1,721,699 research outputs found
Interview, Willis Williams and Viola B. Muse, Slave Interview, March 20, 1937
No full textFederal Writers\u27 Project interview with Willis Williams, focusing on his years as a slave near Tallahassee, Florida. Interview conducted by Viola B. Muse, field worker.https://digitalcommons.usf.edu/formerly_enslaved_narratives/1032/thumbnail.jp
Interview, Charles Coates and Viola B. Muse, Slave Interview, December 3, 1936
No full textFederal Writers\u27 Project interview with Father Charles Coates, who was born into slavery in Virginia and later sold to a plantation in Georgia; Coates lived in Jacksonville, Florida later in life. Interview conducted and summarized by Viola B. Muse, field worker.https://digitalcommons.usf.edu/formerly_enslaved_narratives/1009/thumbnail.jp
Viola B. Murphy
No full textViola B. Murphy who attended Howard University has arrived in England to serve the armed forces as an American Red Cross staff assistant.https://dh.howard.edu/arotc_photos/1008/thumbnail.jp
Divertor currents optimization procedure for JET-ILW high flux expansion experiments
No full textThis paper deals with a divertor coil currents optimized procedure to design High Flux Expansion (HFE) configurations in the JET tokamak aimed to study the effects of flux expansion variation on the radiation fraction and radiated power re-distribution. A number of benefits of HFE configuration have been experimentally demonstrated on TCV, EAST, NSTX and DIII-D tokamaks and are under investigation for next generation devices, as DEMO and DTT. The procedure proposed here exploits the linearized relation between the plasma-wall gaps and the Poloidal Field (PF) coil currents. Once the linearized model is provided by means of CREATE-NL code, the divertor coils currents are calculated using a constrained quadratic programming optimization procedure, in order to achieve HFE configuration. Flux expanded configurations have been experimentally realized both in ohmic and heated plasma with and without nitrogen seeding. Preliminary results on the effects of the flux expansion variation on total power radiation increase will be also briefly discussed. © 2018 Elsevier B.V
TECXY Study of a Liquid Lithium Divertor for DEMO
No full textDivertor targets made out of a capillary porous system (CPS) filled with liquid lithium, have been proposed as an alternative to standard, solid state plates. In the current work we simulate the DEMO edge plasma in either a standard single-null or snowflake divertor configuration. Our tool is the 2D code TECXY. Lithium ablated from the target plate surface and released into the plasma is shown here to partially screen the incoming heat flux. Lithium's moderate SOL radiation levels suggest additional seeding to be beneficial. Very high heat fluxes to the divertor need to be avoided, as intensive lithium evaporation might unacceptably pollute the plasma. © 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH & Co. KGaA Weinheim. This is an open access article under the terms of the Creative Commons Attribution License, which permits use, distribution and reproduction in any medium, provided the original work is properly cited
Effect of the relative shift between the electron density and temperature pedestal position on the pedestal stability in JET-ILW and comparison with JET-C
No full textThe electron temperature and density pedestals tend to vary in their relative radial positions, as observed in DIII-D (Beurskens et al 2011 Phys. Plasmas 18 056120) and ASDEX Upgrade (Dunne et al 2017 Plasma Phys. Control. Fusion 59 14017). This so-called relative shift has an impact on the pedestal magnetohydrodynamic (MHD) stability and hence on the pedestal height (Osborne et al 2015 Nucl. Fusion 55 063018). The present work studies the effect of the relative shift on pedestal stability of JET ITER-like wall (JET-ILW) baseline low triangularity (δ) unseeded plasmas, and similar JET-C discharges. As shown in this paper, the increase of the pedestal relative shift is correlated with the reduction of the normalized pressure gradient, therefore playing a strong role in pedestal stability. Furthermore, JET-ILW tends to have a larger relative shift compared to JET carbon wall (JET-C), suggesting a possible role of the plasma facing materials in affecting the density profile location. Experimental results are then compared with stability analysis performed in terms of the peeling-ballooning model and with pedestal predictive model EUROPED (Saarelma et al 2017 Plasma Phys. Control. Fusion). Stability analysis is consistent with the experimental findings, showing an improvement of the pedestal stability, when the relative shift is reduced. This has been ascribed mainly to the increase of the edge bootstrap current, and to minor effects related to the increase of the pedestal pressure gradient and narrowing of the pedestal pressure width. Pedestal predictive model EUROPED shows a qualitative agreement with experiment, especially for low values of the relative shift. © EURATOM 2018
Effect of the relative shift between the electron density and temperature pedestal position on the pedestal stability in JET-ILW
No full textAnalysis of FAST snowflake divertor by EDGE2D/EIRENE
No full textThe snowflake [1,2] divertor is a proposal for solving the heat and particle exhaust problem in fusion grade plasmas. Turning the X-point into a second order null gives the possibility of radially expanding the poloidal flux in the divertor region much more than in a SD, increasing the connection length, redistributing the power load on a larger area and enhancing radiative losses. Since the efforts associated to the design of reactor-relevant configurations, like the snowflake, are large, ENEA is studying this configuration using efficient and flexible numerical tools to design and optimise tokamak equilibrium configurations. Such studies are applied to the Divertor Test Tokamak FAST, a satellite tokamak proposed for the European roadmap towards fusion. © 2014 Elsevier B.V
Comparative study of a conventional, Quasi-snowflake and Liquid Lithium divertor for the DTT (Divertor Test Tokamak) tokamak
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