Showing papers in "Fusion Science and Technology in 2005"

The EIRENE and B2-EIRENE Codes

Abstract: The EIRENE neutral gas transport Monte Carlo code has been developed initially for TEXTOR since the early 1980s. It is currently applied worldwide in most fusion laboratories for a large variety of...

MHD Equilibrium Reconstruction in the DIII-D Tokamak

Abstract: Physics elements and advances crucial for the development of axisymmetric magnetohydrodynamic equilibrium reconstruction to support plasma operation and data analysis in the DIII-D tokamak are revi...

The National Ignition Facility: Laser Performance and First Experiments

Abstract: The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory is a stadium-sized facility containing a 192-beam, 1.8-Megajoule, 500-Terawatt, ultraviolet laser system together with a 10-meter diameter target chamber with room for nearly 100 experimental diagnostics. NIF will be the world's largest and most energetic laser experimental system, providing a scientific center to study inertial confinement fusion (ICF) and matter at extreme energy densities and pressures. NIF's energetic laser beams will compress fusion targets to conditions required for thermonuclear burn, liberating more energy than required to initiate the fusion reactions. Other NIF experiments will study physical processes at temperatures approaching 10{sup 8} K and 10{sup 11} bar, conditions that exist naturally only in the interior of stars, planets and in nuclear weapons. NIF has successfully activated, commissioned, and utilized the first four beams of the laser system to conduct over 300 shots between November 2002 and August 2004. NIF laser scientists have established that the laser meets nearly all performance requirements on a per beam basis for energy, uniformity, timing, and pulse shape. Using these four beams, ICF and high-energy-density physics researchers have conducted a number of experimental campaigns resulting in high quality data that could not be reachedmore » on any other laser system. We discuss the successful NIF Early Light Program including details of laser performance, examples of experiments performed to date, and recent advances in the ICF Program that enhance prospects for successful achievement of fusion ignition on NIF.« less

U.S. Plans and Strategy for ITER Blanket Testing

Abstract: Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation in the ITER Test Blanket Module (TBM) Program. A US strategy for ITER-TBM has evolved that emphasizes international collaboration. A study was initiated to select the two blanket options for the US ITER-TBM in light of new R and D results from the US and world programs over the past decade. The study is led by the Plasma Chamber community in partnership with the Materials, PFC, Safety, and physics communities. The study focuses on assessment of the critical feasibility issues for candidate blanket concepts and it is strongly coupled to R and D of modeling and experiments. Examples of issues are MHD insulators, SiC insert viability and compatibility with PbLi, tritium permeation, MHD effects on heat transfer, solid breeder 'temperature window' and thermomechanics, and chemistry control of molten salts. A dual coolant liquid breeder and a helium-cooled solid breeder blanketmore » concept have been selected for the US ITER-TBM.« less

Design Features of the Tokamak TEXTOR

Abstract: TEXTOR is the Tokamak Experiment for Technology Oriented Research in the field of plasma-wall interaction The scope includes a detailed analysis of particle and energy exchange between the plasma and the surrounding chamber as well as active measures to optimize the first wall and the plasma boundary region TEXTOR is a medium-sized tokamak belonging to the class of moderate-field but large-volume devices having a circular cross section of the plasma and an iron core The plasma major radius is 175 m, and the minor radius is 047 m The maximum plasma current is 08 MA, the maximum field is 3 T, and the maximum pulse length is 10 s TEXTOR is fed directly from the 110-kV grid using an installed converter power of {approx}300 MVA The inner wall of TEXTOR is equipped with several specially shaped limiters being partly remotely movable Special design features of TEXTOR are excellent access for diagnostics to domains near the wall, large portholes suitable for implementing methods to control the plasma boundary, facilities to heat the vacuum vessel and the liner, and provisions for exchange of the liner TEXTOR has been upgraded with auxiliary heating systems (neutral beam injection, radio-frequency heating, and microwave heating ofmore » 9 MW in total), a toroidal pumped limiter, an upgraded magnetization coil, and recently the dynamic ergodic divertor (DED) The DED is a novel flexible tool to influence transport parameters at the plasma edge and to study the resulting effects on heat exhaust, edge cooling, impurity screening, plasma confinement, and stability The number of special features and the flexibility of TEXTOR provide excellent opportunities for important contributions to fusion research« less

Axial electron heat loss from mirror devices revisited

Abstract: An issue of the axial electron heat loss is of a significant importance for mirror-based fusion devices. This problem has been considered in a number of publications but it is still shrouded in misconceptions. In this paper we revisit it once again. We discuss the following issues: 1) Formation of the electron distribution function in the end tank at large expansion ratios; 2) The secondary emission from the end plates and the ways of suppressing it (if needed); 3) Ionization and charge exchange in the presence of neutrals in the end tanks; 4) Instabilities caused by the peculiar shape of the electron distribution function and their possible impact on the electron heat losses; 5) Electron heat losses in the pulsed mode of operation of mirror devices.

Development Path for Z-Pinch IFE

Abstract: The long-range goal of the Z-Pinch IFE program is to produce an economically-attractive power plant using high-yield z-pinch-driven targets (~3GJ) with low rep-rate per chamber (~0.1 Hz). The prese...

Effective thermal conductivity of lithium ceramic pebble beds for fusion blankets: a review

Abstract: The use of lithium ceramic pebble beds has been considered in many blanket designs for the fusion reactors. Lithium ceramics have received a significant interest as tritium breeders for the fusion blankets during the last three decades. The thermal performance of the lithium ceramic pebble beds plays a key role for the fusion blankets. In order to study the heat transfer in the blanket, the effective thermal conductivity of the lithium ceramics pebble beds has to be well measured and characterized. The data of effective thermal conductivity of lithium ceramic pebble beds is important for the blanket design. Several studies have been dedicated to investigate the effective conductivity of the lithium ceramics pebble beds. The objective of this work is to review and compare the available data, presented by various studies, of effective conductivity of lithium ceramic pebble beds in order to address the current status of these data.

Limiter Lock Systems at TEXTOR: Flexible Tools for Plasma-Wall Investigation

Abstract: Limiter lock systems on the top and the bottom of the TEXTOR vessel are essential elements for experimental investigations of plasma-wall interaction in a tokamak. The lock systems are designed as user facilities that allow the insertion of wall elements (limiter) and tools for diagnostic (electrical probes, gas injection) without breaking the TEXTOR vacuum. The specially designed holder on top of the central carrier and a powerful vacuum pump system permit the exchange of components within {approx}1 h. Up to ten electrical signals, four thermocouples, and a gas supply can be connected at the holder interface. Between discharges, the inserted component can be positioned radially and turned with respect to the toroidal magnetic field. Additionally, the central carrier is electrically isolated to apply bias voltages and currents up to 1 kV and 1 kA, respectively.An important feature of the lock system is the good access for optical spectroscopic observation of the inserted components in the vicinity of the edge plasma. The whole spectrum from ultraviolet to infrared is covered by spectrometers and filters combined with cameras. Toroidally and poloidally resolved measurements are obtained from the view on top of the probes while the tangential poloidal view delivers radially and toroidallymore » resolved information.A programmable logic controller (Simatic S5) that is operated inside the TEXTOR bunker and from remote locations outside the concrete wall drives all possible features of the lock system.« less

Status of Tritium Permeation Barrier Development in the EU

Abstract: Tritium permeation can be significantly reduced by a suitable barrier on the structural materials of a future fusion power plant Since alumina has the capability of tritium permeation reduction, t

Thermofluid Magnetohydrodynamic Issues for Liquid Breeders

Abstract: This paper provides a description of the most promising liquid breeder blankets currently proposed for testing in ITER. The critical MHD issues for selfcooled and dual coolant LM systems are the MH...

Development of steady-state advanced tokamak research in the DIII-D tokamak

Abstract: Research into the feasibility of steady-state operation of high-fusion-gain tokamak plasmas is one of the central elements of the DIII-D program. Realization of such discharges has progressed to the point of demonstrating well-aligned noninductive current profiles for a resistive time at 90% of the total current with plasma pressure and confinement consistent with fusion gain >5 in an ITER-sized tokamak. Full current drive discharges with poorer alignment have been obtained for shorter duration. The design methodology and the path to integrating the various elements necessary for full noninductive operation on DIII-D are discussed in detail.

Observation of hydrogen distribution around non-metallic inclusions in steels with tritium microautoradiography

Abstract: Hydrogen distributions around non-metallic inclusions in steels are successfully characterized with high-resolution tritium autoradiography. The autoradiographs show that hydrogen accumulation characteristics around the inclusions depend on types of the inclusions. In the case of MnS, hydrogen was inhomogeneously distributed in the ferrite matrix surrounding the MnS inclusion, probably because hydrogen is trapped in defects formed around MnS. The inhomogeneous distribution of hydrogen may be originated from the asymmetric stress field produced by a contraction of the MnS phase in the heat treatment, i.e. the inhomogeneous volumetric change of MnS owing to its larger thermal expansion than that of the ferrite phase. In the case of Al 2 O 3 , hydrogen was intensely localized at boundary layers of the ferrite matrix surrounding the Al 2 O 3 inclusion. This could be attributed to hydrogen trapping at defects introduced by a residual stress in the boundary layers of the ferrite matrix due to larger contraction of the ferrite phase than that of the Al 2 O 3 phase on cooling. Similarly hydrogen was accumulated in the surrounding ferrite matrix but more widely distributed around Cr carbide probably because difference in the thermal expansion between the Cr carbide and ferrite phases is less than that between the Al 2 O 3 and ferrite phases.

Measurement of Neutron Emission Spectra in Li(d,xn) Reaction with Thick and Thin Targets for 40-MeV Deuterons

Abstract: To improve the data accuracy of the neutron emission spectra of the natLi(d,xn) reaction that will be used as the neutron source in the International Fusion Materials Irradiation Facility, the auth...

Successful Experimental Verification of the Tokamak Exhaust Processing Concept of ITER with the CAPER Facility

Abstract: One of the design targets for the Tokamak Exhaust Processing (TEP) system of ITER is not to lose more than 10 -5 gh -1 into the Normal Vent Detritiation system of the Tritium Plant The plasma exhaust gas therefore needs to be processed in a way that a tritium removal efficiency of about 10 8 with respect to the flow rate is achieved Expressed in terms of tritium concentrations this corresponds to a decontamination from about 130 gm -3 down to about 10 -4 gm -3 (about 1 Cim -3 = 37*10 10 Bqm -3 ) The three step reference process for the TEP system of ITER is called CAPER and has been developed and realized at the Tritium Laboratory Karlsruhe (TLK) After the successful commissioning of the PERMCAT reactor as the key component of the third step detailed parametric tritium testing of the 3 steps involving the processing of more than 300 g tritium has been carried out and decontamination factors beyond the design requirements have been demonstrated for each process step and for the process as a whole Not only the decontamination factor of 10 8 as required by ITER, but also the operational mode of TEP as a waste dump for gases from diverse sources has been experimentally validated with the CAPER facility

TEXTOR: A Pioneering Device for New Concepts in Plasma-Wall Interaction, Exhaust, and Confinement

Abstract: As an introduction to this special issue, this paper gives a brief overview of the key activities and outstanding results obtained on the TEXTOR tokamak.

Progress on the Multimirror Trap GOL-3

Abstract: Main results of researches on plasma heating and confinement of dense plasma in the multimirror trap GOL-3 are presented.Recently magnetic system of the facility was converted into completely multimirror one. This results in further improvement of energy confinement time of plasma with ion temperature {approx}1 keV. Collective plasma heating by {approx}120 kJ ({approx}8 {omega}s) relativistic electron beam results in T{sub e} {approx} 2 keV at {approx}10{sup 21} m{sup -3} density. High T{sub e} exists for {approx}10 {mu}s. To this time T{sub i} reaches {approx}2 keV. Ion temperature keeps at the high level during {approx}1 ms. The energy confinement time sufficiently increases and a value of n{tau}{sub E} = (1.5-3).10{sup 18} m{sup -3}s.

Overview of University of Wisconsin Inertial-Electrostatic Confinement Fusion Research

Abstract: In Inertial Electrostatic Confinement (IEC) devices, a voltage difference between concentric, nearly transparent spherical grids accelerates ions to fusion-relevant velocities. The University of Wisconsin (UW) operates two IEC devices: a cylindrical aluminum chamber and a spherical, water-cooled, stainless-steel chamber, with a power supply capable of 75 mA and 200 kV. The research program aims to generate fusion reaction products for various applications, including protons for creating radioisotopes for nuclear medicine and neutrons for detecting clandestine materials. Most IEC devices worldwide, including the UW devices, presently operate primarily in a pressure range (1-10 mtorr) that allows ions to make only a few passes through the core before they charge exchange and lose substantial energy or they collide with cathode grid wires. It is believed that fusion rates can be raised by operating at a pressure where neutral gas does not impede ion flow, and a helicon ion source has been developed t...

Axisymmetric tandem mirrors: Stabilization and confinement studies

Abstract: The 'Kinetic Stabilizer' has been proposed as a means of MHD stabilizing an axisymmetric tandem mirror system. The K-S concept is based on theoretical studies by Ryutov, confirmed experimentally in the Gas Dynamic Trap experiment in Novosibirsk. In the K-S beams of ions are directed into the end of an 'expander' region outside the outer mirror of a tandem mirror. These ions, slowed, stagnated, and reflected as they move up the magnetic gradient, produce a low-density stabilizing plasma. At the Lawrence Livermore National Laboratory we have been conducting theoretical and computational studies of the K-S Tandem Mirror. These studies have employed a low-beta code written especially to analyze the beam injection/stabilization process, and a new code SYMTRAN (by Hua and Fowler) that solves the coupled radial and axial particle and energy transport in a K-S TM. Also, a 'legacy' MHD stability code, FLORA, has been upgraded and employed to benchmark the injection/stabilization code and to extend its results to high beta values. The FLORA code studies so far have confirmed the effectiveness of the K-S in stabilizing high-beta (40%) plasmas with stabilizer plasmas the peak pressures of which are several orders of magnitude smaller than those of the confined plasma.more » Also the SYMTRAN code has shown D-T plasma ignition from alpha particle energy deposition in T-M regimes with strong end plugging. Our studies have confirmed the viability of the K-S-T-M concept with respect to MHD stability and radial and axial confinement. We are continuing these studies in order to optimize the parameters and to examine means for the stabilization of possible residual instability modes, such as drift modes and 'trapped-particle' modes. These modes may in principle be controlled by tailoring the stabilizer plasma distribution and/or the radial potential distribution. In the paper the results to date of our studies are summarized and projected to scope out possible fusion-power versions of the K-S T-M« less

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

Abstract: As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant lead-eutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R&D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

The Electron Cyclotron Resonant Heating System on the DIII-D Tokamak

Abstract: In the DIII-D electron heating and current drive installation, up to six gyrotron microwave generators in the 1-MW class at pulse lengths up to 5 s have been operated simultaneously. The frequency for all the gyrotrons is 110 GHz, corresponding to the second harmonic of the electron gyrofrequency at 2 T. The peak generated power has been >4 MW with peak injected power slightly greater than 3 MW. The radio frequency (rf) generators are located remotely and are connected to the tokamak by up to 100 m of evacuated circular corrugated waveguide carrying the HE 1,1 mode with overall transmission efficiency, including coupling to the waveguide, of up to 75%. Ancillary equipment for polarization control, beam switching, power monitoring, control of launch direction, and system protection has been developed. The system has been used to support a wide variety of physics experiments, including control of magnetohydrodynamic modes, current density profile modifications, basic plasma heating and current drive, transport studies, and rf-assisted start-up. The gyrotron complex is being upgraded by the acquisition of additional tubes with 5- to 10-s pulse length capability.

Multi-Component Adsorption Behavior of Hydrogen Isotopes on Zeolite 5A and 13X at 77.4 K

Abstract: Equilibrium isotherms for the adsorption of H 2 , HD, HT, D 2 , DT and T 2 on synthetic zeolite type 5A or 13X at 77.4 K are estimated by using a theoretical formula, where the isotopic difference in adsorption depends on the zero-point energy difference between hydrogen isotopes. The formula agrees with the experimental isotherms for H 2 and D 2 on the zeolites. Adsorption of H 2 -D 2 and H 2 -HD-D 2 mixtures on the same adsorbents is experimentally examined. The experiments are performed using a volumetric apparatus and a quadra-pole-type mass spectrograph. The experimental adsorption behavior of H 2 , D 2 and HD shows agreement of separation factors with results calculated according to the ideal adsorbed solution theory describing multi-component behavior, where the equilibrium isotherms estimated for H 2 , HD and D 2 are used. Based on the theoretical adsorption model, the multi-component behavior of HT, DT and T 2 is predicted here.

Optimizing Neutron Production Rates from D-D Fusion in an Inertial Electrostatic Confinement Device

Abstract: Detection of explosives has been identified as a near term commercial opportunity for using a fusion plasma. Typical explosive compositions contain low Z material (C, N, O) which are not easily det...

Electron cyclotron resonance heating on TEXTOR

Abstract: TEXTOR is equipped with two gyrotrons at 110 and 140 GHz, respectively. Both share a single power supply and a confocal quasi-optical transmission line. They cannot be operated simultaneously. The 110-GHz gyrotron with limited power and pulse length (300 kW; 200 ms) has been used in a first series of experiments on electron cyclotron resonance heating (ECRH) and electron cyclotron current drive (ECCD) and for collective Thomson scattering (CTS) diagnostics of energetic ions. In the future the 110-GHz gyrotron will be operated exclusively for CTS diagnostics, while for ECRH and ECCD, the newly installed 140-GHz, high-power (800-kW), long-pulse (>3-s) gyrotron is now available. The highlights of first ECRH experiments with the 110-GHz gyrotron are reported. These include observations of internal transport barriers with ECRH on various target plasmas: in the current plateau phase of both ohmic and radiation improved mode (RI-mode) discharges. In addition, sawtooth control by localized ECRH is demonstrated. First results on CTS include the observation of the slowing down of energetic ions and of the redistribution of energetic ions in sawtooth crashes.

The Dynamics of Tritium - Including OBT - in the Aquatic Food Chain

Abstract: Tritiated water spills by nuclear installations result in uptake in aquatic organisms. The radionuclide uptake model BURN (developed by NRG, modified), considers not only tritium as tritiated water (HTO) but also the conversion into organically bound tritium (OBT). Comparison with the original BURN mode showed that the modified model gave more realistic results in terms of concentration levels, and consequently for dose assessment as result of ingestion of fishery products. For more accurate modelling, seasonal effects and half-life estimates asa function of body weight and water temperature must be taken into account. A first attempt is given, although limited empirical data gives reason to further investigation of this significant effect.

The Closed Tritium Cycle of the Tritium Laboratory Karlsruhe

Abstract: The Tritium Laboratory Karlsruhe (TLK) was commissioned with tritium in 1994 and since then has continuously improved its infrastructure and has expanded its experimental activities. With a fully closed tritium cycle and the license to handle 40 g of tritium TLK is an almost unique experimental facility. More than 10 glove box systems with a total volume of about 125 m 3 are operated to house experiments and infrastructure facilities on an area of more than 1000 m 2 . Today TLK has about 23 g of tritium on site. The paper describes the closed tritium loop of the TLK infrastructure and its links to different experiments. Some experience gained during operation of TLK is also presented.

Overview of the ALPS Program

Abstract: The US Advanced Limiter-divertor Plasma-facing Systems (ALPS) program is developing the science of liquid metal surface divertors for near and long term tokamaks. These systems may help solve the d...

Theory and Simulation Basis for Magnetohydrodynamic Stability in DIII-D

Abstract: Theory and simulation have provided one of the critical foundations for many of the significant achievements in magnetohydrodynamic (MHD) stability in DIII-D over the past two decades. Early signature achievements included the validation of tokamak MHD stability limits, beta and performance optimization through cross-section shaping and profiles, and the development of new operational regimes. More recent accomplishments encompass the realization and sustainment of wall stabilization using plasma rotation and active feedback, a new understanding of edge stability and its relation to edge-localized modes, and recent successes in predicting resistive tearing and interchange instabilities. The key to success has been the synergistic tie between the theory effort and the experiment made possible by the detailed equilibrium reconstruction data available in DIII-D and the corresponding attention to the measured details in the modeling. This interaction fosters an emphasis on the important phenomena and leads to testable theoretical predictions. Also important is the application of a range of analytic and simulation techniques, coupled with a program of numerical tool development. The result is a comprehensive integrated approach to fusion science and improving the tokamak approach to burning plasmas.

Overview of the DIII-D Fusion Science Program

Abstract: This overview of the DIII-D fusion program provides an introduction to the research program carried out on the DIII-D tokamak since its inception in 1986. It serves as the introduction and summary ...

Innovative Liquid Breeder Blanket Design Activities in Japan

Abstract: In order to clarify key engineering issues and to enhance key R&D activities for D-T fusion blankets, many design activities on innovative liquid blanket systems are on going as collaboration studi...