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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Lagrangian particle simulation of hydrogen pellets and SPI into runaway electron beam in ITER

Numerical studies of the ablation of pellets and shattered pellet injection (SPI) fragments into a runaway electron beam in ITER have been performed using a time-dependent pellet ablation code [Samulyak et al., Nucl. Fusion, 61(4), 046007 (2021)]. The code resolves detailed ablation physics near pellet fragments and large-scale expansion of ablated clouds. The study of a single-fragment ablation quantifies the influence of various factors, in particular, the impact ionization by runaway electrons and cross-field transport models, on the dynamics of ablated plasma and its penetration into the runaway beam. Simulations of SPI performed using different numbers of pellet fragments study the formation and evolution of the ablation clouds and their large-scale dynamics in ITER. In conclusion, the penetration depth of the ablation clouds is found to be of the order of 50 cm.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Theoretical investigation of braking of tearing mode rotation by resistive walls in ITER

The locking of the 2/1 tearing mode to the resistive wall in the ITER tokamak (15 MA inductive scenario 2) is investigated theoretically using a cylindrical asymptotic matching model. The model takes into account the fact that ITER plasmas will effectively be surrounded by two walls; the inner blanket module layer with a time constant of about 23 ms, and the outer vacuum vessel with a time constant of about 380 ms. The model also takes cognizance of the fact that neither the blanket module layer nor the vacuum vessel can be accurately described as “thin” walls (in the ordinarily accepted sense). The model incorporates changes in both the plasma poloidal and the toroidal angular velocity profiles, in response to the electromagnetic braking torque that develops at the rational surface, because it turns out that neoclassical poloidal flow-damping is not strong enough to completely suppress changes in the poloidal velocity. Finally, the model accurately calculates changes in the poloidal and toroidal plasma angular velocity profiles by evolving the full angular equations of motion, taking the electromagnetic braking torque, plasma inertia, plasma viscosity, and poloidal flow-damping into account. The time required for the 2/1 tearing mode to grow from a small amplitude to a sufficient one to lock to the walls is found to be about 3.5 s. As a result, the critical full radial island width at which wall locking is triggered is found to be about 9% of the plasma minor radius.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Chaos as the cause of randomness in tearing mode onset times in DIII-D ITER baseline scenario plasmas

The first evidence of widespread chaotic plasma dynamics relevant to the onset of disruptive tearing modes is found in DIII-D tokamak discharges. This evidence is obtained by calculating positive maximal Lyapunov exponents from Mirnov signals using the Rosenstein algorithm, implying the chaotic growth and/or rotation of magnetohydrodynamic modes in DIII-D [Luxon, Nucl. Fusion 42, 614 (2002)] plasmas. Such chaotic dynamics offer an explanation for the random tearing mode onset times previously observed in ITER baseline scenario discharges. Further, the Lyapunov timescales associated with this chaos could inhibit reliable prediction of tearing mode onset beyond the angular momentum confinement timescale; this constraint should inform the design of plasma control systems on ITER.

Chaotic dynamics↗
DOE OSTI2022

HyKKT: a hybrid direct-iterative method for solving KKT linear systems

Here, we propose a solution strategy for the large indefinite linear systems arising in interior methods for nonlinear optimization. The method is suitable for implementation on hardware accelerators such as graphical processing units (GPUs). The current gold standard for sparse indefinite systems is the LBLT factorization where L is a lower triangular matrix and B is 1×1 or 2×2 block diagonal. However, this requires pivoting, which substantially increases communication cost and degrades performance on GPUs. Our approach solves a large indefinite system by solving multiple smaller positive definite systems, using an iterative solver on the Schur complement and an inner direct solve (via Cholesky factorization) within each iteration. Cholesky is stable without pivoting, thereby reducing communication and allowing reuse of the symbolic factorization. We demonstrate the practicality of our approach on large optimal power flow problems and show that it can efficiently utilize GPUs and outperform LBL T factorization of the full system.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of the coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Electromagnetic Analysis of ITER Electron Cyclotron Emission Components and Support Structures for the Preliminary Design Review

The ITER electron cyclotron emission (ECE) diagnostic system is located at Design Shield Module (DSM) 2, Equatorial Port (EP) 9 to measure electron temperature profile and electron temperature fluctuations and also to assess nonthermal electron distributions via the oblique view. Therefore, ECE has both radial and oblique views with two couples of mirrors for two different optical views. This ECE diagnostic system shall be exposed to significant power due to unabsorbed electron cyclotron heating power in the plasma. It shall also receive large electromagnetic (EM) loads up to 100 MN/m 3 force density due to the eddy currents generated by the short 16-ms transient plasma disruption. The global EM models of EP12 and EP11 with the worst case of plasma disruption Major Disruption Downward Exponential 16 ms Category III (MD_DW_EXP16MS_CATIII) have been provided by ITER. However, the support structure for these critical components, including four mirrors, is still under the design and development phase, especially during the period of preliminary design reviews (PDRs). For those small in-vessel components that are not modeled explicitly for the PDRs, we can extract the magnetic field (B) data and flux variations (dB/dt) at those locations from the global model and then use the B and dB/dt method to calculate the eddy currents, EM forces, and torques on those components in the local submodels. In conclusion, this paper will present the detailed PDR EM analysis results of the ECE components and support structures.

Electromagnetic↗
DOE OSTI2015

Iterative Implicit Monte Carlo

Here, we describe a new Monte Carlo thermal radiation transport method that is fully implicit in the value of matter temperature used to calculate thermal emission. This method involves iterating on the matter temperature until convergence at the end of the time-step value of the matter temperature is obtained. Because the method is fully implicit, it eliminates the violations of the maximum principle that can occur in Implicit Monte Carlo (IMC) simulations on the problems on which we have tested it. The method has the drawback that it is considerably more expensive than IMC for simulations with cold opaque regions. We discuss the reasons for this, and discuss some ways that the number of iterations may be reduced.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Material migration studies with an ITER first wall panel proxy on EAST

The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8 µm is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. In conclusion, excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

The three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

Results from three-dimensional modeling of plasma edge transport and plasma–wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q = 10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q95 = 4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP coil current yield a reduction of the width of the divertor flux spreading to about 20–25 cm and cause increased peak heat fluxes back to values similar to those in the axisymmetric case. The dependencies of these features on the divertor recycling regime and the perpendicular transport assumptions, as well as toroidal averaged effects mimicking rotation of the RMP field, are discussed in the paper.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Development of a tomographic reconstruction method for axisymmetric D α emission profiles in the ITER plasma boundary

A tomographic method for reconstructing the axially symmetric D α emission profile in the ITER scrape-off layer and divertor from D α intensity measurements in the fields of view (FoV) of Vis/IR TV and Divertor Impurity Monitor diagnostics is under development. The method takes into account the strong background signal due to stray light from divertor produced by reflections from the metal wall of the vacuum chamber. The method allows to filter out the reflections in D α signals measured in the main chamber. Here, the accuracy of this method is tested within the framework of a synthetic diagnostic, which uses the Raysect and Cherab numerical codes for ray-tracing simulations. The D α emission profiles, simulated with the SOLPS and OEDGE codes, are used as synthetic experimental data. Uncertainty in light reflection properties of the first wall (FW) is the major factor affecting the accuracy of the D α emission reconstruction. The dependence on the uncertainty in the FW surface roughness and on the ratio of diffusive to specular reflection of the error in recovering useful D α signals (the direct light without reflections) in the FoV of the ITER main chamber H-alpha and Visible Spectroscopy Diagnostic is obtained.

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 OSTI2017

Statistical analysis of the Nb 3 Sn strand production for the ITER toroidal field coils

The ITER toroidal field (TF) strand procurement initiated the largest Nb 3 Sn superconducting strand production hitherto. The industrial-scale production started in Japan in 2008 and finished in summer 2015. Six ITER partners (so-called Domestic Agencies, or DAs) are in charge of the procurement and involved eight different strand suppliers all over the world, of which four are using the bronze route (BR) process and four the internal-tin (IT) process. In total more than 500 tons have been produced including excess material covering losses during the conductor manufacturing process, in particular the cabling. The procurement is based on a functional specification where the main strand requirements like critical current, hysteresis losses, Cu ratio and residual resistance ratio are specified but not the strand production process or layout. We introduce the analysis on the data acquired during the quality control (QC) process that was carried out to ensure the same conductor performance requirements are met by the different strand suppliers regardless of strand design. The strand QC is based on 100% billet testing and on applying statistical process control (SPC) limits. Throughout the production, samples adjacent to the strand pieces tested by the suppliers are cross-checked ('verified') by their respective DAs reference labs. The level of verification was lowered from 100% at the beginning of the procurement progressively to approximately 25% during the final phase of production. Based on the complete dataset of the TF strand production, an analysis of the SPC limits of the critical strand parameters is made and the related process capability indices are calculated. In view of the large-scale production and costs, key manufacturing parameters such as billet yield, number of breakages and piece-length distribution are also discussed. Furthermore, the results are compared among all the strand suppliers, focusing on the difference between BR and IT processes. Following the completion of the largest Nb 3 Sn strand production, our experience gained from monitoring the execution of the QC activities and from auditing the results from the measurements is summarised for future superconducting strand material procurement activities.

Vostner, A.↗
DOE OSTI2017

Overview of the preliminary design of the ITER plasma control system

An overview of the Preliminary Design of the ITER Plasma Control System (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Antenna-plasma coupling calculations for the ITER low-field side reflectometer

Here, the antenna-plasma coupling for the ITER low-field side reflectometer system (LFSR) was studied using a hybrid 3D reflectometer simulation code. Reflections from equilibrium profiles were simulated to optimize the power that is coupled the receive antenna. At the optimal coupling the effects of density fluctuations were also calculated, showing that the density fluctuations decrease the average coupling by as much as 6~dB. This study indicates that after some minor changes to the proposed design of this reflectometer system, the antenna-plasma coupling of the ITER LFSR is sufficient to design and build a system that meets the required specifications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Analysis of Alfven eigenmode destabilization in ITER using a Landau closure model

Alfvén eigenmodes (AE) can be destabilized during ITER discharges driven by neutral beam injection (NBI) energetic particles (EP) and alpha particles. The aim of the present study is to analyze the AE stability of different ITER operation scenarios considering multiple energetic particle species. We use the reduced magneto-hydrodynamic (MHD) equations to describe the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations of density and parallel velocity moments for the EP species including the effect of the acoustic modes. Furthermore, the AEs driven by the NBI EP and alpha particles are stable in the configurations analyzed, only MHD-like modes with large toroidal couplings are unstable, although both can be destabilized if the EP increases above a threshold. The threshold is two times the model value for the NBI EP and alpha particles in the reverse shear (RS) case, leading to the destabilization of Beta induced AE (BAE) near the magnetic axis with a frequency of kHz and toroidal or elliptical AE (TAE/EAE) in the RS region with a frequency of kHz, respectively. On the other hand, the hybrid and steady state configurations show a threshold 3 times larger with respect to the model for the alpha particle and 40 times for the NBI EP, also destabilizing BAE and TAE between the inner and middle plasma region. In addition, a extended analysis of the RS scenario where the of both alpha particles and NBI EP are above the AE threshold, multiple EP damping effects are also identified as well as optimization trends regarding the resonance properties of the alpha particle and NBI EP with the bulk plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Advances in plasma–wall interaction control for H-mode operation over 100 s with ITER-like tungsten divertor on EAST

A total power injection up to 0.3 GJ has been achieved in EAST long pulse H-mode operation of 101.2 s with an ITER-like water-cooled tungsten (W) mono-block divertor, which has steady-state power exhaust capability of 10 MWm –2 . The peak temperature of W target saturated at 12 s to the value T ~ 500 °C with a heat flux ~3.3 MW m –2 being maintained during the discharge. By tailoring the 3D divertor plasma footprint through edge magnetic topology change, the heat load was broadly dispersed and thus peak heat flux and W sputtering were well controlled. Active feedback control of H-mode detachment with D 2 fuelling or divertor impurity seeding has been achieved successfully, with excellent compatibility with the core plasma performance. Active feedback control of radiative power utilizing neon seeding was achieved with f rad = 18%–41% in H-mode operation, exhibiting potential for heat flux reduction with divertor and edge radiation. This has been further demonstrated in DIII-D high β P H-mode scenario within the joint DIII-D/EAST experiment using impurity seeding from the divertor volume. Steady-state particle control and impurity exhaust has been achieved for long pulse H-mode operation over 100 s with the W divertor by leveraging the effect of drifts and optimized divertor configuration, coupled with strong pumping and extensive wall conditioning. Methods toward the reduction of divertor W sourcing, which is of crucial importance for a metal-wall tokamak, are also explored. These advances provide vital experimental information on favourable core-edge integration for high power, long-pulse H-mode operation in EAST, ITER and CFETR.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Overview of the JET preparation for deuterium–tritium operation with the ITER like-wall

For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfvèn eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

L – H transition trigger physics in ITER-similar plasmas with applied n = 3 magnetic perturbations

In this work, the L – H transition power threshold Ρ LH is observed to increase with applied n = 3 Resonant Magnetic Perturbations (RMP) in ITER-similar-shape (ISS) plasmas with balanced neutral beam torque injection in DIII-D. The increase is most pronounced with added ECH: Ρ LH increases with decreasing edge plasma collisionality as Ρ LH / Ρ LH-08 ~ ( v *) -0.5 , where Ρ LH-08 is the 2008 ITPA multi-machine power threshold scaling. This result raises concerns for H-mode access at low edge collisionality in ITER, where resonant magnetic perturbations (RMP) may have to be applied before the L – H transition to safely suppress the first ELM. Non-axisymmetric modifications with RMP include a simultaneous reduction of the radial electric field ( E r ) well depth and E × B shear. This can be attributed to increasing edge toroidal co-current rotation, and is consistent with substantially increased local long-wavelength turbulence [measured via Beam Emission Spectroscopy (BES)]. At high RMP perturbation strength the edge electric field E r reverses sign locally (becomes positive), with changes in dominant turbulence modes. Edge magnetic stochasticity provides an attractive explanation of the observed modifications, and the observed changes in toroidal rotation and E r are consistent with a simple fluid model describing radial electron current flow along stochastic fieldlines. The observed collisionality dependence of the L-mode edge electric field with applied RMP is also qualitatively consistent with this model. Reflectometry data indicate a significant reduction of the normalized L-mode radial density gradient a/L n at high RMP field with simultaneous increase in radial particle flux and electron thermal flux from power balance analysis. We conjecture that the increase of Ρ LH with RMP results from the combined effects of reduced E × B flow shear (increasing turbulent transport levels) and toroidal/poloidal flow modulation due to edge stochasticity. Initial experiments indicate that non-resonant n = 3 magnetic perturbations lead only to relatively small changes in E r , E × B shear and fluctuation characteristics, and have less impact on the L – H transition power threshold. In conclusion, this motivates further exploration of the RMP spectrum dependence of Ρ LH for possible mitigation of the observed threshold increase.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗