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Conceptual Engineering Method for Attenuating He Ion Interactions on First Wall Components in the Fusion Test Facility (FTF) Employing a Low-Pressure Noble Gas

Description: It has been shown that post detonation energetic helium ions can drastically reduce the useful life of the (dry) first wall of an IFE reactor due to the accumulation of implanted helium. For the purpose of attenuating energetic helium ions from interacting with first wall components in the Fusion Test Facility (FTF) target chamber, several concepts have been advanced. These include magnetic intervention (MI), deployment of a dynamically moving first wall, use of a sacrificial shroud, designing the target chamber large enough to mitigate the damage caused by He ions on the target chamber wall, and the use of a low pressure noble gas resident in the target chamber during pulse power operations. It is proposed that employing a low-pressure (~ 1 torr equivalent) noble gas in the target chamber will thermalize energetic helium ions prior to interaction with the wall. The principle benefit of this concept is the simplicity of the design and the utilization of (modified) existing technologies for pumping and processing the noble ambient gas. Although the gas load in the system would be increased over other proposed methods, the use of a "gas shield" may provide a cost effective method of greatly extending the first wall of the target chamber. An engineering study has been initiated to investigate conceptual engineering metmethods for implementing a viable gas shield strategy in the FTF.
Date: September 21, 2009
Creator: C.A.Gentile, W.R.Blanchard, T.Kozub, C.Priniski, I.Zatz, S.Obenschain
Partner: UNT Libraries Government Documents Department

Correlation of Neutral Beam Injection Parameters and Core B with Anomalous First-Wall Heating During QH-Mode

Description: Anomalous first-wall heating has been observed far from the divertor strike point during QH-mode in DIII-D, with measured heat flux comparable to that at the outer strike point. The data are consistent with deuterium ions of approximately the pedestal energy carrying the anomalous heat flux. Although an instability has not been identified that is correlated with the anomalous heat flux, two classes of behavior have been observed: one in which the anomalous heat flux depends linearly on core {beta}, and another class with no {beta}-dependence. The anomalous heat flux depends strongly on the injected beam energy of the non-tangentially-injected neutral beams but not that of the tangential beams.
Date: May 15, 2006
Creator: Lasnier, C; Burrell, K; deGrassie, J; Rhodes, T; VanZeeland, M & Watkins, J
Partner: UNT Libraries Government Documents Department

Lifetime survivability of contaminated target-chamber optics

Description: Target chambers used for Inertial Confinement Fusion (ICF) expose laser optics to a very hostile environment, not only from high-fluence laser irradiation but also x-ray irradiation and particulate debris from targets and chamber wall materials. Expendable debris shields provide the first line of defense to more costly optics upstream in the laser beam path to contaminants generated within the target chamber. However, the replacement of a large number of debris shields is also an expensive proposition so that extending their usable lifetime within the chamber is important. We have conducted tests to show that optics can both be cleaned and damaged by laser irradiation at 355 nm after being contaminated with potential chamber-wall materials such as B{sub 4}C and Al{sub 2}O{sub 3}. Such optics can survive from one to hundreds of laser shots, depending on degree of contamination and laser fluence levels. Similarly, we have studied the survivability of optics that have been exposed to direct contamination from representative target materials irradiated in the target chamber. We have also studied the effects on optics that were not directly exposed to targets, yet received secondary exposure from the above directly-exposed samples.
Date: November 1, 1996
Creator: Rainer, F.; Anderson, A.; Burnham, A.; Milam, D. & Turner, R.
Partner: UNT Libraries Government Documents Department

Evaluation of US demo helium-cooled blanket options

Description: A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed.
Date: October 1, 1995
Creator: Wong, C.P.C.; McQuillan, B.W. & Schleicher, R.W.
Partner: UNT Libraries Government Documents Department

Foil deposition alpha collector probe for TFTR`s D-T phase

Description: A new foil deposition alpha collector sample probe has been developed for TFTR`s D-T phase. D-T fusion produced alpha particles escaping from the plasma are implanted in nickel foils located in a series of collimating ports on the detector. The nickel foils are removed from the tokamak after exposure to one or more plasma discharges and analyzed for helium content. This detector is intended to provide improved alpha particle energy resolution and pitch angle coverage over existing lost alpha detectors, and to provide an absolutely calibrated cross-check with these detectors. The ability to resolve between separate energy components of alpha particle loss is estimated to be {approx} 20%. A full 360{degree} of pitch angle coverage is provided for by 8 channels having an acceptance range of {approx} 53{degree} per channel. These detectors will be useful in characterizing classical and anomalous alpha losses and any collective alpha instabilities that may be excited during the D-T campaign of TFTR.
Date: March 1, 1995
Creator: Hermann, H.W.; Darrow, D.S.; Timberlake, J.; Zweben, S.J.; Chong, G.P.; Pitcher, C.S. et al.
Partner: UNT Libraries Government Documents Department

Proceedings of the 4th International Workshop on Tritium Effects in Plasma Facing Components

Description: The 4th International Workshop on Tritium Effects in Plasma Facing Components was held in Santa Fe, New Mexico on May 14-15, 1998. This workshop occurs every two years, and has previously been held in Livermore/California, Nagoya/Japan, and the JRC-Ispra Site in Italy. The purpose of the workshop is to gather researchers involved in the topic of tritium migration, retention, and recycling in materials used to line magnetic fusion reactor walls and provide a forum for presentation and discussions in this area. This document provides an overall summary of the workshop, the workshop agenda, a summary of the presentations, and a list of attendees.
Date: February 1, 1999
Creator: Causey, R. A.
Partner: UNT Libraries Government Documents Department

A comparison of stresses in armor joints with and without interlayers

Description: Reliable joining of armor to heat sinks for plasma facing components has been a persistent problem in fusion and a concern for the International Thermonuclear Experimental Reactor (ITER). Post-fabrication and operating stresses in heat sinks with a 1mm compliant layer (or no interlayer) between tungsten armor and a CuCrZr channel were analyzed with a 2-D finite element model with temperature dependent properties, generalized plane strain, and strain hardening.
Date: November 1, 1997
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department

Fusion Ignition Research Experiment System Integration

Description: This paper describes the current status of the FIRE configuration and the integration of the major subsystem components. FIRE has a major radius of 2 m, a field on axis of 10T, a plasma current of 6.4 MA. It is capable of 18 second pulses when operated with DT and 26 s when operated with DD. The general arrangement consists of sixteen wedged TF coils that surround a free standing central solenoid, a double wall vacuum vessel and internal plasma facing components that are segmented for maintenance through horizontal ports. Large rings located outside the TF coils are used to obtain a load balance between wedging of the intercoil case structure and wedging at the upper/lower inboard corners of the TF coil winding. The magnets are liquid nitrogen cooled and the entire device is surrounded by a thermal enclosure. The double wall vacuum vessel integrates cooling and shielding in a shape that maximizes shielding of ex-vessel components. Within the vacuum vessel, plasma-facing components frame the plasma. First wall tiles are attached directly to inboard and outboard vacuum vessel walls. The divertor is designed for a high triangularity, double-null plasma with a short inner null point-to-wall distance and near vertical outer divertor flux line. The FIRE configuration has been developed to meet the physics objectives and subsystem requirements in an arrangement that allows remote maintenance of in-vessel components and hands-on maintenance of components outside the TF boundary.
Date: October 17, 2000
Creator: Brown, T.
Partner: UNT Libraries Government Documents Department

Effect of Boronization on Ohmic Plasmas in NSTX

Description: Boronization of the National Spherical Torus Experiment (NSTX) has enabled access to higher density, higher confinement plasmas. A glow discharge with 4 mTorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B+C areal density of 10 to the 18 (B+C) cm to the -2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by x15, carbon emission reduced by two and copper reduced to undetectable levels. After boronization, the plasma current flattop time increased by 70% enabling access to higher density, higher confinement plasmas.
Date: March 27, 2001
Creator: Skinner, C.H.; Kugel, H.; Maingi, R.; Wampler, W.R.; Blanchard, W.; Bell, M. et al.
Partner: UNT Libraries Government Documents Department

Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

Description: The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed.
Date: June 1, 2000
Partner: UNT Libraries Government Documents Department

Determination of wall reflectivity for ECE frequencies in DIII-D

Description: The significance of cyclotron radiation losses in next-generation tokamaks depends on the reflectivity of first wall materials. An experimental study of the effective reflectivity for electron cyclotron frequencies in the graphite-walled DIII-D tokamak is reported. Measurements of optically-thin harmonics ({omega} = n{omega}{sub ce}, n > 4) are made for two polarizations from thermal plasma discharges using an absolutely calibrated Michelson interferometer. The reflectivity r and polarization transfer fraction p are obtained by matching measured spectra to simulations from an ECE radiation transport code with adjustable wall parameters. For the frequency range 150-400 GHz average values of r = 0.76 and p = 0.19 are found.
Date: May 1, 1997
Creator: Austin, M.E.; Ellis, R.F. & Luce, T.C.
Partner: UNT Libraries Government Documents Department

Overview of impurity control and wall conditioning in NSTX

Description: The National Spherical Torus Experiment (NSTX) started plasma operations in February 1999, In the first extended period of experiments, NSTX achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. As expected, discharge reproducibility and performance were strongly affected by wall condition. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results.
Date: May 23, 2000
Creator: Kugel, H.W.; Maingi, R.; Wampler, W.; Berry, R.E. & al, et
Partner: UNT Libraries Government Documents Department

Liquid first walls for magnetic fusion energy

Description: Liquids ({approximately}7 neutron mean free paths thick) with certain restrictions can probably be used in magnetic fusion designs between the burning plasma and the structural materials of the plant. If this works there are a number of profound advantages: lower the cost of electricity by more than 35%; remove the need to develop first wall materials saving over 4B$ in development costs; reduce the amount and kind of wastes generated in the plant; and permit a wider choice of materials. Evaporated liquid must be efficiently ionized in an edge plasma to prevent penetrating into the burning plasma and diminishing the burn rate. The fraction of evaporated material ionized is estimated to be 0.993 for Li, 0.98 for Flibe and 0.9999 for Li{sub 17}Pb{sub 83}. This ionized vapor would be swept along open field lines into a remote burial chamber. The most practical systems would be those with topological open field lines on the outer surface as is the case of a field reversed configuration (FRC), a Spheromak, a Z-pinch, or a mirror machine. In a Tokamak, including the Spherical Tokamak, the field lines outside the separatrix are restricted to a small volume inside the toroidal coil making for difficulties in introducing the liquid and removing the ionized vapor.
Date: March 28, 1996
Creator: Moir, R.W.
Partner: UNT Libraries Government Documents Department

Fusion Ignition Research Experiment System Integration

Description: The FIRE (Fusion Ignition Research Experiment) configuration has been designed to meet the physics objectives and subsystem requirements in an arrangement that allows remote maintenance of in-vessel components and hands-on maintenance of components outside the TF (toroidal-field) boundary. The general arrangement consists of sixteen wedged-shaped TF coils that surround a free-standing central solenoid (CS), a double-wall vacuum vessel and internal plasma-facing components. A center tie rod is used to help support the vertical magnetic loads and a compression ring is used to maintain wedge pressure in the inboard corners of the TF coils. The magnets are liquid nitrogen cooled and the entire device is surrounded by a thermal enclosure. The double-wall vacuum vessel integrates cooling and shielding in a shape that maximizes shielding of ex-vessel components. The FIRE configuration development and integration process has evolved from an early stage of concept selection to a higher level of machine definition and component details. This paper describes the status of the configuration development and the integration of the major subsystem components.
Date: November 1, 1999
Creator: Brown, T.
Partner: UNT Libraries Government Documents Department

Divertor particle exhaust and wall inventory on DIII-D

Description: Many tokamaks achieve optimum plasma performance by achieving low recycling; various wall conditioning techniques including helium glow discharge cleaning (HeGDC) are routinely applied to help achieve low recycling. Many of these techniques allow strong, transient wall pumping, but they may not be effective for long-pulse tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), the Tokamak Physics Experiment (TPX), Tore Supra Continu, and JT-60SU. Continuous particle exhaust using an in-situ pumping scheme may be effective for wall inventory control in such devices. Recent particle balance experiments on the Tore Supra and DIII-D tokamaks demonstrated that the wall particle inventory could be reduced during a given discharge by use of continuous particle exhaust. In this paper we report the first results of wall inventory control and good performance with the in-situ DIII-D cryopump, replacing the HeGDC normally applied between discharges.
Date: June 1, 1995
Creator: Maingi, R.; Wade, M.R. & Mioduszewski, P.K.
Partner: UNT Libraries Government Documents Department

Leakage analysis of the evolve first wall.

Description: Leakage of lithium through cracks in the first wall of EVOLVE was analyzed for two limiting cases, which are simplified versions of the real case, where the lithium enters the cracks as liquid and flashes to vapor phase within the first wall. Leakage rates were calculated for the cases of liquid lithium flow and lithium vapor flow. Inasmuch as the coolant pressure is close to the saturation pressure, the limiting case of lithium vapor flow should be closer to reality. The impact of lithium leakage on first-wall cooling and plasma contamination is discussed.
Date: March 8, 2002
Creator: Majumdar, S.
Partner: UNT Libraries Government Documents Department

Focus Magnet and Vessel Interface Issues in HYLIFE-II

Description: The present Heavy Ion Driver design for HYLIFE-II calls for 96 beams from each side, or a total of 192 beams. The beams are separated from each other, at present, by an angle of 4.25 degrees. This report shows the focus magnet locations and a magnet build that leads to the minimum angle of 4.25 degrees between beams. Beam line and first wall shielding for the oscillating flow version of HYLIFE-II is accomplished by a series of horizontal and vertical jets. Ideally the horizontal jets would not deviate from a straight line but this is not feasible due to the force of gravity. Methods of altering the beam line array pattern to accommodate the curved ''horizontal'' jets are addressed.
Date: January 18, 2000
Creator: House, P.A.
Partner: UNT Libraries Government Documents Department

Analyses in Support of Z-IFE LLNL Progress Report for FY-05

Description: The FY04 LLNL study of Z-IFE [1] proposed and evaluated a design that deviated from SNL's previous baseline design. The FY04 study included analyses of shock mitigation, stress in the first wall, neutronics and systems studies. In FY05, the subject of this report, we build on our work and the theme of last year. Our emphasis continues to be on alternatives that hold promise of considerable improvements in design and economics compared to the base-line design. Our key results are summarized here.
Date: October 17, 2005
Creator: Moir, R W; Abbott, R P; Callahan, D A; Latkowski, J F; Meier, W R & Reyes, S
Partner: UNT Libraries Government Documents Department

Experimental Study of High-Z Gas Buffers in Gas-Filled ICF Engines

Description: ICF power plants, such as the LIFE scheme at LLNL, may employ a high-Z, target-chamber gas-fill to moderate the first-wall heat-pulse due to x-rays and energetic ions released during target detonation. To reduce the uncertainties of cooling and beam/target propagation through such gas-filled chambers, we present a pulsed plasma source producing 2-5 eV plasma comprised of high-Z gases. We use a 5-kJ, 100-ns theta discharge for high peak plasma-heating-power, an electrode-less discharge for minimizing impurities, and unobstructed axial access for diagnostics and beam (and/or target) propagation studies. We will report on the plasma source requirements, design process, and the system design.
Date: December 3, 2010
Creator: Rhodes, M A; Kane, J; Loosmore, G; DeMuth, J & Latkowski, J
Partner: UNT Libraries Government Documents Department

Integrated Chamber Design for the Laser Inertial Fusion Energy (LIFE) Engine

Description: The Laser Inertial Fusion Energy (LIFE) concept is being designed to operate as either a pure fusion or hybrid fusion-fission system. A key component of a LIFE engine is the fusion chamber subsystem. The present work details the chamber design for the pure fusion option. The fusion chamber consists of the first wall and blanket. This integrated system must absorb the fusion energy, produce fusion fuel to replace that burned in previous targets, and enable both target and laser beam transport to the ignition point. The chamber system also must mitigate target emissions, including ions, x-rays and neutrons and reset itself to enable operation at 10-15 Hz. Finally, the chamber must offer a high level of availability, which implies both a reasonable lifetime and the ability to rapidly replace damaged components. An integrated LIFE design that meets all of these requirements is described herein.
Date: December 7, 2010
Creator: Latkowski, J F; Kramer, K J; Abbott, R P; Morris, K R; DeMuth, J; Divol, L et al.
Partner: UNT Libraries Government Documents Department

An Assessment of the Penetrations in the First Wall Required for Plasma Measurments for Control of an Advanced Tokamak Plasma Demo

Description: A Demonstration tokamak (Demo) is an essential next step toward a magnetic-fusion based reactor. One based on advanced-tokamak (AT) plasmas is especially appealing because of its relative compactness. However, it will require many plasma measurements to provide the necessary signals to feed to ancillary systems to protect the device and control the plasma. This note addresses the question of how much intrusion into the blanket system will be required to allow the measurements needed to provide the information required for plasma control. All diagnostics will require, at least, the same shielding designs as planned for ITER, while having the capability to maintain their calibration through very long pulses. Much work is required to define better the measurement needs and the quantity and quality of the measurements that will have to be made, and how they can be integrated into the other tokamak structures.
Date: February 22, 2010
Creator: Young, Kenneth M.
Partner: UNT Libraries Government Documents Department

Magnetic and thermal energy flow during disruptions in DIII-D

Description: The authors present results from disruption experiments where they measure magnetic energy flow across a closed surface surrounding the plasma using a Poynting flux analysis to measure the electromagnetic power, bolometers to measure radiation power and IR scanners to measure radiation and particle heat conduction to the divertor. The initial and final stored energies within the volume are found using the full equilibrium reconstruction code EFIT. From this analysis they calculate an energy balance and find that they can account for all energy deposited on the first wall and the divertor to within about 10%.
Date: July 1, 1996
Creator: Hyatt, A.W.; Lee, R.L.; Humphreys, D.A.; Kellman, A.G.; Taylor, P.L.; Cuthbertson, J.W. et al.
Partner: UNT Libraries Government Documents Department

Neutral Beam Ion Loss Modeling for NSTX

Description: A numerical model, EIGOL, has been developed to calculate the loss rate of neutral beam ions from NSTX and the resultant power density on the plasma facing components. This model follows the full gyro-orbit of the beam ions, which can be a significant fraction of the minor radius. It also includes the three-dimensional structure of the plasma facing components inside NSTX. Beam ion losses from two plasma conditions have been compared: {beta} = 23%, q{sub 0} = 0.8, and {beta} = 40%, q{sub 0} = 2.6. Global losses are computed to be 4% and 19%, respectively, and the power density on the rf antenna is near the maximum tolerable levels in the latter case.
Date: June 1, 1999
Creator: Mikkelsen, D.; Darrow, D.S.; Grisham, L.; Akers, R. & Kaye, S.
Partner: UNT Libraries Government Documents Department

Characterization of wall conditions in DIII-D

Description: Wall conditioning in DIII-D is one of the most important factors in achieving reproducible high confinement discharges. For example, the very high confinement mode (VH-mode) was only discovered after boronization, a CVD technique to deposit a thin boron film over the entire surface of the tokamak. In order to evaluate wall conditions and provide a data base to correlate these wall conditions with tokamak discharge performance, a series of nominally identical reference VH-mode discharges (1.6 MA, 2.1 T, double-null diverted) were taken at various times during a series of experimental campaigns with evolving wall conditions. These reference discharges have allowed a quantitative determination of how the wall conditions have evolved. For instance, core carbon and oxygen levels in the VH-mode phase remains at historically low levels during the 1995 run year and there was also a steady decrease in the oxygen levels at plasma initiation during this period. The authors discuss the long term changes in low Z impurities and the effect of wall conditioning techniques such as boronization and baking on these impurities. In addition, the evolution of the deuterium recycling rates will be discussed.
Date: October 1, 1996
Creator: Holtrop, K.L.; Jackson, G.L.; Kellman, A.G.; Lee, R.L.; West, W.P.; Wood, R.D. et al.
Partner: UNT Libraries Government Documents Department