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Title: Thermal Management of Tungsten Leading Edges in DIII-D and ITER.

Abstract

Abstract not provided.

Authors:
; ;
Publication Date:
Research Org.:
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sandia National Laboratories, San Diego, CA
Sponsoring Org.:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
OSTI Identifier:
1380224
Report Number(s):
SAND2016-8653C
647130
DOE Contract Number:
AC04-94AL85000
Resource Type:
Conference
Resource Relation:
Conference: Proposed for presentation at the 29th Symposium on Fusion Technology held September 5-9, 2016 in Prague, Czech Republic.
Country of Publication:
United States
Language:
English

Citation Formats

Nygren, Richard E., Watkins, Jonathan G., and Barton, Joseph L. Thermal Management of Tungsten Leading Edges in DIII-D and ITER.. United States: N. p., 2016. Web.
Nygren, Richard E., Watkins, Jonathan G., & Barton, Joseph L. Thermal Management of Tungsten Leading Edges in DIII-D and ITER.. United States.
Nygren, Richard E., Watkins, Jonathan G., and Barton, Joseph L. 2016. "Thermal Management of Tungsten Leading Edges in DIII-D and ITER.". United States. doi:. https://www.osti.gov/servlets/purl/1380224.
@article{osti_1380224,
title = {Thermal Management of Tungsten Leading Edges in DIII-D and ITER.},
author = {Nygren, Richard E. and Watkins, Jonathan G. and Barton, Joseph L},
abstractNote = {Abstract not provided.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Conference:
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  • Abstract not provided.
  • The DiMES materials probe exposed tungsten blocks with 0.3 and 1 mm high leading edges to DIII-D He plasmas in 2015 and 2016 viewed with high resolution IRTV. The 1-mm edge may have reached >2400┬░ C in a 3-s shot with a (parallel) heat load of ~50 MW/m 2 and ~10 MW/m 2 on the surface based on modeling. The experiments support ITER. Leading edges were also a concern in the DIII-D Metal Tile Experiment in 2016. Two toroidal rings of divertor tiles had W-coated molybdenum inserts 50 mm wide radially. This study presents data and thermal analyses.
  • Pellet injection is the primary fueling technique planned for central fueling of the ITER burning plasma, which is a requirement for achieving high fusion gain. Injection of pellets from the inner wall has been shown on present day tokamaks to provide efficient fueling and is planned for use on ITER [1,2]. Significant development of pellet fueling technology has occurred as a result of the ITER R&D process. Extrusion rates with batch extruders have reached more than 1/2 of the ITER design specification of 1.3 cm3/s [3] and the ability to fuel efficiently from the inner wall by injecting through curvedmore » guide tubes has been demonstrated on several fusion devices. Modeling of the fueling deposition from inner wall pellet injection has been done using the Parks et al. ExB drift model [4] shows that inside launched pellets of 3mm size and speeds of 300 m/s have the capability to fuel well inside the separatrix. Gas fueling on the other hand is calculated to have very poor fueling efficiency due to the high density and wide scrape off layer compared to current machines. Isotopically mixed D/T pellets can provide efficient tritium fueling that will minimize tritium wall loading when compared to gas puffing of tritium. In addition, the use of pellets as an ELM trigger has been demonstrated and continues to be investigated as an ELM mitigation technique. During the ITER CDA and EDA the U.S. was responsible for ITER fueling system design and R&D and is in good position to resume this role for the ITER pellet fueling system. Currently the performance of the ITER guide tube design is under investigation. A mockup is being built that will allow tests with different pellet sizes and repetition rates. The results of these tests and their implication for fueling efficiency and central fueling will be discussed. The ITER pellet injection technology developments to date, specified requirements, and remaining development issues will be presented along with a plan to reach the design goal in time for employment on ITER.« less
  • Characteristics of the H-mode pedestal are studied in Type 1 ELM discharges with ITER cross-sectional shape and aspect ratio. The scaling of the width of the edge step gradient region, {delta}, which is most consistent with the data is with the normalized edge pressure, ({beta}{sub POL}{sup PED}){sup 0.4}. Fits of {delta} to a function of temperature, such as {rho}{sub POL}, are ruled out in divertor pumping experiments. The edge pressure gradient is found to scale as would be expected from infinite n ballooning mode theory; however, the value of the pressure gradient exceeds the calculated first stable limit by moremore » than a factor of 2 in some discharges. This high edge pressure gradient is consistent with access to the second stable regime for ideal ballooning for surfaces near the edge. In lower q discharges, including discharges at the ITER value of q, edge second stability requires significant edge current density. Transport simulations give edge bootstrap current of sufficient magnitude to open second stable access in these discharges. Ideal kink analysis using current density profiles including edge bootstrap current indicate that before the ELM these discharges may be unstable to low n, edge localized modes.« less
  • The maximum beta which can be sustained for a long pulse in ITER-shaped plasmas in DIII-D with q{sub 95} {approx_gt} 3, ELMs, and sawteeth is found to be limited by resistive tearing modes, particularly m/n = 3/2 and 2/1. At low collisionality comparable to that which will occur in ITER, the beta limit is a factor of two below the usually expected n = {infinity} ballooning and n = 1 kink ideal limits.