Relevant bibliographies by topics / Deuterium-Tritium fueling

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Conference papers

Academic literature on the topic 'Deuterium-Tritium fueling'

Author: Grafiati
Published: 8 June 2024
Last updated: 31 July 2025

Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles

Select a source type:

Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'Deuterium-Tritium fueling.'

Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.

You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.

Journal articles on the topic "Deuterium-Tritium fueling"

1

Graber, V., and E. Schuster. "Nonlinear burn control in ITER using adaptive allocation of actuators with uncertain dynamics." Nuclear Fusion 62, no. 2 (2022): 026016. http://dx.doi.org/10.1088/1741-4326/ac3cd8.

Full text
Abstract:
Abstract ITER will be the first tokamak to sustain a fusion-producing, or burning, plasma. If the plasma temperature were to inadvertently rise in this burning regime, the positive correlation between temperature and the fusion reaction rate would establish a destabilizing positive feedback loop. Careful regulation of the plasma’s temperature and density, or burn control, is required to prevent these potentially reactor-damaging thermal excursions, neutralize disturbances and improve performance. In this work, a Lyapunov-based burn controller is designed using a full zero-dimensional nonlinear
APA, Harvard, Vancouver, ISO, and other styles
2

King, D. B., R. Sharma, C. D. Challis, et al. "Tritium neutral beam injection on JET: calibration and plasma measurements of stored energy." Nuclear Fusion 63, no. 11 (2023): 112005. http://dx.doi.org/10.1088/1741-4326/acee97.

Full text
Abstract:
Abstract Neutral beam injection (NBI) is a flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to the various plasma configurations required in the Tritium and mixed deuterium-tritium experimental campaign on the Joint European Torus (JET) device. High NBI power was required for high fusion yield and alpha particle studies and to provide mixed deuterium-tritium (D-T) fuelling in the plasma core, it was necessary to operate the JET NBI systems in both deuterium and tritium. Further, the pure tritium experiments performed required T NBI for high isotopic p
APA, Harvard, Vancouver, ISO, and other styles
3

Tala, T., A. E. Järvinen, C. F. Maggi, et al. "Isotope mass scaling and transport comparison between JET Deuterium and Tritium L-mode plasmas." Nuclear Fusion 63, no. 11 (2023): 112012. http://dx.doi.org/10.1088/1741-4326/acea94.

Full text
Abstract:
Abstract The dimensionless isotope mass scaling experiment between pure Deuterium and pure Tritium plasmas with matched ρ ∗ , ν ∗ , β n , q and T e / T i has been achieved in JET L-mode with dominant electron heating (NBI+ohmic) conditions. 28% higher scaled energy confinement time B t τ E , t h / A is found in favour of the Tritium plasma. This can be cast in the form of the dimensionless energy confinement scaling law as Ω i τ E , t h ∼ A 0.48 ± 0.16 . This significant isotope mass scaling is consequently seen in the scaled one-fluid heat diffusion coefficient A χ e f f / B t which is around
APA, Harvard, Vancouver, ISO, and other styles
4

Militello Asp, E., G. Corrigan, P. da Silva Aresta Belo, et al. "JINTRAC integrated simulations of ITER scenarios including fuelling and divertor power flux control for H, He and DT plasmas." Nuclear Fusion 62, no. 12 (2022): 126033. http://dx.doi.org/10.1088/1741-4326/ac90d4.

Full text
Abstract:
Abstract We have modelled self-consistently how to most efficiently fuel ITER hydrogen (H), helium (He) and deuterium–tritium (DT) plasmas with gas and/or pellets with the integrated core and 2D SOL/divertor suite of codes JINTRAC. This paper presents the first overview of full integrated simulations from core to divertor of ITER scenarios following their evolution from X-point formation, through L-mode, L–H transition, steady-state H-mode, H–L transition and current ramp-down. Our simulations respect all ITER operational limits, maintaining the target power loads below 10 MW m−2 by timely gas
APA, Harvard, Vancouver, ISO, and other styles
5

Wauters, T., D. Matveev, D. Douai, et al. "Isotope removal experiment in JET-ILW in view of T-removal after the 2nd DT campaign at JET." Physica Scripta 97, no. 4 (2022): 044001. http://dx.doi.org/10.1088/1402-4896/ac5856.

Full text
Abstract:
Abstract A sequence of fuel recovery methods was tested in JET, equipped with the ITER-like beryllium main chamber wall and tungsten divertor, to reduce the plasma deuterium concentration to less than 1% in preparation for operation with tritium. This was also a key activity with regard to refining the clean-up strategy to be implemented at the end of the 2nd DT campaign in JET (DTE2) and to assess the tools that are envisaged to mitigate the tritium inventory build-up in ITER. The sequence began with 4 days of main chamber baking at 320 °C, followed by a further 4 days in which Ion Cyclotron
APA, Harvard, Vancouver, ISO, and other styles
6

Lennholm, M., L. Piron, D. Valcarcel, et al. "Fusion Burn Regulation via Deuterium Tritium Mixture Control in the Joint European Torus." PRX Energy 4, no. 2 (2025). https://doi.org/10.1103/prxenergy.4.023007.

Full text
Abstract:
The first generation of nuclear fusion reactors is expected to operate using a mixture of deuterium (D) and tritium (T) fuel. Controlling the D:T ratio is a promising option to control the fusion burn rate. The Joint European Torus (JET), as the only operational tokamak that can use tritium, is uniquely placed to test the feasibility of such control. Experiments carried out in 2023, during the third JET D-T campaign, have demonstrated effective feedback control of the D:T ratio under H-mode conditions. The D:T ratio was measured using visible spectroscopy and tritium was injected via gas valve
APA, Harvard, Vancouver, ISO, and other styles
7

Chaban, Ryan A., Saskia Mordijck, Aaron Michael Rosenthal, et al. "The role of isotope mass on neutral fueling and density pedestal structure in the DIII-D tokamak." Nuclear Fusion, January 22, 2024. http://dx.doi.org/10.1088/1741-4326/ad2113.

Full text
Abstract:
Abstract Experimental measurements on DIII-D of hydrogen neutral penetration lengths (λn0 ) on the high field side are longer by a factor of √2 than for deuterium consistent with the thermal velocity ratio for neutrals at the same temperature (vth H / vth H = √2). This ratio is constant for both low and high pedestal electron density. At low pedestal density (ne ∽4 × 1019m-3), the neutral penetration length is greater than the density pedestal width for both isotopes, and the additional 41% increase of neutral penetration in hydrogen widens the pedestal by the same amount. As the density pedes
APA, Harvard, Vancouver, ISO, and other styles
8

Hegna, C. C., D. T. Anderson, E. C. Andrew, et al. "The Infinity Two Fusion Pilot Plant baseline plasma physics design." Journal of Plasma Physics, March 26, 2025, 1–44. https://doi.org/10.1017/s0022377825000364.

Full text
Abstract:
We provide an assessment of the Infinity Two Fusion Pilot Plant (FPP) baseline plasma physics design. Infinity Two is a four-field period, aspect ratio A = 10, quasi-isodynamic stellarator with improved confinement appealing to a max-J approach, elevated plasma density and high magnetic fields (⟨B⟩ = 9 T). At the envisioned operating point [800 MW deuterium-tritium (DT) fusion], the configuration has robust magnetic surfaces based on magnetohydrodynamic (MHD) equilibrium calculations and is stable to both local and global MHD instabilities. The configuration has excellent confinement propertie
APA, Harvard, Vancouver, ISO, and other styles
9

Graber, Vincent, and Eugenio Schuster. "Divertor-safe nonlinear burn control based on a SOLPS parameterized core-edge model for ITER." Nuclear Fusion, May 30, 2024. http://dx.doi.org/10.1088/1741-4326/ad521b.

Full text
Abstract:
Abstract For ITER operations, the range of desirable burning-plasma regimes with high fusion power output will be restricted by various operational constraints. These constraints include the saturation of ITER’s various heating and fueling actuators such as the neutral beam injectors, the ion and electron cyclotron heating systems, the gas puffing system, and the deuterium-tritium pellet injectors. In addition to these actuator constraints, the H-mode power threshold, divertor detachment, and the heat load on the divertor targets may apply limitations to ITER’s operational space. In this work,
APA, Harvard, Vancouver, ISO, and other styles
10

Valovic, Martin, Spyridon Aleiferis, Peter Blatchford, et al. "Fuelling of deuterium-tritium plasma by peripheral pellets in JET experiments." Nuclear Fusion, April 24, 2024. http://dx.doi.org/10.1088/1741-4326/ad42b2.

Full text
Abstract:
Abstract A baseline scenario of deuterium-tritium (D-T) plasma with peripheral high field side fuelling pellets has been produced on JET in order to mimic the situation in ITER. The isotope mix ratio is controlled in order to target the value of 50%-50% by combination of tritium gas puffing and deuterium pellet injection. Multiple factors controlling the fuelling efficiency of individual pellets are analysed with following findings: (1) prompt particle losses due to pellet triggered ELMs are detected, (2) plasmoids drift velocity might be smaller than predicted by simulation, (3) post-pellet p
APA, Harvard, Vancouver, ISO, and other styles

Dissertations / Theses on the topic "Deuterium-Tritium fueling"

1

Geulin, Eléonore. "Contribution to the modeling of pellet injection : from the injector to ablation in the plasma." Electronic Thesis or Diss., Aix-Marseille, 2023. http://www.theses.fr/2023AIXM0066.

Full text
Abstract:
La méthode privilégiée d'alimentation des machines à fusion est l'utilisation de glaçons de D et/ou T injectés dans le plasma. Ils sont utilisés actuellement, mais les résultats ne sont pas extrapolables aux futures machines de plus grande taille où le design du système d'injection et la construction de scenarii seront surtout basés sur les simulations. II est donc important de combler les vides dans les modèles existants allant de la fabrication des glaçons au dépôt de matière dans le plasma. Deux manques apparaissent : la modélisation du transport du glaçon dans le tuyau d'injection et la va
APA, Harvard, Vancouver, ISO, and other styles

Conference papers on the topic "Deuterium-Tritium fueling"

1

Martinez, Oscar, Sumalatha Yaski, Sarah Smith, Kara Godsey, David Rasmussen, and Gary Lovett. "Thermal-Structural and Shock Event Evaluations of the Fueling Pellet Injection System for ITER." In ASME 2024 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/pvp2024-123447.

Full text
Abstract:
Abstract The United States is among seven partner nations in a collaborative effort to design, build, and demonstrate fusion’s ability to become a large-scale carbon-free energy source. Each country has its own domestic agencies that contribute directly to the ITER project. US ITER, which is a US Department of Energy Office of Science project managed by Oak Ridge National Laboratory, is developing world-class engineering solutions to the design, construction, and assembly of the burning plasma experiment that can demonstrate the scientific and technological feasibility of fusion. US ITER’s sco
APA, Harvard, Vancouver, ISO, and other styles
2

Couso, Daniel, Jose´ Fano, Felicidad Ferna´ndez, et al. "Development of Codes and Standards for ITER In-Vessel Components." In ASME 2011 Pressure Vessels and Piping Conference. ASMEDC, 2011. http://dx.doi.org/10.1115/pvp2011-57611.

Full text
Abstract:
This paper describes the changes made to existing version of the Structural Design Criteria for In-vessel Components (SDC-IC) within the ITER project, as a result of the revision and update process carried out recently. Several ITER components, referred to as In-vessel Components, are located inside the ITER Vacuum Vessel: (a) Blanket System: shields the Vessel and Magnets from heat and neutron fluxes; (b) Divertor: extracts heat, helium ash and impurities from the plasma; (c) Fuelling: gas injection system to introduce fuel into the Vacuum Vessel; (d) Ion Cyclotron Heating & Current Drive
APA, Harvard, Vancouver, ISO, and other styles
We offer discounts on all premium plans for authors whose works are included in thematic literature selections. Contact us to get a unique promo code!

To the bibliography