By Topic

Fusion Engineering 2005, Twenty-First IEEE/NPS Symposium on

Date Sept. 2005

Filter Results

Displaying Results 1 - 25 of 152
  • Korean ITER Participation and Activities in ITER Transitional Arrangements

    Publication Year: 2005 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (7715 KB) |  | HTML iconHTML  

    Korean ITER participation impacts huge things on her fusion research as an energy development program. The ITER Transitional Arrangements (ITA) is conducted by the International Team with supports from participant teams. The Korean participant team (KO-PT) has contributed to the ITA by sharing several technical tasks to verify feasibilities of fabrication and quality control method in procuring ITER equipments and facilities. Technical data to confirm validity of the present design have also been developed on some issues. Completing these technical preparations will make it possible to finalize the specification of ITER procurements. This presentation provides an overview focusing on critical activities needed for the successes of ITER project and fusion energy development View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Current Status of Experimental Study and Device Modifications in JT-60U

    Publication Year: 2005 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (8555 KB) |  | HTML iconHTML  

    Since tokamak magnetic fusion research has just made a step forward to an international collaborative project ITER, the existing tokamaks including JT-60 are expected to explore more advanced operation scenarios. To test those scenarios in the JT-60 experiment, the discharge pulse length and the duration time of additional NBI/RF heating were extended to 65 s and 30 s/60 s, respectively, in 2003 with modification of the corresponding control systems for power supplies and heating devices. The experimental campaign in 2003-2004 after the above modifications has ended up with the following significant results: (a) The high bootstrap current ratio of 75 % was sustained for 7.4 s in an R/S plasma, (b) Normalized beta value of 2.3 was also done for 22.3 s in a high-beta H-mode plasma, (c) The quasi-steady state beta value was increased to 3.0 with a pulse of 6.2 s with NTM suppression by ECCD. For further exploration toward high performance plasmas, the following modifications will be or has been conducted: (1) To minimize the power loss from a plasma at the region of toroidal field ripple, the 8Cr ferritic steel tiles, having a similar magnetic property to the low activation ferritic material for a DEMO reactor, are being equipped on the first wall of the JT-60 vacuum vessel. (2) Since plasma shape and current profile in the poloidal cross-section are expected to be reproduced in real time to optimize a plasma performance with suppressing the plasma instabilities, a precise reproduction method has been installed in the plasma control system. In this report, the current status of plasma experimental study will be presented together with on-going device modifications in JT-60U View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • The National Ignition Facility: The World's Largest Laser

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4376 KB) |  | HTML iconHTML  

    The National Ignition Facility (NIF) is a 192-beam laser facility presently under construction at LLNL. When completed, NIF will be a 1.8-MJ, 500-TW ultraviolet laser system. Its missions are to obtain fusion ignition and to perform high energy density experiments in support of the U.S. nuclear weapons stockpile. Four of the NIF beams have been commissioned to demonstrate laser performance including target and beam alignment. During this time, NIF demonstrated on a single-beam basis that it will meet its performance goals and demonstrated its precision and flexibility for pulse shaping, pointing, timing and beam conditioning. It also performed four important experiments for inertial confinement fusion and high energy density science. Presently, the project is installing production hardware to complete the project in 2009 with the goal to begin ignition experiments in 2010. An integrated plan has been developed including the NIF operations, user equipment such as diagnostics and cryogenic target capability, and experiments and calculations to meet this goal View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • JET Enhancement Programmes: key steps in the preparation of ITER

    Publication Year: 2005 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (8571 KB) |  | HTML iconHTML  

    Since 2000, the JET Facilities have been exploited under the European Fusion Development Agreement (EFDA) and have played a key role in the preparation of ITER, in particular by operating in deuterium/tritium plasmas, optimising remote handling and consolidating ITER operating scenarios with increasing heating power. JET has undergone significant enhancement during this period. A first enhancement programme is now approaching fruition as the present shutdown nears an end. The experimental campaigns are due to resume in November 2005 with an improved divertor, increased heating power and around 20 new diagnostics. A second enhancement programme, to be completed by the end of 2008, is already underway, and will enable major experiments in support of ITER. This includes the installation of a new first wall with an ITER-like materials combination, a further upgrade of the neutral beams, an upgraded plasma control system, a high frequency pellet injector and a number of diagnostics key for ITER technology and physics View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Wendelstein 7-X, Overview and Status of Construction

    Publication Year: 2005 , Page(s): 1 - 6
    Cited by:  Papers (4)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (8178 KB) |  | HTML iconHTML  

    The line of the Wendelstein stellarators developed in IPP is being continued with a superconducting device, Wendelstein 7-X. This fully optimised stellarator which shall demonstrate the reactor potential of the Helias-type stellarator, is presently under construction in the Greifswald branch institute of IPP. Manufacturing of the W7-X components has progressed well over the last years, and assembly of the device has started early in 2005 View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Status of the KSTAR Tokamak Construction

    Publication Year: 2005 , Page(s): 1 - 6
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (7123 KB) |  | HTML iconHTML  

    The KSTAR is a superconducting tokamak under construction at the Korea Basic Science Institute (KBSI) in Daejeon, Korea. The project mission aims at a steady-state operation and advanced tokamak physics. At present, the project is in the peak of fabrication and assembly phase. The fabrication of the major tokamak structures such as vacuum vessel, cryostat, port system, thermal shields, and gravity support, is completed. The manufacture and testing of the 30 superconducting magnets are rigorously being progressed. As of Sep. 2005, 16 toroidal field coils and 4 large poloidal field coils are completed. To verify the operational feasibility of the KSTAR coils, cool-down and current charging tests of a real sized prototype TF coil and a pair of CS model coil have been carried out. The assembly of the device has begun from beginning of 2004. Now, the vacuum vessel body, thermal shields and 8 toroidal field magnets are assembled on the tokamak position. Assembly finish is scheduled for August 2007. This paper describes the manufacture and assembly progress of the KSTAR tokamak View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • NCSX Construction Progress and Research Plans

    Publication Year: 2005 , Page(s): 1 - 6
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (6509 KB) |  | HTML iconHTML  

    Stellarators use 3D plasma and magnetic field shaping to produce a steady-state disruption-free magnetic confinement configuration. Compact stellarators have additional attractive properties-quasi-symmetric magnetic fields and low aspect ratio. The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL) to test the physics of a high-beta compact stellarator with a low-ripple, tokamak-like magnetic configuration. The engineering challenges of NCSX stem from its complex geometry requirements. These issues are addressed in the construction project through manufacturing R&D and system engineering. As a result, the fabrication of the coil winding forms and vacuum vessel are proceeding in industry without significant technical issues, and preparations for winding the coils at PPPL are in place. Design integration, analysis, and dimensional control are functions provided by system engineering to ensure that the finished product will satisfy the physics requirements, especially accurate realization of the specified coil geometries. After completion of construction in 2009, a research program to test the expected physics benefits will start View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • DIII-D: Recent Physics Results, Implemented and Planned Hardware Upgrades

    Publication Year: 2005 , Page(s): 1 - 7
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (8819 KB) |  | HTML iconHTML  

    During the last two years, the DIII-D tokamak has been operated for a total of 34 run weeks, during which significant advances have been achieved in many areas of plasma physics. This progress was only possible because of the improvements in the tools available to the DIII-D program to control and manipulate the plasma core and edge conditions. The important systems in this effort include the electron cyclotron (EC) system, the fast wave system (restarted after being sidelined for four years), 12 new internal coils, an upgraded plasma control system and a comprehensive set of turbulence diagnostics. The EC system's versatility was demonstrated by the various roles it played in the physics research program. It was used as a probe to demonstrate the "hybrid" plasmas are regulated by m/n=3/2 tearing modes, it was used to suppress the m/n=2/1 Neoclassical Tearing Mode, which allowed the plasma pressure to be raised to new heights, and in an active feedback mode EC power was used to control the q-profile using real-time equilibrium reconstructions based upon motional Stark effect data. The fast wave system was used in conjunction with the EC system for current profile experiments. The internal control coil system was used to investigate suppression of the resistive wall modes and reduction and/or elimination of ELMs. During this year, the DIII-D facility will implement major changes and upgrades to expand the frontiers in several of the areas of tokamak plasma physics research. One of the four neutral beams will be rotated from the co-to the counter-injection mode so that heated plasmas with little or no rotation can be studied. The present lower divertor will be removed and a new extended shelf divertor will be installed to provide the capability of pumping high triangular double null plasmas. Three new long pulse one MW class gyrotron systems will be brought on line, which will double the long pulse capability of the EC system. Two of the three aging cooling towers wi- ll be replaced with two new high efficiency towers that can handle the higher heat loads expected in the future from 10 s pulse operation. These and other improvements to the facility will be discussed and presented View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Recent Accomplishments and Future Plans of ASDEX Upgrade

    Publication Year: 2005 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (8375 KB) |  | HTML iconHTML  

    ASDEX Upgrade has demonstrated clear progress in ITER relevant plasma operation scenarios. Successful operation with about 70% Tungsten clad first wall is shown. The operation space of the improved H-mode (advanced tokamak scenario) is enlarged in safety factor 3lesq95 les5 and density 0.4lesne/nGWles0.9. The performance can clearly exceed the ITER reference value. At high densities type-II ELMs mitigate the peak power load in the divertor. A similar effect is achieved by increasing the type-I ELM frequency by shallow pellet injection or cyclic vertical plasma motion. Off-axis neutral beam current drive leads at low powers to a clear current density profile modification. Sawtooth tailoring by electron cyclotron current drive (ECCD) can prevent the growth of neoclassical tearing modes (NTM). (3, 2) and (2, 1) NTMs excited at high plasma pressure can be stabilized by ECCD at moderate power levels. Narrow ECCD absorption profiles are more effective than broad ones View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Recent Results in Large Helical Device

    Publication Year: 2005 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (6505 KB) |  | HTML iconHTML  

    This paper presents recent experimental results in the Large Helical Device (LHD), which is the world largest helical system (Rax~3.6 m, aav~0.6 m, Bax~2.75 T) and has capability of steady state operation because of the superconducting magnets. LHD confines the plasma with Tio~2 keV and ne ~0.8times1019 m-3 for 1905 sec using the ICRF heating. In this case, the total plasma heating energy was ~1.3 GJ. This indicates that high energy ions generated by ICRF is well confined. Large magnetic island in the vacuum field due to the error field is vanished in plasma with higher temperature or with higher beta, so the error field is not an issue in helical systems. The obtained averaged beta is more than 4.3 % in LHD. Since the beta is not limited by the MHD events or magnetic island but the beta is limited by the lack of heating power, further improvement of beta can be expected. Internal and external transport barrier have been also observed, and the electron temperature and the ion temperature exceed 10 keV in the improved confinement mode in LHD. Besides the potential of the disruption free steady stated operation, LHD experiment accumulates evidences so that fusion reactor becomes more realistic View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Alcator C-Mod Status, Recent Accomplishments, and Future Plans

    Publication Year: 2005 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (6319 KB) |  | HTML iconHTML  

    A number of major hardware items have been installed and/or modified on the Alcator C-Mod tokamak for the most recent run campaign, and additional near-term upgrades will be installed shortly. Topics that are covered in this paper include the changeover to all-metal plasma-facing components, the titanium lower hybrid waveguide grill, the gas jet disruption mitigation system, and the results these made to plasma operation. The design of a new cryopump, which will be installed soon, is also presented View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Engineering Design Status of the Quasi-Poloidal Stellarator (QPS)

    Publication Year: 2005 , Page(s): 1 - 5
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4400 KB) |  | HTML iconHTML  

    The engineering design status of the quasi-poloidal stellarator experiment (QPS) is presented. The overall configuration and the design, manufacturing R&D and assembly techniques of the core components are described View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Progress and Plans of the Proto-Sphera Program

    Publication Year: 2005 , Page(s): 1 - 5
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (3454 KB) |  | HTML iconHTML  

    PROTO-SPHERA is a proposed spherical torus (ST) where a hydrogen plasma arc, in a form of a Screw Pinch (SP), fed by electrodes, replaces the central rod of the standard ST experiments. This machine, with a longitudinal pinch current Ie=60 kA, will produce an ST of diameter 2Rsph=75 cm, aspect ratio A=1.2-1.3, carrying a toroidal current Ip=120-240 kA. The plasma will be magnetically shaped as a disk near each modular annular electrode. Theoretical calculations (using a 3D ideal MHD stability code) show that such a configuration should be ideally stable up to a total beta ranging between 15-25%, depending upon the ratio Ip/Ie. The ST toroidal current should be sustained by helicity injection from the Screw Pinch. In order to demonstrate the feasibility of the SP discharge and the basic electrode characteristics, a preliminary linear electrode test bed, Proto-Pinch, has been built and operated successfully. The detail mechanical engineering design of PROTO-SPHERA has now been completed. The machine design philosophy, basic geometry and operating conditions with the major components like the vacuum vessel (VV), coils, electrodes, protection components and divertor will be analysed. The thermal and electromagnetic behaviour as well as the predicted and permitted heat stresses will be discussed in order to demonstrate that the design, construction and reliable operation of the machine are feasible. Reference is also made to the proposed Multi-Pinch experiment using the START VV to demonstrate the feasibility and stability of the PROTO-SPHERA configuration. The START VV has now been transported and disassembled in ENEA-Frascati. The detailed design of Multi-Pinch, initial phase of PROTO-SPHERA, has also been completed and the first contract related to the PF coils has been placed View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Project Management in Large Collaborations: SNS Lessons Learned for ITER

    Publication Year: 2005 , Page(s): 1 - 5
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (7086 KB) |  | HTML iconHTML  

    Collaborations of research institutions and industry have been increasingly employed to accomplish the design, procurement, fabrication, assembly, installation, testing and commissioning of complex science facilities to support enhanced research capabilities in many areas. The large cost and significant breadth of technical knowledge, skills and abilities needed to bring into successful operation such complex facilities makes it likely that collaborations among National Institutions and Nations will become the norm for future facilities projects of this nature. The spallation neutron source (SNS), a $1.4 billion accelerator-based facility for neutron science nearing completion at Oak Ridge National Laboratory (ORNL) in Tennessee, is a major collaboration among six US national laboratories and an industrial construction partner whose objective has been to design, construct and operate the world's most powerful neutron source to support world-class materials research. Some of the more important factors that have contributed to the success of the SNS collaboration include the development of an effective project management organization across institutional boundaries, a project focus on integration, involvement of partners to oversee procurements closest to the work, and top-level risk management to include a centrally-controlled reserve for unforeseen events. The lessons learned in planning, executing and managing this successful, multi-partner collaborative project have significance for the International Thermonuclear Experimental Reactor (ITER) project. ITER is a planned partnership among six National organizations (China, the European Union, Japan, the Russian Federation, South Korea and the United States) that is coming together to design, construct and operate a full-scale technology demonstration facility for producing power from fusion energy. The ITER project presents unique challenges for project management with its mix of "in-kind" and "in-cash" deliverables- , the risks associated with the division of scope among the participants and the government-to-government agreements required. SNS lessons learned that can benefit ITER include the early assignment of experienced project leadership, a project-directed risk assessment and technical integration review, development of a realistic integrated project schedule, the creation of a reserve fund under project control to mitigate unforeseen risks, and the development and acceptance of a means for periodic, thorough review of overall project performance. Instituting successful project management within the ITER collaboration will require aligning the project management philosophies, accounting for cultural influences, understanding the participants' political environments, selecting and implementing useful management systems, and successfully incorporating the project management strengths and experience of the ITER partners View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Application, Design and Project Implementation of a Plasma Mass Separator for Enhanced High Level Waste Processing

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (5199 KB) |  | HTML iconHTML  

    The Archimedes Technology Group has developed a partially ionized, low-temperature plasma mass separator that is ideally suited for enhancing high-level waste (HLW) processing prior to incorporation into borosilicate glass for immobilization. The plasma mass separator, referred to as the Archimedes Filter, passes light mass ions to biased electrodes at either end of the plasma chamber, but ejects heavy mass ions to the center of the chamber. A "mass cutoff" is established by adjusting the solenoidal magnetic field surrounding the plasma and biasing the concentric ring electrodes at both ends of the plasma. Material is introduced into the plasma as a vapor and ionized by RF helicon wave heating. Under steady state operation the Archimedes Filter is capable of processing 1.1 metric tons of feed material per day. A full-scale Archimedes Filter was designed and constructed in 2002 and achieved first plasma in February 2003 at the Archimedes Development Center in San Diego. This Demonstration Filter (DEMO) has been used to demonstrate heavy and light metal mass separation in a variety of plasma conditions to support a proposed implementation plan for Hanford HLW processing. The Archimedes Filter is ideally suited for Hanford HLW because 99.9% of the radioactivity is in less than 10% of the mass that is above atomic mass 89(90Sr). Thus, 90% of the HLW is non-radioactive and is an unnecessary load on the very expensive vitrification process for immobilization prior to disposal in Yucca Mountain. Theoretical predictions for heavy mass elements indicated decontamination factors (DF) of >1000 would be possible and DEMO tests have confirmed DF>100, limited only by the diagnostic detection lower limits for heavy mass elements. A Hanford implementation plan was developed in parallel with DEMO by forming an international team of industry and national laboratory participants, including Oak Ridge National Laboratory. The team completed a conceptual design of a tw- o-filter unit Archimedes Filter Plant (AFP) by the end of 2003. This conceptual design enabled a detailed cost estimate and a preliminary Process Hazards Analysis to establish the feasibility and safety of the plant. The AFP is highly flexible in how it could be interfaced to the Hanford Waste Treatment Plant (WTP) currently under construction. A site plan was also developed to establish infrastructure needs and to ensure no significant interferences with WTP View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Fabrication Capabilities for Spherical Foam Targets Used in ICF Experiments

    Publication Year: 2005 , Page(s): 1 - 5
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (5953 KB) |  | HTML iconHTML  

    This paper reviews the processes developed at General Atomics (GA) in the past several years to fabricate a variety of spherical foam targets at various densities for the inertial confinement fusion (ICF) community. The two most common chemical systems used to produce spherical foam targets have been resorcinol-formaldehyde (R/F) aerogel and divinylbenzene (DVB). Spherical targets have been made in the form of shells and beads with diameters ranging from approximately 0.5 to 4.0 mm, and densities from 100 mg/cc to 250 mg/cc, with typical high yield of intact shells or beads of 90%-95%. Permeation barriers have been developed and deposited on both R/F and DVB shells. We have also made R/F foam shells with higher pore size (0.10-0.50 mum) in order to increase the cryo-fill fraction when these shells are cryogenically layered with D2. Another spherical target that is currently under development that will also be discussed is silica aerogels shells and beads. Other foam target materials currently under development, such metal doped R/F aerogel beads for extreme ultra violet (EUV) source experiments will also be discussed View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Shock Mitigation using Compressible Two-Phase Jets for Z-Pinch IFE Reactor Applications

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (5078 KB) |  | HTML iconHTML  

    Compressible liquid/gas jets offer the opportunity to limit and mitigate the mechanical consequences of rapid heating/evaporation of the protective jets in a high-yield, low repetition rate Inertial Fusion Energy (IFE) system such as the Z-Pinch IFE reactor. In this investigation, experiments have been conducted to quantify the extent by which a two-phase jet can attenuate a shock wave. The experiments have been conducted using annular two-phase (water/air) jets with different velocities, void fractions, and initial shock strength. The shock is produced using an exploding wire located along the jet axis. Three different confinement geometries (i.e. boundary conditions) have been used in the experiments, namely, "unconfined" shocks, "radially-confined" shocks, and "radially-and-axially-confined" shocks. A total of 738 experiments corresponding to 39 different test conditions have been conducted. Quantitative data for the transient pressure history at the confinement wall with and without an intervening jet have been obtained; both single-phase (liquid) and two-phase (liquid-gas) jets at different velocities and void fractions have been tested. The data shows that the experiments are highly repeatable, and that two-phase jets at moderate void fractions (~1%) can attenuate a shock wave significantly more than a single-phase (liquid) jet View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • The Dynamic Response of Thick-Liquid Shielding in Z-IFE Reactors

    Publication Year: 2005 , Page(s): 1 - 3
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2378 KB) |  | HTML iconHTML  

    A major concern in the design of thick-liquid protected inertial fusion reactors of all types is the dynamic response of the shielding liquid to the pulsed explosions. Induced liquid motion can stress and damage solid chamber structures such as the first-wall. In a Z-pinch based inertial fusion (Z-IFE) reactor this issue becomes particularly critical due to the relatively large proposed target yields of several GJ. In this paper we summarize an analysis of the liquid response taking into account ablation of target facing surfaces, pocket venting, and neutron isochoric heating. The impact of varying several reactor parameters is also discussed View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Ablation and Its Product of the Fast Ignition Inertial Confinement Fusion Reactor Chamber

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (3394 KB) |  | HTML iconHTML  

    One of the major problems in the technological feasibility of the inertial confinement fusion reactor is the chamber that should accept pulsed load of radiation, ion particles and debris and be pumped out in the repeated pulsed operation cycle of several Hz. Especially in a fast ignition scenario, lead (or lithium lead) cone attached to the target fuel pellets adds specific material transfer issues such as deposited on the chamber wall, ablation, and formation of clusters that is suspected to affect the pumping characteristics. Ablated metal particles from the wall are suspected to form various sizes of clusters that fly slower and more difficult to evacuate. Therefore, it is necessary to investigate behavior of the chamber wall under simulated laser fusion condition by the experiments as well as numerical studies. In this study, we have made the preliminary experiments to investigate formation of cluster of plastic and metals. The YAG laser (2J, 10ns) was irradiated on the targets, and the ablated particles are measured using the Thomson parabola, the charge particle collector, and the quadrupole mass spectrometer. Results show formations cluster of ethylene (-CH2 -)n and metals, and ablation of metals by bombardments by ethylene debris View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Shock Mitigation Studies of Solid Foams for Z-Pinch Chamber Protection

    Publication Year: 2005 , Page(s): 1 - 4
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4606 KB) |  | HTML iconHTML  

    Solid open-cell Al foams are subjected to dynamic compression testing in a vertical shock tube to model the metallic foam being considered for use in an inertial fusion energy reactor. High porosity (0.89) foam samples (stack of two, 25.4 cm square, 10.2 cm high) of three different cell sizes (10, 20, and 40 pores per inch) are compressed with a strong shock (M=6) in a 25 kPa atmosphere of air and SF6. The post-shock samples are highly compressed (strains up to 0.8 for the smallest cell size) and have a wavy upper surface indicating structural anisotropy. Energy absorption is found to vary with cell-size (smaller cells, more absorption) while the impulse of the shock wave is independent of cell size. Pressure data indicate the incident shock wave becomes a compression wave with a non-discontinuous gradient inside of the foam. Using an array of pressure transducers with a vertical spacing of 2.54 cm, the wave speed in the foam sample is reduced to 25% of its value without the presence of the foam View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Foamed Liquidand Cellular Metal Shock Attenuation Analysis for Z-IFE

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4778 KB) |  | HTML iconHTML  

    Foamed liquid and cellular materials are being investigated as potential materials for shock mitigation and energy absorption in the IFE Z-pinch power plant concept. These materials are a critical part of the plant design because they absorb the energy that will be converted to electricity, generate tritium, and protect the containment structure. Foamed materials are being considered for shock mitigation because of their low densities. However, for inhomogeneous materials, such as a liquid with gas bubbles or cellular materials, a low density mixture can be obtained using various configurations, morphologies, or cell sizes. For example, cellular metals with different cell sizes, or a different number of pores per inch (ppi), can be fabricated with the same porosity and density. A series of experiments and ALEGRA simulations were conducted to investigate the effect of material configuration on shock mitigation View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Z-Pinch (LiF)2-BeF2 (flibe) Preliminary Vaporization Estimation Using the BUCKY 1-D Radiation Hydrodynamics Code

    Publication Year: 2005 , Page(s): 1 - 4
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (3486 KB) |  | HTML iconHTML  

    The post-explosion material vaporization characteristics of the proposed Z-pinch reactor design were simulated using the BUCKY 1-D radiation hydrodynamics code. To model the 3-D variations in the reactor chamber design, three separate BUCKY simulations were performed - one in each of the cylindrical coordinate geometries: +z, -z, and r. The simulations were run to a time of 80 mus and the chamber material characteristics were analyzed. These results were compared to a simple analytical model to verify the vaporization radii in each of the three modeled directions. The +z material vaporization has been estimated to be at a radius 53.71 cm, compared to an analytic result of 79.00 cm. The -z material vaporization has been estimated to be at a radius of 101.92 cm, compared to an analytic result of 102.78 cm. The r material vaporization has been estimated to be at a radius of 73.86 cm, compared to an analytic result of 77.63 cm. These simulation results confirm the idea that we can model the exploding Z-Pinch target and its resulting thermal effects on the reactor chamber using the BUCKY 1-D radiation hydrodynamics code. This model is appropriate for analysis of the Z-Pinch reactor because it is a massive structure and because most of the energy coupling to the surrounding structure is via X-rays (30%) rather than expanding ionic debris (4%). Furthermore, we have confirmed the viability of performing three different 1-D simulations in each of the +z, -z, and r directions and merging the three results. Such an approximation to a 3-D phenomenon is valid for times where the outward blast and energy transfer remain nearly spherical View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Design and Numerical Stress Analysis of Silicon Membrane Hibachi Windows

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4327 KB) |  | HTML iconHTML  

    A silicon membrane windows is developed for a KrF laser system. The main function of the windows is to pass electron beam into the laser cell and to hold pressurized gas in the cell. 150 micro-meter thick silicon windows successfully survived from heated electron beam bombardment and shows 80% of electron beam transmission rate. The single silicon windows endured 250,000 cycles of electron beam shots. The arrayed windows did not show sufficient performance due to excessive heat generation. To enhance the longevity, the cooling system has to be improved and a study of the arc generated by the cathode is necessary View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • A New High Current Fast 100ns LTD Based Driver for Z-pinch IFE at Sandia

    Publication Year: 2005 , Page(s): 1 - 4
    Cited by:  Papers (6)  |  Patents (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4377 KB) |  | HTML iconHTML  

    Sandia is actively pursuing the development of new accelerators based on the novice technology of linear transformer driver (LTD). LTD based drivers are currently considered for many applications including future very high current Z-pinch drivers like ZX and z-pinch IFE (inertial fusion energy). LTD is a new method for constructing high-current, high-voltage pulsed accelerators. The salient feature of the approach is switching and inductively adding the pulses at low voltage straight out of the capacitors through low inductance transfer and soft iron core isolation. High currents can be achieved by feeding each cavity core with many capacitors connected in parallel in a circular array. High voltage is obtained by inductively adding the output voltage of many cavities in series. Utilizing the presently available capacitors and switches we can envision building the next generation of fast Z-pinch drivers without the usage of large deionized-water and oil tanks, as it is the case with the present technology drivers. The most significant advantage of all is that the LTD drivers can be rep-rated. They can be multipulsed with a repetition rate, in principle, up to the capacitor specifications and up to 10 Hz. The later makes LTD the driver of choice for z-pinch IFE where the required repetition rate is of the order of 0.1 Hz. Presently we have in rep rated operation in Sandia a one 500-kA, 100-kV LTD cavity. Our goal is to establish the maximum possible frequency of repetition rate and test the longevity of the utilized dry air gas switches. The compact fast (<100 ns pulse rise time) LTD technology was suggested and its development is funded and being monitored by Sandia at the High Current Electronic Institute (HCEI) in Tomsk, Russia, where an additional number of larger and stackable 1-MA cavities are under construction to be utilized as building blocks for a 1-MA, 1-MV voltage adder test module. This module will serve as a prototype for longer higher voltage module- s, a number of which, connected in parallel, could become the driver of the z-pinch IFE reactor. In this paper we briefly describe the basic theory underlying the LTD operation, present the device and give performance results of our 500-kA and 1-MA cavities currently in operation and finally describe a first cut design of an IFE driver utilizing the 1-MA, 100-kV cavity as a building block View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.
  • Study of Ion Stripping and Charge Exchange for Heavy Ion Fusion

    Publication Year: 2005 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (4715 KB) |  | HTML iconHTML  

    To accurately model transport in the heavy ion fusion chamber, it is necessary to know the evolution of beam charge state due to ionization or recombination. At present there is no accurate or complete information for electron loss and capture from these fast, low-charge heavy ions, and the theoretical model in the simulation codes, such as LSP is not accurate enough to describe the processes, especially multi-electron ionizations. Thus, it is very important to explore better theoretical approaches to calculate cross sections of multi-electron losses especially ion stripping, charge exchange for near-term and driver scale HIF research. Based on the investigation of different approaches on the heavy ion-atom collisions, with emphasis on the features related with low-charged heavy ions, especially multi-electron processes and correlation effects, an improved classic trajectory Monte Carlo simulation model will be set up to solve the n-body ion-atom ionization problem in HIF's interest. The calculation results will be benchmarked with the data obtained through finished experiments from the literature as well. Finally, the improved code will be implemented into LSP code as a subroutine to become an integral part of the code. In current stage, a subroutine with a database of typical ionization stripping cross sections and change exchange for low-charge-state, heavy ions mainly in the experimental energy level will be developed for LSP code to serve ongoing experiments and simulations in current heavy ion beam fusion project, which is shown in this paper. More data including for the energy in drive scale can be integrated into the subroutine in the future expecting to be used in HIF and high energy density physics as well View full abstract»

    Full text access may be available. Click article title to sign in or learn about subscription options.