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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Increased accuracy and signal-to-noise ratio through recent improvements in infra-red video bolometer fabrication and calibration

The infra-red video bolometer (IRVB) is a diagnostic equipped with an infra-red camera that measures the total radiated power in thousands of lines of sight within a large field of view. Recently validated in MAST-U [Fderici et al., Rev. Sci. Instrum. 94, 033502 (2023)], it offers a high spatial resolution map of the radiated power in the divertor region, where large gradients are expected. The IRVB’s sensing element comprises a thin layer of high Z absorbing material, typically platinum, usually coated with carbon to reduce reflections [Peterson et al., Rev. Sci. Instrum. 79, 10E301 (2008)].Here, the possibility of using a relatively inert material such as titanium, is explored that can be produced in layers up to 1 μm compared to 2.5 μm for Pt and then coat it with Pt of the desired thickness (0.3 μm per side here) and carbon. This leads to a higher temperature signal (about 3 times) and better spatial resolution (about 4 times), resulting in higher accuracy in the measured power [Peterson et al., Rev. Sci. Instrum. 79, 10E301 (2008)]. This assembly is also expected to improve foil uniformity, as the Pt layer is obtained via deposition rather than mechanical processes [Mukai et al., Rev. Sci. Instrum. 87, 2014 (2016)].Given its multi-material composition, measuring the thermal properties of the foil assembly is vital. Various methods using a calibrated laser as a heat source have been developed, analyzing the temperature profile shape [Sano et al., Plasma and Fusion Res. 7, 2405039 (2012)] and [Mukai et al., Rev. Sci. Instrum. 89, 10E114 (2018)] or fitting the calculated laser power for different intensities and frequencies [Fderici et al., Rev. Sci. Instrum. 94, 033502 (2023)]. Here, a simpler approach is presented, which relies on analyzing the separate components of the foil heat equation for a single laser exposure in a given area. This can then be iterated over the entire foil to capture local deviations.

Magnetic Fusion Energy↗
DOE OSTI2021

Predicted Heat Flux Performance of Actively Cooled Tungsten-Armored Graphitic Foam Monoblocks

Tungsten (W)–armored graphitic foam monoblocks were developed for applications requiring high-Z plasma-facing material in long-pulse fusion experiments and ultimately deuterium-tritium fusion reactors. The monoblocks are an integrated material system combining the advantages of a chemical vapor deposited (CVD) W coating with a high-conductivity graphitic foam. The W is a high-melting-point, high-Z material with low tritium retention. The graphitic foam coupled to a swirl tube serves as a high-thermal-conductivity heat sink that cannot melt, although it can sublime at much higher temperatures than copper melts. Together, they comprise a robust plasma-facing component (PFC) weighing roughly 5% of an all-W component or 17% of a traditional W-coated copper heat sink.A single-channel mock-up consisting of four graphitic foam monoblocks equipped with a water-cooled swirl tube was fabricated for eventual testing in the 60-kW, EB-60, rastered electron beam at the Applied Research Laboratory of The Pennsylvania State University. Two monoblocks have a thin 50-μm-thick coating of pure W chemically vapor deposited over NbC and pure Nb interlayers. Two others have a 2-mm-thick pure W coating CVD on graphitic monoblocks using the same interlayers. The mock-up will be cooled with available 10 m/s, 0.7 MPa water with a 22°C inlet temperature and subjected to varying uniform heat loads up to 20 MW/m 2 . It is equipped with type-K thermocouples at various depths, and calibrated infrared thermography and spot pyrometry will be used to characterize the heated surface. Real-time water calorimetry will be used to ascertain the absorbed steady-state power and infer the heat flux during testing. Since testing cannot be done under prototypic divertor flow conditions, it is necessary to predict the thermal response of this novel PFC system and investigate the power sharing between radiation and convection at divertor heat flux levels and its inherent ability to avoid critical heat flux. Results are reported for predictions obtained from computational fluid dynamics models up to 30 MW/m 2 of steady-state uniform heat flux. Leading-edge heat loads of 30 MW/m 2 on a 2-mm-wide side strip were also investigated to ascertain if coating delamination is likely.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Identification and control of the exhaust using gas perturbations in the DIII-D tokamak

This paper presents perturbative experiments that enable the validation and development of control-oriented models for exhaust control. We identify the response of the divertor plasma and scrape-off layer in the DIII-D tokamak to deuterium and nitrogen multi-sine perturbations, in favorable and unfavorable field directions for H-mode access. We obtained good signal-to-noise ratios in the 1–10 Hz frequency range by measuring Balmer-alpha, Lyman-alpha, and N 4+ line emission, radiated power, and neutral pressure. We find a similar phase response across gas species and magnetic field directions, while the gain response is nonlinear. With these experiments, we identify a control-oriented model to design a divertor radiated power controller to track specified reference waveforms in conjunction with resonant magnetic perturbations. Although the physics basis for compatibility between detachment and resonant magnetic perturbation edge-localized mode suppression remains to be demonstrated, the present results provide a robust controller that represents a promising step toward future joint control strategies.

detachment control↗
DOE OSTI2025

The infrared imaging video bolometer at Wendelstein 7-X

The radiated power distribution is a crucial aspect of heat transport, heat load mitigation, and plasma exhaust performance. An imaging video bolometer camera has been installed to measure the plasma radiated power in the divertor region of the Wendelstein 7-X (W7-X) stellarator. This diagnostic offers a wide-angle (40° × 68°) sampling of the plasma volume in both the poloidal and toroidal directions. The field-of-view is covered with a large number (> 500) of bolometer channels, providing imaging capability. The diagnostic design is introduced here together with its data analysis procedure. A set of laboratory experiments is performed to assess the thermal properties of the gold absorber foil and their spatial uniformity. Following installation, a heat source originating from the inertially cooled front of the diagnostic is identified and filtered out. The discharge data indicate a satisfactory signal-to-noise ratio as well as spatiotemporal resolution. These represent the first toroidally resolved images of the line-integrated radiated power in the W7-X island divertor. The diagnostic was then upgraded with a thinner platinum absorber and adjusted mirrors. Early data from the most recent experimental campaign employing the upgraded design show a considerable improvement in the diagnostic performance with more bolometer channels (> 1400), extended coverage, and increased spatial resolution.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Investigation of divertor detachment induced through neon seeding and density ramp on HL-3

A new self-consistent 1D scrape-off layer model has been recently developed in BOUT++ framework, named SD1D, which includes equations for various particle species (e.g. main plasma, neutrals and impurities) and couples open databases like ADAS and AMJUEL. It is able to quickly and effectively simulate divertor detachment experiments. In this work, a typical detachment experiment (shot #6270) on HL-3 with neon seeding is simulated using the SD1D code. It is found that the target electron temperature and the target ion saturation current in the simulations are consistent with experimental results measured by Langmuir probes on the target plate. The variation of D α radiation intensity in the divertor is qualitatively similar to the measured D α signal. Following the experimental validations, different upstream densities are set in the simulations to study the impurity distribution under different plasma density conditions. It is found that increasing upstream density can be helpful for the control of the neon radiation front (closer to the target). In this work we also compare two detachment regimes in simulations. Based on the same initial experimental parameters (shot #6270) on HL-3, a scan of upstream density and a scan of neon seeding rate are carried out respectively. It is found that the role of atomic and molecular processes is different in the two detachment regimes. The current density roll-over is ascribed to a drop in the divertor ion source, and the variation of D α radiation intensity via different excitation channels is associated with the relevant collisional reaction sources.

BOUT++↗
DOE OSTI2021

Development of an integrated core–edge scenario using the super H-mode

An optimized pedestal regime called the super-H (SH) mode is leveraged to couple a fusion relevant core plasma with a high density scrape-off layer appropriate for realistic reactor power exhaust solutions. Recent DIII-D experiments have expanded the operating space of the SH regime using advanced control algorithms and investigated optimization of impurity seeding, deuterium gas puffing, and 3D magnetic perturbations. Simultaneous real-time control of the pedestal density and radiated power with in-vessel coils and nitrogen seeding enable optimal coupled divertor and pedestal conditions. Four case studies are analysed with varied levels of radiated power in the divertor volume ranging from 0 (no seeding) to 8.5 MW radiated from carbon and nitrogen emission. Plasmas with a 4.5 MW radiated power target establish a radiative mantle, leading to divertor temperatures of ~16 eV while maintaining SH-mode, and with only marginal impact on the pedestal and core performance. Increased levels of N 2 seeding with a 7.5 MW radiated power target facilitate detachment onset and divertor temperatures <5 eV, with no degradation in stored energy and the operational point remaining inside the SH-mode channel for $ > 2.5{\tau }_{\text{E}}$. Finally, a 8.5 MW radiated power target leads to partial detachment, which is so far associated with the loss of access to SH-mode pedestal conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Validation of the strike point position estimation with the local expansion method for MAST upgrade on the DIII-D tokamak

A local expansion method has been proposed for estimating the strike point position for the advanced divertor configuration on the MAST Upgrade tokamak. The paper discusses the application and assesses the performance of the technique on a long-legged divertor plasma configuration on an operating device - the DIII-D tokamak. A comparison of the spatial location of the outer strike point estimated with the local expansion method against the plasma boundary reconstruction and divertor diagnostics on the DIII-D tokamak is reported. A good agreement with the equilibrium reconstruction and diagnostic data is achieved with respect to estimation of the spatial location of the outer strike point for the long-legged divertor plasma discharge.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Impact of gas injection location and divertor surface material on ITER fusion power operation phase divertor performance assessed with SOLPS-ITER *

Abstract The ITER divertor design and performance assessment, primarily based on the SOLPS-4.3 burning plasma database (Pitts R. et al 2019 Nucl. Mater. Energy 20 100696), assumes the use of beryllium (Be) as the divertor surface material and the injection of gas from the main chamber top. However, the current ITER baseline favors gas injection from the more toroidally symmetric sub-divertor region. This paper evaluates the implications of these assumptions for divertor performance in the ITER fusion power operation phase. The impact of the divertor surface material and the gas injection location on the main ions mirrors the hydrogen only low power phase scenario shown in Park J.-S. et al (2020 Nucl. Fusion 61 016021). However, during burning plasma operation, extrinsic impurity seeding will be required. In the case of neon (Ne), studied here, impurity retention is influenced by both the divertor surface material and the fueling location. Neon leakage increases due to more energetic reflection from tungsten than beryllium, but equivalent divertor performance can be achieved by adjusting the neon seeding rate. While the impurity seeding location does not affect the distributions of impurity or radiation, the fueling location does. Top fueling provides local ionization sources mainly in the mid-SOL under detached conditions, enhancing divergences of the flux there (source-driven flow), bringing stagnation points close to the fueling location, and equilibrating flows towards both targets. In contrast, the global flow pattern (in the absence of fluid drifts) in the case of sub-divertor fueling is biased towards the inner target. Impurity flows, driven by force balance, largely mirror those of the main ion flow, including the stagnation point. The case with top fueling enhances Ne retention and corresponding radiation in the outer divertor, effectively reducing the total and peak target heat fluxes by 20%–40%, compared to the case with divertor fueling. Meanwhile, the case with outer target fueling also achieves similar reductions by enhancing plasma-neutral interactions. These results suggest the possibility that the selection of the fueling location and throughput can be used as an actuator to control impurity divertor retention and divertor radiation asymmetry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

New proposed ITER divertor design using carbon insert on tungsten to mitigate ELMs and secondary radiation effects on nearby components

Abstract Building a successful device for the magnetic fusion energy production is a great challenge. ITER is an international project of the tokamak based magnetic fusion design being developed for the demonstration of the feasibility of thermonuclear technologies for future realization of successful commercial fusion energy. A key obstacle to a successful magnetic fusion energy production is however, the performance during abnormal events including plasma disruptions and edge-localized modes (ELMs). A credible reactor design must tolerate at least a few of these transient events without serious consequences such as melting of the structure. This paper investigates and compares the performance of the current ITER tokamak design during two types of transient events, i.e., ELMs occurring at normal operation and disruptions during abnormal operation. We simulated the divertor components response using our integrated 3D HEIGHTS package. The simulations include self-consistent modeling of the interaction of the released core plasma particles with the initial solid divertor material, energy deposition processes, vaporization of divertor material, secondary plasma formation and MHD evolution, incident core particles collisions and scattering from this dense secondary plasma, photon radiation of secondary plasma, and the resulting heat loads on nearby components. Our simulations showed that using a small carbon insert around the strike point can significantly reduce the overall expected damage on the tungsten dome structure, reflector plates, and prevent tungsten vaporization and its potential core plasma contamination.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

2D coherence imaging measurements of C 2+ ion temperatures in the divertor of Wendelstein 7-X

For the first time, 2D ion temperature values are derived from coherence imaging spectroscopy (CIS) fringe contrast measurements by taking Zeeman line broadening effects into account during the analysis procedure of a spatial-heterodyne CIS instrument. This allowed 2D images of C 2+ ion temperatures (T i ) across the 3D-shaped island divertor of the Wendelstein 7-X stellarator. Ion temperatures ranging from 10 to 20 eV are observed for the C 2+ impurity species in the region above the divertor targets. During the transition from the attached to the detached plasma state, the C 2+ radiation zone moves from close to the divertor target towards the last closed flux surface. Within this radiation zone, C 2+ temperature does not decrease significantly. Experimentally, the coherence imaging measurements were cross-calibrated at one poloidal cross-section using a high resolution Echelle spectrometer, that shared its sightlines with the coherence imaging diagnostic. The spectra demonstrated that, apart from Doppler broadening, the Zeeman effect significantly contributes to the spectral line broadening and cannot be neglected when analyzing the CIS contrast data for T i extraction in the edge and scrape-off-layer of Wendelstein 7-X (W7-X), due to the relatively low temperatures (T i < 100 eV) and high magnetic fields (B ≈ 2.5 T).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Multi-X-point radiation and its dynamics in up/down asymmetry in plasma detachment regimes in Wendelstein 7-X

In Wendelstein 7-X (W7-X), stable plasma detachment in ECRH hydrogen plasmas is routinely achieved with the standard magnetic configuration, containing five magnetic islands in the scrape-off layer (SOL). Plasma detachment induced by intrinsic carbon impurities is characterized by a high radiation fraction (ƒ rad ) with significantly reduced divertor heat load and particle flux. In this work, we present the radiation dynamics during the detached plasma (DP) phase, such as the shift of radiation zones in radial and poloidal directions towards the X-points as ƒ rad increases (∼0.6–0.9), as well as the penetration of the radiation layer into the confinement region in the deep DP phase with ƒ rad approaching unity. In particular, a structure of multi-X-point radiation (multi-XPR) with an up/down asymmetry in the DP phase is highlighted, which is revealed by bolometer tomography and is further confirmed by video diagnostics. The multi-XPR structure forms helical 3D bands aligned with W7-X field periodicity. Field reversal experiments show that the brightest XPR displaces between the upper and lower SOL regions, suggesting the E× B drift effects. This paper presents the multi-XPR structure observed in the W7-X plasma in detail for the first time. A simplified model considering the influence of the poloidal E× B drift ( V d ) on the impurity flow in the SOL shows that the downstream drift toward the target or target shadow region ( V d < 0) decreases the impurity content, while the upstream drift toward the LCFS ( V d > 0) increases the impurity content. On this basis, the poloidal drift potentially leads to an up/down asymmetry of impurity density in the SOL despite the symmetry magnetic topology. The dynamics of the up/down asymmetry in the multi-XPR structure is also related to the magnitude V d /D (normalized to the impurity diffusivity), with an additional effect owing to the radial inward shift of the emission zone. These results provide new insights into impurity-induced detachment dynamics and provide a basis for improving 3D modeling of impurity transport, which typically does not consider drifts.

E × B drift↗