCD) system supports a wide range of experiments at the DIII-D National Fusion Facility and is pioneering innovative new techniques for future fusion reactors. The system currently includes eight sockets with four installed gyrotrons, 31.75-mm circular corrugated transmission lines, and four dual real-time steerable launchers plus two top launchers (TLs). After the successful testing of the first TL, this injects the EC waves downward nearly parallel to the resonance layer with double current drive efficiency and a long absorption path, and a second new TL was designed and installed with an upgraded design. Some of the operational limitations encountered for the 110-GHz X-mode second harmonic resonance scenario launched in overdense plasmas were overcome with the alternative use of O-mode second harmonic launch. Two new gyrotrons in a new series manufactured by Communications and Power Industries (CPI) were installed and tested, bringing up the total installed power to 3 MW. The installation of repaired and new gyrotrons will increase the number of installed gyrotrons to 7 in 2023. An addition to the DIII-D building will be completed in early 2022 to provide the space to expand the ECH/ECCD system to ten gyrotron systems plus an extra gyrotron test socket.

Doppler-shift compensated spatial heterodyne spectroscopy for rapidly moving sources.

High resolution luminosity product measurements of neutral beam emission in magnetized plasmas are severely limited by the artificial Doppler broadening inherent to the use of large diameter collection optics. In this paper, a broadening compensation method is developed for the spatial heterodyne spectroscopy interferometric technique. The compensation technique greatly reduces the artificial broadening, thereby enabling high resolution measurements at a significantly higher photon flux than previously available. Compensated and uncompensated measurements of emission generated by impact excitation of 61 keV deuterium neutrals in a tokamak plasma at the DIII-D National Fusion Facility are presented. The spectral width of the compensated measurement is ${\sim}0.13 \;{\rm{nm}}$, which is comparable to the instrument resolution. This width is ${\sim}4 \times$ smaller than the uncompensated width, which for the 20 cm diameter collection lens system utilized in this study is ${\sim}0.5 \;{\rm{nm}}$.

Thermogravimetric Oxidation Analyses of Carbon Tokamak Codeposits and Flakes

The erosion and redeposition of first-wall armor materials is a problem in nuclear fusion devices with carbon walls, where deuterium, tritium, and (eroded) carbon present in the plasma are deposited on the walls of the device, trapping the expensive and radiologically hazardous tritium. Thermo-oxidation, in which vessel surfaces are heated and oxygen containing gas is injected, is a possible solution. It results in the production of carbon oxides and tritiated water vapor, which can be pumped out by the vacuum pumps and recycled in a tritium recycling facility. In the present study, thermogravimetric analysis was used to measure the mass loss (or gain) of codeposited specimens from the General Atomics DIII-D National Fusion Facility under thermo-oxidation, in addition to laser thermal desorption spectroscopy. X-ray photo-electron spectroscopy was also used in this work to examine the tile’s surface composition pre and post oxidation. Dust scraped from the specimen was also studied, as this is a surrogate for dust that naturally falls from the tile codeposits and builds up in the tile gaps. One key conclusion is that boron oxides form where boron is present in the codeposit as an impurity, and these oxides dominate the weight-change behavior of the codeposit specimens for long exposures.

Keras2c: A library for converting Keras neural networks to real-time compatible C

With the growth of machine learning models and neural networks in measurement and control systems comes the need to deploy these models in a way that is compatible with existing systems. Existing options for deploying neural networks either introduce very high latency, require expensive and time consuming work to integrate into existing code bases, or only support a very limited subset of model types. We have therefore developed a new method called Keras2c, which is a simple library for converting Keras/TensorFlow neural network models into real-time compatible C code. It supports a wide range of Keras layers and model types including multidimensional convolutions, recurrent layers, multi-input/output models, and shared layers. Keras2c re-implements the core components of Keras/TensorFlow required for predictive forward passes through neural networks in pure C, relying only on standard library functions considered safe for real-time use. The core functionality consists of ∼1500 lines of code, making it lightweight and easy to integrate into existing codebases. Keras2c has been successfully tested in experiments and is currently in use on the plasma control system at the DIII-D National Fusion Facility at General Atomics in San Diego.

Multiple Analytical Approach to Isotopic Transport Analysis in Magnetic Fusion Devices

High-Z impurities released from plasma-material interactions have been shown to limit the performance of fusion plasmas, and understanding these impurity transport mechanisms throughout the plasma scrape-off layer is a major challenge. Presented herein is a study of tungsten (W) erosion and transport by uniquely measuring absolute quantities of isotopic W in order to determine the source of natural and enriched 182W isotopes that have traveled throughout the tokamak discharges on the DIII-D National Fusion Facility at General Atomics. Two primary analysis methods have been implemented to characterize this W on graphite collector probes that were inserted into DIII-D’s outboard midplane. Results from experiments using Rutherford backscattering spectrometry (RBS) have measured W particle areal densities down the centerline of the probes as high as 6E14 atoms/cm2 with a detection limit of 1E12 atoms/cm2. Laser ablation inductively coupled plasma mass spectrometry (LAMS) has confirmed the elemental trends found with RBS and has provided additional insight into collector probe surface profiles. Two-dimensional elemental and isotopic maps from LAMS are used to reveal new collector probe features and further refine the source of collected W. Variations in isotopic profiles and total W content are coupled to (a) the face of the probe being analyzed, (b) the dimensions of the probe, and (c) the plasma pulse parameters that were used during probe exposure. These results provide one-of-a-kind empirical evidence that is now being utilized for validation of tokamak impurity transport through theoretical models and in codes such as 3D-LIM and OEDGE.

Erosion characterization of SiC and Ti3SiC2 on DIII-D using focused ion beam micro-trenches

Abstract Plasma-facing materials in future large-scale fusion reactors must be designed to withstand high heat fluxes from extreme off-normal events such as edge localized modes and unmitigated plasma disruptions. The erosion rates of possible tungsten-alternative materials are tested under high heat flux conditions at the DIII-D National Fusion Facility. High-purity β-3C CVD silicon carbide was exposed alongside MAX phase ceramic Ti3SiC2 to both L- and H-mode plasma discharges in the DIII-D divertor. Samples survived average heat fluxes ranging from 2–10 MW/m2 over 16 s. A new micro-trench erosion measurement technique was successfully implemented and measured Ti3SiC2 and SiC erosion rates of 0–9 nm/s and 27–73 nm/s, respectively. Additionally, average ion impact angle estimates for an incident B-field angle of ∼1.5° from surface parallel were made using micro-trench impact patterns. Measurements ranged from θ = 24o–34o with respect to Bt and ϕ = 51.5o–55o below the surface normal.

DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants

We review progress made on the advanced tokamak path to fusion energy by the DIII-D National Fusion Facility (Luxon et al. in Nucl Fusion 42:614, 2002). The advanced tokamak represents a highly attractive approach for a future steady state fusion power plant. In this concept, there is a natural alignment between high pressure operation, favorable stability and transport properties, and a highly self-driven (‘bootstrap’) plasma current to sustain operation efficiently and without disruptions. Research on DIII-D has identified several promising plasma configurations for fully non-inductive operation with potential applications to a range of future devices, from ITER to nuclear science facilities, to compact or large scale fusion power plants. Significant progress has been made toward realizing these scenarios, with the demonstration of high β access, off-axis current drive techniques, model based profile control, and stability and ELM control in reactor relevant physics regimes. Radiative techniques have also been pioneered to develop improved compatibility with divertor requirements, and simultaneous access to high performance pedestals. Research has also developed major advances in physics understanding, validating concepts of kinetic damping of ideal MHD instabilities that enable high β operation, identifying how current profile and β influence plasma turbulence in order to validate and improve turbulent transport models, and understanding the physics of energetic particle redistribution due to Alfvénic and other instabilities. These advances have been partnered with development of a rigorous integrated modeling framework used to interpret and validate individual physics models of the various aspects of plasma behavior, and to guide development of improved regimes and upgrades. These tools are also being used to develop and validate concepts for future reactors directly. Having established these foundations, DIII-D is now undergoing a substantial upgrade to raise power, current drive, electron heating and 3-D field capabilities in order to validate this physics and test conceptual solutions in reactor-relevant physics regimes, with a goal to resolve the key scientific and technology questions to enable a decision on a future steady state fusion power plant.

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

An imminent gyrotron body current fault can now be detected and avoided in less than 10 µs using a high-resolution, real-time, high-voltage reference waveform generator with signal analyzer subroutines. The gyrotron is restarted and full power resumed in a little over 10 ms.

Upgrade to DIII-D National Fusion Facility PCS and Neutral Beam Systems: In-Shot Variation of Neutral Beam Particle Energy

On the DIII-D National Fusion Facility tokamak plasma diagnostics continue to improve and experiments increase in complexity. Hence the utility of dynamic control of the beam energy (and therefore also the injected torque, ion heating fraction, etc.) has become apparent. Here we report on upgrades that have been incorporated into the DIII-D Plasma Control System (PCS) and Neutral Beam Injection (NBI) systems in order to allow the beam acceleration voltage (Vaccel) to be varied continuously in a ≤20 kV range during a shot for the first time, generating new capabilities such as smooth plasma transitions and controllable interactions with Alfvén waves.

The Between-Pulse Data Analysis Infrastructure at the DIII-D National Fusion Facility

The DIII-D National Fusion Facility is a national and international collaboration among approximately 100 institutions with an overall goal of establishing the scientific basis for the optimization of the tokamak approach to fusion energy. A key enabler of the DIII-D mission is its extensive diagnostic set (>50), which measures relevant equilibrium parameters as well as turbulence fields. The ability to access, analyze, visualize, and assimilate data between DIII-D pulses that enables real-time decision making by an international team is a critical infrastructure component of the successful operation of the DIII-D facility. This paper examines the computer science issues associated with deploying this infrastructure in a geographically distributed environment where near-real-time support of control room decision making is required. The implication of this work on the operation of future experimental machines such as ITER is also presented.

Diagnostic Developments for the DIII-D National Fusion Facility

The DIII-D National Fusion Facility has long been a center of innovation and development of diagnostics for magnetic fusion devices. The DIII-D device, a moderate size tokamak, with a high flexibility shaping coil set, neutral beam injection (NBI), electron cyclotron heating (ECH) and ion cyclotron heating (ICH), supports a very broad research program infusion science, including critical aspects related to burning plasmas expected to be encountered in ITER. This scientific program is supported by a large set of diagnostics (approximately 50), which is the product of a highly collaborative program between universities, national laboratories and industry. Although many diagnostic systems are now routinely employed to measure a wide range of plasma parameters, such as temperature, rotation, density and current profiles, there are many areas that are inherently difficult or prohibitively expensive to diagnose. Such areas include the measurements associated with energetic ion populations or with the characterization of plasma flows in the divertor/edge area. In addition, the study of burning plasmas will require the development of new and updated techniques, which need to be developed and tested in existing devices in relevant plasma conditions.

Distributed MDSplus database performance with Linux clusters

The staff at the DIII-D National Fusion Facility, operated for the USDOE by General Atomics, are investigating the use of grid computing and Linux technology to improve performance in our core data management services. We are in the process of converting much of our functionality to cluster-based and grid-enabled software. One of the most important pieces is a new distributed version of the MDSplus scientific data management system that is presently used to support fusion research in over 30 countries worldwide. To improve data handling performance, the staff is investigating the use of Linux clusters for both data clients and servers. The new distributed capability will result in better load balancing between these clients and servers, and more efficient use of network resources resulting in improved support of the data analysis needs of the scientific staff.

Resistive Wall Mode Stabilization Studies at DIII-D

The effort to understand the physics of the resistive wall mode (RWM) and develop methods to control this magnetohydrodynamic mode to allow achievement of higher pressure in advanced tokamak plasmas has been an example of successful multi-institutional collaboration at the DIII-D National Fusion Facility in San Diego, California. DIII-D research in this area has produced several advances and breakthroughs following a coordinated research plan involving a sequence of measurements, development of new analysis tools, and the installation of new diagnostic and feedback stabilization hardware: Suppression of the RWM by active magnetic feedback has been demonstrated using the DIII-D six-element error field correction coil, rotational stabilization of the RWM has been demonstrated and sustained for all values of the plasma pressure from the no-wall to the ideal-wall stability limits, improved RWM feedback stabilization has been shown using a new set of 12 internal control coils, and newly developed models of feedback have shown good agreement with the measurements. By so doing, the DIII-D work on RWM stabilization has become a cornerstone of the long-term advanced tokamak program and is having impact on the world fusion program. Presently both ITER and FIRE are including plans for RWM stabilization in their programs.

Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available.

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

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.

A general purpose data analysis monitoring system with case studies from the National Fusion Grid and the DIII–D MDSplus between pulse analysis system

Abstract As computing infrastructures become more complex, it is important to centralize information in order to efficiently monitor and maintain computing processes. A general monitoring system has been developed at the DIII–D National Fusion Facility, that combines code run status, data analysis tracking, logfile access, complex error detection, and expert system capabilities. The monitoring system’s flexibility and ease of deployment have enabled it to be successfully applied to two significantly different computing environments. At DIII–D, the system is being used as the data analysis monitor ( http://nssrv1.gat.com:8000/dam ) to allow both application scientists and computer scientists to monitor the status of between pulse MDSplus dispatched data analysis codes. The monitoring system is also being used by the National Fusion Collaboratory Project, as the Fusion Grid monitor ( http://nssrv1.gat.com:8000/fgm ), to track multiple asynchronous complex code runs on the Fusion Grid.

Status of the Linux PC cluster for between-pulse data analyses at DIII-D

Some analyses that survey experimental data are carried out at a sparse sample rate between pulses during tokamak operation and/or completed as a batch job overnight because the complete analysis on a single fast workstation cannot fit in the narrow time window between two pulses. Scientists therefore miss the opportunity to use these results to guide experiments quickly. With a dedicated Beowulf type cluster at a cost less than that of a workstation, these analyses can be accomplished between pulses and the analyzed data made available for the research team during the tokamak operation. A Linux PC cluster comprised of 12 processors was installed at DIII-D National Fusion Facility in CY00 and expanded to 24 processors in CY01 to automatically perform between-pulse magnetic equilibrium reconstructions using the efit code written in fortran, cer analyses using cerquick code written in interactive data language ( idl) and full profile fitting analyses ( n e, T e, T i, V r, Z eff) using idl code zipfit. This paper reports the current status of the system, the details of the between-pulse profile fitting analyses, and discusses some problems and concerns raised during the implementation and expansion of the system.

DIII-D data management

The DIII-D tokamak at the DIII-D National Fusion Facility routinely acquires ∼500MB of raw data per pulse of the experiment through a centralized data management system. It is expected that in FY01 (October 2001 to September 2002), nearly 1TB of data will be acquired. In addition there are several diagnostics, which are not part of the centralized system, which acquire hundreds of MB of raw data per pulse. There is also a growing suite of codes running between pulses that produce analyzed data, which add ∼10MB per pulse with total disk usage of about 100GB. A relational database system has been introduced which further adds to the overall data load. In recent years there has been an order of magnitude increase in magnetic disk space devoted to raw data and a hierarchical storage management system was implemented to allow 24 h/day, 7 days/week unattended access to raw data. The management of all of the data is a significant and growing challenge as the quantities of both raw and analyzed data are expected to continue to increase in the future. This paper will examine the experiences of the approaches that have been taken in management of the data and plans for the continued growth of the data quantity.