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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Developing the hybrid scenario in DIII-D and KSTAR for W-compatible long-pulse operation

Long-pulse, high-performance operation with tungsten plasma-facing components is a crucial challenge for achieving steady-state (SS) goals in ITER and future fusion devices. This paper presents a joint international effort between the DIII-D and KSTAR tokamaks aimed at demonstrating the compatibility of the high q 95 ~ 6 hybrid scenario with a tungsten divertor and superconducting coils. A DIII-D hybrid scenario recipe is successfully tailored to KSTAR’s operational constraints, which facilitates cross-device studies. The results show successful 30 s long-pulse operation in KSTAR at a normalized beta (β N = 2.4) with a benign m/n = 4/3 mode and effective mitigation of tungsten accumulation. Despite this success, the KSTAR scenario exhibits a greater than 25% reduction in thermal confinement compared to the DIII-D reference. To investigate this discrepancy, we performed a predictive TGYRO validation study, which suggests that the difference is likely caused by a weaker temperature pedestal in the KSTAR discharge due to a higher and narrower density pedestal. In addition, such lower thermal confinement leads to higher impurity accumulation, possibly due to a change in neoclassical transport with increased inward pinch, which further degrades global confinement. We also report that stationary, long-pulse operation with sustained thermal confinement requires consistent fueling and wall conditioning. These findings highlight the importance of optimized fueling and real-time wall conditioning in achieving a high temperature pedestal and strong global confinement during stable long-pulse operation. We propose potential approaches toward fully SS, long-pulse operation, including real-time wall conditioning using an impurity dropper in KSTAR and future tokamaks.

DIII-D↗
DOE OSTI2023

PFPO plasma scenarios for exploration of long pulse operation in ITER

Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils $⩽$30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q $⩾$ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Design and analysis of actively-cooled, edge-transport diagnostic for long-pulsed operation in WEST

Next step fusion devices that will operate in steady-state will require complex plasma-facing components (PFCs) that can survive the harsh environment over long timescales not common in current devices. This will require robust plasma facing surfaces that are integrated with active cooling systems. In a collaboration between CEA and ORNL, a plasma-interacting diagnostic is being designed for the W Environment in Steady-state Tokamak (WEST) in Cadarache, France which requires plasma-facing protection like those needed for steady-state PFCs. This integrated diagnostic studies edge transport and impurity migration within WEST, and is planned to include imbedded temperature Langmuir probe sensors, as well as removeable sample slots for ex-situ surface analysis of plasma-material interactions. The entire assembly is expected to move into the plasma edge for periods up to 1000 s having an energy removeable capability of ∼6 kW and seeing a peak heat flux of more than 7 MW/m2. The assembly compliments WEST high fluence campaigns that plan for multiple 1000 s pulses. Within these specifications, the assembly will require a refractory metal plasma facing surface and integral cooling in order to function within the limited space allotted for such diagnostics. Because of the limited space and linear actuator needs, additive manufacturing of the high heat flux working end of the assembly is being considered which could allow for precision cooling-channels and lighter weight designs. Conceptual design along with simulation and analysis results will be presented for this complex diagnostic with novel PFCs.

Lumsdaine, Arnold↗
DOE OSTI2014

Cryogenic pellet production developments for long-pulse plasma operation

Long pulse plasma operation on large magnetic fusion devices require multiple forms of cryogenically formed pellets for plasma fueling, on-demand edge localized mode (ELM) triggering, radiative cooling of the divertor, and impurity transport studies. The solid deuterium fueling and ELM triggering pellets can be formed by extrusions created by helium cooled, twin-screw extruder based injection system that freezes deuterium in the screw section. A solenoid actuated cutter mechanism is activated to cut the pellets from the extrusion, inserting them into the barrel, and then fired by the pneumatic valve pulse of high pressure gas. Fuel pellets are injected at a rate up to 10 Hz, and ELM triggering pellets are injected at rates up to 20 Hz. The radiative cooling and impurity transport study pellets are produced by introducing impurity gas into a helium cooled section of a pipe gun where it deposits in-situ. A pneumatic valve is opened and propellant gas is released downstream where it encounters a passive punch which initially accelerates the pellet before the gas flow around the finishes the pellet acceleration. This paper discusses the various cryogenic pellet production techniques based on the twin-screw extruder, pipe gun, and pellet punch designs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Fuel recycling control in long pulse operation with full tungsten divertors in EAST tokamak

Fuel recycling control in long pulse discharges is studied in EAST tokamak with tungsten divertors by particle balance. Double null configuration, proper Resonant Magnetic Perturbation (RMP), divertor cryopumps and realtime lithium injection, are all very effective for recycling control. High heating power discharge leads to low recycling due to the enhancement of particle exhausting. High external gas injection is required in a low fuel recycling discharge to keep a constant density, and Dα emission is increased by external injection, therefore a higher Dα emission may indicate low fuel recycling for the same plasma density. Reproducible long pulse discharges of over 100 s with core electron temperature >10 keV, and 1056 s discharge with ~1.5 MW heating power were obtained in the EAST 2021 campaign, fuel recycling was well controlled by divertor cryopumps and realtime lithium injection. The results provide a valuable reference for the high power and long pulse operation in future fusion reactors, such as ITER and CFETR.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Advances in long pulse operation at high radio frequency power in Tore Supra

The lower hybrid current drive (LHCD) system of Tore Supra has been upgraded for long pulse operation at higher power (7–8 MW). The two launchers have coupled on plasma 3.8 MW and 2.7 MW separately. This new power capability allows extending the operational domain of Tore Supra for long pulses at higher current and density. 38 long (20 s –155 s) discharges with very low loop voltage (V{sub L} = 30-60 mV) were performed with combined LHCD (5-5.7 MW) and ICRH (1–3 MW) powers, with up to 1 GJ of injected energy. Higher LHCD efficiency, with respect to the previous long discharges, is reported. MHD stability of these discharges is very sensitive to the LHCD power and parallel wave index, in particular in the preforming phase. For theses evanescent loop voltage plasmas, the ICRH power, in excess of 1 MW, is found to have a beneficial effect on the MHD stability.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

First DIII-D-West hybrid scenario similarity experiments for iter-relevant long-pulse operation

For the first time, similarity experiments between DIII-D and WEST were performed in the ITER "hybrid-like" regime during dedicated campaigns in April and May 2025. The matched parameters include elongation, triangularity, ion ∇B drift direction toward the X-point, qprofile, and core normalized physics quantities in terms of normalized pressure, normalized gyroradius, electron collisionality, ratio of ion to electron temperature, T i /T e . Core transport physics is explored with different aspect ratio (R/a) values (typically 3 at DIII-D and 5 on WEST). DIII-D explored high-beta conditions (electromagnetic effect) with low torque injection (~0 ± 0.5 N•m) using high heating power (up to 6 MW NBI and 2 MW ECRH powers), while scanning the heating mix (ion vs electron), beta, T i /T e , core radiation via controlled tungsten injection using the Laser Blow-Off system. WEST extended operation toward long-duration pulses using its actively cooled tungsten divertor, achieving dominated electron heating regimes with reduced tungsten contamination. Boron impurity injection were scanned on WEST to control edge conditions and core performance. It is found that core confinement improves-manifested by higher electron temperature, total energy content, neutron rate, and ion temperatureunder conditions of low separatrix density, consistent with previous observations [Bourdelle et al., Nucl. Fusion 63 (2023) 056021]. Conditions for Hmode access and for ion heating in electron-dominated regimes in both WEST and DIII-D will be discussed and compared. The ratio of the thermal energy confinement time (τ E ) to the volume-averaged electron-ion collisional heat exchange time (τ e-i ) is a key parameter to enhance ion heating and potentially facilitate H-mode access in electron-heated regimes. These first-of-a-kind coordinated DIII-D and WEST experiments provide a unique multi-machine dataset to validate predictive models and to optimize ITER hybrid-scenario performance under diverse core and edge conditions.

DIII-D↗
DOE OSTI2014

Effects of magnetic configuration on divertor power and particle deposition for long pulse operation in EAST

The magnetic configuration exhibits a strong influence on the dynamics of Edge Localized Modes (ELMs), as demonstrated in the EAST superconducting tokamak. Here we find that poloidal drifts play an important role in particle deposition during the ELMs, leading to a strong up/down asymmetry in the double null divertor configuration, favoring the upper divertor for normal toroidal field, Bt, i.e., with the ion ∇B drift towards the bottom, while the heat flux distribution appears to be rather uniform during ELMs. These observations are well reproduced by the boundary plasma turbulence code, BOUT++. As divertor pumping was only available at the bottom, the preferential particle flow towards the bottom divertor associated with reverse Bt led to a preferred scenario for long pulse operation in EAST.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Tungsten control in long pulse operation: feedback from WEST to ITER

The full W environment that is now foreseen for ITER puts strong emphasis on experimental results obtained in present devices in similar conditions. In this context, the WEST tokamak is well equipped to bring key contributions to the preparation of ITER operation, thanks to its capability to perform long pulses in dominant electron heating, torque-free scheme based on RF systems, and its ITER-grade actively cooled divertor. Recent results of interest cover the understanding of tungsten contamination and the evaluation of conditioning methods, the feedback on the current ramp-up phase, and the ageing of the ITER divertor components in L-mode attached condition. Further, new directions are being explored, in particular low divertor temperature operation that should be the normal operating regime of ITER. We also report on tungsten peaking physics, and the role of light impurity content.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Cesium dynamics in long pulse operation of negative hydrogen ion sources for fusion

Large scale negative hydrogen ion sources operating stable for 1 h (cw mode) are required for the neutral beam heating system of the fusion experiment ITER. The formation of negative hydrogen ions relies on the surface effect for which cesium is evaporated into the source. In order to monitor the cesium dynamics the laser absorption spectroscopy technique is applied to the long pulse test facility MANITU. In the vacuum phase, without plasma operation the evaporation of cesium and the built-up of the cesium in the source are measured. Typical neutral cesium densities are 10{sup 15} m{sup -3}. During plasma operation and after the plasma phase a high cesium dynamics is observed, showing also depletion of cesium during long pulses with low cesium amount. The co-extracted electron current decreases with the cesium amount to a certain level whereas the ion current indicates an optimum density range.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Experience with Long-Pulse Operation of the PIP2IT Warm Front End

The warm front end of the PIP2IT accelerator, assem-bled and commissioned at Fermilab, consists of a 15 mA DC, 30 keV H- ion source, a 2 m long Low Energy Beam Transport (LEBT) line, and a 2.1 MeV, 162.5 MHz CW RFQ, followed by a 10 m long Medium Energy Beam Transport (MEBT) line. A part of the commissioning efforts involves operation with the average beam power emulating the operation of the proposed PIP-II accelera-tor, which will have a duty factor of 1.1% or above. The maximum achieved power is 5 kW (2.1 MeV x 5 mA x 25 ms x 20 Hz). This paper describes the difficulties encoun-tered and some of the solutions that were implemented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

Spectral broadening measurement of the lower hybrid waves during long pulse operation in Tore Supra

On many tokamaks (C-Mod, EAST, FTU, JET, HT-7, TS), a decrease in current drive efficiency of the Lower Hybrid (LH) waves is observed in high electron density plasmas. The cause of this behaviour is believed to be: Parametric Instabilities (PI) and Scattering from Density Fluctuations (SDF). For the ITER LH system, our knowledge must be improved to avoid such effects and to maintain the LH current drive efficiency at high density. The ITPA IOS group coordinates this effort [1] and all experimental data are essential to validate the numerical codes in progress. Usually the broadening of the LH wave frequency spectrum is measured by a probe located in the plasma edge. For this study, the frequency spectrum of a reflected power signal from the LH antenna was used. In addition, the spectrum measurements are compared with the density fluctuations observed on RF probes located at the antenna mouth. Several plasma currents (0.6 to 1.4 MA) and densities up to 5.2 × 10{sup 19} m−3 have been realised on Tore Supra (TS) long pulses and with high injected RF power, up to 5.4 MW-30s. This allowed using a spectrum analyser to make several measurements during the plasma pulse. The side lobe amplitude, shifted by 20-30MHz with respect to the main peak, grows with increasing density. Furthermore, for an increase of plasma current at the same density, the spectra broaden and become asymmetric. Some parametric dependencies are shown in this paper.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2015

Initial operation of high power ICRF system for long pulse in EAST

The ICRF heating system on EAST upgraded by active cooling aims for long pulse operation. In this paper, the main technical features of the ICRF system are described. One of a major challenges for long pulse operation is RF-edge interactions induced impurity production and heat loading. In EAST, ICRF antenna protections and Faraday screen bars damaged due to LH electron beam are found. Preliminary results for the analysis of the interaction between LHCD and ICRF antenna are discussed. Increase of metal impurities in the plasma during RF pulse and in a larger core radiation are also shown. These RF-edge interactions at EAST and some preliminary results for the optimizing RF performance will be presented.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Long plasma duration operation analyses with an international multi-machine (tokamaks and stellarators) database

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international Coordination on International Challenges on Long duration OPeration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

WEST full tungsten operation with an ITER grade divertor

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

A long-pulse repetitive operation magnetically insulated transmission line oscillator

The improved magnetically insulated transmission line oscillator (MILO) is a gigawatt-class L-band high power microwave tube. It has allowed us to generate 3.1 GW pulse of 40 ns duration in the single-pulse operation and 500 MW pulse of 25 ns duration in the repetition rate operation. However, because of the severe impedance mismatch, the power conversion efficiency is only about 4% in the repetition rate operation. In order to eliminate the impedance mismatch and obtain repetitive long-pulse high-power microwave (HPM), a series of experiments are carried out and the recent progress is presented in this paper. In the single-pulse operation, when the diode voltage is 466 kV and current is 41.6 kA, the radiated microwave power is above 2.2 GW, the pulse duration is above 102 ns, the microwave frequency is about 1.74 GHz, and the power conversion efficiency is about 11.5%. In the repetition rate operation, under the condition of the diode voltage about 400 kV, beam current about 38 kA, the radiated microwave power is about 1.0 GW, the pulse duration is about 85 ns. Moreover, the radiated microwave power and the pulse duration decline little by little when the shot numbers increase gradually. The experimental results show that the impedance matching is a vital factor for HPM systems and one of the major technical challenges is to improve the cathode for the repetition rate operation MILO.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

The R and D progress of 4 MW EAST-NBI high current ion source

A high current ion source, which consists of the multi-cusp bucket plasma generator and tetrode accelerator with multi-slot apertures, is developed and tested for the Experimental Advanced Superconducting Tokamak neutral beam injector. Three ion sources are tested on the test bed with arc power of 80 kW, beam voltage of 80 keV, and beam power of 4 MW. The arc regulation technology with Langmuir probes is employed for the long pulse operation of ion source, and the long pulse beam of 50 keV @ 15.5 A @ 100 s and 80 keV @ 52A @ 1s are extracted, respectively.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Potential Impacts of Liquid-Metal Plasma-Facing Components on Heating and Current Drive Actuators for a Fusion Nuclear Science Facility

This paper addresses the potential impact of liquid metal (LM) plasma-facing components (PFCs) for the heating and current drive (H&CD) actuators on the Fusion Nuclear Science Facility (FNSF) fusion reactor. Fulfilling the high neutron fluence mission of the FSNF requires steady-state operation for extremely long pulses (months to years) between maintenance opportunities. The use of LM as a surface material is one strategy for extending the lifetime of the PFCs for long pulse operation in a high heat flux, high neutron flux environment like that of the FNSF. Liquid metal PFCs provide possible pathways forward on many difficult aspects of a fusion reactor; however, the LM PFCs also bring new challenges and unknowns with respect to the H&CD actuators needed to provide steady-state operation. The development of LM-compatible materials for radio-frequency (RF) antennas will be critical, as well as strategies for minimizing contamination of antenna surfaces and the core plasma. Successful deployment of LM PFCs on the FNSF will require operational experience with RF in a LM environment both on test stands and in an integrated toroidal environment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗