Letteratura scientifica selezionata sul tema "Deuterium-Tritium fueling"

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 model. An adaptive estimator manages destabilizing uncertainties in the plasma confinement properties and the particle recycling conditions (caused by plasma–wall interactions). The controller regulates the plasma density with requests for deuterium and tritium particle injections. In ITER-like plasmas, the fusion-born alpha particles will primarily heat the plasma electrons, resulting in different electron and ion temperatures in the core. By considering separate response models for the electron and ion energies, the proposed controller can independently regulate the electron and ion temperatures by requesting that different amounts of auxiliary power be delivered to the electrons and ions. These two commands for a specific control effort (electron and ion heating) are sent to an actuator allocation module that optimally maps them to the heating actuators available to ITER: an electron cyclotron heating system (20 MW), an ion cyclotron heating system (20 MW), and two neutral beam injectors (16.5 MW each). Two different actuator allocators are presented in this work. The first actuator allocator finds the optimal mapping by solving a convex quadratic program that includes actuator saturation and rate limits. It is nonadaptive and assumes that the mapping between the commanded control efforts and the allocated actuators (i.e. the effector model) contains no uncertainties. The second actuator allocation module has an adaptive estimator to handle uncertainties in the effector model. This uncertainty includes actuator efficiencies, the fractions of neutral beam heating that are deposited into the plasma electrons and ions, and the tritium concentration of the fueling pellets. Furthermore, the adaptive allocator considers actuator dynamics (actuation lag) that contain uncertainty. This adaptive allocation algorithm is more computationally efficient than the aforementioned nonadaptive allocator because it is computed using dynamic update laws so that finding the solution to a static optimization problem is not required at every time step. A simulation study assesses the performance of the proposed adaptive burn controller augmented with each of the actuator allocation modules.

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 purity and reduced 14 MeV neutron yield. Accurate power calibrations are also essential to machine safety. Previously on JET there have been a number of questions raised on the NBI power calibration, in particular following the Trace Tritium Experiments (TTEs). Operator activities on the tokamak NBI system, including calibrations, were performed in 2020. Following these activities, a series of plasma experiments were devised to further corroborate the T NBI power by comparing the plasma response to the D NBI power. A series of stationary, L-mode plasmas were performed on JET with different beam combinations used in different phases of the same pulse. By comparing the plasma response for D and T NBI it was possible to corroborate the T NBI power calibration using the D NBI power calibration. The stored energy as measured by magnetic diagnostics, corrected for fast particle stored energy, show that the uncertainty in NBI power calibration in T is comparable to that in D.

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 50% lower in the Tritium plasma throughout the whole plasma radius. The isotope mass dependence in the particle transport channel is negligible, supported also by the perturbative particle transport analysis with gas puff modulation. The comparison of the edge particle fuelling or ionisation profiles from the EDGE2D-EIRENE simulations show that the absolute density differences that are necessary for the dimensionless match in the confined plasma dominate over any isotope mass dependencies of particle fuelling and ionization profiles at the plasma edge. Local GENE simulation results indicate a mild anti-gyroBohm effect at ρ t o r = 0.6 and thereby a small isotope mass dependence in favour of Tritium on heat transport and a negligible effect on particle transport. A significant fraction of the isotope scaling and reduced heat transport observed in the Tritium plasma is not captured in the GENE and ASTRA-TGLF-SAT2 simulations by simply changing the isotope mass for the same input profiles.

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 fuelling or Ne seeding. For the pre-fusion plasma operation (PFPO) phase our aim was to develop robust scenarios and our simulations show that commissioning and operation of the ITER neutral beam (NB) to full power should be possible in 15 MA/5.3 T L-mode H plasmas with pellet fuelling and 20 MW of ECRH. For He plasmas gas fuelling alone allows access to H-mode at 7.5 MA/2.65 T with 53–73 MW of additional heating, since after application of NB and during the L–H transition, the modelled density build-up quickly reduces the NB shine-through losses to acceptable levels. This should allow the characterisation of ITER H-mode plasmas and the demonstration of ELM control schemes in PFPO-2. In ITER DT plasmas we varied the fuelling and heating schemes to achieve a target fusion gain of Q = 10 and to exit the plasma from such conditions with acceptable divertor loads. The use of pellets in DT can provide a faster increase of the density in L-modes, but it is not essential for unrestricted NB operation due to the lower shine-through losses compared to H. During the H–L transition and current ramp-down, gas fuelling and Ne seeding are required to keep the divertor power loads under the engineering limits but accurate control over radiation is crucial to prevent the plasma becoming thermally unstable.

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 Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) were applied with hydrogen fuelling, still at 320 °C, followed by more ICWC while the vessel cooled gradually from 320 °C to 225 °C on the 4th day. While baking alone is very efficient at recovering fuel from the main chamber, the ICWC and GDC sessions at 320 °C still removed slightly higher amounts of fuel than found previously in isotopic changeover experiments at 200 °C in JET. Finally, GDC and ICWC are found to have similar removal efficiency per unit of discharge energy. The baking week with ICWC and GDC was followed by plasma discharges to remove deposited fuel from the divertor. Raising the inner divertor strike point up to the uppermost accessible point allowed local heating of the surfaces to at least 800 °C for the duration of this discharge configuration (typically 18 s), according to infra-red thermography measurements. In laboratory thermal desorption measurements, maintaining this temperature level for several minutes depletes thick co-deposit samples of fuel. The fuel removal by 14 diverted plasma discharges is analysed, of which 9, for 160 s in total, with raised inner strike point. The initial D content in these discharges started at the low value of 3%–5%, due to the preceding baking and conditioning sequence, and reduced further to 1%, depending on the applied configuration, thus meeting the experimental target.

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 pedestal height increases (ne ∽6 × 1019m-3), the neutral penetration lengths drop below the density pedestal widths for both isotopes, and the increased penetration of hydrogen has no increased effect on the pedestal width compared to deuterium. Extrapolating to future reactor-relevant high electron density pedestals, the isotope-mass increase on neutral fueling on the high field side in hydrogen will be negligible (0.2-0.4cm) in comparison to estimates of the width of the density pedestal (6-8.5 cm). Extrapolating to other isotopes compared to deuterium, while hydrogen is an increase of 41% \ (√2 \sim 1.41), moving from deuterium to tritium the neutral penetration will decrease 19% (√(2/3) \sim 0.81) implying the isotope mass effect on neutral fueling in the pedestal will be negligible in a D-T reactor.

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, Plasma Operation Contour (POPCON) plots that map the aforementioned constraints to the temperature-density space are used to investigate which constraints are most limiting towards accessing regimes with high fusion power output. The presented POPCON plots are based on a control-oriented core-edge model that couples the nonlinear density and energy response models for the core-plasma region with SOLPS4.3 parameterizations for conditions in the edge-plasma regions (scrape-off-layer and divertor). Using this control-oriented core-edge model, a nonlinear burn controller, which aims to regulate the plasma temperature and density in the core-plasma region, is constructed in this work. This controller is augmented with an online optimization scheme that governs the control references such that the plasma can be guided towards regimes with high fusion powers while protecting the divertor targets from dangerously high heat loads. A closed-loop simulation study illustrates the capability of this burn control scheme.

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 particle loss is controlled by transient ELMy phases. Overall pellet particle flux normalised to heat flux is similar to that in previous pellet fuelling experiments on AUG and JET.

Abstract The use of spin-polarized fusion fuels would provide a significant boost towards the ignition of a burning plasma. The cross section for D+T→ α+n, would be increased by 1.5 if the fuels were injected with parallel polarization. Furthermore, our simulations demonstrate additional non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. Such benefits require the survival of spin polarizations for periods comparable to the particle confinement time. During the 1980s, calculations predicted that polarizations could survive a plasma environment, although concerns persisted regarding the cumulative impacts of wall recycling. In that era, technical challenges prevented direct tests and left the large scale fueling of a power reactor beyond reach. Over the last decades, this situation has dramatically changed. Detailed simulations of ITER have predicted negligible wall recycling in a high-power reactor, and recent advances in laser-driven sources project the capability of producing large quantities of ~100% polarized D and T. The remaining crucial step is an in-situ demonstration of polarization survival in a plasma. For this, we outline a measurement strategy using the isospin-mirror reaction, D+3He→ α+p. Polarized 3He avoids the complexities of handling tritium, while encompassing the same spin-physics. We evaluate two methods of delivering deuterium, using dynamically polarized Lithium-Deuteride (with vector polarization PV D of 70%) or frozen-spin Hydrogen-Deuteride (with PV D of 40%), together with a method of injecting optically-pumped 3He (with 65% polarization). Pellets of these materials all have long polarization decay times (~6 minutes for LiD at 2K, ~2 months for HD at 2K, and ~3 days for 3He at 77K), all far greater than a plasma shot in a research tokamak such as DIII-D (~20 s). Both species can be propelled from a single cryogenic injection gun. We review plasma requirements and strategies for detecting polarization survival. Polarization alters both fusion yields and the angular distribution of fusion products, and each of these provides a potential signal. In this paper we simulate a selection of shots with similar characteristics in a future high-Tion H plasma, and find ratios of yields from shots with fuel spins parallel and antiparallel reaching 1.3 (HD+3He) to 1.6 (LiD+3He) over a wide range of poloidal angles. (A companion paper finds sensitivity to fusion product angular distributions as reflected in the pitch angles of protons and alphas reaching the plasma facing wall.)

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 validation du processus d'ablation. Ce travail vise à combler ces vides et comporte 3 parties.- Décrire la physique du dépôt de matière, puis l'état de l'art des principaux résultats et enfin la description des systèmes d'injection de glaçons prévus pour les prochaines machines.- Modéliser le transport du glaçon dans le tuyau d'injection. Les effets pris en compte dans le modèle sont la fragilisation de la glace lors des rebonds, l'augmentation de sa température et son érosion. Le modèle donne notamment le ralentissement et la perte de masse du glaçon au cours du trajet, ainsi que l'énergie élastique stockée lié à son intégrité au sortir du tube.- Contribuer à la validation du code d'ablation HPI2, en comparant ses prédictions aux données mesurées dans les nuages d'ablation. La méthode utilisée est un calcul de jeu de données synthétiques à partir des simulations et en les comparant aux mesures. Cette méthode a permis de valider les hypothèses et approximations du modèle d'ablation susmentionné
The preferred method of fueling fusion device is the use of D and/or T pellets injected into the plasma. They are currently used, but the results cannot be extrapolated to future larger reactors where the design of the injection system and the construction of scenarios will be mainly based on simulations. It is therefore important to fill in the gaps in the existing models from the manufacture of pellets to the deposition of material in the plasma. Two lacks of knowledge appear: the modeling of the pellet transport in the injection pipe and the validation of the ablation process. This work aims to fill these gaps and consists of 3 parts.- Describe the physics of material deposition, then the state of the art of the main results and finally the description of the pellet injection systems planned for the next machines.- Model the transport of the pellet in the injection pipe. The effects taken into account in the model are the weakening of the ice during rebounds, the increase in its temperature and its erosion. The model gives in particular the slowing down and the loss of mass of the pellet during the journey, as well as the stored elastic energy linked to its integrity on leaving the tube.- Contribute to the validation of the HPI2 ablation code, by comparing its predictions to data measured in ablation clouds. The method used is a calculation of synthetic data sets from simulations and comparing them to measurements. This method made it possible to validate the assumptions and approximations of the ablation model

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 System: transfers energy to the plasma by electromagnetic radiation; (e) Electron Cyclotron Heating & Current Drive System: uses radio waves to heat to the plasma; (f) Neutral Beam Heating & Current Drive System: accelerates Deuterium particles into the plasma; (g) Lower Hybrid Heating & Current Drive System: drives electric current into the plasma; (h) Diagnostics: measurement systems to control plasma performance, and further understand plasma physics; (i) Test Blankets: demonstrate techniques for ensuring tritium production within the tokamak. ITER In-vessel Components will be subjected to special operating and environmental conditions (neutron radiation, high heat fluxes, electromagnetic forces, etc.). The effects of irradiation on them, including embrittlement, swelling and creep, are not addressed in the existing commercial codes. These conditions are different from conditions in fission reactors and create challenging issues related to the design of these components. For this reason the Structural Design Criteria for ITER In-vessel Components (SDC-IC) [1] was developed for design purposes. SDC-IC was based mainly on the RCC-MR [2] code, and included rules for assessment of effect of neutron irradiation. In 2008 some issues were identified: (1) Some parts had not been fully prepared to cover all needed areas for design; (2) Some important topics needed to be improved; (3) New editions of codes on pressure equipment had been published; (4) No manufacturing rules were included, so consistency between manufacturing rules to be used and design rules in SDC-IC needed to be demonstrated; (5) Compliance with the ESP (French Decree concerning the Pressure Equipment Directive 97/23/EC for non-nuclear pressure vessels) [3] and ESPN (French Order applicable for pressure vessels intended for nuclear facilities) [4] needed to be addressed. The work carried out for Fusion For Energy (European Union’s Joint Undertaking for ITER) is: (a) Modification of design rules, incorporating rules from recently developed codes, and development of specific design rules to cover ITER specific issues and operational conditions; (b) Demonstration of consistency between design rules in SDC-IC and european standards used for manufacturing, in particular EN 13445 [5]; identifying areas where consistency is not provided; (c) Assessment of the compliance with the Essential Safety Requirements of the French Regulations (ESP and ESPN).