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The Fusion Energy Program: The Role of TPX and Alternate Concepts

Description: This report focuses on the following two questions for the U.S. fusion energy program. First, what is the role of the Tokamak Physics Experiment (TPX), an approximately $700 million fusion reactor currently awaiting a congressional decision to begin construction? Second, what is the role of alternatives to the tokamak concept in a broad-based fusion energy program?
Date: unknown
Creator: United States. Congress. Office of Technology Assessment.
Partner: UNT Libraries Government Documents Department

Progress in Developing the K-DEMO Device Configuration

Description: K-DEMO is being studied by South Korean researchers as a follow-on to ITER and the next step toward the construction of a commercial fusion power plant. The K-DEMO mission defines a staged approach targeting operation with an initial testing phase for plasma facing components and critical operating systems to be followed by a second phase which centers on upgrading the in-vessel components for operation at 200 to 600 MWe with a planned 70% availability.
Date: June 27, 2013
Creator: Brown, Tom
Partner: UNT Libraries Government Documents Department

Tritium Processing and Containment Technology for Fusion Reactors, Annual Report: July 1975-June 1976

Description: The hydrogen permeabilities of selected metals, alloys, and multiplex preparations that are of interest to fusion reactor technology are being characterized. A high-vacuum hydrogen-permeation apparatus has been constructed for this purpose. A program of studies has been initiated to develop design details for the tritium-handling systems of near-term fusion reactors. This program has resulted in a better definition of reactor-fuel-cycle and enrichment requirements and has helped to identify major research and development problems in the tritium-handling area. The design and construction of a 50-gallon lithium-processing test loop (LPTL) is well under way. Studies in support of this project are providing important guidance in the selection of hardware for the LPTL and in the design of a molten-salt processing test section.
Date: 1976?
Creator: Maroni, V. A.; Calaway, W. F.; Misra, B.; Van Deventer, E. H.; Wegton, J. R.; Yonco, R. M. et al.
Partner: UNT Libraries Government Documents Department

Comparison of Options for a Pilot Plant Fusion Nuclear Mission

Description: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.
Date: August 27, 2012
Creator: Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S et al.
Partner: UNT Libraries Government Documents Department

Prospects for inertial fusion energy based on a diode-pumped solid-state laser (DPSSL) driver: Overview and development path

Description: It is now known with certainty that the type of fusion known as inertial fusion will work with sufficient energy input, so inertial fusion is really beyond the ``scientific breakeven`` point in many respects. The most important question that remains for inertial fusion energy (IFE) is whether this type of fusion can operate with sufficiently low input energy to make it economically feasible for energy production. The constraint for low input energy demands operation near the inertial fusion ignition threshold, and such operation creates enormous challenges to discover a target design that will produce sufficient energy gain. There are also multiple issues relating to the scientific feasibility of using a laboratory-type ``driver`` to energize a target, such as those concerning bandwidth and beam smoothing for ``direct drive,`` and extension of hohlraum plasma physics to the IFE scale for ``indirect drive.`` One driver that appears as though it will be able to meet the IFE requirements, assuming modest development and sufficient target gain, is a diode-pumped solid-state laser (DPSSL). We give an overview of this type of laser system, and explain what development remains for the economic production of electricity using this type of driver for IFE.
Date: March 1, 1997
Creator: Orth, C.D.
Partner: UNT Libraries Government Documents Department

Tritium-Management Requirements for D-T Fusion Reactors (ETF, INTOR, FED)

Description: The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design.
Date: October 1981
Creator: Finn, P. A.; Clemmer, Robert G. & Misra, B.
Partner: UNT Libraries Government Documents Department

Starpower: the U.S. and the international quest for fusion energy

Description: This report reviews the status of magnetic confinement fusion research and compares its progress with the requirements for development of a useful energy technology. The report does not analyze inertial confinement fusion research, which is overseen by the House and Senate Armed Services Committees.
Date: October 1987
Creator: United States. Congress. Office of Technology Assessment.
Partner: UNT Libraries Government Documents Department

Microtearing Instability In The ITER Pedestal

Description: Unstable microtearing modes are discovered by the GS2 gyrokinetic siimulation code, in the pedestal region of a simulated ITER H-mode plasma with approximately 400 WM DT fusion power. Existing nonlinear theory indicates that these instabilities should produce stochastic magnetic fields and broaden the pedestal. The resulted electron thermal conductivity is estimated and the implications of these findings are discussed.
Date: December 1, 2010
Creator: Wong, K. L.; Mikkelsen, D. R.; Rewoldt, G. M. & Budny, R.
Partner: UNT Libraries Government Documents Department

Novel Use of Water Soluble "Aquapour" As A Temporary Spacer During Coil Winding For The NSTX-U Centerstack

Description: A major facility upgrade to the National Spherical Torus eXperiment (NSTX-U) is currently underway at Princeton Plasma Physics Laboratory (PPPL). A key component of NSTX-U is the fabrication of a new, higher field centerstack (CS). In order to simultaneously provide robust joints between the inner and outer legs of the Toroidal Field Coils (TF) and minimize radial build, the NSTX-U CS design requires that the Ohmic Heating solenoid (OH) be wound directly on the inner TF bundle. To protect the OH against thermal expansion stress during scenarios where the inner TF bundle is hot but the OH is relatively cool, the completed CS will have a 0.100 inch annular gap between the outer diameter of the TF bundle and the inner diameter of the OH solenoid. "Aquapour", a proprietary material produced by the Advanced Ceramics Manufacturing Company will be used during manufacture to produce this gap. After the TF bundle is vacuum pressure impregnated and cured, a cylindrical "clam shell" mold will be assembled around it, and a slurry of powdered Aquapour and water will be pumped into the annular space between the mold and TF bundle. Subsequent baking will turn the Aquapour solid, and a protective layer of wet lay-up fiberglass and resin will be added. The OH solenoid will be wound directly on this wet lay-up shell. After vacuum pressure impregnation of the OH, the water soluble Aquapour will be washed away, leaving the required radial clearance between the TF and OH. This paper will describe prototyping and testing of this process, and plans for use on the actual CS fabrication.
Date: July 1, 2013
Creator: Mardenfeld, Michael
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

Disruptions, Disruptivity, and Safer Operating Windows in the High-β Spherical Torus NSTX

Description: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. This paper reviews the AT pilot plant design, detailing the selected maintenance approach, the device arrangement and sizing of the in-vessel components. Details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will also be discussed.
Date: September 26, 2012
Creator: Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S et al.
Partner: UNT Libraries Government Documents Department

Quasi-spherical direct drive fusion.

Description: The authors present designs of quasi-spherical direction drive z-pinch loads for machines such as ZR at 28 MA load current with a 150 ns implosion time (QSDDI). A double shell system for ZR has produced a 2D simulated yield of 12 MJ, but the drive for this system on ZR has essentially no margin. A double shell system for a 56 MA driver at 150 ns implosion has produced a simulated yield of 130 MJ with considerable margin in attaining the necessary temperature and density-radius product for ignition. They also represent designs for a magnetically insulated current amplifier, (MICA), that modify the attainable ZR load current to 36 MA with a 28 ns rise time. The faster pulse provided by a MICA makes it possible to drive quasi-spherical single shell implosions (QSDD2). They present results from 1D LASNEX and 2D MACH2 simulations of promising low-adiabat cryogenic QSDD2 capsules and 1D LASNEX results of high-adiabat cryogenic QSDD2 capsules.
Date: January 1, 2007
Creator: VanDevender, J. Pace; Abbott, Lucas M.; Langston, William L.; McDaniel, Dillon Heirman; Nash, Thomas J.; Roderick, Norman Frederick et al.
Partner: UNT Libraries Government Documents Department

Quasi-spherical direct drive fusion simulations for the Z machine and future accelerators.

Description: We explored the potential of Quasi-Spherical Direct Drive (QSDD) to reduce the cost and risk of a future fusion driver for Inertial Confinement Fusion (ICF) and to produce megajoule thermonuclear yield on the renovated Z Machine with a pulse shortening Magnetically Insulated Current Amplifier (MICA). Analytic relationships for constant implosion velocity and constant pusher stability have been derived and show that the required current scales as the implosion time. Therefore, a MICA is necessary to drive QSDD capsules with hot-spot ignition on Z. We have optimized the LASNEX parameters for QSDD with realistic walls and mitigated many of the risks. Although the mix-degraded 1D yield is computed to be {approx}30 MJ on Z, unmitigated wall expansion under the > 100 gigabar pressure just before burn prevents ignition in the 2D simulations. A squeezer system of adjacent implosions may mitigate the wall expansion and permit the plasma to burn.
Date: November 1, 2007
Creator: VanDevender, J. Pace; McDaniel, Dillon Heirman; Roderick, Norman Frederick & Nash, Thomas J.
Partner: UNT Libraries Government Documents Department

Preliminary Physics Motivation and Engineering Design Assessment of the National High Power Torus

Description: In April 2006, Dr. Ray Orbach, Director of the DOE Office of Science, challenged the fusion community to "propose a new facility... which will put the U.S. at the lead in world fusion science." Analysis of the gaps between expected ITER performance and the requirements of a demonstration power plant (Demo) pointed to the critical and urgent need to develop fusion-relvant plasma-material interface (PMI) solutions consistent with sustained high plasma performance. A survey of world fusion program indicated that present and planned experimental devices do not advance the PMI issue beyond ITER, and a major dedicated experimental facility is warranted. Such a facility should provide the flexibility and access needed to solve plasma boundary challenges related to divertor heat flux and particle exhaust while also developing methods to minimize hydrogenic isotope retention and remaining compatible with high plasma performance.
Date: June 11, 2009
Creator: Woolley, Robert D.
Partner: UNT Libraries Government Documents Department

Proliferation Risks of Fusion Energy: Clandestine Production, Covert Production, and Breakout

Description: Nuclear proliferation risks from fusion associated with access to weapon-usable material can be divided into three main categories: 1) clandestine production of fissile material in an undeclared facility, 2) covert production of such material in a declared and safeguarded facility, and 3) use of a declared facility in a breakout scenario, in which a state begins production of fissile material without concealing the effort. In this paper we address each of these categories of risk from fusion. For each case, we find that the proliferation risk from fusion systems can be much lower than the equivalent risk from fission systems, if commercial fusion systems are designed to accommodate appropriate safeguards.
Date: August 13, 2009
Creator: R.J. Goldston, A. Glaser, A.F. Ross
Partner: UNT Libraries Government Documents Department

A Midsize Tokamak As Fast Track To Burning Plasmas

Description: This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.
Date: July 14, 2010
Creator: Mazzucato, E.
Partner: UNT Libraries Government Documents Department

Evaluation of containment failure and cleanup time for Pu shots on the Z machine.

Description: Between November 30 and December 11, 2009 an evaluation was performed of the probability of containment failure and the time for cleanup of contamination of the Z machine given failure, for plutonium (Pu) experiments on the Z machine at Sandia National Laboratories (SNL). Due to the unique nature of the problem, there is little quantitative information available for the likelihood of failure of containment components or for the time to cleanup. Information for the evaluation was obtained from Subject Matter Experts (SMEs) at the Z machine facility. The SMEs provided the State of Knowledge (SOK) for the evaluation. There is significant epistemic- or state of knowledge- uncertainty associated with the events that comprise both failure of containment and cleanup. To capture epistemic uncertainty and to allow the SMEs to reason at the fidelity of the SOK, we used the belief/plausibility measure of uncertainty for this evaluation. We quantified two variables: the probability that the Pu containment system fails given a shot on the Z machine, and the time to cleanup Pu contamination in the Z machine given failure of containment. We identified dominant contributors for both the time to cleanup and the probability of containment failure. These results will be used by SNL management to decide the course of action for conducting the Pu experiments on the Z machine.
Date: February 1, 2010
Creator: Darby, John L.
Partner: UNT Libraries Government Documents Department

A Compact Quasi-axisymmetric Stellarator Reactor

Description: We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils.
Date: October 20, 2003
Creator: Ku, L. P.
Partner: UNT Libraries Government Documents Department

Radiation drive in laser heated hohlraums

Description: Nearly 10 years of Nova experiments and analysis have lead to a relatively detailed quantitative and qualitative understanding of radiation drive in laser heated hohlraums. Our most successful quantitative modelling tool is 2D Lasnex numerical simulations. Analysis of the simulations provides us with insight into the details of the hohlraum drive. In particular we find hohlraum radiation conversion efficiency becomes quite high with longer pulses as the accumulated, high Z blow-off plasma begins to radiate. Extensive Nova experiments corroborate our quantitative and qualitative understanding.
Date: November 3, 1995
Creator: Suter, L.J.; Kauffman, R.L. & Darrow, C.B.
Partner: UNT Libraries Government Documents Department

A systems assessment of the five Starlite tokamak power plants

Description: The ARIES team has assessed the power-plant attractiveness of the following five tokamak physics regimes: (1) steady state, first stability regime; (2) pulsed, first stability regime; (3) steady state, second stability regime; (4) steady state, reversed shear; and (5) steady state, low aspect ratio. Cost-based systems analysis of these five tokamak physics regimes suggests that an electric power plant based upon a reversed-shear tokamak is significantly more economical than one based on any of the other four physics regimes. Details of this comparative systems analysis are described herein.
Date: July 1, 1996
Creator: Bathke, C.G.
Partner: UNT Libraries Government Documents Department