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DOE OSTI2025

The spherical tokamak advanced reactor (STAR) fusion power plant design

Scientific and technical advancements have been made that improve fusion’s prospects to provide a new energy source, showing enhanced plasma confinement conditions with plasma temperatures reaching or exceeding 100 million degrees. Overshadowing this progress is the challenge involved in developing an economically viable fusion power plant design. Many proposed next-step DEMO and pilot plant designs are extensions of existing physics-focused experimental devices defined to understand and control plasma operations to achieve and sustain a fusion reaction. Transitioning scientific and technical advancements into a functional power plant requires a dedicated focus on architectural designs that integrate diverse technologies, while optimizing physics conditions, with a focus on economic viability. This holistic approach is essential in turning the promise of fusion energy into a reality. The Spherical Tokamak Advanced Reactor (STAR) is a fusion power plant conceptual design with the architectural focus that strives to balance physics, engineering, and cost considerations. In conclusion, it has been set up to introduce relevant physics, engineering and concept features that an intermediate pilot plant might follow, with the goal of meeting system performances and economic requirements that lead to a commercially competitive fusion power plant.

Blanket segmentation↗

NASA NTRS2001

Conceptual Design of an MTF Space Propulsion System

Most fusion propulsion concepts that have been investigated in the past employ some form of inertial or magnetic confinement separately, and are encumbered by the need for advanced drivers (e.g. laser) or steady-state magnetic confinement systems (e.g. superconductors) that have historically resulted in large, massive spacecraft designs. Here we present a comparatively new approach, Magnetized Target Fusion (MTF), which offers a nearer-term avenue for realizing the tremendous performance benefits of fusion propulsion. MTF attempts to combine the favorable attributes of both inertially and magnetically confined fusion to achieve both efficient and low-cost compressional plasma heating and energy confinement. The key advantage of MTF is its less demanding requirements for driver energy and power processing. Additional features include: 1) very low system masses and volumes, 2) high gain and relatively low waste heat, 3) substantial utilization of energy from product neutrons, 4) efficient, low peak-power drivers based on existing pulsed power technology, and 5) very high I(sub sp), specific power and thrust. MTF overcomes many of the problems associated with traditional fusion techniques, thus making it particularly attractive for space applications. I(sub sp) greater than 50,000 seconds and specific powers greater than 20 kilowatts/kilogram appear feasible using relatively near-term pulse power and plasma gun technology.

Thio, Y. C. Francis↗

NASA NTRS2002

Conceptual Design of an MTF Space Propulsion System

Most fusion propulsion concepts that have been investigated in the past employ some form of inertial or magnetic confinement separately, and are encumbered by the need for advanced drivers (e.g. laser) or steady-state magnetic confinement systems (e.g. superconductors) that have historically resulted in large, massive spacecraft designs. Here we present a comparatively new approach, Magnetized Target Fusion (MTF), which offers a nearer-term avenue for realizing the tremendous performance benefits of fusion propulsion. MTF attempts to combine the favorable attributes of both inertially and magnetically confined fusion to achieve both efficient and low-cost compressional plasma heating and energy confinement. The key advantage of MTF is its less demanding requirements for driver energy and power processing. Additional features include: 1) very low system masses and volumes, 2) relatively low waste heat, 3) substantial utilization of energy from product neutrons, 4) efficient, low peak-power drivers based on existing pulsed power technology, 5) very high Isp , specific power and thrust, and 6) relatively affordable R&D pathways. MTF overcomes many of the problems associated with traditional fusion techniques, thus making it particularly attractive for space applications. Isp greater than 50,000 seconds and specific powers greater than 20 kilowatts/kilogram appear feasible using relatively near-term pulse power and plasma gun technology.

Thio, Y. C. F.↗

DOE OSTI2016

Apparatus for magnetic and electrostatic confinement of plasma

An apparatus and method for containing plasma and forming a Field Reversed Configuration (FRC) magnetic topology are described in which plasma ions are contained magnetically in stable, non-adiabatic orbits in the FRC. Further, the electrons are contained electrostatically in a deep energy well, created by tuning an externally applied magnetic field. The simultaneous electrostatic confinement of electrons and magnetic confinement of ions avoids anomalous transport and facilitates classical containment of both electrons and ions. In this configuration, ions and electrons may have adequate density and temperature so that upon collisions ions are fused together by nuclear force, thus releasing fusion energy. Moreover, the fusion fuel plasmas that can be used with the present confinement system and method are not limited to neutronic fuels only, but also advantageously include advanced fuels.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

LDRD FY25 Program Overview

As Lawrence Livermore National Laboratory’s (LLNL’s) Laboratory Directed Research and Development (LDRD) program enters its fifth decade of leading-edge research and development, its impact and importance have never been stronger. The program continues to advance strategic investments in pioneering science, technology, and engineering, ensuring LLNL will be ready to deliver on our mission as it evolves over the coming decades. Investing in LDRD research, and the people who perform this critical work, gives LLNL the ability to sustain our role as a leader in the Department of Energy and National Nuclear Security Administration enterprise. The LDRD program enables high-risk, high-payoff research that anticipates emerging threats and future mission needs. By nurturing the ingenuity of the Lab’s greatest asset, its people, LDRD funding advances not only our research but also grows and nurtures our workforce: engaging future innovators with student mentoring, challenging postdoctoral researchers to apply their skills to support national security, and strengthening the leadership skills of early career staff. This annual report documents how LDRD investments advance LLNL’s science, technology, and engineering across our mission space. To assess LDRD’s impact we track both short and long-term metrics such as peer-reviewed publications, number of students, or professional fellows. In addition to reviewing these metrics, I encourage you to delve deeper into the breadth of science and technology that illustrate the strategic value of this research portfolio. For instance, a recent exploratory research project used advanced manufacturing to construct miniaturized three-dimensional ion traps for a quantum computer with reduced quantum error rates to enable applications that address national security missions and support basic science. Another project has delved into studying detonation by examining deflagration to enhance the safety and security of the nuclear weapons stockpile. LDRD researchers are also deploying AI agents on two of the world’s most powerful supercomputers to automate and accelerate inertial confinement fusion experiments. Other teams are delivering more accurate optical constants to enable improved validation for aluminum to advance atomic and molecular physics models. LDRD-driven discoveries of how metals deform under extreme conditions strengthen our ability to model and design materials for demanding national security environments. National security challenges are increasingly complex and continuously evolving. LDRD focuses our most innovative science and technology on these challenges, ensuring the Laboratory is developing creative, forward-leaning solutions for our nation and the world. The following pages feature highlights of published scientific advances, patents, and honors that stem from LDRD investments. As you read this report, I hope you will understand how these investments position the Laboratory, and our partners, to meet the demands of the decades ahead.

36 MATERIALS SCIENCE↗

DOE OSTI2010

Ideal MHD stability of a spherical tokamak power plant and a component test facility.

We have investigated ideal MHD stability of two advanced spherical tokamak confinement concepts: the spherical tokamak power plant (STPP), a 3 GW concept fusion power plasma producing 1 GW of electric power, and the component test facility (CTF), a concept designed for in situ materials testing for ITER and beyond. Detailed stability studies for toroidal mode number n = 1, 2, 3 displacements are presented as a function of conformal wall radius R{sub w} and on-axis safety factor q{sub 0}. For the STPP marginal stability scans held the current profile fixed, but varied the total plasma current. For the CTF we have extended and parallelized earlier marginal stability scans to scan over both the plasma beta and q{sub 0} by varying the current profile to preserve the total plasma current. These confirm that both concepts are stable provided that the wall is sufficiently close and q{sub 0} sufficiently large (q{sub 0} > 2.8 for the power plant and q{sub 0} > 2.1 for the CTF). Both power plant and CTF configurations are found to be ballooning stable.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2013

Apparatus for magnetic and electrostatic confinement of plasma

42 ENGINEERING↗

DOE OSTI2014

Preface to Special Topic: Advances in Radio Frequency Physics in Fusion Plasmas

It has long been recognized that auxiliary plasma heating will be required to achieve the high temperature, high density conditions within a magnetically confined plasma in which a fusion “burn” may be sustained by copious fusion reactions. Consequently, the application of radio and microwave frequency electromagnetic waves to magnetically confined plasma, commonly referred to as RF, has been a major part of the program almost since its inception in the 1950s. These RF waves provide heating, current drive, plasma profile control, and Magnetohydrodynamics (MHD) stabilization. Fusion experiments employ electromagnetic radiation in a wide range of frequencies, from tens of MHz to hundreds of GHz. The fusion devices containing the plasma are typically tori, axisymmetric or non, in which the equilibrium magnetic fields are composed of a strong toroidal magnetic field generated by external coils, and a poloidal field created, at least in the symmetric configurations, by currents flowing in the plasma. The waves are excited in the peripheral regions of the plasma, by specially designed launching structures, and subsequently propagate into the core regions, where resonant wave-plasma interactions produce localized heating or other modification of the local equilibrium profiles. Experimental studies coupled with the development of theoretical models and advanced simulation codes over the past 40+ years have led to an unprecedented understanding of the physics of RF heating and current drive in the core of magnetic fusion devices. Nevertheless, there are serious gaps in our knowledge base that continue to have a negative impact on the success of ongoing experiments and that must be resolved as the program progresses to the next generation devices and ultimately to “demo” and “fusion power plant.” A serious gap, at least in the ion cyclotron (IC) range of frequencies and partially in the lower hybrid frequency ranges, is the difficulty in coupling large amount of power to the plasma while minimizing the interaction between the plasma and launching structures. These potentially harmful interactions between the plasma and the vessel and launching structures are challenging: (i) significant and variable loss of power in the edge regions of confined plasmas and surrounding vessel structures adversely affect the core plasma performance and lifetime of a device; (ii) the launcher design is partly “trial and error,” with the consequence that launchers may have to be reconfigured after initial tests in a given device, at an additional cost. Over the broader frequency range, another serious gap is a quantitative lack of understanding of the combined effects of nonlinear wave-plasma processes, energetic particle interactions and non-axisymmetric equilibrium effects on determining the overall efficiency of plasma equilibrium and stability profile control techniques using RF waves. This is complicated by a corresponding lack of predictive understanding of the time evolution of transport and stability processes in fusion plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

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

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

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Broadband laser output by application of stimulated rotational Raman scattering during high-intensity propagation of laser pulses with modified spectral profiles

Research at the Naval Research Laboratory pursues advanced laser technologies that can mitigate laser plasma instabilities (LPI) in applications such as inertial confinement fusion. Increased laser bandwidth is currently being developed as a strategy to suppress LPI. In particular, stimulated rotational Raman scattering (SRRS) has been demonstrated using the Nike krypton fluoride (KrF) laser facility as a viable method for enhancing bandwidth. The technique is applicable to other large laser facilities. Through a combination of high-intensity propagation and optimization of the laser spectrum in the low-energy stages of the laser, the output spectrum of Nike has been broadened with SRRS from its standard operating value of 1 THz full width at half-maximum (FWHM) bandwidth to span a range greater than 4.5 THz. This amount of bandwidth has been shown in simulations to be effective for mitigation of slow-growth laser plasma instabilities, such as cross-beam energy transport (CBET).

Optics↗

DOE OSTI2011

Change in Inertial Confinement Fusion Implosions upon Using an Ab Initio Multiphase DT Equation of State

Improving the description of the equation of state (EOS) of deuterium-tritium (DT) has recently been shown to change significantly the gain of an inertial confinement fusion target [S. X. Hu et al., Phys. Rev. Lett. 104, 235003 (2010)]. Here we use an advanced multiphase EOS, based on ab initio calculations, to perform a full optimization of the laser pulse shape with hydrodynamic simulations starting from 19 K in DT ice. The thermonuclear gain is shown to be a robust estimate over possible uncertainties of the EOS. Two different target designs are discussed, for shock ignition and self-ignition. In the first case, the areal density and thermonuclear energy can be recovered by slightly increasing the laser energy. In the second case, a lower in-flight adiabat is needed, leading to a significant delay (3 ns) in the shock timing of the implosion.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2026

Advances in 3D transient plasma dynamics and control through MHD and hybrid fluid-kinetic simulations with JOREK

Transient phenomena and their control are of high relevance in magnetic confinement fusion plasmas to guarantee a stable and safe plasma operation. Interpretative simulations can maximize the insights gained from experiments on present machines and predictive simulations can help in the preparation of design, mitigation techniques and operational scenarios for future devices. In this article, we provide an overview of recent advances and novel scientific results obtained with the 3D non-linear hybrid fluid-kinetic code JOREK, covering physics of plasma transients from the core to the scrape-off layer (SOL) both for tokamak and stellarator devices. Substantial progress was made in the physics understanding, model validation with experiments and experiment interpretation, thus, giving confidence for predictions to devices like DTT, ITER and DEMO. The topics addressed comprise a wide range: the edge physics of new operation scenarios and edge localized mode suppression; major disruptions with a focus on runaway electrons and vertical displacement events as well as disruption mitigation by shattered pellet injection; the physics mechanisms and operational limits of the flux pumping regime for sawtooth control; MHD limits of stellarators and work towards incorporating advanced edge/SOL/exhaust dynamics; continuing improvements of the code for more efficient hybrid simulations on conventional and accelerated high performance computing architectures.

disruptions↗

DOE OSTI2019

Initial results from the hotspot detection scheme for protection of plasma facing components in Wendelstein 7-X

One of the main aims of Wendelstein 7-X (W7-X), an advanced stellarator, is to investigate the quasi-steady state operation of magnetic confinement devices for nuclear fusion, for which power exhaust is an important issue. A dominant fraction of the energy leaving from the confined plasma region will be removed by 10 so-called island divertor units, which are designed to sustain a maximum heat flux of up to 10 MWm –2 . An essential prerequisite for the safe operation of a steady-state device is automatic detection of hot spots and other abnormal events. Simple temperature limits in infrared (IR) thermographic images will not be enough on their own, because of plasma-generated surface coatings and other effects summarized in the following. To protect divertor elements from overheating, and to monitor power deposition onto the divertor elements, near real-time hotspot detection algorithms for the analysis of carbon plasma facing components (PFCs) were implemented and tested in the GLADIS facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2017

Electron Temperature Fluctuation Measurements and Transport Model Validation at Alcator C-Mod

The tokamak is a type of toroidal device used to confine a fusion plasma using large magnetic fields. Tokamaks and stellarators the leading devices for confining plasmas for fusion, and the capability to predict performance in these magnetically confined plasmas is essential for developing a sustainable fusion energy source. The magnetic configuration of tokamaks and stellarators does not exist in Nature, yet, the fundamental processes governing transport in fusion plasmas are universal – turbulence and instabilities, driven by inhomogeneity and asymmetry in the plasma, conspire to transport heat and particles across magnetic field lines and can play critical roles in impurity confinement and generation of intrinsic rotation. Turbulence exists in all plasmas, and in neutral fluids as well. The study of turbulence is essential to developing a fundamental understanding of the nature of the fourth state of matter, plasmas. Experimental studies of turbulence in tokamaks date back to early scattering observations from the late 1970s. Since that time, great advances in turbulence diagnostics have been made, all of which have significantly enhanced our knowledge and understanding of turbulence in tokamaks. Through comparisons with advanced gyrokinetic theory and turbulent-transport models a great deal of evidence exists to implicate turbulent-driven transport as an important mechanism determining transport in all channels: heat, particle and momentum However, prediction and control of turbulent-driven transport remains elusive. Key to development of predictive transport models for magnetically confined fusion plasmas is validation of the nonlinear gyrokinetic transport model, which describes transport due to turbulence. Validation of gyrokinetic codes must include detailed and quantitative comparisons with measured turbulence characteristics, in addition to comparisons with inferred transport levels and equilibrium profiles. For this reason, advanced plasma diagnostics for studying core turbulence are needed in order to assess the accuracy of gyrokinetic models for turbulent-driven particle, heat and momentum transport. New core turbulence diagnostics at the world-class tokamaks Alcator C-Mod at MIT and ASDEX Upgrade at the Max Planck Institute for Plasma Physics have been designed, developed, and operated over the course of this project. These new instruments are capable of measuring electron temperature fluctuations and the phase angle between density and temperature fluctuations locally and quantitatively. These new data sets from Alcator C-Mod and ASDEX Upgrade are being used to fill key gaps in our understanding of turbulent transport in tokamaks. In particular, this project has results in new results on the topics of the Transport Shortfall, the role of ETG turbulence in tokamak plasmas, profile stiffness, the LOC/SOC transition, and intrinsic rotation reversals. These data are used in a rigorous process of “Transport model validation”, and this group is a world-leader on using turbulence models to design new hardware and new experiments at tokamaks. A correlation electron cyclotron emission (CECE) diagnostic is an instrument used to measure micro-scale fluctuations (mm-scale, compared to the machine size of meters) of electron temperature in magnetically confined fusion plasmas, such as those in tokamaks and stellarators. These micro-scale fluctuations are associated with drift-wave type turbulence, which leads to enhanced cooling and mixing of particles in fusion plasmas and limits achieving the required temperatures and densities for self-sustained fusion reactions. A CECE system can also be coupled with a reflectometer system that measured micro-scale density fluctuations, and from these simultaneous measurements, one can extract the phase between the density (n) and temperature (T) fluctuations, creating an nT phase diagnostic. Measurements of the fluctuations and the phase angle between them are extremely useful for testing and validating predictive models for the transport of heat and particles in fusion plasmas due to turbulence. Once validated, the models are used to predict performance in ITER and other burning plasmas, such as the MIT ARC design. Most recently, data from the newly developed, so-called “CECE diagnostic” [Cima 1995, White 2008] and “nT phase angle measurements” [Haese 1999, White 2010] ]will be combined with data from density fluctuation diagnostics at ASDEX Upgrade to support a long-term program of physics research in turbulence and transport that will allow for more stringent testing and validation of gyrokinetic turbulent-transport codes. This work directly impacts the development of predictive transport models in the U.S. FES program, such as TGLF, developed by General Atomics, which are used to predict performance in ITER and other burning plasma devices as part of advancing the development of fusion energy sciences.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Non-linear MHD modelling of transients in tokamaks: a review of recent advances with the JOREK code

Transient magneto-hydrodynamic (MHD) events like edge localized modes (ELMs) or disruptions are a concern for magnetic confinement fusion power plants. Research with the MHD code JOREK towards understanding control of such instabilities is reviewed here in a concise way to provide a complete overview, while we refer to the original publications for details. Experimental validation for unmitigated vertical displacement events progressed. The mechanism of vertical force mitigation by impurity injection was identified. Two-way eddy current coupling to CARIDDI was completed. Shattered pellet injection was simulated in JET, KSTAR, ASDEX Upgrade (AUG) and ITER. Benign runaway electron beam termination in JET and ITER was studied. Coupling of kinetic REs to the MHD is ongoing and a virtual RE synchrotron radiation diagnostic was developed. Regarding pedestal physics, regimes devoid of large ELMs in AUG were simulated and predictive JT60-SA simulations are ongoing. For ELM suppression by resonant magnetic perturbations (RMPs), AUG, ITER and EAST simulations were performed. A free boundary RMP model was validated against experiments. Evidence for penetrated magnetic islands at the pedestal top based on AUG experiments and simulations was found. Simulations of the naturally ELM-free quiescent H-mode in AUG and HL-3 show external kink mode formation prevents pedestal build-up towards an ELM within windows of the edge safety factor. With kinetic neutral particles, high field side high density formation in ITER was simulated and with kinetic impurities, tungsten transport in AUG RMP plasmas was studied. To capture turbulent transport, electro-static full-f particle in cell models for ion temperature gradient and trapped electron modes were established and benchmarked. Application to RMP plasmas shows enhanced turbulence in comparison to unperturbed states. Energetic particle interactions with MHD were studied. Flux pumping that prevents the safety factor on axis from dropping below unity was simulated. First non-linear stellarator applications include current relaxation in $l$ = 2 stellarators, while verification for advanced stellarators progresses.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

High-fidelity kinetic modeling of plasmas in Z-pinch and pulsed power configurations (18-ERD-048)

Plasma dynamics can be strongly influenced by fine-scale kinetic physics, which is not captured by standard fluid treatments, is challenging to simulate, and is poorly understood. The aim of this research was to investigate the fundamental fine-scale properties of collisionless plasmas in Z-pinch and pulsed power configurations by deploying novel noisefree high-accuracy computational algorithms. Significant advances were achieved through the development of robust numerical capabilities to exploit axisymmetry in kinetic simulations, the development of new methodologies to construct kinetic equilibria, and the theoretical and computational study of kinetic sheared-flow-driven transport in pulsed power systems. These new capabilities shed light on the anisotropies that can develop in Z-pinches and the transport processes that can lead to parasitic currents in pulsed power inertial confinement fusion experiments. The use of cutting-edge massively-parallel numerical algorithms to address outstanding scientific questions in plasma physics advances the mission of Lawrence Livermore National Laboratory in the areas of high energy density science and high performance computing. By bringing together kinetic theory, high-order numerical methods, high performance computing, and experimental applications, this research advances the state of the art in plasma physics modeling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Progress in Extending the High-Temperature Plasma Regime in the LHD

In this paper, we show the expansion of the high-temperature regime and related physical phenomena in the Large Helical Device (LHD). At the LHD, we have developed a high-temperature operational regime for both ion (T i ) and electron (T e ) temperatures through understanding the physical characteristics of high-temperature plasma, the enhancement of plasma heating, and the development of technical methods such as wall recycling control and impurity control. In particular, the formation of the ion internal transport barrier (i-ITB) and electron internal transport barrier (e-ITB) in the plasma core region is the key to improving plasma performance. In addition, the deuterium experiment that began in 2017 showed that, due to the isotope effect, the thermal transport of both ions and electrons is suppressed in deuterium plasma compared to light hydrogen plasma, and that a high central temperature can be achieved efficiently. Furthermore, the combination of neutral beam injection (NBI) and electron cyclotron heating (ECH) has extended the operating conditions that simultaneously maintain high T i and T e , and comprehensive research on plasma confinement characteristics with an eye to future fusion reactor conditions has progressed.

Large Helical Device↗

DOE OSTI2026

Neural-Network-Enhanced COTSIM: Advancing Predictive Capabilities for Fast DIII-D Simulations

Sustaining fusion reactions in tokamaks requires heating plasma to thermonuclear temperatures while maintaining confinement and stability. Neutral beam injection (NBI) provides heating, current drive, torque, and fueling, while electron cyclotron (EC) waves are widely used for heating and current drive; together, these actuators shape the plasma current, temperature, and density profiles. The control-oriented tokamak simulator (COTSIM), a predictive, control-oriented code, has been enhanced with neural-network surrogates for transport and sources. Turbulent transport is predicted by MMMnet—a neural-network version of the updated multimode model (MMM 9.0.10)—with significantly reduced computation time relative to MMM; neoclassical transport follows the Chang–Hinton model. NUBEAMnet, a surrogate of the Monte Carlo NUBEAM module, predicts beam-driven heating, current, and torque. EC heating and current drive use a control-oriented, empirically scaled source model; plasma resistivity follows the Spitzer formulation; bootstrap current uses the Sauter model. Equilibrium is computed using both prescribed and fixed-boundary solvers (FBSs), and the pedestal structure is modeled with an empirical pedestal model. For a representative DIII-D discharge, COTSIM predicts electron and ion temperature and safety-factor profiles in close agreement with TRANSP predictive and interpretive simulations while extending predictions through the pedestal region to the plasma edge (versus 80% of the minor radius in TRANSP). Furthermore, the equivalent COTSIM simulation runs in under 3 min compared to about 2 h for TRANSP, enabling rapid scenario planning, optimization of tokamak operation, and between-pulse control design.

Control-oriented tokamak simulator (COTSIM)↗