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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Impact of a narrow limiter SOL heat flux channel on the ITER first wall panel shaping

We report the inboard limiters for ITER were initially designed on the assumption that the parallel heat flux density in the scrape-off layer (SOL) could be approximated by a single exponential with decay length λ q . This assumption was found not to be adequate in 2012, when infra-red (IR) thermography measurements on the inner column during JET limiter discharges clearly revealed the presence of a narrow heat flux channel adjacent to the last closed flux surface. This near-SOL decay occurs with λ q ~ few mm, much shorter than the main SOL λ q , and can raise the heat flux at the limiter apex a factor up to ~4 above the value expected from a single, broader exponential. The original logarithmically shaped ITER inner wall first wall panels (FWPs) would be unsuited to handling the power loads produced by such a narrow feature. A multi-machine study involving the C-Mod, COMPASS, DIII-D and TCV tokamaks, employing inner wall IR measurements and/or inner wall reciprocating probes, was initiated to investigate the narrow limiter SOL heat flux channel. This paper describes the new results which have provided an experimental database for the narrow feature and presents an ITER inner wall FWP toroidal shape optimized for a double-exponential profile with λ q = 4 (narrow feature) and 50 mm (main-SOL), the latter also derived from a separate multi-machine database constituted recently within the International Tokamak Physics Activity. It is shown that the new shape allows the power handling capability of the original shape design to be completely recovered for a wide variety of limiter start-up equilibria in the presence of a narrow feature, even taking assembly tolerances into account. Additionally, it is further shown that the new shape has the interesting property of both mitigating the impact of the narrow feature and resulting in only a very modest increase in heat load, compared to the current design, if the narrow feature is not eventually found on ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

The interaction of the ITER first wall with magnetic perturbations

Mitigation of the multiple risks associated with disruptions and runaway electrons in tokamaks involves competing demands. Success requires that each risk be understood sufficiently that appropriate compromises can be made. Here the focus is on the interaction of short timescale magnetic-perturbations with the structure in ITER that is closest to the plasma, blanket modules covered by separated beryllium tiles. The effect of this tiled surface on the perturbations and on the forces on structures is subtle. Indeterminacy can be introduced by tile-to-tile shorting. A determinate subtlety is introduced because electrically separated tiles can act as a conducting surface for magnetic perturbations that have a normal component to the surface. A practical method for including this determinate subtlety into plasma simulations is developed. The shorter the timescales and the greater the localization, particularly in the toroidal direction, the more important the magnetic effects of the tiles become.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Comparison of plasma start-up with high Z and low Z first wall in WEST

The choice of first wall material is of paramount importance for the plasma start-up conditions in ITER and future fusion power plants. In this context, the present work focuses on the correlations between first wall impurity sources and total radiated power during plasma start-up in the tungsten (W) Environment in Steady-state Tokamak (WEST). The objective is to highlight experimental indications for a preferable combination of start-up plasma scenario and first wall materials. Until 2019, WEST featured a full high Z first wall with all limiters exposing only W surfaces to the plasma. To study the impact of a low Z first wall in WEST, boron nitride tiles were installed in the central part of the inner and outer limiters in 2020. Although visible spectroscopy and bolometry measurements show respectively a strong weakening of the WI line intensity on the limiters and a reduction of radiated power after the changeover, a degradation occurs with the accumulation of plasma exposure. In addition, the different plasma facing elements of the main chamber do not influence equally the radiated power during start-up. In both high Z and low Z environments, a clear non-linear dependence is found between the start-up radiated power and the outer limiter W impurity source. Since W seems to be the main cause for core radiation, correlation between outer limiter W sources and other impurity sources are investigated. Finally, analysis of the legacy of B powder drops on a number of start-up plasmas suggests that it is less effective at reducing radiated power when the first wall is covered with W.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Models implemented in the methodological approach to design the initial STEP first wall contour

The official Spherical Tokamak for Energy Production mission aims to demonstrate the ability to generate net electricity from fusion with the STEP Prototype Power plant. One of the key technological and engineering challenges in fusion power plants is managing the loads on the first wall within acceptable limits. Therefore, the conceptual design development of the STEP Prototype Power plant needs to be based on load estimates derived using legitimate plasma physics assumptions through dynamic and flexible tools. The current design foresees the STEP main chamber first wall to withstand steady-state heat loads of up to ~1 MW/m 2 , excluding critical regions expected to receive higher heat loads such as the baffle regions approaching the divertors. These critical areas will require ad hoc assessments and will be designed with the presence of limiters. This article focuses on the models and methodology adopted for designing the 2-D poloidal contour of the STEP first wall, based on the anticipated charged particle and radiation heat loads during normal operation. Firstly, the models adopted for calculating the charged particle and radiation heat loads are introduced. The first model is validated through benchmarking against the particle tracing code SMARDDA, while the second model is verified by comparing it with data from the MAST-U experiment. Secondly, the model used to design the 2-D first wall contour according to the heat loads is explained. We acknowledge that this preliminary design stage assumes certain simplifications, notably an axisymmetric geometry, for computational efficiency and clarity in presentation. It is understood that subsequent design phases will address the complexities of real-world engineering, including non-axisymmetric effects, transient plasma scenarios, and the impact of disruptions on the first wall design. Finally, an automatic procedure based on these models is presented for defining the 2-D poloidal contour of the STEP first wall to minimize heat loads, taking into account the need to radiate most of the alpha-particle and auxiliary heating power. Here, by providing an overview of the models, methodology, and an automatic procedure, this paper contributes to the design process of the STEP first wall, addressing the engineering challenges associated with fusion power plant development.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

First Wall Design of a Tokamak Pilot Plant Using a Monte Carlo Model for 3-D Heat Flux Deposition

We present a method for calculating the heat fluxes deposited on nonaxisymmetric tokamak first wall components, allowing for a first-of-its-kind model for power handling in the tokamak far scrape-off layer (SOL). The DIV3D Monte Carlo model features strict global power conservation and can calculate the finite cross-field plasma transport into magnetically-shadowed regions, which is significant when dealing with meter-scale shadows introduced by components such as poloidal limiters or antennas. As a case study, we apply the DIV3D model to inform the distribution of first wall poloidal limiters in an ARC-class reactor device. We demonstrate that discrete protection limiters can efficiently reduce peak heat fluxes on recessed breeder wall components in the presence of significant far-SOL plasma fluxes. By varying the toroidal periodicity and radial standoff depth of the limiters, we demonstrate one of the tradeoffs that must be considered in first wall design: more limiters provide greater protection, but at the cost of reduced breeding performance. We also present the impact that radial misalignments between limiters would have on first wall power loading.

Monte Carlo methods↗
DOE OSTI2025

Simulations of Heat Transfer Using Tight-Fitting Twisted Tape Inserts for First Wall Cooling in Molten Salt Breeder Blankets

One of the major components in fusion energy systems is the fusion blanket, which has a vacuum vessel to contain the plasma. As part of the fusion blanket/vacuum vessel, the first wall and plasma-facing components require sufficient cooling to prevent material degradation during operation from the superheated plasma. Most fusion blanket concepts involve first wall and divertor coolant channels with heat transfer enhancements (HTEs) that are intended to withstand the incident high heat fluxes of 1 to 5 MW/m 2 . Twisted tape inserts are a proposed HTE that have been investigated previously for first wall cooling and monoblock divertor cooling channels and in other nonfusion heat transfer components. By inserting twisted tapes into straight pipes, the amount of turbulence in the system can be increased at lower Reynolds numbers by swirling the flow. This results in better heat transfer characteristics with marginal increases in frictional pressure losses. In particular, simulations of high-Prandtl-number fluids such as the proposed molten salt FLiBe in twisted tapes, which is prototypic to liquid immersion blankets, have not been previously explored. Here, in this study, we simulate various Prandtl numbers in pipes with twisted tape inserts using large eddy simulations to determine the effects of increasing Prandtl numbers on heat transfer performance. The quantities of particular interest are the Nusselt number and the friction factor, which were recovered using data reduction techniques to determine impacts on heat transfer and pressure losses. This work serves as a starting point for determining the feasibility of twisted tape inserts for liquid immersion blanket concepts.

LES↗
DOE OSTI2025

First-wall fluxes from large and small ELMs

Experimental analysis at DIII-D shows that small edge localized modes (ELMs) deposit a larger fraction of their energy to the first wall, compared to type-I ELMs in similar magnetic configuration and input power. The energy (λQ) and particle (λΓ) flux decay lengths in the scrape-off layer (SOL) are up to 3 and 5 times larger, respectively, for small ELMs than for larger type-I ELMs. Transport dynamics of ELM filaments in the SOL are found to be related to divertor conditions, where high divertor collisionality, typical for partially detached plasmas, is associated with increased cross field ELM radial fluxes. Results show that a sufficiently large outer wall gap and/or limiters might be needed in future scenarios to protect the first wall, if operating with small ELMs and a cold divertor. This might also have implications for RF heating in future devices, where the coupling efficiency is dependent to some degree to the outer-wall gap.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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 OSTI2015

Numerical Analysis of Coolant Flow and Heat Transfer in ITER Diagnostic First Wall

We performed numerical simulations of the ITER Diagnostic First Wall (DFW) using ANSYS workbench. During operation DFW will include solid main body as well as liquid coolant. Thus thermal and hydraulic analysis of the DFW was performed using conjugated 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. This approach includes interface between solid and liquid part of the systemAnalysis was performed using ANSYS CFX software. CFX software allows solution of heat transfer equations in solid and liquid part, and solution of the flow equations in the liquid part. Coolant flow in the DFW was assumed turbulent and was resolved using Reynolds averaged Navier-Stokes equations with Shear Stress Transport turbulence model. Meshing was performed using CFX method available within ANSYS. The data cloud for thermal loading consisting of volumetric heating and surface heating was imported into CFX Volumetric heating source was generated using Attila software. Surface heating was obtained using radiation heat transfer analysis. Our results allowed us to identify areas of excessive heating. Proposals for cooling channel relocation were made. Additional suggestions were made to improve hydraulic performance of the cooling system.

42 ENGINEERING↗
DOE OSTI2018

Study of the Interaction of Plasma with Beryllium That Is a Candidate Material for the First Wall of a Fusion Reactor

The experimental conditions for treatment of a candidate material of the first wall, beryllium, by the plasma have been determined. The plasma parameters obtained using the simulator with a plasma-beam setup during the irradiation of samples as well as the results of material studies obtained by scanning electron microscopy and energy-dispersion analysis of the beryllium surface layer have been considered.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Integrated Liquid Metal Flowing First Wall and Open-Surface Divertor for Fusion Nuclear Science Facility: Concept, Design, and Analysis

The Fusion Energy System Studies (FESS) Fusion Nuclear Science Facility (FNSF) project team in the United States is examining the use of liquid metals (LMs) for plasma-facing components (PFCs). Our approach has been to utilize an already established fusion design, FESS-FNSF, which is a tokamak-based machine with 518 MW fusion power, a 4.8-m major radius, a 1.2-m minor radius, and a machine average neutron wall loading of ~1 MW/m 2 . For this design, we propose a PFC concept that integrates a flowing LM first wall (FW) and an open-surface divertor. The flowing LM first removes the surface heat flux from the FW and then proceeds to the lower section of the vacuum chamber to form a large area LM surface for absorbing high peak surface heat flux in the divertor region. In pursuing the application of large open LM surfaces in the FNSF, two new computer codes have been developed and then applied to the analysis of free-surface magnetohydrodynamic flows and heat transfer, including fast thin flowing liquid layers over the solid FW (liquid wall), a tublike divertor, and a fast flow divertor. The analysis is aimed at optimization of the liquid wall design by matching certain proposed design criteria and also at evaluation of the maximum heat fluxes, using liquid lithium (Li) as a working fluid. It was demonstrated that the flowing Li FW (at ~2 cm and ~10 m/s) can tolerate a surface heat flux of ~1 MW/m 2 , while the open-surface Li divertor can remove a maximum high peak heat flux of 10 MW/m 2 . Furthermore, the paper also focuses on the underlying science. One such example is the evaluation and characterization of heat transfer mechanisms and heat transfer intensification in the tublike Li divertor.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Interfacial characterization by TEM and nanoindentation of W-Eurofer brazed joints for the first wall component of the DEMO fusion reactor

Highlights: • Interfacial characterization of W-Eurofer joints by TEM and nanoindentation has been accomplished. • Microstructural and mechanical properties were in complete agreement. • The combination of both nano-techniques allows to obtain valuable information about braze interfaces. - Abstract: The interfacial characterization of W-Eurofer brazed joints, which could be implemented in the first wall component of the future DEMO nuclear fusion reactor, has been carried out by TEM and nanoindentation. The Eurofer-braze interface is characterized by the formation of two reaction layers formed during the melting stage of the filler in the brazing process, which reported high hardness and modulus values. The obtained microstructure supposes a homogenous transition between the base material and the braze. However, in the W-braze interface a more defined interface is formed, where the presence of an intermetallic transition compound has been identified. The combination of both techniques allows to study the resultant microstructure and to analyze the mechanical response of the phases, which constitute the interface and could be helpful to control the brazing conditions avoiding undesirable brazeability problems.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2019

Multiple striated heat fluxes patterns on the EAST first wall generated by lower hybrid wave absorption in the scrape-off layer

Multiple striated heat fluxes on the EAST first wall are observed with lower hybrid wave (LHW) heating on the EAST device. Our analysis indicates that the heat source for the striated heat flux is from LHW absorption in the scrape-off layer (SOL). Magnetic field line tracing from the front of the LHW antenna grills is in good agreement with the striated heat flux profile. In addition the relative intensity of the LHW absorption in the SOL is consistent with the measured striated heat flux. However, the heat flux at inner strike points and outer strike points significantly decreased when the ion cyclotron resonance frequency (ICRF) wave was switched off, while there was no change for the striated heat flux, suggests that the striated heat flux is independent on the injected power to the core plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Simulation of Channel Flow with Square Ribs for Blanket First-Wall Cooling: Geometry-Specific Tuning of k-ω Model Using Adjoint Method

Cooling of the plasma-facing first wall is challenging in the design of blanket components because of the high heat flux (on the order of 𝑀𝑊/𝑚2) from the plasma, especially when a low thermal mass medium like helium is chosen as the coolant. Therefore, heat transfer enhancement in which the convective heat transfer rate is augmented by the addition of turbulence-promoting structures becomes a key initiative for providing sufficient cooling capability with helium. Previously, computational fluid dynamics simulations had been performed on pipe flows with different transverse and longitudinal ribbed geometries at Oak Ridge National Laboratory to compare the enhancement performance among different ribbed geometries. Rib shape morphing had been conducted to obtain an optimized rib profile. In the work presented here, the adjoint method is adopted in the ANSYS Fluent solver for turbulence model augmentation, and the Generalized k-ω (GEKO) turbulence model is employed because of its ability of tuning the turbulence model. The Nusselt number and pressure drop obtained from the channel flow with bottom ribbed wall experiments are used as the targets. Sensitivity analysis provides information as guidance to improve the turbulence model accuracy. The augmented GEKO model is tuned for the studied ribbed channel geometry and flow conditions, providing improved predictive accuracy within this context. Extension to other configurations offers potential but may require additional tuning and validation.

Xu, Tracy [ORNL] (ORCID:0009000193700887)↗
DOE OSTI2016

Effect of Temperature on the Desorption of Lithium from Molybdenum(110) Surfaces: Implications for Fusion Reactor First Wall Materials

Determining the strength of Li binding to Mo is critical to assessing the survivability of Li as a potential first wall material in fusion reactors. Here, we present the results of a joint experimental and theoretical investigation into how Li desorbs from Mo(110) surfaces, based on what can be deduced from temperature-programmed desorption measurements and density functional theory (DFT). Li desorption peaks measured at temperatures ranging from 711 K (1 monolayer, ML) to 1030 K (0.04 ML), with corresponding desorption onsets from 489 to 878 K, follow a trend similar to predicted Gibbs free energies for Li adsorption. Bader charge analysis of DFT densities reveals that repulsive forces between neighboring positively charged Li atoms increase with coverage and thus reduce the bond strength between Mo and Li, thereby lowering the desorption temperature as the coverage increases. In addition, DFT predicts that Li desorbs at higher temperatures from a surface with vacancies than from a perfect surface, offering an explanation for the anomalously high desorption temperatures for the last Li to desorb from Mo(110). Analysis of simulated local densities of states indicates that the stronger binding to the defective surface is correlated with enhanced interaction between Li and Mo, involving the Li 2s electrons and not only the Mo 4d electrons as in the case of the pristine surface, but also the Mo 5s electrons in the case with surface vacancies. We suggest that steps and kinks present on the Mo(110) surface behave similarly and contribute to the high desorption temperatures. These findings imply that roughened Mo surfaces may strengthen Li film adhesion at higher temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ~400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α'), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R&D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗