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Sample records for compact tokamak reactors

  1. The ignition physics study group supports the compact ignition tokamak and engineering test reactor programs

    SciTech Connect

    Sheffield, J.

    1987-01-01

    This report presents a collection of Vugraphs dealing with the Compact Ignition Tokamak (CIT) and the Engineering Test Reactor (ETR). The role of the Ignition Physics Study Group is defined. Several design goals are presented. (JDH)

  2. The conceptual design of a robust, compact, modular tokamak reactor based on high-field superconductors

    NASA Astrophysics Data System (ADS)

    Whyte, D. G.; Bonoli, P.; Barnard, H.; Haakonsen, C.; Hartwig, Z.; Kasten, C.; Palmer, T.; Sung, C.; Sutherland, D.; Bromberg, L.; Mangiarotti, F.; Goh, J.; Sorbom, B.; Sierchio, J.; Ball, J.; Greenwald, M.; Olynyk, G.; Minervini, J.

    2012-10-01

    Two of the greatest challenges to tokamak reactors are 1) large single-unit cost of each reactor's construction and 2) their susceptibility to disruptions from operation at or above operational limits. We present an attractive tokamak reactor design that substantially lessens these issues by exploiting recent advancements in superconductor (SC) tapes allowing peak field on SC coil > 20 Tesla. A R˜3.3 m, B˜9.2 T, ˜ 500 MW fusion power tokamak provides high fusion gain while avoiding all disruptive operating boundaries (no-wall beta, kink, and density limits). Robust steady-state core scenarios are obtained by exploiting the synergy of high field, compact size and ideal efficiency current drive using high-field side launch of Lower Hybrid waves. The design features a completely modular replacement of internal solid components enabled by the demountability of the coils/tapes and the use of an immersion liquid blanket. This modularity opens up the possibility of using the device as a nuclear component test facility.

  3. Transport Comparisons between ITER-FEAT and Compact High-Field Tokamak Reactors

    NASA Astrophysics Data System (ADS)

    Zhu, Ping; Horton, Wendell; Bateman, Glenn; Kritz, Arnold H.; Porcelli, Franco

    2000-10-01

    The next generation magnetic fusion reactors under active consideration are ITER-FEAT and compact high-field (CHF) tokamaks. These represent two rather different concepts for achieving a burning plasma. We use the BALDUR predictive transport code with the MMM99 transport model and the alternative OHE transport model(Aaron J. Redd, et al.,) Phys. Plasmas 5, (1998) 1369. with ITG and new ETG physics. We simulate the proposed reactor designs and compare their fusion performance parameters. ITER-FEAT is a next step large NBI driven reactor with the goal of achieving Q=10 and thus represents a straightforward scaling of our recent study of the transport in JET and DIIID.(P. Zhu, et al.,) Phys. Plasmas 7 (2000) 2898. Comparisons will be made with CHF reactor designs as well as the original ITER design in order to clarify issues identified in the comparison of these machines types. The off-diagonal turbulent particle pinch will be studied as a key issue in the CHF reactor design simulations.

  4. Activation analysis of the compact ignition tokamak

    SciTech Connect

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak.

  5. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  6. ARIES tokamak reactor study

    SciTech Connect

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein.

  7. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Hoffman, H. A.; Logan, B. G.; Campbell, R. B.

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boiling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection temperatures, and only a relatively small natural-draft heat exchanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

  8. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  9. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  10. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    NASA Astrophysics Data System (ADS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Garcia, J.; Arimoto, H.; Shoji, T.

    2009-05-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  11. Compact torsatron reactors

    SciTech Connect

    Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

    1988-05-01

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R/sub 0/ = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R/sub 0/ approx. = 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs.

  12. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  13. D-D tokamak reactor studies

    SciTech Connect

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.

  14. Compact reactor design automation

    NASA Technical Reports Server (NTRS)

    Nassersharif, Bahram; Gaeta, Michael J.

    1991-01-01

    A conceptual compact reactor design automation experiment was performed using the real-time expert system G2. The purpose of this experiment was to investigate the utility of an expert system in design; in particular, reactor design. The experiment consisted of the automation and integration of two design phases: reactor neutronic design and fuel pin design. The utility of this approach is shown using simple examples of formulating rules to ensure design parameter consistency between the two design phases. The ability of G2 to communicate with external programs even across networks provides the system with the capability of supplementing the knowledge processing features with conventional canned programs with possible applications for realistic iterative design tools.

  15. Physics evaluation of compact tokamak ignition experiments

    SciTech Connect

    Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

    1985-01-01

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/S/q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs.

  16. Cooldown of the Compact Ignition Tokamak

    SciTech Connect

    Keeton, D.C.

    1987-08-01

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.

  17. Plasma-gun fueling for tokamak reactors

    SciTech Connect

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment.

  18. Recent progress on the Compact Ignition Tokamak (CIT)

    SciTech Connect

    Ignat, D.W.

    1987-01-01

    This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule.

  19. Compact Ignition Tokamak Program: status of FEDC studies

    SciTech Connect

    Flanagan, C.A.

    1985-01-01

    Viewgraphs on the Compact Ignition Tokamak Program comprise the report. The technical areas discussed are the mechanical configuration status, magnet analysis, stress analysis, cooling between burns, TF coil joint, and facility/device layout options. (WRF)

  20. Plasma diagnostics for the compact ignition tokamak

    SciTech Connect

    Medley, S.S.; Young, K.M.

    1988-06-01

    The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10/sup 21/m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs.

  1. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  2. Development of tokamak reactor system analysis code NEW-TORSAC

    NASA Astrophysics Data System (ADS)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  3. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  4. Maintenance concept development for the Compact Ignition Tokamak

    SciTech Connect

    Macdonald, D.

    1988-01-01

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs.

  5. Equilibrium calculations for plasma control in CIT (Compact Ignition Tokamak)

    SciTech Connect

    Strickler, D.J.; Peng, Y-K.M. . Fusion Engineering Design Center); Pomphrey, N.; Jardin, S.C. . Plasma Physics Lab.)

    1990-01-01

    The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this report, VEQ and its role in the TSC analysis of the CIT PF system are described. Equilibrium and coil current calculations are discussed, an overview of the CIT PF system is presented, a set of reference equilibria for modeling a complete discharge in CIT is described, and the concept of a plasma shape control matrix is introduced. 9 refs., 8 figs., 7 tabs.

  6. (Injection of compact toroids for tokamak fueling and current drive)

    SciTech Connect

    Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.

    1991-01-01

    The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.

  7. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  8. Liquid nitrogen cooling considerations of the Compact Ignition Tokamak

    SciTech Connect

    Dabiri, A.E.

    1986-10-01

    A simple model was developed to estimate the cooldown time between pulses of toroidal field (TF) coils of the Compact Ignition Tokamak (CIT) using liquid nitrogen. Good agreement was obtained between the analysis results and those measured in the early fusion experimental devices. A cooldown time of about 1 h would reduce the TF coil temperature to about 80 K. An R and D experimental program is required to determine the actual cooldown time between pulses, an issue in the conceptual design of the CIT.

  9. Natural fueling of a tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Wan, Weigang; Parker, Scott E.; Chen, Yang; Perkins, Francis W.

    2010-04-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is presented. In H-mode plasmas dominated by ion-temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward toward the core. This mechanism is due to the quasineutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection is augmented by an inward pinch of cold DT fuel. The natural fueling mechanism is investigated using the gyrokinetic turbulence code GEM [Y. Chen and S. E. Parker, J. Comput. Phys. 220, 839 (2007)] and is analyzed using quasilinear theory. Profiles similar to those used for conservative International Thermonuclear Experimental Reactor [R. Aymar et al., Nucl. Fusion 41, 1301 (2001)] transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rate and energy transport. Natural fueling requires a two-component plasma and ion-ion and charge-exchange collisions set limits on this favorable effect.

  10. The insulation irradiation test program for the Compact Ignition Tokamak

    SciTech Connect

    McManamy, T.J. ); Kanemoto, G. ); Snook, P. )

    1990-01-01

    The electrical insulation for the toroidal field coils of the Compact Ignition Tokamak (CIT) is expected to be exposed to radiation doses on the order of 10{sup 10} rad with {approx}90% of the dose from neutrons. The coils are cooled to liquid nitrogen temperature and then heated during the pulse to a peak temperature >300 K. In a program to evaluate the effects of radiation exposure on the insulators, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1, and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to approximately 5 {times} 10{sup 9} rad and 3 {times} 10{sup 10} rad with 35 to 40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was done by cycling the shear load for up to 30,000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths on the order of 120 MPa were measured. The behavior of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed nearly identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. 9 refs., 5 figs.

  11. Insulation irradiation test programme for the Compact Ignition Tokamak

    NASA Astrophysics Data System (ADS)

    McManamy, T. J.; Kanemoto, G.; Snook, P.

    The electrical insulation for the toroidal field coils of the Compact Ignition Tokamak (CIT) is expected to be exposed to radiation doses of the order of 10 10 rad with ≈ 90% of the dose from neutrons. The coils are cooled to liquid nitrogen temperature and then heated during the pulse to a peak temperature ; 300 K. In a programme to evaluate the effects of radiation exposure on the insulators, three types of boron-free insulation were irradiated at room temeprature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1 and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to ≈ 5 × 10 9 and 3 × 10 10 rad with 35-40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was performed by cycling the shear load for up to 30 000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths of the order of 120 MPa were measured. The behaviour of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed almost identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. No swelling was measured; however, the epoxy samples did twist slightly. Shear tests of bonded samples without

  12. Resistive demountable toroidal-field coils for tokamak reactors

    SciTech Connect

    Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

    1981-07-01

    Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments.

  13. Remote maintenance concepts for the Compact Ignition Tokamak

    SciTech Connect

    Davis, F.C.; Hager, E.R.

    1988-01-01

    Because deuterium-tritium fuel will be used in the Compact Ignition Tokamak (CIT), remote handling technology is needed to carry out some maintenance operations on the machine. In keeping with the compact, low-cost nature of CIT, remote maintenance is provided only for systems with the highest probability of failure. Remote operations include removing, repairing (if feasible), and replacing such components as thermal protection tiles on the first wall, radio-frequency (rf) heating modules, and diagnostic modules. For maintenance inside the vacuum vessel, major pieces of equipment under development include an articulated boom manipulator with servomanipulators, an inspection manipulator, and special tooling. For maintenance outside the cryostat, remote equipment includes a bridge-mounted manipulator system, equipment for decontamination and hot cell activities, and for handling and packaging solid radioactive waste. The conceptual design phase of the CIT project is nearing completion; research and development activities in support of the project include demonstrations of remote maintenance operations on full-size partial mock-ups. 9 figs.

  14. Remote maintenance of Compact Ignition Tokamak ex-vessel systems

    SciTech Connect

    DePew, R.E.; Macdonald, D.

    1989-01-01

    The use of deuterium-tritium (D-T) fuel in the Compact Ignition Tokamak (CIT) will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion device's auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. Throughout the CIT remote maintenance (RM) studies conducted to date, computer modeling has been used extensively to investigate manipulator access in these complex, tightly packed, and cluttered surroundings. A recent refinement of computer modeling involves the use of an intelligent engineering work station for realtime interactive display of task simulations. This paper discusses the use of three-dimensional (3-D) kinematic computer models of the CIT machines that are proving to be powerful tools in our efforts to evaluate RM requirements. This presentation includes a video-taped simulation of remote replacement of a plasma viewing assembly. The simulation illustrates some of the constraints associated with typical RM activities and the ways in which computer modeling enhances the design process. 1 ref., 3 figs.

  15. The impact of improved physics on commercial tokamak reactors

    SciTech Connect

    Galambos, J.D.; Perkins, L.J.; Haney, S.; Mandrekas, J.

    1994-01-01

    Improvements in the confinement and beta capability of tokamak devices have long been a goal of the fusion program. We examine the impact of improvements in present day confinement and beta capabilities on commercial tokamak reactors. We characterize confinement with the achievable enhancement factor (H) over the ITER89 Power scaling confinement time, and beta by the Troyon coefficient g. A surprisingly narrow range of plasma confinement and beta are found to be useful in minimizing the cost of electricity for a tokamak reactor. Improvements in only one of these quantities is not useful beyond some point, without accompanying improvements in the other. For the plasma beta limited by a Troyon coefficient (g) near 4.3 (%mT/MA), confinement levels characterized by H factor enhancements of only 2 are useful for our nominal steady-state driven tokamak. These confinement levels are similar to those observed in present day experiments. If the permissible Troyon beta coefficient is near 6, the useful H factor confinement range increases to 2.5, still close to present day confinement levels. Inductively driven, pulsed reactors have somewhat increased useful ranges of confinement, relative to the steady-state cases. For a Troyon beta limit coefficient g near 4.3, H factors up to 2.5 are useful, and for g near 6, H factors up to 3 are useful.

  16. Preliminary design of the CIT (Compact Ignition Tokamak) cryostat

    SciTech Connect

    Goins, M.L.

    1989-01-01

    For the Compact Ignition Tokamak (CIT) to achieve the performance goals set forth, the toroidal field (TF) and poloidal field (PF) coil systems must operate in a cryogenic temperature regime. The cryostat has been designed to provide and maintain this environment. The preliminary design activity is addressing the design issues and interfaces necessary to provide a cryogenic vessel that will maintain a maximum temperature differential of 8{degree}C between the outer vessel wall and the ambient test cell conditions; operate in a pressure range of +5 psig to {minus}2 psig; accommodate numerous penetrations, including cooling, diagnostic, and gravity support items; and maintain a maximum leak rate of gaseous nitrogen at 1 l/s at 1 atm. Conceptually, the cryostat consists of thermal insulation sandwiched between an inner primary stainless steel pressure vessel and a thin outer stainless steel wall. Design activities have concentrated on determining the size and shape of the primary vessel wall and selecting the best candidate thermal insulation materials for future irradiation testing. The following shapes of the upper and lower cryostat structure were analyzed: a standard ASME torispherical domed top and bottom; a nonstandard domed top and bottom; and a 2{degree} sloped conical top and bottom contour. Screening of candidate insulation materials was based on lowest thermal conductivity over the range of temperatures anticipated in the CIT environment; low material cost and apparent ease of assembly; and survivability of material in the CIT irradiation environment. This paper presents the configuration development of the cryostat used to maintain the cryogenic temperature environment for CIT. 3 refs., 3 figs., 3 tabs.

  17. Resistive toroidal-field coils for tokamak reactors

    SciTech Connect

    Kalnavarns, J.; Jassby, D.L.

    1980-11-01

    This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized.

  18. Compact reactor/ORC power source

    SciTech Connect

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500/sup 0/C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370/sup 0/C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs.

  19. Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

    SciTech Connect

    Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

    1990-05-01

    This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements.

  20. WILDCAT: a catalyzed D-D tokamak reactor

    SciTech Connect

    Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

    1981-11-01

    WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.

  1. Materials needs for compact fusion reactors

    SciTech Connect

    Krakowski, R.A.

    1983-01-01

    The economic prospects for magnetic fusion energy can be dramatically improved if for the same total power output the fusion neutron first-wall (FW) loading and the system power density can be increased by factors of 3 to 5 and 10 to 30, respectively. A number of compact fusion reactor embodiments have been proposed, all of which would operate with increased FW loadings, would use thin (0.5 to 0.6 m) blankets, and would confine quasi-steady-state plasma with resistive, water-cooled copper or aluminum coils. Increased system power density (5 to 15 MWt/m/sup 3/ versus 0.3 to 0.5 MW/m/sup 3/), considerably reduced physical size of the fusion power core (FPC), and appreciably reduced economic leverage exerted by the FPC and associated physics result. The unique materials requirements anticipated for these compact reactors are outlined against the well documented backdrop provided by similar needs for the mainline approaches. Surprisingly, no single materials need that is unique to the compact systems is identified; crucial uncertainties for the compact approaches must also be addressed by the mainline approaches, particularly for in-vacuum components (FWs, limiters, divertors, etc.).

  2. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  3. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; K. Rule; M. Viola; M. Williams; R. Strykowsky

    1999-11-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling.

  4. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

  5. A Compact Quasi-axisymmetric Stellarator Reactor

    SciTech Connect

    L.P. Ku; the ARIES-CS Team

    2003-10-20

    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.

  6. Plasma Physics Regimes in Tokamaks with Li Walls

    SciTech Connect

    L.E. Zakharo; N.N. Gorelenkov; R.B. White; S.I. Krasheninnikov; G.V. Pereverzev

    2003-08-21

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors.

  7. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  8. Operation of a tokamak reactor in the radiative improved mode

    NASA Astrophysics Data System (ADS)

    Morozov, D. Kh.; Mavrin, A. A.

    2016-03-01

    The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ E with an increase in the introduced power has been analyzed.

  9. Radiation analysis of the CIT (Compact Ignition Tokamak) pellet injector system and its impact on personnel access

    SciTech Connect

    Selcow, E.C.; Stevens, P.N.; Gomes, I.C.; Gomes, L.M.

    1987-01-01

    Conceptual design of the Compact Ignition Tokamak (CIT) is near completion. This short-pulse ignition experiment is planned to follow the operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The high neutron wall loadings, /approximately/4-5 MW/m/sup 2/, associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of device components and personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure, and the entire experiment is housed in a circular test cell facility with a radius of /approximately/12 m. The most critical radiation concern in the CIT design process relates to the numerous penetrations in the device. This paper discusses the impact of a major penetration on the design and operations of the CIT pellet injection system. The pellet injector is a major component, which has a line-of-sight penetration through the igloo and test cell wall. All current options for maintenance of the injector require personnel access. A nuclear analysis has been performed to determine the feasibility of hands-on access. Results indicate that personnel access to the pellet injector glovebox is possible. 10 refs., 3 figs., 3 tabs.

  10. Tangential and Vertical Compact Torus Injection Experiments on the STOR-M Tokamak

    NASA Astrophysics Data System (ADS)

    Xiao, Chijin; D, Liu; S, Livingstone; A, K. Singh; E, Zhang; A, Hirose

    2005-04-01

    This paper describes the setup and results of compact torus (CT) injection experiments on the STOR-M tokamak. Tangential CT injection into STOR-M induced H-mode-like phenomena including doubling the electron density, reduction in the Hα radiation level, suppression of the floating potential fluctuations, suppression of the m = 2 Mirnov oscillations, and increase in the global energy confinement time. Experimental setup, bench-test results, and some preliminary injection data for vertical CT injection experiments on STOR-M will be shown. In addition, numerical simulations of the CT trajectories in tokamak discharges for both tangential and vertical injection geometries will be discussed.

  11. Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas

    NASA Astrophysics Data System (ADS)

    Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira

    2010-11-01

    As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.

  12. Plasma engineering studies for Tennessee Tokamak (TENTOK) fusion power reactor

    SciTech Connect

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Uckan, N.A.; Shannon, T.E.

    1984-02-01

    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses deuterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. The major parameters are R/sub 0/ = 6.4 m, a = 1.6 m, sigma (elongation) = 1.65, (n) = 1.5 x 10/sup 20/ m/sup -3/, (T) = 15 keV, (..beta..) = 6%, B/sub T/ (on-axis) = 5.6 T, I/sub p/ = 8.5 MA, and wall loading = 3 MW/m/sup 2/. Detailed analyses are performed in the areas of (1) transport simulation using the one-and-one-half-dimensional (1-1/2-D) WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control systems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

  13. Exploratory studies of tokamaks as fusion test reactors

    NASA Astrophysics Data System (ADS)

    Mau, T. K.; Conn, R. W.

    1982-06-01

    Studies have been performed to explore various plasma burn scenarios for a tokamak test reactor which could follow the next generation of large tokamak experiments. Tradeoffs between an ignited burning plasma and a sub-ignited driven plasma are examined in terms of device size and performance as a fusion engineering test facility. It is found that plasma performance levels, measured by ignition margin, amplification factor Q, and fusion power output, increase with device size, more optimistic transport scaling laws, lower magnetic field ripple, and higher Β. The performance of a generally low stress ( B 0=4 T) reference device, with major radius R=4.5 m and minor radius a=1.3 m in a D-shaped ( κ=1.6) plasma has been evaluated over a wide range of operating parameters. In particular, a moderate fusion power output of 300 MW is obtained, the driven plasma having Q≅ 10, an edge ripple of 1%, and a density ranging between 1.0 and 1.5×1014 cm-3. The same device operated at a higher general level of stress ( B 0=5.3 T) is predicted to achieve ignition, but is not required for the mission of an engineering test facility and would entail greater technical risk.

  14. Transient getter scheme for the Tokamak Fusion Test Reactor

    SciTech Connect

    Cecchi, J.L.; Cohen, S.A.; Sredniawski, J.J.

    1980-01-01

    The ability of the Tokamak Fusion Test Reactor (TFTR) to attain the largest fusion power gain depends critically on minimizing plasma contamination and controlling the densities of the reacting deuterium and tritium. Experiments on a number of tokamaks have demonstrated that gettering over an appreciable surface area (greater than or equal to 10%) of the vacuum vessel greatly facilitates both of these objectives. One particular problem in implementing a surface pumping system in TFTR, however, is a restriction on the maximum allowable tritium content of the getter. This restriction could require regeneration of the absorbed tritium after as few as 50 machine pulses. We have developed a scheme utilizing SAES Zr/Al getter modules which obviates the need for such frequent interruptions of machine operation by taking advantage of the pulsed operation of TFTR. With the Zr/Al getter at temperatures between 500/sup 0/C to 600/sup 0/C it is possible to achieve a quasi-steady state in the tritium loading where the quantity of tritium desorbed between pulses is equal to the quantity which is absorbed during a pulse. Since frequent thermal cycling is not required, this scheme also reduces the possibility of Zr/Al getter material fatigue.

  15. A non-inductively driven steady state tokamak reactor

    SciTech Connect

    Fenstermacher, M.E.; Devoto, R.S.; Bulmer, R.H.; Lee, J.D.; Miller, J.R.; Schultz, J.

    1988-09-20

    The physics and engineering guidelines for the ITER device are shown to lead to viable and attractive operating points for a steady state tokamak power reactor. Non-inductive current drive is provided in steady state by high energy neutral beam injection in the plasma core, lower hybrid slow waves in the outer regions of the plasma and bootstrap current. Plasma gain Q (/equivalent to/fusion power/input power) in excess of 20 and average neutron wall loading, approx. 2.0 MW/m/sup 2/ are predicted in a device with major radius, R/sub 0/ = 7.5 m and minor radius, a = 2.8 m. 15 refs., 3 figs., 3 tabs.

  16. Tritium pellet injector design for tokamak fusion test reactor

    SciTech Connect

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.; Combs, S.K.; Easterly, C.E.; Lunsford, R.V.; Milora, S.L.; Schuresko, D.D.; White, J.A.; Williamson, D.H.

    1985-01-01

    A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper.

  17. The spheromak as a compact fusion reactor

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.

    1987-03-01

    After summarizing the economic and utility-based rationale for compact, higher-power-density fusion reactors, the gun-sustained spheromak concept is explored as one of a number of poloidal-field-dominated confinement configurations that might improve the prospects for economically attractive and operationally simplified fusion power plants. Using a comprehensive physics/engineering/costing model for the spheromak, guided by realistic engineering constraints and physics extrapolation, a range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported. The results presented herein provide the basis for conceptual engineering designs of key fusion-power-core (FPC) subsystems and more detailed plasma modeling of this promising, high mass-power-density concept, which stresses single-piece FPC maintenance, steady-state current drive through electrostatic magnetic helicity injection, a simplified co-axial electrode-divertor, and efficient resistive-coal equilibrium-field coils. The optimal FPC size and the cost estimates project a system that competes aggressively with the best offered by alternative energy sources while simplifying considerably the complexity that has generally been associated with most approaches to magnetic fusion energy.

  18. Physical and engineering constraints for tokamak reactors with helical coils

    SciTech Connect

    Potok, R.E.; Becker, H.

    1983-09-01

    The authors present an analytical and numerical analysis of a tokamak reactor with a set of helical coils added in order to eliminate plasma disruptions. The optimal helical configuration was found to be a set of continuous, l = 2 stellarator coils which are made of copper and are internal to the toroidal field coils, l being the number of poloidal field periods. (The optimization process did not include evaluation of the viability of a modular stellarator reactor). Scaling laws were developed for this optimal configuration, and a series of parametric scans are performed with varying assumptions for the forces on the helical coils and the ratio of helical coil transform to plasma transform (M). The option space available for attractive reactor designs is strongly constrained and involves large forces on the helical coils, low q plasma operation (q being the plasma safety factor), and moderately low M (3 to 5). Numerical calculations showed that M must be > 3 in order to obtain well defined flux surfaces. This is in agreement with results from the JIPP-T-II Tokomak.

  19. Demountable Toroidal Field Magnets for Use in a Compact Modular Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Mangiarotti, F. J.; Goh, J.; Takayasu, M.; Bromberg, L.; Minervini, J. V.; Whyte, D.

    2014-05-01

    A concept of demountable toroidal field magnets for a compact fusion reactor is discussed. The magnets generate a magnetic field of 9.2 T on axis, in a 3.3 m major radius tokamak. Subcooled YBCO conductors have a critical current density adequate to provide this large magnetic field, while operating at 20 K reduces thermodynamic cooling cost of the resistive electrical joints. Demountable magnets allow for vertical replacement and maintenance of internal components, potentially reducing cost and time of maintenance when compared to traditional sector maintenance. Preliminary measurements of contact resistance of a demountable YBCO electrical joint between are presented.

  20. Conceptual design study of a superconducting spherical tokamak reactor with a self-consistent system analysis code

    NASA Astrophysics Data System (ADS)

    Hong, B. G.; Hwang, Y. S.; Kang, J. S.; Lee, D. W.; Joo, H. G.; Ono, M.

    2011-11-01

    In a spherical tokamak (ST) reactor, the radial build of toroidal field coil and the shield play a key role in determining the size of the reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with a one-dimensional radiation transport code. A conceptual design study of a compact superconducting ST reactor with an aspect ratio of up to 2.0 is conducted and the optimum radial build is identified. It is shown that the use of an improved shielding material and high-temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at a low aspect ratio, and that by using an inboard neutron reflector instead of a breeding blanket, tritium self-sufficiency is possible with an outboard blanket only and thus a compact-sized all superconducting coil ST reactor is viable.

  1. A Conceptual Design of Superconducting Spherical Tokamak Reactor

    NASA Astrophysics Data System (ADS)

    Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi

    This paper presents a fusion reactor concept named “JUST (Japanese Universities’ Super Tokamak reactor)”. From the plasma confinement system to the power generation system is evaluated in this work. JUST design has features as follows: the superconducting magnet, the steady state operation with high bootstrap current fraction, the easy replacement of neutron damaged first wall, the high heat flux in the divertor, and the low cost (or high β). By winding the OH solenoid over the center stack of toroidal field coil, we have the low aspect ratio and the 80cm thick neutron shield to protect the superconducting center stack. JUST is designed by using the 0-D transport code under the assumption that the energy confinement time is 1.8 times of the IPB98(y,2) scaling. Main parameters are as follows: the major radius of 4.5m, the aspect ratio of 1.8, the elongation ratio of 2.5, the toroidal field of 2.36T, the plasma current of 18MA, the toroidal beta of 22%, the central electron and ion temperature of 15keV and the fusion thermal power of 2.4GW. By using the mercury heat exchanger and the steam turbine, the heat efficiency is 33% and the electric power is 0.74GW.

  2. Radiation analysis of the CIT (Compact Ignition Tokamak) pellet injector system and its impact on personnel access

    SciTech Connect

    Selcow, E.C.; Stevens, P.N.; Gomes, I.C.; Gomes, L.M.

    1988-08-01

    The conceptual design of the Compact Ignition Tokamak (CIT) is nearing completion. The CIT is a short-pulse ignition experiment, which is planned to follow the operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). The high neutron wall loadings, 4--5 MW/m/sup 2/, associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of device components as well as personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure, and the entire experiment is housed in a circular test cell facility that has a radius of 12 m. The most critical radiation concerns in the CIT design process relate to the numerous penetrations in the device. This report discusses the impact of a major penetration on the design and operation of the pellet injection system in the CIT. The pellet injector is a major component, and it has a line-of-sight penetration through the igloo and test cell wall. All current options for maintenance of the injector require hands-on-access. A nuclear analysis has been performed to establish the feasibility of hands-on-access. A coupled Monte Carlo/discrete-ordinates methodology was used to perform the analysis. This problem is characterized by deep penetration and streaming with very large length-to-diameter ratios. Results from this study indicate that personnel access to the pellet injector glovebox is possible. 14 refs., 3 figs., 3 tabs.

  3. Comparing Linear Microinstability of the National Compact Stellarator Expriment and a Shaped Tokamak

    SciTech Connect

    J.A. Baumgaertel, G.W. Hammett and D.R. Mikkelsen

    2012-11-20

    One metric for comparing con nement properties of di erent magnetic fusion energy con gurations is the linear critical gradient of drift wave modes. The critical gradient scale length determines the ratio of the core to pedestal temperature when a plasma is limited to marginal stability in the plasma core. The gyrokinetic turbulence code GS2 was used to calculate critical temperature gradients for the linear, collisionless ion tem- perature gradient (ITG) mode in the National Compact Stellarator Experiment (NCSX) and a prototypical shaped tokamak, based on the pro les of a JET H-mode shot and the stronger shaping of ARIES-AT. While a concern was that the narrow cross section of NCSX at some toroidal locations would result in steep gradients that drive instabilities more easily, it is found that other stabilizing e ects of the stellarator con guration o set this so that the normalized critical gradients for NCSX are competitive with or even better than for the tokamak. For the adiabatic ITG mode, NCSX and the tokamak had similar critical gradients, though beyond marginal stability, NCSX had larger growth rates. However, for the kinetic ITG mode, NCSX had a higher critical gradient and lower growth rates until a/LT ≈ 1:5 a/LT;crit, when it surpassed the tokamak's. A discussion of the results presented with respect to a/LT vs R/LT is included.

  4. Current generation by helicons and lower hybrid waves in modern tokamaks and reactors ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect

    Vdovin, V. L.

    2013-02-15

    The innovative concept and 3D full-wave code modeling the off-axis current drive by radio-frequency (RF) waves in large-scale tokamaks, ITER and DEMO, for steady-state operation with high efficiency is proposed. The scheme uses the helicon radiation (fast magnetosonic waves at high (20-40) ion cyclotron frequency harmonics) at frequencies of 500-700 MHz propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by helicons, in conjunction with the bootstrap current, ensure the maintenance of a given value of the total current in the stability margin q(0) {>=} 2 and q(a) {>=} 4, and will help to have regimes with a negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure {beta}{sub N} > 3 (the so-called advanced scenarios) of interest for the commercial reactor. Modeling with full-wave three-dimensional codes PSTELION and STELEC showed flexible control of the current profile in the reactor plasmas of ITER and DEMO, using multiple frequencies, the positions of the antennae and toroidal wave slow down. Also presented are the results of simulations of current generation by helicons in the DIII-D, T-15MD, and JT-60AS tokamaks. Commercially available continuous-wave klystrons of the MW/tube range are promising for commercial stationary fusion reactors. The compact antennae of the waveguide type are proposed, and an example of a possible RF system for today's tokamaks is given. The advantages of the scheme (partially tested at lower frequencies in tokamaks) are a significant decline in the role of parametric instabilities in the plasma periphery, the use of electrically strong resonator-waveguide type antennae, and substantially greater antenna-plasma coupling.

  5. [Injection of compact toroids for tokamak fueling and current drive]. Progress report, 1990--1991

    SciTech Connect

    Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.

    1991-12-31

    The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.

  6. Scale length study in TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Hiroe, S.; Goldston, R.J.; Bitter, M.; Bush, C.E.; Efthimion, P.C.; Grek, B.; Johnson, D.W.; Murakami, M.; Schivell, J.; Towner, H.H.

    1988-12-01

    The scale lengths of the electron density (L/sub n//sub e/), temperature (L/sub T//sub e/), and pressure (L/sub p//sub e/) gradients were investigated during the 1985 operating period of the Tokamak Fusion Test Reactor (TFTR) for gas-fueled plasmas with neutral beam injection and movable limiter. Although the global energy confinement time degrades as the heating power increases or the plasma current decreases, the radial profiles of the scale lengths (L/sub T//sub e/ and L/sup p//sub e/) remain unchanged. Especially, the electron pressure profile is constrained not to change. This trend appears to hold over a fairly wide range of TFTR operational regimes. The radial profiles of L/sub n//sub e/ and /eta//sub e/ (= L/sub n//sub e//L/sub T//sub e/) also appear to remain unchanged, although the uncertainties of the experimental data for these quantities are greater than those for L/sub T//sub e/ and L/sub p//sub e/. The experimental parameters are used to evaluate theoretical predictions of the electron thermal diffusivity, and the results are compared with the empirical thermal diffusivity. 34 refs., 18 figs., 2 tabs.

  7. Compact formulas for bounce/transit averaging in axisymmetric tokamak geometry

    SciTech Connect

    Duthoit, F.-X.; Brizard, A. J.; Hahm, T. S.

    2014-12-15

    Compact formulas for bounce and transit orbit averaging of the fluctuation-amplitude eikonal factor in axisymmetric tokamak geometry, which is frequently encountered in bounce-gyrokinetic description of microturbulence, are given in terms of the Jacobi elliptic functions and elliptic integrals. These formulas are readily applicable to the calculation of the neoclassical susceptibility in the framework of modern bounce-gyrokinetic theory. In the long-wavelength limit for axisymmetric electrostatic perturbations, we recover the expression for the Rosenbluth-Hinton residual zonal flow [M. N. Rosenbluth and F. L. Hinton, Phys. Rev. Lett. 80, 724 (1998)] accurately.

  8. Problems in modeling TF ripple loss of fast alphas from a tokamak reactor

    SciTech Connect

    Hively, L.M.

    1987-01-01

    The present status of modeling TF ripple loss of fast alphas from tokamaks is summarized. The modeling issues are discussed, and several new aspects of this problem are described, including gyromotion, radial electric field, and sawtoothing. Existing models predict that TF ripple loss of fast alphas will have a low-to-moderate impact on the design of a tokamak engineering test reactor (ETR). 52 refs., 3 figs., 2 tabs.

  9. Summary of TFTR (Tokamak Fusion Test Reactor) diagnostics, including JET (Joint European Torus) and JT-60

    SciTech Connect

    Hill, K.W.; Young, K.M.; Johnson, L.C.

    1990-05-01

    The diagnostic instrumentation on TFTR (Tokamak Fusion Test Reactor) and the specific properties of each diagnostic, i.e., number of channels, time resolution, wavelength range, etc., are summarized in tables, grouped according to the plasma parameter measured. For comparison, the equivalent diagnostic capabilities of JET (Joint European Torus) and the Japanese large tokamak, JT-60, as of late 1987 are also listed in the tables. Extensive references are given to publications on each instrument.

  10. Design of an ion cyclotron resonance heating system for the Compact Ignition Tokamak

    SciTech Connect

    Yugo, J.J.; Goranson, P.L.; Swain, D.W.; Baity, F.W.; Vesey, R.

    1987-01-01

    The Compact Ignition Tokamak (CIT) requires 10-20 MW of ion cyclotron resonance heating (ICRH) power to raise the plasma temperature to ignition. The initial ICRH system will provide 10 MW of power to the plasma, utilizing a total of six rf power units feeding six current straps in three ports. The systems may be expanded to 20 MW with additional rf power units, antennas, and ports. Plasma heating will be achieved through coupling to the fundamental ion cyclotron resonance of a /sup 3/He minority species (also the second harmonic of tritium). The proposed antenna is a resonant double loop (RDL) structure with vacuum, shorted stubs at each end for tuning and impedance matching. The antennas are of modular, compact construction for installation and removal through the midplane port. Remote maintainability and the reactorlike operating environment have a major impact on the design of the launcher for this machine. 6 refs., 7 figs., 5 tabs.

  11. A high speed compact microwave interferometer for density fluctuation measurements in Sino-UNIted Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Zhong, H.; Tan, Y.; Liu, Y. Q.; Xie, H. Q.; Gao, Z.

    2016-11-01

    A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector, without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer's capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.

  12. Compact Reversed-Field Pinch Reactors (CRFPR): preliminary engineering considerations

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Embrechts, M.J.; Schnurr, N.M.; Battat, M.E.; LaBauve, R.J.; Davidson, J.W.

    1984-08-01

    The unique confinement physics of the Reversed-Field Pinch (RFP) projects to a compact, high-power-density fusion reactor that promises a significant reduction in the cost of electricity. The compact reactor also promises a factor-of-two reduction in the fraction of total cost devoted to the reactor plant equipment (i.e., fusion power core (FPC) plus support systems). In addition to operational and developmental benefits, these physically smaller systems can operate economically over a range of total power output. After giving an extended background and rationale for the compact fusion approaches, key FPC subsystems for the Compact RFP Reactor (CRFPR) are developed, designed, and integrated for a minimum-cost, 1000-MWe(net) system. Both the problems and promise of the compact, high-power-density fusion reactor are quantitatively evaluated on the basis of this conceptual design. The material presented in this report both forms a framework for a broader, more expanded conceptual design as well as suggests directions and emphases for related research and development.

  13. Plasma-material Interaction Studies On Lithium And Lithiated Substrates During Compact Tokamak Operation

    SciTech Connect

    Nieto, M.; Allain, J. P.; Hassanein, A.; Titov, V.; Hendricks, M.; Gray, T.; Kaita, R.; Kugel, H.; Majeski, R.; Mansfield, D.; Spaleta, J.; Timberlake, J.

    2006-12-04

    The role of lithium on the modification of recycling regimes in fusion reactors has renewed interest of previous lithium supershot experiments carried out in TFTR. There is a need to understand the interaction between edge plasmas and lithiated plasma-facing components (PFCs), which have the potential of enabling fusion reactors to operate at low-recycling regimes. The Interaction of Materials with Particles and Components Testing (IMPACT) facility at Argonne National Laboratory is currently collaborating with Princeton Plasma Physics Laboratory (PPPL) to conduct lithiated surface studies for the National Spherical Tokamak Experiment (NSTX) and the Current Drive eXperiment - Upgrade (CDX-U). IMPACT has the necessary tools to perform experiments that diagnose the surface dynamics of lithium thin films on metallic and non-metallic substrates, and can be monitored with multiple in-situ techniques (LEISS, AES, QMS and XPS) capturing real-time surface dynamics. Therefore, these techniques are available during He+ and D+ irradiation. Surface sputtering measurements can be performed using a quartz crystal microbalance -- dual crystal unit (QCM-DCU) with very high sensitivity.Initial results suggest that lithium intercalation into graphite occurs quite rapidly and only a fraction lithium can be kept on the surface. On metallic substrates this intercalation is absent. Additional results of Li/metal systems show lithium surface self-healing with temperature. It was also found that the presence of lithium seems to inhibit hydrocarbon formation during D+ bombardment of graphite. Experiments in CDX-U have tested the effect of both solid and liquid lithium PFCs on tokamak plasmas, and significant changes in tokamak operation are observed. These include a strong reduction in both recycling and impurity levels in the gas phase, lowered loop voltage during ohmic operation, and an increased electron temperature at the edge.

  14. Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

    SciTech Connect

    Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.

    1994-07-01

    Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES&H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power.

  15. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Program

    NASA Astrophysics Data System (ADS)

    McGuire, Thomas

    2015-11-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. An overview of the concept and its diamagnetic, high beta magnetically encapsulated linear ring cusp confinement scheme will be given. The analytical model of the major loss mechanisms and predicted performance will be discussed, along with the major physics challenges. Key features of an operational CFR reactor will be highlighted. The proposed developmental path following the current experimental efforts will be presented. ©2015 Lockheed Martin Corporation. All Rights Reserved.

  16. Elastic-plastic analysis of the toroidal field coil inner leg of the compact ignition tokamak

    SciTech Connect

    Horie, T.

    1987-07-01

    Elastic-plastic analyses were made for the inner leg of the Compact Ignition Tokamak toroidal field (TF) coil, which is made of copper-Inconel composite material. From the result of the elastic-plastic analysis, the effective Young's moduli of the inner leg were determined by the analytical equations. These Young's moduli are useful for the three-dimensional, elastic, overall TF coil analysis. Comparison among the results of the baseline design (R = 1.324 m), the bucked pressless design, the 1.527-m major radius design, and the 1.6-m major radius design was also made, based on the elastic-plastic TF coil inner leg analyses.

  17. Plans for the CIT (Compact Ignition Tokamak) instrumentation and control system

    SciTech Connect

    Preckshot, G.G.

    1987-10-07

    Extensive experience with previous fusion experiments (TFTR, MFTF-B and others) is driving the design of the Instrumentation and Control System (I and C) for the Compact Ignition Tokamak (CIT) to be built at Princeton. The new design will reuse much equipment from TFTR and will be subdivided into six major parts: machine control, machine data acquisition, plasma diagnostic instrument control and instrument data acquisition, the database, shot sequencing and safety interlocks. In a major departure from previous fusion experiment control systems, the CIT machine control system will be a commercial process control system. Since the machine control system will be purchased as a completely functional product, we will be able to concentrate development manpower in plasma diagnostic instrument control, data acquisition, data processing and analysis, and database systems. We will discuss the issues driving the design, give a design overview and state the requirements upon any prospective commercial process control system.

  18. Ecological environment of the proposed site for the Compact Ignition Tokamak at Princeton Plasma Physics Laboratory

    SciTech Connect

    Not Available

    1987-12-01

    This report gives a description of the exological environment of D-site and the surrounding area at Princeton Plasma Physics Laboratory (PPPL) near Princeton, New Jersey. D-site at PPL is the proposed location for construction of a new fusion test facility, the Compact Ignition Tokamak (CIT). This report was prepared as supplemental information for an Environmental Assessment for the proposed CIT at PPL. The report characterizes the vegetation and wildlife occuring at and near the site and describes the water quality and aquatic ecology of Bee Brook. No threatened or endangered plant or animal species are known to occur in the area, although suitable habitat exists for some species. The occurrence of a forested wetland north of the site is discussed. 9 refs., 2 figs.

  19. The development of a universal diagnostic probe system for Tokamak fusion test reactor

    NASA Technical Reports Server (NTRS)

    Mastronardi, R.; Cabral, R.; Manos, D.

    1982-01-01

    The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

  20. Reactors Power Balance Based on Compact Toroid

    NASA Astrophysics Data System (ADS)

    Romadanov, I.

    2013-10-01

    The power balance of the plasma source system based on compact toroid with a pulse mode of formation is considered. Developed model takes into account the time dependence of the processes, in a pulsed mode of operation of the system. Also magnetic configuration shape and nuclei energy distribution fluency were considered. Analytical solution of Grad-Shafranov equation was taken to determine the shape of the separatrix and magnetic fields into the configuration. For practical calculation, program was written. Code is able to calculates volume power reactions in the confined plasma, using as input the geometry of the magnetic field, the cross section of reaction rates and energy distribution of the nuclei.

  1. UCLA program in reactor studies: The ARIES tokamak reactor study. Progress report, December 1, 1990--November 30, 1991

    SciTech Connect

    Not Available

    1991-12-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ``modest`` extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  2. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  3. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  4. Alpha particle losses from Tokamak Fusion Test Reactor deuterium{endash}tritium plasmas

    SciTech Connect

    Darrow, D.S.; Zweben, S.J.; Batha, S.; Budny, R.V.; Bush, C.E.; Chang, Z.; Cheng, C.Z.; Duong, H.H.; Fang, J.; Fisch, N.J.; Fischer, R.; Fredrickson, E.D.; Fu, G.Y.; Heeter, R.F.; Heidbrink, W.W.; Herrmann, H.W.; Herrmann, M.C.; Hill, K.; Jaeger, E.F.; James, R.; Majeski, R.; Medley, S.S.; Murakami, M.; Petrov, M.; Phillips, C.K.; Redi, M.H.; Ruskov, E.; Spong, D.A.; Strait, E.J.; Taylor, G.; White, R.B.; Wilson, J.R.; Wong, K.; Zarnstorff, M.C.

    1996-05-01

    Because alpha particle losses can have a significant influence on tokamak reactor viability, the loss of deuterium{endash}tritium alpha particles from the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire {ital et} {ital al}., Phys. Plasmas {bold 2}, 2176 (1995)] has been measured under a wide range of conditions. In TFTR, first orbit loss and stochastic toroidal field ripple diffusion are always present. Other losses can arise due to magnetohydrodynamic instabilities or due to waves in the ion cyclotron range of frequencies. No alpha particle losses have yet been seen due to collective instabilities driven by alphas. Ion Bernstein waves can drive large losses of fast ions from TFTR, and details of those losses support one element of the alpha energy channeling scenario. {copyright} {ital 1996 American Institute of Physics.}

  5. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  6. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    SciTech Connect

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  7. Radioactive waste disposal characteristics of candidate tokamak demonstration reactors

    SciTech Connect

    Hoffman, E.A.; Stacey, W.M.; Hertel, N.E.

    1998-08-01

    Results from the current physics, materials and blanket R and D programs are combined with physics and engineering design constraints to characterize candidate tokamak demonstration plant (DEMO) designs. Blanket designs based on the principal structural materials, breeding materials and coolants being developed for the DEMO were adapted from the literature. Neutron flux and activation calculations were performed, and several radioactive waste disposal indices were evaluated, for each design. Of the primary low-activation structural materials under development in the US, it appears that vanadium and ferritic steel alloys, and possibly silicon carbide, could lead to DEMO designs which could satisfy realistic low-level waste (LLW) criteria, provided that impurities can be controlled within plausible limits. Allowable LLW concentrations are established for the limiting alloying and impurity elements. All breeding materials and neutron multipliers considered meet the LLW criterion.

  8. TRAIL: a Tokamak Rail Gun Limiter for fusion reactors

    SciTech Connect

    Powell, J R; Yu, W S; Fillo, J A; Usher, J L

    1980-01-01

    An attractive new limiter concept is investigated. The Tokamak Rail Gun Limiter (TRAIL) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state-of-the-art (several km/sec). Accelerators injecting pellets at approx. 1 km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  9. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

  10. Compact, battery powered, wireless digitizers for in situ data acquisitions in the sino-united spherical tokamak.

    PubMed

    Liu, Yangqing; Tan, Yi; Ke, Rui; Yang, Hao; Wang, Wenhao; Gao, Zhe

    2015-07-01

    Potential isolation and long cable drive are very important in acquiring certain signals from tokamak diagnostics. Compact, battery powered, wireless digitizers for in situ data acquisition have been developed and routinely used in the sino-united spherical tokamak to solve the problems of isolation and long cables. The wireless digitizers utilize the integrated analog to digital converters and the static random access memory of microcontrollers but transfer data wirelessly. They consist of simple and concise circuits but have considerable performances of 12-16 bit in resolution and 500-1000 kS/s in sample rate. Wireless triggering and energy saving are two major challenges of the wireless digitizers. Wireless transceivers in the data link layer are used as trigger and can reduce the trigger jitters to be smaller than 1 μs. In order to reduce the energy consumption, the wireless digitizers are waken only when the tokamak is about to discharge. After discharges, they turn to a periodic checking mode with current consumption smaller than 200 μA. Because of low duty cycle, the wireless digitizers have a battery life of up to four weeks. In general, the wireless digitizers have better performance than normal isolation amplifiers and can greatly simplify the cable connections. They are very suitable for the data acquisition of dangerous and/or susceptible analog signals in tokamaks.

  11. Compact, battery powered, wireless digitizers for in situ data acquisitions in the sino-united spherical tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Yangqing; Tan, Yi; Ke, Rui; Yang, Hao; Wang, Wenhao; Gao, Zhe

    2015-07-01

    Potential isolation and long cable drive are very important in acquiring certain signals from tokamak diagnostics. Compact, battery powered, wireless digitizers for in situ data acquisition have been developed and routinely used in the sino-united spherical tokamak to solve the problems of isolation and long cables. The wireless digitizers utilize the integrated analog to digital converters and the static random access memory of microcontrollers but transfer data wirelessly. They consist of simple and concise circuits but have considerable performances of 12-16 bit in resolution and 500-1000 kS/s in sample rate. Wireless triggering and energy saving are two major challenges of the wireless digitizers. Wireless transceivers in the data link layer are used as trigger and can reduce the trigger jitters to be smaller than 1 μs. In order to reduce the energy consumption, the wireless digitizers are waken only when the tokamak is about to discharge. After discharges, they turn to a periodic checking mode with current consumption smaller than 200 μA. Because of low duty cycle, the wireless digitizers have a battery life of up to four weeks. In general, the wireless digitizers have better performance than normal isolation amplifiers and can greatly simplify the cable connections. They are very suitable for the data acquisition of dangerous and/or susceptible analog signals in tokamaks.

  12. Design and Implementation of a Compact Lithium Evaporator to Minimize Edge Neutral Drag on the HBT-EP Tokamak

    NASA Astrophysics Data System (ADS)

    Maurer, D. A.; Shiraki, D.; Hanson, J. M.; James, R.; Mauel, M. E.; Navratil, G. A.; Pedersen, T. S.

    2007-11-01

    A candidate for the dissipation mechanism responsible for rotational stabilization of the resistive wall mode (RWM) on the HBT-EP tokamak is neutral damping via charge exchange reactions with cold Deuterium neutrals in the edge plasma. Lithium has been used successfully on the CDX-U experiment to substantially reduce recycling as a plasma fueling mechanism [1]. To study charge exchange drag and its effect on the RWM we have designed and plan to implement a small, compact Lithium evaporator that will getter neutral recycled Deuterium atoms to minimize the edge neutral population and as a result reduce charge exchange reactions. Evaporator design details, bench measurements of Li effusion rate performance, and plans for implementation of the evaporator during daily tokamak operation will be outlined. [1] R. Majeski, et al., Phys. Rev. Lett. 97, 7, 075002, 2006.

  13. Model for collisional fast ion diffusion into Tokamak Fusion Test Reactor loss cone

    SciTech Connect

    Chang, C.S. |; Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S.

    1994-08-01

    An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium-tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR). A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the TRANSP code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing-down path due to reduction in banana thickness, which enables us to understand, for the first time. the dependence of the collisional loss rate on Z{sub eff}.

  14. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    SciTech Connect

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  15. Tokamak reactor for treating fertile material or waste nuclear by-products

    DOEpatents

    Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

    2012-10-02

    Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

  16. Articulated limiter blade for a tokamak fusion reactor

    DOEpatents

    Doll, D.W.

    1982-10-21

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  17. Articulated limiter blade for a tokamak fusion reactor

    DOEpatents

    Doll, David W.

    1985-01-01

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  18. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

  19. RF-driven tokamak reactor with sub-ignited, thermally stable operation

    SciTech Connect

    Harten, L.P.; Bers, A.; Fuchs, V.; Shoucri, M.M.

    1981-02-01

    A Radio-Frequency Driven Tokamak Reactor (RFDTR) can use RF-power, programmed by a delayed temperature measurement, to thermally stabilize a power equilibrium below ignition, and to drive a steady state current. We propose the parameters for such a device generating approx. = 1600 MW thermal power and operating with Q approx. = 40 (= power out/power in). A one temperature zero-dimensional model allows simple analytical formulation of the problem. The relevance of injected impurities for locating the equilibrium is discussed. We present the results of a one-dimensional (radial) code which includes the deposition of the supplementary power, and compare with our zero-dimensional model.

  20. Experimental study of thermal crisis in connection with Tokamak reactor high heat flux components

    NASA Astrophysics Data System (ADS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G. P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-04-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

  1. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  2. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    SciTech Connect

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-12-31

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

  3. Local tests of parallel electrical resistivity in the Tokamak Fusion Test Reactor

    SciTech Connect

    Batha, S.H.; Levinton, F.M.; Ramsey, A.T.; Schmidt, G.L.; Zarnstorff, M.C.

    1997-01-01

    The motional Stark effect (MSE) polarimeter measures the local magnetic field pitch angle, proportional to the ratio of the poloidal to toroidal magnetic fields, in the Tokamak Fusion Test Reactor (TFTR). The authors have used the polarimeter to measure the temporal evolution of the local value of the magnetic field pitch angle during large changes in the current profile such as during a current ramp or discharge initiation. The measured evolution is compared to the evolution predicted by classical and neoclassical resistivity models. The neoclassical resistivity model is a better predictor of the local pitch angle temporal evolution than the classical model.

  4. Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

    SciTech Connect

    Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

    1987-11-01

    The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability. 23 refs., 9 figs., 4 tabs.

  5. Near-term tokamak-reactor designs with high-performance resistive magnets

    SciTech Connect

    Cohn, D.R.; Bromberg, L.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Advanced Fusion Test Reactors (AFTR) designs have been developed using BITTER type magnets which are capable of steady state operation. The goals of compact AFTR designs (with major radii R approx. 2.5 - 4 m), include DT ignition with large physics margins; high duty cycle, long pulse operation; and DD-DT operation with low tritium concentration. Larger AFTR designs (R approx. 5 m), have the additional goal of early demonstration of self sufficiency in tritium production. The AFTR devices could also serve as prototypes for commercial reactors. Compact ignition test reactors have also been designed (R approx. 1 - 2 m). These designs use BITTER magnets that are inertially cooled starting at liquid nitrogen temperature. A detailed engineering design was developed for ZEPHYR.

  6. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    SciTech Connect

    Rule, K.; Scott, J.; Larson, S.

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  7. Safety in the ARIES-III D- sup 3 He tokamak reactor design

    SciTech Connect

    Herring, J.S.; Dolan, T.J.

    1991-01-01

    The ARIES-3 reactor study is an extensive examination of the viability of a D-{sup 3}He-fueled commercial tokamak power reactor. Because neutrons are produced only through side reactions, the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The ARIES-3 reactor uses an organic coolant to cool the first wall, shield and divertor. The organic coolant has a low vapor pressure at the operating temperature required for good thermal efficiency. Radiation damage requires processing the coolant to remove and crack radiolytic products that would otherwise foul cooling surfaces. The cracking process produces waste, which must be disposed of through incineration or burial. We estimated the offsite doses due to incineration at five candidate locations. The plasma confinement requirements for a D-{sup 3}He reactor are much more challenging than those for a D-T reactor. Thus, the demands on the divertor are more severe, particularly during a disruption. We explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. We also modeled a loss-of-cooling accident in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. We analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps. For our reference design, the tritium will be burned in the plasma. These results re-emphasize the need for low activation materials and advanced divertor designs, even in reactors using advanced fuels.

  8. Compact and Lightweight Sabatier Reactor for Carbon Dioxide Reduction

    NASA Technical Reports Server (NTRS)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

    2011-01-01

    The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA s cabin Atmosphere Revitalization System and In-Situ Resource Utilization architectures for both low-earth orbit and long-term manned space missions. In the current International Space Station (ISS) and other low orbit missions, metabolically-generated CO2 is removed from the cabin air and vented into space, resulting in a net loss of O2. This requires a continuous resupply of O2 via water electrolysis, and thus highlights the need for large water storage capacity. For long-duration space missions, the amount of life support consumables is limited and resupply options are practically nonexistent, thus atmosphere resource management and recycle becomes crucial to significantly reduce necessary O2 and H2O storage. Additionally, the potential use of the Martian CO2-rich atmosphere and Lunar regolith to generate life support consumables and propellant fuels is of interest to NASA. Precision Combustion, Inc. (PCI) has developed a compact, lightweight Microlith(Registered TradeMark)-based Sabatier (CO2 methanation) reactor which demonstrates the capability of achieving high CO2 conversion and near 100% CH4 selectivity at space velocities of 30,000-60,000 hr-1. The combination of the Microlith(Registered TradeMark) substrates and durable, novel catalyst coating permitted efficient Sabatier reactor operation that favors high reactant conversion, high selectivity, and long-term durability. This paper presents the reactor development and performance results at various operating conditions. Additionally, results from 100-hr durability tests and mechanical vibration tests are discussed.

  9. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    SciTech Connect

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  10. Long- and short-term trends in vessel conditioning of TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.; Boody, F.P.; Bush, C.E.; Groebuer, R.J.; Hawryluk, R.J.; Hill, K.W.; Mueller, D.; Owens, D.K.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150/sup 0/C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area.

  11. A magnetohydrodynamic stability study of reverse shear equilibria in the Tokamak Fusion Test Reactor

    SciTech Connect

    Phillips, M.W.; Zarnstorff, M.C.; Manickam, J.; Levinton, F.M.; Hughes, M.H.

    1996-05-01

    A study is presented of the low-{ital n} ({ital n}=1,2,3) magnetohydrodynamic stability of equilibria with reverse shear safety factor profiles. The low-{ital n} stability boundaries are found to be characterized by resonance structures due to internal so-called {open_quote}{open_quote}infernal{close_quote}{close_quote} mode types of instabilities. The parametric dependence of shear reversal width and depth, current, and pressure gradient on the beta limit are determined by using profile models that allow each parameter to be varied independently. Reverse magnetic shear is found to have a stabilizing influence for modes with toroidal mode numbers {ital n}{ge}2 leading to the possibility of improved {beta} limits in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. {bold 26}, 11 (1984)]. {copyright} {ital 1996 American Institute of Physics.}

  12. Anomalous Loss of DT Alpha Particles in the Tokamak Fusion Test Reactor

    SciTech Connect

    Herrmann, Hans W.

    1997-06-01

    Princeton's Tokamak Fusion Test Reactor (TFTR) is the first experimental fusion device to routinely use tritium to study the deuterium-tritium (DT) fusion reaction,allowing the first systematic study of DT alpha particles in tokamak plasmas. A crucial aspect of alpha-particle physics is the fraction of alphas that escape from the plasma, particularly since these energetic particles can do severe damage to the first wall of a reactor. An escaping alpha collector probe has been developed for TFTR's DT phase. Energy distributions of escaping alphas have been determined by measuring the range of alpha-particles implanted into nickel foils located within the alpha collector. Results at 1.0 MA of plasma current are in good agreement with predictions for first orbit alpha loss. Results at 1.8 MA, however, show a significant anomalous loss of partially thermalized alphas (in addition to the expected first orbit loss), which is not observed with the lost alpha scintillator detectors in DT plasmas, but does resemble the anomalous "delayed" loss seen in DD plasmas. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations. An experiment designed to study the effect of plasma major radius shifts on alpha-particle loss has led to a better understanding of alpha-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing alpha-particles forced to move toward higher magnetic field during an inward major radius shift (i.e. compression) would mirror and become trapped particles, leading to increased alpha loss. Such an effect was looked for during the shift experiment, however, no significant changes in alpha loss to the 90 degree lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an alpha-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized alpha-particles near the

  13. Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2015-11-01

    The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.

  14. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    SciTech Connect

    Bell, M.G.; Beer, M.; Batha, S.

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} {approx} 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions.

  15. Deuterium--tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    SciTech Connect

    Bell, M.G.; Batha, S.; Beer, M.; Bell, R.E.; Belov, A.; Berk, H.; Bernabei, S.; Bitter, M.; Breizman, B.; Bretz, N.L.; Budny, R.; Bush, C.E.; Callen, J.; Cauffman, S.; Chang, C.S.; Chang, Z.; Cheng, C.Z.; Darrow, D.S.; Dendy, R.O.; Dorland, W.; Duong, H.; Efthimion, P.C.; Ernst, D.; Evenson, H.; Fisch, N.J.; Fisher, R.; Fonck, R.J.; Fredrickson, E.D.; Fu, G.Y.; Furth, H.P.; Gorelenkov, N.N.; Goloborodko, V.Y.; Grek, B.; Grisham, L.R.; Hammett, G.W.; Hawryluk, R.J.; Heidbrink, W.; Herrmann, H.W.; Herrmann, M.C.; Hill, K.W.; Hogan, J.; Hooper, B.; Hosea, J.C.; Houlberg, W.A.; Hughes, M.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Kaita, R.; Kaye, S.; Kesner, J.; Kim, J.S.; Kissick, M.; Krasilnikov, A.V.; Kugel, H.; Kumar, A.; Lam, N.T.; Lamarche, P.; LeBlanc, B.; Levinton, F.M.; Ludescher, C.; Machuzak, J.; Majeski, R.P.; Manickam, J.; Mansfield, D.K.; Mauel, M.; Mazzucato, E.; McChesney, J.; McCune, D.C.; McKee, G.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mirnov, S.V.; Mueller, D.; Nagayama, Y.; Navratil, G.A.; Nazikian, R.; Okabayashi, M.; Osakabe, M.; Owens, D.K.; Park, H.K.; Park, W.; Paul, S.F.; Petrov, M.P.; Phillips, C.K.; Phillips, M.; Phillips, P.; Ramsey, A.T.; Rice, B.; Redi, M.H.; Rewoldt, G.; Reznik, S.; Roquemore, A.L.; Rogers, J.; Ruskov, E.; Sabbagh, S.A.; Sasao, M.; Schilling, G.; Schmidt, G.L.; Scott, S.D.; Semenov, I.; Senko, T.; Skinner, C.H.; Stevenson, T.; Strait, E.J.; Stratton, B.C.; Strachan, J.D.; Stodiek, W.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Thompson, M.E.; von Goeler, S.; Von Halle, A.; Walters, R.T.; Wang, S.; White, R.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.L.; Wurden, G.A.; Yamada, M.; Yavorski, V.; Young, K.M.; Zakharov, L.; Zarnstorff, M.C.; Zweben, S.J.

    1997-05-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas {bold 2}, 2176 (1995)] have explored several novel regimes of improved tokamak confinement in deuterium{endash}tritium (D--T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through {ital in situ} deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a}{approx}4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D--T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D--T plasmas with q{sub 0}{gt}1 and weak magnetic shear in the central region, a toroidal Alfvn eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions. {copyright} {ital 1997 American Institute of Physics.}

  16. The Compact Ignition Tokamak and electron cyclotron heating: Description of need; assessment of prospects

    SciTech Connect

    Ignat, D.W.; Cohn, D.R.; Woskov, P.P.

    1989-01-01

    The CIT will benefit from auxiliary heating of 10 to 40 MW. The schedules of both the CIT construction project and the operating plan contain adequate time to develop and implement ECH systems based on the gyrotron and the induction free electron laser (IFEL). Each approach has advantages and is the object of R and D at the level of many millions of dollars per year. While the gyrotron is further advanced in terms of power and pulse length achieved, rapid progress is scheduled for the IFEL, including experiments on tokamaks. Plans of CIT, gyrotron, and IFEL make 1992 an appropriate time frame to commit to one or both systems. 12 refs., 8 figs., 2 tabs.

  17. MIRI: A multichannel far-infrared laser interferometer for electron density measurements on TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Mansfield, D.K.; Park, H.K.; Johnson, L.C.; Anderson, H.M.; Chouinard, R.; Foote, V.S.; Ma, C.H.; Clifton, B.J.

    1987-07-01

    A ten-channel far-infrared laser interferometer which is routinely used to measure the spatial and temporal behavior of the electron density profile on the TFTR tokamak is described and representative results are presented. This system has been designed for remote operation in the very hostile environment of a fusion reactor. The possible expansion of the system to include polarimetric measurements is briefly outlined. 13 refs., 8 figs.

  18. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  19. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    SciTech Connect

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-19

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  20. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  1. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  2. Spent fuel utilization in a compact traveling wave reactor

    SciTech Connect

    Hartanto, Donny; Kim, Yonghee

    2012-06-06

    In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

  3. High poloidal beta long-pulse experiments in the Tokamak Fusion Test Reactor*

    NASA Astrophysics Data System (ADS)

    Kesner, J.; Mauel, M. E.; Navratil, G. A.; Sabbagh, S. A.; Bell, M.; Budny, R.; Bush, C.; Fredrickson, E.; Grek, B.; Janos, A.; Johnson, D.; Mansfield, D.; McCune, D.; McGuire, K.; Park, H.; Ramsey, A.; Synakowski, E.; Taylor, G.; Zarnstorff, M.; Batha, S. H.; Levinton, F. M.

    1993-07-01

    Experiments have been performed in the Tokamak Fusion Test Reactor [D. M. Meade et al. in Plasma Physics Controlled Nuclear Fusion Research, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] with neutral beam injection of up to 4 sec. duration, which is comparable to the time scale for resistive redistribution of the plasma current profile. These plasmas were created using a rapid decrease of the plasma current which initially created a plasma with enhanced stability and confinement. As the current profile evolved, a significantly reduced beta limit was observed. The high ɛβp plasmas had up to 90% of the current driven noninductively which significantly broadened the current profile during the long pulse lengths. These experiments demonstrated that high βN plasmas could not be sustained for times longer than the resistive relaxation of the outer current region which at early times after the current ramp-down carried negative current. At later times in lower βN discharges, beta collapses were sometimes observed as the current profile broadened at βN˜1.5. The appearance of disruptions was consistent with the predictions of ideal magnetohydrodynamics (MHD) stability analyses.

  4. Energy confinement time and electron density profile shape in TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Park, H.K.; Bell, M.G.; Goldston, R.J.; Hawryluk, R.J.; Johnson, D.W.; Scott, S.D.; Wieland, R.M.; Zarnstorff, M.C.; Bitter, M.; Bretz, N.; Budny, R.; Dylla, H.F.; Grek, B.; Howell, R.B.; Hsuan , H.; Johnson, L.C.; Mansfield, D.K.; Ramsey, A.T.; Schivell, J.; Taylor, G.; Ulrickson, M.

    1989-11-01

    The electron density profiles of intense deuterium neutral-beam- heated plasmas (P{sub tot}/P{sub ohm} {gt} 10) are characterized as a peakedness parameter (F{sub ne} = n{sub eo}/{l angle}n{sub e}{r angle}) in the Tokamak Fusion Test Reactor (TFTR). The gross energy confinement time ({tau}{sub E} = E{sub tot}/P{sub tot}) at the time of maximum stored energy is found to be a weak function of the plasma current and total heating power but depends strongly on the peakedness parameter. A regression study showed {tau}{sub E} = 2.4 {times} 10{sup {minus}3}F{sub ne}{sup 0.76}I{sub P}{sup 0.18}P{sub tot}{sup {minus}0.12} for a data set of 561 discharges in the TFTR. Also {tau}{sub E} can be represented as {tau}{sub E} = {tau}{sub E}{sup L}f(F{sub ne}), where {tau}{sub E}{sup L} is the empirical L-mode scaling result. A similar scaling applies to an appropriately defined incremental energy confinement time ({tau}{sub inc} = dE{sub tot}/dP{sub tot}{vert bar}{sub F{sub ne} = constant}). 14 refs., 4 figs.

  5. Design of deuterium and tritium pellet injector systems for Tokamak Fusion Test Reactor

    SciTech Connect

    Wysor, R.B.; Baylor, L.R.; Bryan, W.E.; Combs, S.K.; Fisher, P.W.; Lunsford, R.V.; Maxon, B.S.; Milora, S.L.; Webster, D.J.; Schmidt, G.L.

    1985-01-01

    Three pellet injector designs developed by the Oak Ridge National Laboratory (ORNL) are planned for the Tokamak Fusion Test Reactor (TFTR) to reach the goal of a tritium pellet injector by 1988. These are the Repeating Pneumatic Injector (RPI), the Deuterium Pellet Injector (DPI) and the Tritium Pellet Injector (TPI). Each of the pellet injector designs have similar performance characteristics in that they deliver up to 4-mm-dia pellets at velocities up to 1500 m/s with a dsign goal to 2000 m/s. Similar techniques are utilized to freeze and extrude the pellet material. The injector systems incorporate three gun concepts which differ in the number of gun barrels and the method of forming and chambering the pellets. The RPI, a single barrel repeating design, has been operational on TFTR since April 1985. Fabrication and assembly are essentially complete for DPI, and TPI is presently on hold after completing about 80% of the design. The TFTR pellet injector program is described, and each of the injector systems is described briefly. Design details are discussed in other papers at this symposium.

  6. Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

    SciTech Connect

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja

    1994-05-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.

  7. Anomalous fast ion losses at high β on the tokamak fusion test reactor

    SciTech Connect

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R.

    2015-03-15

    This paper describes experiments carried out on the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] to investigate the dependence of β-limiting disruption characteristics on toroidal field strength. The hard disruptions found at the β-limit in high field plasmas were not found at low field, even for β's 50% higher than the empirical β-limit of β{sub n} ≈ 2 at high field. Comparisons of experimentally measured β's to TRANSP simulations suggest anomalous loss of up to half of the beam fast ions in the highest β, low field shots. The anomalous transport responsible for the fast ion losses may at the same time broaden the pressure profile. Toroidal Alfvén eigenmodes, fishbone instabilities, and Geodesic Acoustic Modes are investigated as possible causes of the enhanced losses. Here, we present the first observations of high frequency fishbones [F. Zonca et al., Nucl. Fusion 49, 085009 (2009)] on TFTR. The interpretation of Axi-symmetric Beam-driven Modes as Geodesic Acoustic Modes and their possible correlation with transport barrier formation are also presented.

  8. Review of deuterium--tritium results from the Tokamak Fusion Test Reactor*

    SciTech Connect

    McGuire, K. M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, C. W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M; Bell, M. G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C. S.; Cheng, C. Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D. S; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P. C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Fujita, T.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Hawryluk, R. J.; Heidbrink, W.; Herrmann, H. W.; Hill, K. W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Loughlin, M. J.; Machuzak, J.; Majeski, R.; Mansfield, D. K.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M. P.; McChesney, J.; McCormack, B.; McCune, D. C.; McKee, G.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Norris, M.; O`Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Parks, P.; Paul, S. F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G.; Scillia, R.; Scott, S. D.; Semenov, I.; Senko, T.; Sesnic, S.; Sissingh, R.; Skinner, C. H.; Snipes, J.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Tighe, W.; Timberlake, J. R.; Tobita, K.; Towner, H. H.; Tuszewski, M.; Halle, A. Von; Vannoy, C.; Viola, M.; Goeler, S. Von; Voorhees, D.; Walters, R. T.; Wester, R.; White, R.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Winston, J.; Wright, K.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zavereev, V.; Zweben, S. J.

    1995-01-01

    The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≈ 2.8 MW m₋3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni (0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

  9. Mode Conversion of High-Field-Side-Launched Fast Waves at the Second Harmonic of Minority Hydrogen in Advanced Tokamak Reactors

    NASA Astrophysics Data System (ADS)

    Sund, R.; Scharer, J.

    2003-12-01

    Under advanced tokamak reactor conditions, the Ion-Bernstein wave (IBW) can be generated by mode conversion of a fast magnetosonic wave incident from the high-field side on the second harmonic resonance of a minority hydrogen component, with near 100% efficiency. IBWs have the recognized capacity to create internal transport barriers through sheared plasma flows resulting from ion absorption. The relatively high frequency (around 200 MHz) minimizes parasitic electron absorption and permits the converted IBW to approach the 5th tritium harmonic. It also facilitates compact antennas and feeds, and efficient fast wave launch. The scheme is applicable to reactors with aspect ratios < 3 such that the conversion and absorption layers are both on the high field side of the magnetic axis. Large machine size and adequate separation of the mode conversion layer from the magnetic axis minimize poloidal field effects in the conversion zone and permit a 1-D full-wave analysis. 2-D ray tracing of the IBW indicates a slightly bean-shaped equilibrium allows access to the tritium resonance.

  10. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade

    NASA Astrophysics Data System (ADS)

    Arredondo Parra, R.; Moreno Quicios, R.; Ploeckl, B.; Birkenmeier, G.; Herrmann, A.; Kocsis, G.; Laggner, F. M.; Lang, P. T.; Lunt, T.; Macian-Juan, R.; Rohde, V.; Sellmair, G.; Szepesi, T.; Wolfrum, E.; Zeidner, W.; Neu, R.

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 1019 atoms (0.21 mg) to 1.64 × 1020 atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet.

  11. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade.

    PubMed

    Arredondo Parra, R; Moreno Quicios, R; Ploeckl, B; Birkenmeier, G; Herrmann, A; Kocsis, G; Laggner, F M; Lang, P T; Lunt, T; Macian-Juan, R; Rohde, V; Sellmair, G; Szepesi, T; Wolfrum, E; Zeidner, W; Neu, R

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 10(19) atoms (0.21 mg) to 1.64 × 10(20) atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet. PMID:26931850

  12. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade.

    PubMed

    Arredondo Parra, R; Moreno Quicios, R; Ploeckl, B; Birkenmeier, G; Herrmann, A; Kocsis, G; Laggner, F M; Lang, P T; Lunt, T; Macian-Juan, R; Rohde, V; Sellmair, G; Szepesi, T; Wolfrum, E; Zeidner, W; Neu, R

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 10(19) atoms (0.21 mg) to 1.64 × 10(20) atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet.

  13. Development of Compact Quasi-Axisymmetric Stellarator Reactor Configurations

    SciTech Connect

    L.P. Ku; M. Zarnstorff; R.B. White; W.A. Cooper; R. Sanchez; H. Neilson; J.A. Schmidt

    2003-09-19

    We have started to examine the reactor potential of quasi-axisymmetric (QA) stellarators with an integrated approach that includes systems evaluation, engineering considerations, and plasma and coil optimizations. In this paper, we summarize the progress made so far in developing QA configurations with reduced alpha losses while retaining good MHD stability properties. The minimization of alpha losses is achieved by directly targeting the collisionless orbits to prolong the average resident times. Configurations with an overall energy loss rate of {approx}10% or less, including collisional contributions, have been found. To allow remotely maintaining coils and machine components in a reactor environment, there is a desire to simplify to the extent possible the coil design. To this end, finding a configuration that is optimized not only for the alpha confinement and MHD stability but also for the good coil and reactor performance, remains to be a challenging task.

  14. A compact breed and burn fast reactor using spent nuclear fuel blanket

    SciTech Connect

    Hartanto, D.; Kim, Y.

    2012-07-01

    A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

  15. Natural Fueling of the Core and Edge in a Tokamak Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Wan, Weigang

    2010-11-01

    A natural fueling mechanismootnotetextW. Wan, S. E. Parker, Y. Chen and F. W. Perkins, Phys. Plasmas 17, 040701 (2010). that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is presented. In H-mode plasmas dominated by ion-temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is investigated using the gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport. Additionally, it is shown that the Helium ash diffuses radially outward as the cold fuel moves radially inward. The natural fueling effect may also apply to the edge pedestal density buildup. Recent DEGAS 2 calculations indicate the neutrals in the pedestal are colder than the background ions.ootnotetextD. Stotler, International Transport Task Force Meeting, Annapolis, MD (2010). We intend to do further work to determine what cold fuel profiles are needed to fuel the pedestal and if they are consistent with edge neutral source models. Natural fueling (either in the core or edge) requires a two component (hot bulk and cold fuel) plasma and charge exchange collisions tend to equilibrate the ion and neutral source temperature reducing the effect. We will further investigate the relevant collisional time scales and further demonstrate the viability of the natural fueling mechanism for ITER parameters.

  16. End points in discharge cleaning on TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Mueller, D.; Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Bush, C.E.; Gettelfinger, G.; Hawryluk, R.J.; Hill, K.W.; Janos, A.C.; Jobes, F.C.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) /approx/ 3 at /approx/ 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H/sub 2/O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after /approx/ 10/sup 5/ pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150/degree/C to about 250/degree/C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H/sub 2/O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab.

  17. The ARIES-III D- sup 3 He tokamak reactor: Design-point determination and parametric studies

    SciTech Connect

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A. ); Santarius, J.F. )

    1991-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-3 design uses a D-{sup 3}He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-1 is included. 11 refs., 5 figs.

  18. A New Interpretation of Alpha-particle-driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    R. Nazikian; G.J. Kramer; C.Z. Cheng; N.N. Gorelenkov; H.L. Berk; S.E. Sharapov

    2003-03-26

    The original description of alpha-particle-driven instabilities in the Tokamak Fusion Test Reactor (TFTR) in terms of Toroidal Alfvin Eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the anti-ballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time.

  19. Plasma engineering design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    NASA Astrophysics Data System (ADS)

    Bathke, C. G.; Embrechts, M. J.; Hagenson, R. L.; Krakowski, R. A.; Miller, R. L.

    1983-11-01

    The rationale for and the characteristics of the high-power-density Compact Reversed-Field Pinch Reactor (CRFPR) are discussed. Particular emphasis is given to key plasma engineering aspects of the conceptual design, including plasma operations, current drive, and impurity/ash control by means of pumped limiters or magnetic divertors. A brief description of the Fusion-Power-Core integration is given.

  20. FIREBIRD: A conceptual design of a field reversed configuration Compact Torus Fusion Reactor (CTFR)

    NASA Astrophysics Data System (ADS)

    Raman, Roger; Zubrin, Robert M.

    Work carried out by the Nuclear Engineering 512 design team at the University of Washington on a conceptual design study of a Compact-Torus (field-reverse) Fusion Reactor Configuration (CTFR) is summarized. The primary objective was to develop a reactor design for high engineering power density, modest recirculating power, and competitive cost of electrical power. A conceptual design was developed for a translating field-reversed configuration reactor; based on the physics developed by Tuszewski and Lindford at LANL and by Hoffman and Milroy at MSNW. Furthermore, it also appears possible to operate a simplified form of this reactor using a pure D-D fuel cycle after an initial D-T ignition ramp to reach the advanced fuel operating regime. One optimistic reactor so designed has a length of about 35 meters, producing a net electrical power of about 375 MWe.

  1. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  2. ECH tokamak

    SciTech Connect

    Firestone, M.A.; Mau, T.K.; Conn, R.W.

    1985-04-01

    A small steady-state tokamak capable of producing power in the 100 to 300 MWe range and relying on electron cyclotron RF heating (ECH) for both heating and current drive is described. Working in the first MHD stability regime for tokamaks, the approach adheres to the recently discovered maximum beta limit. An appropriate figure of merit is the ratio of the fusion power to absorbed RF power. Efficient devices are feasible at both small and large values of fusion power, thereby pointing to a development path for an attractive commercial fusion reactor.

  3. Compact, high-power nuclear reactor systems based on small diameter particulate fuel

    NASA Astrophysics Data System (ADS)

    Powell, J. R.; Botts, T. E.

    Two compact, high-power nuclear reactor concepts are discussed. Both are gas-cooled cavity-type reactors which utilize particulate fuel of the type now used in HTGR reactors. Unshielded reactor volumes are on the order of one cubic meter. The Fixed Bed Reactor operating temperature is limited to 2500 K and the output power to 250 MW(e). In the Rotating Bed Reactor fuel is held within a rotating porous metal drum as a rotating fluidized bed. Rotating Bed Reactor outlet temperatures up to 3000 K and output power levels up to 1000 MW(e) are achievable. Both reactors can be brought up from stand by to full power in times on the order of a few seconds, due primarily to the short thermal time constant for the fuel particles. Turbine and MHD Brayton are the power conversion cycles of choice. Open cycle operation is generally favored for applications operating at less than 1000 sec of equivalent integrated full power. At power levels above 1 MW(e), the liquid droplet radiator is the favored means of heat rejection. Power system specific power levels of 10 kW(e)/kg (not including shield) appears to be quite feasible.

  4. Criticality-safety analyses of compacted and water-flooded. SP-100 reactors

    SciTech Connect

    Brandon, D.I.; Sapir, J.L.

    1986-01-01

    Reactivity calculations were performed to determine the sensitivity of three liquid metal-cooled, fast reactor designs to various accident environments. The concepts, proposed for the SP-100 Space Nuclear Power Program, included one thermionic and two fuel-pin designs. Numerous models of each core were developed to analyze the effect of core compaction and of water-flooded lattice spreading. Results indicate that those designs incorporating in-core control are least affected by core compaction and that the thermonic concept can best withstand expansion of the flooded fuel element array.

  5. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    NASA Astrophysics Data System (ADS)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  6. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect

    Vdovin, V.

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  7. Development of tokamak experiment technology: A study on the nuclear fusion reactor

    NASA Astrophysics Data System (ADS)

    Lee, Kwang Won; Hwang, Chul Kew; In, Sang Ryul; Kim, Sung Kyu; Oh, Byung Hoon; Jeong, Seung Ho; Hong, Bong Guen; Yoon, Jae Sung; Yoon, Byung Joo; Song, Woo Sup

    1994-07-01

    Overall test operation of KT-1 tokamak has been done with the modified power supplies and constructed plasma position feedback control system. The achieved plasma parameters of KT-1 are Ip (plasma current) greater than 15 kA (9ms) and Td (plasma current sustaining time) greater than 12 ms (7.5 kA) until now. Instrumentations for the carbonization/boronization are also equipped and applied to the first boronization of invessel components of KT-1 tokamak. As a preliminary treatment of the 1st wall of tokamak vacuum torus, RG discharge cleaning has been done. Plasma qualities have been improved by discharge cleaning. In addition, helicon plasma device is designed succeeding exploitation of the first wall conditioning program which are all for the development of conditioning technology of 1st wall facing the plasma. As a part of the development program on the plasma monitoring technologies, conceptual design study on the FIR interferometer- polarimeter is accomplished. Parallel to these experimental activities, an effort has been made in designing the operating scenario of poloidal field system of KT-2 tokamak.

  8. Exchange Flow Characteristics in a Tokamak Vacuum Vessel of Fusion Reactor Under the Loss-of-Vacuum Conditions

    NASA Astrophysics Data System (ADS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi

    1997-06-01

    When a Tokamak vacuum vessel of fusion reactor is broken, buoyancy-driven exchange flows will take place through breaches after the inside pressure of the vacuum vessel (VV) becomes equal to the outside pressure. The exchange flow may bring a mixture of activated dusts and tritium from the inside of the VV to the outside through the breaches. Moreover, the exchange flow may remove decay heat from the plasma-facing components. A preliminary LOVA (Loss Of VAcuum event) apparatus was constructed to investigate quantitative heat transfer characteristics of the exchange flows through the breaches under the LOVA conditions. The results of this study, the relationship between Froude numbers and breach locations in the VV was determined and empirical correlations for the average Froude numbers were derived.

  9. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    SciTech Connect

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.; Chan, A.A.; England, A.C.; Hendel, H.W.; Medley, S.S.; Nieschmidt, E.; Roquemore, A.L.; Scott, S.D.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and /sup 3/He ions, respectively. When the plasma was compressed, the d(d,n)/sup 3/He fusion reaction rate increased a factor of five, and the /sup 3/He(d,p)/sup 4/He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling.

  10. A compact-toroid fusion reactor based on the field-reversed theta pinch

    NASA Astrophysics Data System (ADS)

    Hagenson, R. L.; Krakowski, R. A.

    1981-03-01

    A dynamic plasma model and an overall systems approach were used to examine a compact toroid (CTOR) reactor embodiment that uses a field reversed theta pinch as a plasma source. The field reversed plasmoid was formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high technology plasmoid source from the hostile reactor environment. Stabilization of the field reversed plasmoid was provided by a passive conducting shell located outside the high temperature blanket but within the low field superconducting magnets and associated radiation shielding. A reactor concept was developed with a length below approximately 40 m that generates 300-400 MWe of net electrical power with a recirculating power fraction less than 0.15.

  11. Compact-Toroid Fusion Reactor (CTOR) based on the field reversed theta pinch

    NASA Astrophysics Data System (ADS)

    Hagenson, R. L.; Krakowski, R. A.

    Scoping studies of a translating compact torus reactor (CTOR) made on the basis of a dynamic plasma model and an overall systems approach are presented. This CTOR embodiment uses a field reversed theta pinch as a plasma source. The field reversed plasmoid would be formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high technology plasmoid source from the hostile reactor environment. Stabilization of the field reversed plasmoid would be provided by a passive conducting shell located outside the high temperature blanket but within the low field superconducting magnets and associated radiation shielding. On the basis of this batch burn but thermally steady state approach, a reactor concept emerges with a length below approximately 40 m that generates 300 to 400 MWe of net electrical power with a recirculating power fraction less than 0.15.

  12. Compact-Toroid fusion reactor based on the field-reversed theta pinch

    NASA Astrophysics Data System (ADS)

    Hagenson, R. L.; Krakowski, R. A.

    1981-03-01

    Early scoping studies based on approximate, analytic models were extended on the basis of a dynamic plasma model and an overall systems approach to examine a compact toroid (CTOR) reactor embodiment that uses a field-reversed theta pinch as a plasma source. The field-reversed plasmoid would be formed and compressionally heated to ignition prior to injection into and translation through a linear burn chamber, thereby removing the high-technology plasmoid source from the hostile reactor environment. Stabilization of the field-reversed plasmoid would be provided by a passive conducting shell located outside the high temperature blanket but within the low field superconducting magnets and associated radiation shielding. On the basis of this batch-burn but thermally steady-state approach, a reactor concept emerges with a length below approx. 40 m that generates 300 to 400 MWe of net electrical power with a recirculating power fraction less than 0.15.

  13. Development and analysis of a compact low-conversion ratio fast burner reactor.

    SciTech Connect

    Smith, M. A.; Hill, R. N.; Nuclear Engineering Division

    2006-05-12

    This report explores design options for compact fast burner reactors that can achieve low conversion ratios. Operational characteristics and whole-core reactivity coefficients are generated and contrasted with low conversion ratio designs of previous studies. A compact core point design is then selected and detailed reactivity coefficients are displayed and discussed. The effectiveness of fast spectrum systems for actinide transmutation has been well documented. The key advantage of the fast spectrum resides in the severely reduced capture/fission ratios. this inhibits the production of the higher actinides that dominate the long-term radiotoxicity of nuclear waste. In conventional fast burner studies, the transmutation rate was limited by constraints placed on the fuel composition. In an earlier phase of this study the entire range of fuel compositions (including non-uranium fuel) was explored to assess the performance and safety limits of fast burner reactor systems. In this report, similar fuel compositions are utilized for application in compact configurations to achieve conversion ratios below 0.5.

  14. Visible Spectrometer at the Compact Toroid Injection Experiment, the Sustained Spheromak Plasma Experiment and the Alcator C-Mod Tokamak for Doppler Width and Shift Measurements

    SciTech Connect

    Graf, A; Howard, S; Horton, R; Hwang, D; May, M; Beiersdorfer, P; McLean, H; Terry, J

    2006-05-15

    A novel Doppler spectrometer is currently being used for ion or neutral velocity and temperature measurements on the Alcator C-Mod Tokamak. The spectrometer has an f/No. of {approx}3.1 and is appropriate for visible light (3500-6700 {angstrom}). The full width at half maximum from a line emitting calibration source has been measured to be as small as 0.4 {angstrom}. The ultimate time resolution is line brightness light limited and on the order of ms. A new photon efficient detector is being used for the setup at C-Mod. Time resolution is achieved by moving the camera during a plasma discharge in a perpendicular direction through the dispersion plane of the spectrometer causing a vertical streaking across the camera face. Initial results from C-Mod as well as previous measurements from the Compact Toroid Injection Experiment (CTIX) and the Sustained Spheromak Plasma Experiment (SSPX) are presented.

  15. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  16. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    NASA Astrophysics Data System (ADS)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  17. Electron cyclotron heating of a tokamak reactor at down-shifted frequencies

    SciTech Connect

    Fidone, I.; Giruzzi, G.; Mazzucato, E.

    1985-01-01

    The absorption of electron cyclotron waves in a hot and dense tokamak plasma is investigated for the case of the extraordinary mode for outside launching. It is shown that, for electron temperatures T/sub e/ greater than or equal to 5 keV, strong absorption occurs for oblique propagation at frequencies significantly below the electron gyrofrequency at the plasma center. A new density dependence of the wave absorption is found which is more favorable for plasma heating than the familiar n/sub e//sup -1/ scaling.

  18. High-power-density approaches to magnetic fusion energy: Problems and promise of compact Reversed-Field Pinch Reactors (CRFPR)

    NASA Astrophysics Data System (ADS)

    Hagenson, R. L.; Krakowski, R. A.; Dreicer, H.

    If the cost assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves optimistic, approaches that promise considerably increased system power density and reduced mass utilization are required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique magnetic topology associated with the Reversed Field Pinch (RFP), the compact reactor embodiment of this approach is particularly attractive from the view point of low field resistive coils operating with Ohmic losses that are small relative to the fusion power. The RFP, one example of a high power density (HPD) approach to magnetic fusion energy. A comprehensive system model is described and applied to select a unique, cost optimized design point that is used for a subsequent conceptual engineering design of the Compact RFP Reactor.

  19. Thermionic plasma injection for the Lockheed Martin T4 Compact Fusion Reactor experiment

    NASA Astrophysics Data System (ADS)

    Heinrich, Jonathon

    2015-11-01

    Lockheed Martin's Compact Fusion Reactor (CFR) concept relies on diamagnetic confinement in a magnetically encapsulated linear ring cusp geometry. Plasma injection into cusp field configurations requires careful deliberation. Previous work has shown that axial injection via a plasma gun is capable of achieving high-beta conditions in cusp configurations. We present a pulsed, high power thermionic plasma source and the associated magnetic field topology for plasma injection into the caulked-cusp magnetic field. The resulting plasma fueling and cross-field diffusion is discussed.

  20. Final Report for Monitoring of Reactor Antineutrinos with Compact Germanium Detectors

    SciTech Connect

    Orrell, John L.; Collar, J. I.

    2009-07-01

    This 2008 NCMR project has pursued measurement of the antineutrino-nucleus coherent scattering interaction using a low-energy threshold germanium gamma-ray spectrometer of roughly one-half kilogram total mass. These efforts support development of a compact system for monitoring the antineutrino emission from nuclear reactor cores. Such a monitoring system is relevant to nuclear safeguards and nuclear non-proliferation in general by adding a strong method for assuring quantitative material balance of special nuclear material in the nuclear fuel cycle used in electricity generation.

  1. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    SciTech Connect

    Schriener, T. M.; El-Genk, M. S.

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  2. Ignition in near term D-/sup 3/He tokamak reactors: Appendix B

    SciTech Connect

    Emmert, G.A.; Deng, B.Q.

    1987-01-01

    The prospects for achieving breakeven and ignition in near term ETR type tokamaks under D/He-3 relevant conditions are considered. Using present scaling laws for beta in the first stability regime, it is found that CIT may be close to breakeven with the presently planned toroidal magnetic field system, if the ASDEX H-mode scaling law is used. With Kaye-Goldston scaling, Q = .22 can be attained, but this requires an excessive amount of rf heating power. Larger devices, such as NET/INTOR, can ignite with ASDEX H-mode scaling with an increase of the toroidal field by 20% and removal of the blanket and reduction of the inboard shield to that required for D/He-3. 5 refs., 4 figs., 2 tabs.

  3. Method for determining fast-alpha-particle confinement in tokamak plasmas using resonant nuclear reactors

    SciTech Connect

    Cecil, F.E.; Zweben, S.J.; Medley, S.S.

    1986-03-01

    The resonant nuclear reactions D(..cap alpha..,..gamma..)/sup 6/Li, /sup 6/Li(..cap alpha..,..gamma..)/sup 10/B, and /sup 7/Li(..cap alpha..,..gamma..)/sup 11/B are examined as diagnostics of fast-alpha-particle confinement in tokamak plasmas. Gamma rays from these resonant reactions with energies from 2.1 MeV to 9.2 MeV may be used to infer the alpha-particle population between energies of 0.4 MeV and 2.6 MeV. The ratio of these alpha-burnup reactions to the reactions T(D,..gamma..)/sup 5/He and /sup 3/He(D,..gamma..)/sup 5/Li provides a technique for the measurement of alpha confinement.

  4. Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

    SciTech Connect

    Kim, Seokho H; Berry, Jan

    2011-01-01

    U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclear pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.

  5. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    SciTech Connect

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  6. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    SciTech Connect

    Ferrada, Juan J; Reiersen, Wayne T

    2011-01-01

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment

  7. Mode-particle resonances during near-tangential neutral beam injection in the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Kaita, R.; White, R. B.; Morris, A. W.; Fredrickson, E. D.; McGuire, K. M.; Medley, S. S.; Murphy, T. J.; Scott, S. D.

    1990-07-01

    Coherent magnetohydrodynamic modes have been observed previously during neutral beam injection in the PDX tokamak [Phys. Rev. Lett. 50, 891 (1983)] and they have now been seen in the TFTR tokamak [Phys. Fluids 26, 2958 (1983)]. Periodic bursts of oscillations were detected with several plasma diagnostics, and Fokker-Planck calculations show that the populations of trapped particles in both tokamaks are sufficient to account for fishbone destabilization if a resonant interaction, between the mode and the beam ions, is assumed. Estimates of mode parameters are in reasonable agreement with the experiments, and they indicate that the fishbone mode may continue to affect the performance of intensely heated tokamaks.

  8. A comparative study of ideal kink stability in two reactor-relevant tokamak plasma configurations with negative and positive triangularity

    NASA Astrophysics Data System (ADS)

    Ren, Jing; Liu, Yueqiang; Liu, Yue; Medvedev, S. Yu; Wang, Zhirui; Xia, Guoliang

    2016-11-01

    The effects of an ideal/resistive conducting wall, the drift kinetic resonances, as well as the toroidal plasma flow, on the stability of the ideal external kink mode are numerically investigated for a reactor-relevant tokamak plasma with strongly negative triangularity (NTR) shaping. Comparison is made for a similar plasma equilibrium, but with positive triangularity (PTR). It is found that the ideal wall stabilization is less efficient for the kink stabilization in the NTR plasma due to a less ‘external’ eigenmode structure compared to the PTR plasma. The associated plasma displacement in the NTR plasma does not ‘balloon’ near the outboard mid-plane, as is normally the case for the pressure-driven kink-ballooning instability in PTR plasmas, but being more pronounced near the X-points. The toroidal flow plays a similar role for the kink stability for both NTR and PTR plasmas. The drift kinetic damping is less efficient for the ideal external kink mode in the NTR plasma, despite a somewhat larger fraction of the particle trapping near the plasma edge compared to the PTR equilibrium. However, the drift kinetic damping of the resistive wall mode (RWM) in the NTR plasma is generally as efficient as that of the PTR plasma, although the RWM window, in terms of the normalized pressure, is narrower for the NTR plasma.

  9. Dynamic behavior of plasma-facing materials during plasma instabilities in tokamak reactors

    SciTech Connect

    Hassanein, A.; Konkashbaev, I.

    1997-09-01

    Damage to plasma-facing and nearby components due to plasma instabilities remains a major obstacle to a successful tokamak concept. The high energy deposited on facing materials during plasma instabilities can cause severe erosion, plasma contamination, and structural failure of these components. Erosion damage can take various forms such as surface vaporization, spallation, and liquid ejection of metallic materials. Comprehensive thermodynamic and radiation hydrodynamic codes have been developed, integrated, and used to evaluate the extent of various damage to plasma-facing and nearby components. The eroded and splashed materials will be transported and then redeposited elsewhere on other plasma-facing components. Detailed physics of plasma/solid-liquid/vapor interaction in a strong magnetic field have been developed, optimized, and implemented in a self-consistent model. The plasma energy deposited in the evolving divertor debris is quickly and intensely reradiated, which may cause severe erosion and melting of other nearby components. Factors that influence and reduce vapor-shielding efficiency such as vapor diffusion and turbulence are also discussed and evaluated.

  10. Simulation studies on alpha-particle-driven current in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Tani, K.; Azumi, M.

    2008-08-01

    An asymmetry particle trapping caused by the finite banana width of alpha particles and a current produced by the resulting imbalance between transit particles with positive and negative velocities parallel to the magnetic field line (simply referred to as transit current) are numerically demonstrated by using an orbit-following Monte-Carlo code. The transit current is considerably enhanced by a collaborative effect of the finite-banana-width and a non-uniformity of the alpha particle source. The pitch-angle scattering is also very important for it. Banana-trapped particles drive a substantially negative current in the central region. Consequently, the positive transit current is degraded by the negative banana current and the net alpha current is almost flat near the plasma centre. The net alpha-driven total currents in an ITER-like normal aspect-ratio system (aspect ratio = 3.17) and in a low aspect-ratio system VECTOR (aspect ratio = 1.98) estimated with a conventional electron screening effect are about 0.4 MA and 1 MA, respectively. A low-aspect-ratio tokamak has an advantage over a normal or a large-aspect-ratio system in driving current by fusion-produced alpha particles.

  11. Designing a tokamak fusion reactor—How does plasma physics fit in?

    NASA Astrophysics Data System (ADS)

    Freidberg, J. P.; Mangiarotti, F. J.; Minervini, J.

    2015-07-01

    This paper attempts to bridge the gap between tokamak reactor design and plasma physics. The analysis demonstrates that the overall design of a tokamak fusion reactor is determined almost entirely by the constraints imposed by nuclear physics and fusion engineering. Virtually, no plasma physics is required to determine the main design parameters of a reactor: a , R 0 , B 0 , T i , T e , p , n , τ E , I . The one exception is the value of the toroidal current I , which depends upon a combination of engineering and plasma physics. This exception, however, ultimately has a major impact on the feasibility of an attractive tokamak reactor. The analysis shows that the engineering/nuclear physics design makes demands on the plasma physics that must be satisfied in order to generate power. These demands are substituted into the well-known operational constraints arising in tokamak physics: the Troyon limit, Greenwald limit, kink stability limit, and bootstrap fraction limit. Unfortunately, a tokamak reactor designed on the basis of standard engineering and nuclear physics constraints does not scale to a reactor. Too much current is required to achieve the necessary confinement time for ignition. The combination of achievable bootstrap current plus current drive is not sufficient to generate the current demanded by the engineering design. Several possible solutions are discussed in detail involving advances in plasma physics or engineering. The main contribution of the present work is to demonstrate that the basic reactor design and its plasma physics consequences can be determined simply and analytically. The analysis thus provides a crisp, compact, logical framework that will hopefully lead to improved physical intuition for connecting plasma physic to tokamak reactor design.

  12. Particle confinement property in the cusp-mirror field of a compact fusion reactor

    NASA Astrophysics Data System (ADS)

    Zhu, Limin; Liu, Haifeng; Wang, Xianqu

    2016-09-01

    The cusp-mirror magnetic structure in a compact fusion reactor (CFR) is investigated to understand the properties of the particle confinement for the first time. Compared with a cascade magnetic mirror device, its advanced performance is shown by means of test particle simulations. Some interesting results are obtained as follows: the adiabatic region and non-adiabatic region are found in the CFR’s magnetic configuration. In the non-adiabatic region, due to the magnetic field-free space existing, the ions are scattered stochastically and are not directly guided into the loss cone, unlike the particles around the fixed magnetic lines in the adiabatic region, which decrease the ion loss fraction. The CFR’s configuration, combining advantages of cusp-magnetic configuration and mirror-magnetic configuration, leads to confine particles longer than cascade magnetic mirror’s. This phenomenon may be relevant to the construction of advanced magnetic-confinement devices.

  13. Two-dimensional TBR calculations for conceptual compact reversed-field pinch reactor blanket

    NASA Astrophysics Data System (ADS)

    Davidson, J. W.; Battat, M. E.; Dudziak, D. J.

    A detailed two-dimensional nucleonic analysis was performed for a conceptual first wall, blanket, and shield design for the Compact Reversed-Field Pinch Reactor. The design includes significant two-dimensional aspects presented by the limiter, vacuum ducts, and coolant manifolds; these aspects seriously degrade the tritium-breeding reaction (TBR) predicted by one-dimensional calculations. A range of design change to increase the TBR were investigated within the two-dimensional analysis. The results of this investigation indicated that an adequate TBR could be achieved with a thinning copper first wall, a (6)Li enrichment near 90%, the proper selection of reflector, and a small addition to the blanket thickness, determined by the one-dimensional analysis.

  14. Anomalous loss of DT alpha particles in the Tokamak Fusion Test Reactor

    SciTech Connect

    Herrmann, H.W.

    1997-09-01

    An escaping alpha collector probe has been developed for TFTR`s DT phase. Energy distributions of escaping alphas have been determined by measuring the range of {alpha}-particles implanted into nickel foils located within the alpha collector. Results at 1.0 MA of plasma current are in good agreement with predictions for first orbit alpha loss. Results at 1.8 MA, however, show a significant anomalous loss of partially thermalized alphas (in addition to the expected first orbit loss), which is not observed with the lost alpha scintillator detectors in DT plasmas, but does resemble the anomalous delayed loss seen in DD plasmas. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations. An experiment designed to study the effect of plasma major radius shifts on {alpha}-particle loss has led to a better understanding of {alpha}-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing {alpha}-particles forced to move toward higher magnetic field during an inward major radius shift (i.e., compression) would mirror and become trapped particles, leading to increased alpha loss. Such an effect was looked for during the shift experiment, however, no significant changes in alpha loss to the 90{degree} lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an {alpha}-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized {alpha}-particles near the passing/trapped boundary was observed to occur between inward and outward shifts at an intermediate value of plasma current (1.4 MA). This anomalous loss feature is not yet understood.

  15. Tokamak ARC damage

    SciTech Connect

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  16. Real-time loop-mediated isothermal DNA amplification in compact disc micro-reactors.

    PubMed

    Santiago-Felipe, Sara; Tortajada-Genaro, Luis A; Carrascosa, Javier; Puchades, Rosa; Maquieira, Ángel

    2016-05-15

    An integrated device composed of micro-reactors embedded onto compact discs is proposed for real-time targeted DNA determination. The method principle is based on in-disc loop-mediated isothermal amplification (iD-LAMP) and quantitative optical read-out by a disc drive. In the presence of a target, the turbidimetric or colorimetric properties of reaction solution change, and the transmitted intensity of the disc drive laser modifies according to reaction yield. Monitoring real-time curves allowed the quantitative determination of DNA template amounts. The best amplification/detection results were obtained with micro-reactors (2mm diameter and 1.1mm in depth) drilled on a digital video disc (DVD) and detection based on the colorimetric mode. As proof-of-concept, the assay was applied to detect pathogenic bacteria Salmonella spp. and to identify bovine meat in food samples. Ninety-six samples were simultaneously analysed in 15 min, with high selectivity and sensitivity (5 CFU/mL and 10 µg/g for bacteria and meat, respectively). The in-disc results were comparable to those obtained by conventional LAMP or qPCR approaches. The developed device allows low sample and reagent consumption (3 µL of reaction), portability, ease-of-use, and rapid low-cost high-throughput analyses. PMID:26716424

  17. Liquid immersion blanket design for use in a compact modular fusion reactor

    NASA Astrophysics Data System (ADS)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  18. Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

    SciTech Connect

    Bartel, N.; Chen, M.; Utgikar, V. P.; Sun, X.; Kim, I. -H.; Christensen, R.; Sabharwall, P.

    2015-04-04

    A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimum combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.

  19. Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

    DOE PAGES

    Bartel, N.; Chen, M.; Utgikar, V. P.; Sun, X.; Kim, I. -H.; Christensen, R.; Sabharwall, P.

    2015-04-04

    A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

  20. Compact Stellarator Path to DEMO

    NASA Astrophysics Data System (ADS)

    Lyon, J. F.

    2007-11-01

    Issues for a DEMO reactor are sustaining an ignited/high-Q plasma in steady state, avoiding disruptions and large variations in power flux to the wall, adequate confinement of thermal plasma and alpha-particles, control of a burning plasma, particle and power handling, etc. Compact stellarators have key advantages -- steady-state high-plasma-density operation without external current drive or disruptions, stability without a close conducting wall or active feedback systems, and low recirculating power -- in addition to moderate plasma aspect ratio, good confinement, and high-beta potential. The ARIES-CS study established that compact stellarators can be competitive with tokamaks as reactors. Many of the issues for a compact stellarator DEMO can be answered using results from large tokamaks, ITER D-T experiments and fusion materials, technology and component development programs, in addition to stellarators in operation, under construction or in development. However, a large next-generation stellarator will be needed to address some physics issues: size scaling and confinement at higher parameters, burning plasma issues, and operation with a strongly radiative divertor. Technology issues include simpler coils, structure, and divertor fabrication, and better cost information.

  1. High-power-density approaches to magnetic fusion energy: Problems and promise of compact reversed-field pinch reactors (CRFPR)

    NASA Astrophysics Data System (ADS)

    Hagenson, Randy L.; Krakowski, Robert A.; Dreicer, Harry

    1983-03-01

    If the costing assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves overly optimistic, approaches that promise considerably increased system power density and reduced mass utilization will be required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique magnetic topology associated with the Reversed-Field Pinch (RFP), the compact reactor embodiment for this approach is particularly attractive from the viewpoint of low-field resistive coils operating with ohmic losses that can be made small relative to the fusion power. The RFP, therefore, is used as one example of a high-power-density (HPD) approach to magnetic fusion energy. A comprehensive system model is described and applied to select a unique, cost-optimized design point that will be used for a subsequent conceptual engineering design of the compact RFP Reactor (CRFPR). This cost-optimized CRFPR design serves as an example of a HPD fusion reactor that would operate with system power densities and mass utilizations that are comparable to fission power plants, these measures of system performance being an order of magnitude more favorable than the conventional approaches to magnetic fusion energy (MFE).

  2. Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology

  3. A study of hydrogenic retention in a tokamak with reactor-like plasma-facing surfaces; Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lipschultz, Bruce

    2008-11-01

    Tritium retention is an important safety concern for ITER; Operation for 1000 discharges without a major stoppage will require the fraction of ion fluence to Plasma Facing Components (PFCs) that is retained, R, to be < 0.001%. One year operation of a reactor, where tungsten (W) PFCs are envisioned, requires R to be 100x smaller! Co-deposition of H with carbon projects to unacceptably high T retention in ITER. We present the results of the first in-depth study of fuel retention for high-Z PFCs with ITER divertor ne, Te, particle and heat fluxes. We utilize molybdenum (Mo, with a small fraction of W), which is very similar to tungsten in terms of hydrogenic retention. The retention observed in a series of disruption-free C-Mod discharges is high, R˜1%, 1000x than expected from inherent Mo properties. These retention characteristics are exhibited regardless if the Mo surfaces are bare or partially covered by B films; D co-deposition with B is not contributing significantly to retention. Retention appears linear in fluence up to the limit of the discharge sequence, ˜20s, approaching one ITER discharge. Comparison of He- and D-fueled discharges gives support to a model of retention site creation in the lattice (`traps') due to D neutral buildup and accompanying lattice distortion driven by recombination-limited release (D->D2) from the front surface. Disruptions can be used to rapidly heat surfaces, releasing the H/D for recovery, potentially applicable to ITER. Naturally-occurring disruptions appear to balance single-discharge retention reducing the campaign-integrated retention by at least 100. Comparisons to laboratory-based retention studies indicate that the tokamak environment leads to additional enhancements of retention. This work is supported by U.S. Dept. of Energy Coop. Agreement DE-FC02-99ER54512.

  4. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  5. Advanced commercial tokamak study

    SciTech Connect

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs.

  6. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  7. Engineering design of a liquid metal cooled self-pumped limiter for a tokamak reactor

    SciTech Connect

    Brooks, J.N.; Cha, Y.; Hassanein, A.; Majumdar, S.; Mattas, R.F.; Smith, D.L.

    1987-10-01

    A lithium cooled self-pumped limiter has been designed as the impurity control system for the TPSS high-..beta.. power reactor conceptual design. The limiter removes helium by trapping impinging helium ions in freshly deposited vanadium surface layers in a slot region. No hydrogen is removed and no pumps or vacuum penetrations are used, thereby eliminating penetration shielding and reducing tritium handling. The limiter is composed of a vanadium alloy structure with a 2mm tungsten cladding on the front face and leading edges for sputtering control. Up to approx.3cm of vanadium trapping material is deposited in the slot region during 5 years of operation. A key design feature is the use of a calcium oxide electrical insulator which coats the limiter coolant channels to reduce MHD pressure drops. A combination of high lithium coolant velocity, made possible by the insulator, and mid-limiter manifolding has been used to obtain acceptable material temperatures with moderately high heat fluxes (3 to 5 mw/m/sup 2/). Overall, a limiter lifetime of approx.5 years is predicted by stress and lifetime analysis. This would permit maintenance free impurity control operation between first wall/blanket replacement periods. 4 refs., 3 figs., 1 tab.

  8. Progress in Compact Toroid Experiments

    SciTech Connect

    Dolan, Thomas James

    2002-09-01

    The term "compact toroids" as used here means spherical tokamaks, spheromaks, and field reversed configurations, but not reversed field pinches. There are about 17 compact toroid experiments under construction or operating, with approximate parameters listed in Table 1.

  9. Impact of magnetic field inhomogeneity on electron cyclotron radiative loss in tokamak reactors

    SciTech Connect

    Kukushkin, A. B.; Minashin, P. V.; Polevoi, A. R.

    2012-03-15

    The potential importance of electron cyclotron (EC) emission in the local electron power balance in the steady-state regimes of ITER operation with high temperatures, as well as in the DEMO reactor, requires accurate calculation of the one-dimensional (over magnetic surfaces) distribution of the net radiated power density, P{sub EC}({rho}). When the central electron temperature increases to {approx}30 keV, the local EC radiative loss comprises a substantial fraction of the heating power from fusion alphas and is close to the total auxiliary NBI heating power, P{sub EC}(0) Asymptotically-Equal-To 0.3P{sub {alpha}}(0) Asymptotically-Equal-To P{sub aux}(0). In the present paper, the model of EC radiative transport in an axisymmetric toroidal plasma is extended to the case of an inhomogeneous magnetic field B(R, Z). The impact of such inhomogeneity on local and total power losses is analyzed in the framework of this model by using the CYNEQ code. It is shown that, for the magnetic field B, temperature T{sub e}, density n{sub e}, and wall reflection coefficient R{sub w} expected in ITER and DEMO, accurate simulations of the EC radiative loss require self-consistent 1.5D transport analysis (i.e., one-dimensional simulations of plasma transport and two-dimensional simulations of plasma equilibrium). It is shown that EC radiative transport can be described with good accuracy in the 1D approximation with the surface-averaged magnetic field, B({rho}) = Left-Pointing-Angle-Bracket B(R, Z) Right-Pointing-Angle-Bracket {sub ms}. This makes it possible to substantially reduce the computational time required for time-dependent self-consistent 1.5D transport analysis. Benchmarking of the CYNEQ results with available results of the RAYTEC, EXACTEC, and CYTRAN codes is performed for various approximations of the magnetic field.

  10. QUANTITATIVE HOMOGENEITY AND IN-CONTACT PARTICLES OF HIGH TEMPERATURE REACTORS (HTR) COMPACTS DETERMINATION VIA X-RAY TOMOGRAPHY

    SciTech Connect

    Lecomte, G.; Letang, J. M.; Tisseur, D.; Banchet, J.; Vitali, M. P.

    2008-02-28

    In AREVA Nuclear Power's High Temperature Reactor (HTR) design called ANTARES, fuel consists of compacts composed of few thousands millimetric quasi-spherical particles dispersed in a graphite matrix. Compact homogeneity, defined as the homogeneous particles spatial distribution in the matrix, as well as the possibility of obtaining particles in contact, need to be assessed since they condition the thermo-mechanical behavior of the nuclear fuel under irradiation. In this paper, image and data processing algorithms are developed to do so, based on X-Ray tomographic images.

  11. An interim report on the materials and selection criteria analysis for the Compact Ignition Tokamak Toroidal Field Coil Turn-to-Turn Insulation System

    SciTech Connect

    Campbell, V.W.; Dooley, J.B.; Hubrig, J.G.; Janke, C.J.; McManamy, T.J.; Welch, D.E.

    1990-01-01

    Design criteria for the Compact Ignition Tokamak, Toroidal-Field (TF) Coil, Turn-to-Turn Insulation System require an insulation sheet and bonding system that will survive cryogenic cycling in a radiation environment and maintain structural integrity during exposure to the significant compressive and shear loads associated with each operating cycle. For thermosetting resin systems, a complex interactive dependency exists between optimum peak value, in-service property performance capabilities of candidate generic materials; key handling and processing parameters required to achieve their optimum in-service property performance as an insulation system; and suitability of their handling and processing parameters as a function of design configuration and assembly methodology. This dependency is assessed in a weighted study matrix in which two principal programmatic approaches for the development of the TF Coil Subassembly Insulation System have been identified. From this matrix study, two viable approaches to the fabrication of the insulation sheet were identified: use of a press-formed sheet bonded in place with epoxy for mechanical bonding and tolerance take-up and formation of the insulation sheet by placement of dry cloth and subsequent vacuum pressure impregnation. Laboratory testing was conducted to screen a number of combinations of resins and hardeners on a generic basis. These combinations were chosen for their performance in similar applications. Specimens were tested to screen viscosity, thermal-shock tolerance, and cryogenic tolerance. Cryogenic shock and cryogenic temperature proved to be extremely lethal to many combinations of resin, hardener, and cure. Two combinations survived: a heavily flexibilized bisphenol A resin with a flexibilized amine hardener and a bisphenol A resin with cycloaliphatic amine hardener. 7 refs., 12 figs., 6 tabs.

  12. Assessment of torsatrons as reactors

    SciTech Connect

    Lyon, J.F. ); Painter, S.L. )

    1992-12-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

  13. ADX - Advanced Divertor and RF Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  14. Bootstrapped tokamak with oscillating field current drive

    SciTech Connect

    Weening, R.H. )

    1993-07-01

    A magnetic helicity conserving mean-field Ohm's law is used to study bootstrapped tokamaks with oscillating field current drive. The Ohm's law leads to the conclusion that the tokamak bootstrap effect can convert the largely alternating current of oscillating field current drive into a direct toroidal plasma current. This plasma current rectification is due to the intrinsically nonlinear nature of the tokamak bootstrap effect, and suggests that it may be possible to maintain the toroidal current of a tokamak reactor by supplementing the bootstrap current with oscillating field current drive. Steady-state tokamak fusion reactors operating with oscillating field current drive could provide an alternative to tokamak reactors operating with external current drive.

  15. Three-dimensional analysis of tokamaks and stellarators

    PubMed Central

    Garabedian, Paul R.

    2008-01-01

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

  16. Tritium catalyzed deuterium tokamaks

    SciTech Connect

    Greenspan, E.; Miley, G.H.; Jung, J.; Gilligan, J.

    1984-04-01

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the /sup 3/He from the D(D,n)/sup 3/He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general).

  17. A physics study for negative void reactivity in compact supercritical CO{sub 2}-cooled fast reactor

    SciTech Connect

    Kim, Y.; Hartanto, D.; Lee, J. I.

    2013-07-01

    A compact S-CO{sub 2}-cooled fast reactor which has negative Coolant Void Reactivity (CVR) has been investigated. A negative CVR is important for the gas cooled fast reactor as an inherent safety mechanism to prevent the sudden positive reactivity insertion when the loss of coolant accident happens. An alternative solution to reduce the CVR is investigated in this study by using O-17 instead of O-16 in UO{sub 2} fuel. By using O-17 in the fuel, it is found that the CVR can even be negative. Impacts of the radial reflector on the CVR are also evaluated for the small SCO{sub 2} cooled fast reactor in this study. We have considered a pure lead (Pb) reflector and a lead magnesium eutectic (LME) reflector as alternative radial reflectors of the S-CO 2-cooled fast reactor. It has been shown that, with the LME radial reflector, the CVR can be negative, while the pure lead reflector provides a slightly positive CVR. (authors)

  18. Preliminary Design For Conventional and Compact Secondary Heat Exchanger in a Molten Salt Reactor

    SciTech Connect

    Piyush Sabharwall; Mike Patterson; Ali Siahpush; Eung Soo Kim

    2012-07-01

    The strategic goal of the Advance Reactors such as AHTR is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors

  19. Toroidal rotation and radial electric field driven by the lower-hybrid-wave in a tokamak fusion reactor

    SciTech Connect

    Wang Shaojie

    2011-10-15

    A theoretical model is proposed to interpret the counter-current rotation driven by the lower-hybrid-wave observed in the tokamak lower-hybrid-wave parallel current drive experiments. It is found that ions absorb the toroidal momentum indirectly from the wave through collisional friction with the resonant electrons that directly take the momentum from the wave through Landau resonance. This momentum coupling pumps out the ions to produce a negative radial electric field and makes the plasma rotate in the counter-current direction.

  20. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    NASA Astrophysics Data System (ADS)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  1. Calculation of the absolute detection efficiency of a moderated /sup 235/U neutron detector on the Tokamak Fusion Test Reactor

    SciTech Connect

    Ku, L.P.; Hendel, H.W.; Liew, S.L.

    1989-02-01

    Neutron transport simulations have been carried out to calculate the absolute detection efficiency of a moderated /sup 235/U neutron detector which is used on the TFTR as a part of the primary fission detector diagnostic system for measuring fusion power yields. Transport simulations provide a means by which the effects of variations in various shielding and geometrical parameters can be explored. These effects are difficult to study in calibration experiments. The calculational model, benchmarked against measurements, can be used to complement future detector calibrations, when the high level of radioactivity resulting from machine operation may severely restrict access to the tokamak. We present a coupled forward-adjoint algorithm, employing both the deterministic and Monte Carlo sampling methods, to model the neutron transport in the complex tokamak and detector geometries. Sensitivities of the detector response to the major and minor radii, and angular anisotropy of the neutron emission are discussed. A semi-empirical model based on matching the calculational results with a small set of experiments produces good agreement (+-15%) for a wide range of source energies and geometries. 20 refs., 6 figs., 4 tabs.

  2. An evaluation of fusion gain in the compact helical fusion reactor FFHR-c1

    NASA Astrophysics Data System (ADS)

    Miyazawa, J.; Goto, T.; Sakamoto, R.; Sagara, A.; the FFHR Design Group

    2014-01-01

    A new procedure to predict achievable fusion gain in a sub-ignition fusion reactor is proposed. This procedure uses the direct profile extrapolation (DPE) method based on the gyro-Bohm model. The DPE method has been developed to predict the radial profiles in a fusion reactor sustained without auxiliary heating (i.e., in the self-ignition state) from the experimental data. To evaluate the fusion gain in a fusion reactor sustained with auxiliary heating (i.e., in the sub-ignition state), the DPE method is modified to include the influence of the auxiliary heating. The beta scale factor from experiment to reactor is assumed to be 1. Under this assumption, it becomes reasonable to apply the magnetohydrodynamic (MHD) equilibrium (which is calculated to reproduce the experimental data) to the reactor. At the same time, the MHD stability of the reactor plasma is also guaranteed to a certain extent since that beta was already proven in the experiment. The fusion gain in the helical type nuclear test machine FFHR-c1 has been evaluated using this modified DPE method. FFHR-c1 is basically a large duplication of the Large Helical Device (LHD) with a scale factor of 10/3, which corresponds to the major radius of the helical coils of 13.0 m and the plasma volume of ∼1000 m3. Two options with different magnetic field strengths are considered. The fusion gain in FFHR-c1 extrapolated from a set of radial profile data obtained in LHD ranges from 1 to 7, depending on the profiles used together with the assumptions of the magnetic field strength and the alpha heating efficiency.

  3. Evaluation of Torsatrons as reactors

    SciTech Connect

    Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

    1994-03-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

  4. Design and optimization of Artificial Neural Networks for the modelling of superconducting magnets operation in tokamak fusion reactors

    NASA Astrophysics Data System (ADS)

    Froio, A.; Bonifetto, R.; Carli, S.; Quartararo, A.; Savoldi, L.; Zanino, R.

    2016-09-01

    In superconducting tokamaks, the cryoplant provides the helium needed to cool different clients, among which by far the most important one is the superconducting magnet system. The evaluation of the transient heat load from the magnets to the cryoplant is fundamental for the design of the latter and the assessment of suitable strategies to smooth the heat load pulses, induced by the intrinsically pulsed plasma scenarios characteristic of today's tokamaks, is crucial for both suitable sizing and stable operation of the cryoplant. For that evaluation, accurate but expensive system-level models, as implemented in e.g. the validated state-of-the-art 4C code, were developed in the past, including both the magnets and the respective external cryogenic cooling circuits. Here we show how these models can be successfully substituted with cheaper ones, where the magnets are described by suitably trained Artificial Neural Networks (ANNs) for the evaluation of the heat load to the cryoplant. First, two simplified thermal-hydraulic models for an ITER Toroidal Field (TF) magnet and for the ITER Central Solenoid (CS) are developed, based on ANNs, and a detailed analysis of the chosen networks' topology and parameters is presented and discussed. The ANNs are then inserted into the 4C model of the ITER TF and CS cooling circuits, which also includes active controls to achieve a smoothing of the variation of the heat load to the cryoplant. The training of the ANNs is achieved using the results of full 4C simulations (including detailed models of the magnets) for conventional sigmoid-like waveforms of the drivers and the predictive capabilities of the ANN-based models in the case of actual ITER operating scenarios are demonstrated by comparison with the results of full 4C runs, both with and without active smoothing, in terms of both accuracy and computational time. Exploiting the low computational effort requested by the ANN-based models, a demonstrative optimization study has been

  5. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    NASA Astrophysics Data System (ADS)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  6. Energy losses on tokamak startup

    SciTech Connect

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells.

  7. Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

    DOE PAGES

    Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; Roney, T. J.; Morrell, S. R.

    2016-02-05

    In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

  8. A computer model for the transient analysis of compact research reactors with plate type fuel

    SciTech Connect

    Sofu, T.; Dodds, H.L.

    1994-03-01

    A coupled neutronics and core thermal-hydraulic performance model has been developed for the analysis of plate type U-Al fueled high-flux research reactor transients. The model includes point neutron kinetics, one-dimensional, non-homogeneous, equilibrium two-phase flow and beat transfer with provision for subcooled boiling, and spatially averaged one-dimensional beat conduction. The feedback from core regions other than the fuel elements is included by employing a lumped parameter approach. Partial differential equations are discretized in space and the combined equation set representing the model is converted to an initial value problem. A variable-order, variable-time-step time advancement scheme is used to solve these ordinary differential equations. The model is verified through comparisons with two other computer code results and partially validated against SPERT-II tests. It is also used to analyze a series of HFIR reactivity transients.

  9. R&D around a photoneutralizer-based NBI system (Siphore) in view of a DEMO Tokamak steady state fusion reactor

    NASA Astrophysics Data System (ADS)

    Simonin, A.; Achard, Jocelyn; Achkasov, K.; Bechu, S.; Baudouin, C.; Baulaigue, O.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; de Esch, H. P. L.; Fiorucci, D.; Fubiani, G.; Furno, I.; Futtersack, R.; Garibaldi, P.; Gicquel, A.; Grand, C.; Guittienne, Ph.; Hagelaar, G.; Howling, A.; Jacquier, R.; Kirkpatrick, M. J.; Lemoine, D.; Lepetit, B.; Minea, T.; Odic, E.; Revel, A.; Soliman, B. A.; Teste, P.

    2015-11-01

    Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η  >  60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (~3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

  10. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor.

    PubMed

    Singh, M J; De Esch, H P L

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

  11. Shielding analysis of a small compact space nuclear reactor. Final report

    SciTech Connect

    Woodrow, L.L.

    1987-08-01

    The SP-100 reactor concept, currently in its developmental stage, has layered tungsten - lithium hydride shield. Studies indicate that this shield configuration is the lightest weight shield. This configuration and three other shielding concepts were analyzed to determine the lightest shield and to determine the shield configuration with the smallest volume. The other three concepts were a boron carbide - beryllium layered shield, and a lithium hydride - beryllium shield. FEMP2D and FEMP1D codes were used in this analysis. These codes were developed at Sandia National Laboratory (SNL), using the input from another code, RFCC, which produced energy-dependent dose conversion factors, and determined the shield's ability to attenuate the neutron and gamma radiation to permissible dose limits. The results of this analysis show that the lithium hydride - tungsten layered shield was indeed the lightest weight shield. However, a boron carbide - tungsten shield was calculably volume constrained. Therefore volume, not weight, may be the driving factor in determining the shield configuration.

  12. The need and prospects for improved fusion reactors

    NASA Astrophysics Data System (ADS)

    Krakowski, R. A.; Miller, R. L.; Hagenson, R. L.

    1986-09-01

    Conceptual fusion reactor studies over the past 10-15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100-200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.

  13. Pyrolucite fluidized-bed reactor (PFBR): a robust and compact process for removing manganese from groundwater.

    PubMed

    Dashtban Kenari, Seyedeh Laleh; Barbeau, Benoit

    2014-02-01

    The purpose of this paper is to introduce a pyrolucite fluidized-bed reactor (PFBR) as a potential drinking water process to treat groundwater containing high levels of dissolved manganese (Mn(II)) (0.5-3 mg/L) and reduce its concentration to <0.02 mg/L in treated water. A pilot-scale study was conducted under dynamic conditions using synthetic groundwater (SGW), to elucidate the effect of operational conditions and groundwater composition on manganese (Mn) removal achieved by the PFBR. Results demonstrated almost complete Mn removal (close to 100%) in less than 1 min under all tested operational conditions (influent Mn concentration of 0.5-3 mg/L, calcium (Ca(2+)) hardness of 0-200 mg CaCO3/L, pH of 6.2-7.8, temperature of 9 & 23 °C and high hydraulic loading rate (HLR) of 24-63 m/h (i.e., bed expansion of 0-30%)). Improved Mn removal profile was achieved at higher water temperature. Also, the results showed that adsorption of Mn(II) onto pyrolucite and subsequent slower surface oxidation of sorbed Mn(II) was the only mechanism responsible for Mn removal while direct oxidation of Mn(II) by free chlorine did not occur even at high concentrations of Mn(II) and free chlorine and elevated temperatures. Higher average mass transfer coefficient and consequently adsorption rate was achieved at elevated HLR. Increasing effluent free chlorine residuals from 1.0 to 2.0-2.6 mg Cl2/L allowed increasing the operation time needed for media regeneration from 6 days to >12 days. Turbidity was maintained around 0.2 NTU during the entire test periods indicating good capture of MnOx colloids within the PFBR. PMID:24183400

  14. Simulation study of power load with impurity seeding in advanced divertor "short super-X divertor" for a tokamak reactor

    NASA Astrophysics Data System (ADS)

    Asakura, N.; Hoshino, K.; Shimizu, K.; Shinya, K.; Utoh, H.; Tokunaga, S.; Tobita, K.; Ohno, N.

    2015-08-01

    A short super-X divertor (SXD) is proposed as an option for the Demo divertor, where the field line length from the divertor null to the outer target was largely increased compared to a similar-size conventional divertor. Physics and engineering design studies for a 3 GW-level fusion power Demo reactor (SlimCS) (Tobita et al., 2009) have recently progressed. Minimal number of the divertor coils were installed inside the toroidal field coil, i.e. interlink-winding. Arrangement of the poloidal field coils and their currents were determined, taking into account of the engineering design such as vacuum vessel and the neutron shield structures, and the divertor maintenance scenario. Divertor plasma simulation showed that significant radiation region is produced between the super-X null and the target. Radiation loss in the divertor was increased, producing fully detached plasmas efficiently. Advantages of the short SXD were demonstrated, but the total peak heat load was a marginal level (10 MW m-2) for the engineering design.

  15. Tokamak power systems studies, FY 1985

    SciTech Connect

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  16. Driven-current tokamak (DCT) scoping study

    SciTech Connect

    Reid, R.L.

    1983-01-01

    The present Department of Energy (DOE) plan calls for the construction of an Engineering Test Reactor (ETR) that is to be the last major experimental fusion device prior to the commercialization of fusion power. The plasma driver of the ETR is to be either a long-pulse tokamak or a tandem mirror machine. The possibility of using the Tokamak Fusion Test Reactor (TFTR) facility to consolidate the physics and technology database for the tokamak version of the ETR has been considered. This paper addresses two of the options being considered: (1) a superconducting toroidal field (TF) coil-hydrogen plasma alternative, and (2) a superconducting or hybrid TF coil-high Q alternative. Both options assume essentially steady-state operation through the application of rf current drive. The options are evaluated on the basis of performance and cost determined by application of the Fusion Engineering Design Center (FEDC) Tokamak System Code.

  17. Superconducting magnet system for the TPX Tokamak

    SciTech Connect

    Hassenzahl, W.V.; Chaplin, M.R.; Heim, J.R.

    1993-09-15

    The Tokamak Physics Experiment (TPX) will be the first Tokamak using superconducting magnets for both the poloidal and toroidal field. It is designed for advanced Tokamak physics experiments in steady-state and long-pulse operation. The TPX superconducting magnets use an advanced cable-in-conduit conductor (CICC) design similar to that developed in support of the International Thermonuclear Experimental Reactor (ITER). The toroidal field magnets provide 4.0 T at 2.25 m with a stored energy of 1.05 GJ. The poloidal field magnets provide 18.0 V-s to ohmically start and control long burns of a 2.0 MA plasma.

  18. Tokamak pump limiters

    NASA Astrophysics Data System (ADS)

    Conn, Robert W.

    1984-12-01

    edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this "Z-mode" of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described.

  19. Control of Dust Inventory in Tokamaks

    SciTech Connect

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Pitcher, C. S.; Taylor, N.; Furlan, J.

    2008-09-07

    Particles with sizes ranging from 100 nm to 100 {mu}m are produced in tokamaks by the interaction of the plasma with the first wall materials and divertor. Dust has not yet been of a major concern in existing tokamaks mainly because their quantities are small and these devices are not nuclear facilities. However, in ITER and in future reactors, they could represent operational and potential safety issues. The aim of this paper is thus to describe the dust creation processes in the tokamak environment. The diagnostics and removal techniques that are needed to be implemented to measure and minimise the dust inventory are also presented. The integration of these techniques into a tokamak environment is also discussed.

  20. OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS

    SciTech Connect

    LIN-LIU,YR; STAMBAUGH,RD

    2002-11-01

    OAK A271 OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS. The dependence of the ideal ballooning {beta} limit on aspect ratio, A, and elongation {kappa} is systematically explored for nearly 100% bootstrap current driven tokamak equilibria in a wide range of the shape parameters (A = 1.2-7.0, {kappa} = 1.5-6.0 with triangularity {delta} = 0.5). The critical {beta}{sub N} is shown to be optimal at {kappa} = 3.0-4.0 for all A studied and increases as A decreases with a dependence close to A{sup -0.5}. The results obtained can be used as a theoretical basis for the choice of optimum aspect ratio and elongation of next step burning plasma tokamaks or tokamak reactors.

  1. Advanced fusion MHD power conversion using the CFAR (compact fusion advanced Rankine) cycle concept

    SciTech Connect

    Hoffman, M.A.; Campbell, R.; Logan, B.G.; Lawrence Livermore National Lab., CA )

    1988-10-01

    The CFAR (compact fusion advanced Rankine) cycle concept for a tokamak reactor involves the use of a high-temperature Rankine cycle in combination with microwave superheaters and nonequilibrium MHD disk generators to obtain a compact, low-capital-cost power conversion system which fits almost entirely within the reactor vault. The significant savings in the balance-of-plant costs are expected to result in much lower costs of electricity than previous concepts. This paper describes the unique features of the CFAR cycle and a high- temperature blanket designed to take advantage of it as well as the predicted performance of the MHD disk generators using mercury seeded with cesium. 40 refs., 8 figs., 3 tabs.

  2. Hybrid Fusion: The Only Viable Development Path for Tokamaks?

    NASA Astrophysics Data System (ADS)

    Manheimer, Wallace

    2009-03-01

    The world needs a great deal of carbon free energy, and soon, for civilization to continue. Fusion's goal is to develop such a carbon free energy source. For the last 4 decades, tokamaks have been the best magnetic fusion has to offer. But what if its development stops short of commercial fusion? This paper introduces `conservative design principles' for tokamaks. These are very simple, are reasonably based in theory, and have always constrained tokamak operation. Assuming they continue to do so, it is unlikely that tokamaks will ever make it as commercial reactors. This is independent of their confinement properties. However because of the large additional gain in hybrid fusion, tokamaks reactors look like they can make it as hybrid fuel producers, and provide large scale power by mid century or shortly thereafter.

  3. Ripple induced trapped particle loss in tokamaks

    SciTech Connect

    White, R.B.

    1996-05-01

    The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks.

  4. Ideal magnetohydrodynamic simulations of low beta compact toroid injection into a hot strongly magnetized plasma

    SciTech Connect

    Liu, Wei; Hsu, Scott; Li, Hui

    2009-01-01

    We present results from three-dimensional ideal magnetohydrodynamic simulations of low {beta} compact toroid (CT) injection into a hot strongly magnetized plasma, with the aim of providing insight into CT fueling of a tokamak with parameters relevant for ITER (International Thermonuclear Experimental Reactor). A regime is identified in terms of CT injection speed and CT-to-background magnetic field ratio that appears promising for precise core fueling. Shock-dominated regimes, which are probably unfavorable for tokamak fueling, are also identified. The CT penetration depth is proportional to the CT injection speed and density. The entire CT evolution can be divided into three stages: (1) initial penetration, (2) compression in the direction of propagation and reconnection, and (3) coming to rest and spreading in the direction perpendicular to injection. Tilting of the CT is not observed due to the fast transit time of the CT across the background plasma.

  5. Computed tomography of Spheromak-like compact toroid injection

    NASA Astrophysics Data System (ADS)

    Terry, Stephen Donald

    1999-12-01

    The construction and installation of a tomographic imaging system on the Compact Toroid Injection Experiment is described. This system is used to examine the physical processes which are in effect during the penetration of a vacuum magnetic field by a Spheromak-like Compact Toroid (SCT). Compact toroid fueling is an alternative fueling concept to pellet injection. It has the potential to provide central fueling at high repetition rates while minimizing injector size. Several physics issues need to be settled to demonstrate the practicality of this fueling system for fusion power reactors. One of these issues is the interaction between the SCT and the tokamak magnetic field. Several processes which can occur during this interaction are presented. They include compression and heating. Some possible mechanisms for these processes are hypothesized. The Compact Toroid Injection Experiment (CTIX) at the Davis Diverted Tokamak (DDT) is an experiment to study the physics of SCT fueling. The diagnostics available on the machine are described. The theories of SCT formation and accelleration are presented and the operation of the injector is described. A tomographic imaging system has been built and installed in the drift tube section of the injector. This system measures the total radiated power from the SCT. The data is then used to construct a two dimensional plasma emissivity profile. This is done with a software routine based on the Second Order Regularization method. The performance of this algorithm is compared to other reconstruction algorithms and the calibration of the system is described. The results of several measurements of SCT parameters are described. Radial compression of the SCT in the tokamak magnetic field is indicated by the absence of magnetic probe signal for high toroidal field strengths. Heating of the electrons is shown by obtaining the electron energy distribution function. An enhancement of SCT radiation in the tokamak magnetic field has been

  6. Compact stellarators with modular coils

    PubMed Central

    Garabedian, P. R.

    2000-01-01

    Compact stellarator designs with modular coils and only two or three field periods are now available; these designs have both good stability and quasiaxial symmetry providing adequate transport for a magnetic fusion reactor. If the bootstrap current assumes theoretically predicted values a three field period configuration is optimal, but if that net current turns out to be lower, a device with two periods and just 12 modular coils might be better. There are also attractive designs with quasihelical symmetry and four or five periods whose properties depend less on the bootstrap current. Good performance requires that there be a satisfactory magnetic well in the vacuum field, which is a property lacking in a stellarator-tokamak hybrid that has been proposed for a proof of principle experiment. In this paper, we present an analysis of stability for these configurations that is based on a mountain pass theorem asserting that, if two solutions of the problem of magnetohydrodynamic equilibrium can be found, then there has to be an unstable solution. We compare results of our theory of equilibrium, stability, and transport with recently announced measurements from the large LHD experiment in Japan. PMID:10899993

  7. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  8. Tokamak Systems Code

    SciTech Connect

    Reid, R.L.; Barrett, R.J.; Brown, T.G.; Gorker, G.E.; Hooper, R.J.; Kalsi, S.S.; Metzler, D.H.; Peng, Y.K.M.; Roth, K.E.; Spampinato, P.T.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged.

  9. Acceleration of compact toroid plasma rings for fusion applications

    NASA Astrophysics Data System (ADS)

    Hartman, C. W.; Barr, W. L.; Eddleman, J. L.; Gee, M.; Hammer, J. H.; Ho, S. K.; Logan, B. G.; Meeker, D. J.; Mirin, A. A.; Nevins, W. M.

    1988-08-01

    We describe experimental results for a new type of collective accelerator based on magnetically confined compact torus (CT) plasma rings and discuss applications to both inertial and magnetic fusion. We have demonstrated the principle of CT acceleration in the RACE device with acceleration of 0.5 mg ring masses to 400 km/s and 0.02 mg ring masses to 1400 km/s at greater than or equal to 30 percent efficiency. Scaling the CT accelerator to the multi-megajoule level could provide an efficient, economical driver for inertial fusion (ICF) or magnetically insulated inertial fusion. Efficient conversion to X-rays for driving hohlraum-type ICF targets has been modeled using a radiation-hydrodynamics code. At less demanding conditions than required for ICF, a CT accelerator can be applied to fueling and current drive in tokamaks. Fueling is accomplished by injecting CTs at the required rate to sustain the particle inventory and at a velocity sufficient to penetrate to the magnetic axis before CT dissolution. Current drive is a consequence of the magnetic helicity content of the CT, which is approximately conserved during reconnection of the CT fields with the tokamak. Major areas of uncertainty in CT fueling and current drive concern the mechanism by which CTs will stop in a tokamak plasma and the effects of the CT on energy confinement and magnetic stability. Bounds on the required CT injection velocity are obtained by considering drag due to emission of an Alfven-wave wake and rapid reconnection and tilting on the internal Alfven time scale of the CT. Preliminary results employing a 3-D, resistive MHD code show rapid tilting with the CT aligning its magnetic moment with the tokamak field. Requirements for an experimental test of CT injection and scenarios for fueling a reactor will also be discussed.

  10. Magnetic confinement experiment. I: Tokamaks

    SciTech Connect

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 2.5x greater than ELM`ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices.

  11. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  12. Simulation of the alpha particle heating and the helium ash source in an International Thermonuclear Experimental Reactor-like tokamak with an internal transport barrier

    SciTech Connect

    Ye, Lei Guo, Wenfeng; Xiao, Xiaotao; Dai, Zongliang; Wang, Shaojie

    2014-12-15

    A guiding center orbit following code, which incorporates a set of non-singular coordinates for orbit integration, was developed and applied to investigate the alpha particle heating in an ITER-like tokamak with an internal transport barrier. It is found that a relatively large q (safety factor) value can significantly broaden the alpha heating profile in comparison with the local heating approximation; this broadening is due to the finite orbit width effects; when the orbit width is much smaller than the scale length of the alpha particle source profile, the heating profile agrees with the source profile, otherwise, the heating profile can be significantly broadened. It is also found that the stagnation particles move to the magnetic axis during the slowing-down process, thus the effect of stagnation orbits is not beneficial to the helium ash removal. The source profile of helium ash is broadened in comparison with the alpha source profile, which is similar to the heating profile.

  13. Tokamak Physics Experiment (TPX) power supply design and development

    SciTech Connect

    Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

    1995-04-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes.

  14. Characteristics of solid-target charge-exchange analyzers for energetic ion diagnostics on tokamaks

    SciTech Connect

    Beiersdorfer, P.; Roquemore, A.L.; Kaita, R.

    1987-05-01

    Compact electrostatic charge-exchange analyzers have been constructed for installation in areas of high magnetic fields and restricted access near tokamak fusion devices. The analyzers employed carbon stripping foils, and have been calibrated for proton energies between 1 and 70 keV. They have been successfully used to study charge-exchange losses in auxiliary-heated tokamak plasmas.

  15. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  16. Development of a repetitive compact torus injector

    NASA Astrophysics Data System (ADS)

    Onchi, Takumi; McColl, David; Dreval, Mykola; Rohollahi, Akbar; Xiao, Chijin; Hirose, Akira; Zushi, Hideki

    2013-10-01

    A system for Repetitive Compact Torus Injection (RCTI) has been developed at the University of Saskatchewan. CTI is a promising fuelling technology to directly fuel the core region of tokamak reactors. In addition to fuelling, CTI has also the potential for (a) optimization of density profile and thus bootstrap current and (b) momentum injection. For steady-state reactor operation, RCTI is necessary. The approach to RCTI is to charge a storage capacitor bank with a large capacitance and quickly charge the CT capacitor bank through a stack of integrated-gate bipolar transistors (IGBTs). When the CT bank is fully charged, the IGBT stack will be turned off to isolate banks, and CT formation/acceleration sequence will start. After formation of each CT, the fast bank will be replenished and a new CT will be formed and accelerated. Circuits for the formation and the acceleration in University of Saskatchewan CT Injector (USCTI) have been modified. Three CT shots at 10 Hz or eight shots at 1.7 Hz have been achieved. This work has been sponsored by the CRC and NSERC, Canada.

  17. Operating tokamaks with steady-state toroidal current

    SciTech Connect

    Fisch, N.J.

    1981-04-01

    Continuous operation of a tokamak requires, among other things, a means of continuously providing the toroidal current. Various methods have been proposed to provide this current including methods which utilize radio-frequency waves in any of several frequency regimes. Here we elaborate on the prospects of incorporating these current-drive techniques in tokamak reactors, concentrating on the theoretical minimization of the power requirements.

  18. Neoclassical diffusion of heavy impurities in a rotating tokamak plasma

    SciTech Connect

    Wong, K.L.; Cheng, C.Z.

    1987-08-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle simulation is carried out and the results offer a qualitative explanation for some experimental data from the Tokamak Test Reactor (TFTR). 13 refs., 2 figs.

  19. Fast ion effects on magnetic instabilities in the PDX tokamak

    NASA Astrophysics Data System (ADS)

    Buchenauer, D. A.

    Modification and excitation of nondisruptive magnetic instabilities due to near perpendicular neutral beam injection on the PDX tokamak were made to determine the importance of these instabilities at low q. The instabilities consisted of resistive MHD modes, beam driven ideal MHD modes, and beam driven ion cyclotron modes. Evidence of enhanced transport is presented for several of these instabilities as well as comparison of the experimental results with theory. Possible consequences for reactor type tokamaks and high power auxiliary heating systems are discussed.

  20. Compact, Lightweight Adsorber and Sabatier Reactor for CO2 Capture and Reduction for Consumable and Propellant Production

    NASA Technical Reports Server (NTRS)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Busby, Stacy A.; Abney, Morgan B.; Perry, Jay L.; Knox, James C.

    2012-01-01

    The utilization of CO2 to produce (or recycle) life support consumables, such as O2 and H2O, and to generate propellant fuels is an important aspect of NASA's concept for future, long duration planetary exploration. One potential approach is to capture and use CO2 from the Martian atmosphere to generate the consumables and propellant fuels. Precision Combustion, Inc. (PCI), with support from NASA, continues to develop its regenerable adsorber technology for capturing CO2 from gaseous atmospheres (for cabin atmosphere revitalization and in-situ resource utilization applications) and its Sabatier reactor for converting CO2 to methane and water. Both technologies are based on PCI's Microlith(R) substrates and have been demonstrated to reduce size, weight, and power consumption during CO2 capture and methanation process. For adsorber applications, the Microlith substrates offer a unique resistive heating capability that shows potential for short regeneration time and reduced power requirements compared to conventional systems. For the Sabatier applications, the combination of the Microlith substrates and durable catalyst coating permits efficient CO2 methanation that favors high reactant conversion, high selectivity, and durability. Results from performance testing at various operating conditions will be presented. An effort to optimize the Sabatier reactor and to develop a bench-top Sabatier Development Unit (SDU) will be discussed.

  1. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Dickens, J. K.; Hill, N. W.; Hou, F. S.; McConnell, J. W.; Spencer, R. R.; Tsang, F. Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

  2. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0. 8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    SciTech Connect

    Dickens, J.K.; Hill, N.W.; Hou, F.S.; McConnell, J.W.; Spencer, R.R.; Tsang, F.Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

  3. Fuel retention in tokamaks

    NASA Astrophysics Data System (ADS)

    Loarer, T.

    2009-06-01

    Tritium retention constitutes an outstanding problem for ITER operation and future fusion reactors, particularly for the choice of the first wall materials. In present day tokamaks, fuel retention is evaluated by two complementary methods. The in situ gas balance allows evaluation of how much fuel is retained during a discharge and, typically, up to one day of experiments. Post-mortem analysis is used to determine where the fuel is retained, integrated over an experimental campaign. In all the carbon clad devices, using the two methods, the retention is demonstrated to be very closely related to the carbon net erosion. This results from plasma-wall interaction with ion and charge-exchange fluxes, ELMs and is proportional to the pulse duration. The fuel retention by implantation saturates at high wall temperatures and limits the D/C ratio in the deposited layers but, as far as a carbon source exists, the dominant retention process remains the co-deposition of carbon with deuterium. In full metallic device, in the absence of wall conditioning with boron, co-deposition is strongly reduced and fuel retention below 1% can be achieved. Extrapolation to ITER shows that removing the carbon from the plasma-facing components would increase the number of discharges to 2500 before reaching the maximum tritium limit of 700 g.

  4. Broadband measurements of electron cyclotron emission in TFTR (Tokamak Fusion Test Reactor) using a quasi-optical light collection system and a polarizing Michelson interferometer

    SciTech Connect

    Stauffer, F.J.; Boyd, D.A.; Cutler, R.C.; Diesso, M.; McCarthy, M.P.; Montague, J.; Rocco, R.

    1988-04-01

    For the past three years, a Fourier transform spectrometer diagnostic system, employing a fast-scanning polarizing Michelson interferometer, has been operating on the TFTR tokamak at Princeton Plasma Physics Laboratory. It is used to measure the electron cyclotron emission spectrum over the range 2.5 to 18 cm/sup /minus/1/ (75-540 GHz) with a resolution of 0.123 cm/sup /minus/1/(3.7 GHz), at a rate of 72 spectra per second. The quasi-optical system for collecting the light and transporting it through the interferometer to the detector has been designed using the concepts of both Gaussian and geometrical optics in order to produce a system that is efficient over the entire spectral range. The commerical Michelson interferometer was custom-made for this project and is at the state of the art for this type of specialized instrument. Various pre-installation and post-installation tests of the optical system and the interferometer were performed and are reported here. An error propagation analysis of the absolute calibration process is given. Examples of electron cyclotron emission spectra measured in two polarization directions are given, and electron temperature profiles derived from each of them are compared. 34 refs., 17 figs.

  5. ORNL compact loop antenna design for TFTR and Tore Supra

    SciTech Connect

    Taylor, D.J.; Baity, F.W.; Bryan, W.E.; Hoffman, D.J.; McIlwain, R.L. ); Ray, J.M. )

    1987-01-01

    The goal supplemental ion cyclotron resonance heating (ICRH) of fusion plasma is to deliver power at high efficiencies deep within the plasma. The technology for fast-wave ICRH has reached the point of requiring proof-of-performance'' demonstration of specific antenna configurations of specific antenna configurations and their mechanical adequacy for operating in a fusion environment. Oak Ridge National Laboratory (ORNL) has developed the compact loop antenna concept based on a resonant double loop (RDL) configuration for use in both Tokamak Fusion Test Reactor (TFTR) and the Tore Supra ICRH programs. A description and a comparison of the technologies developed in the two designs are presented. The electrical circuit and the mechanical philosophy employed are the same for both antennas, but different operating environments result in substantial differences in the design of specific components. The ORNL TFTR antenna is designed to deliver 4 MW over a 2-s pulse, and the ORNL Tore Supra antenna is designed for 4 MW and essentially steady-state conditions. The TFTR design embodies the first operations compact RDL antenna, and the Tore Supra antenna extends the technology to an operational duty cycle consistent with reactor-relevant applications. 7 refs., 5 figs.

  6. TSC simulation of feedback stabilization of axisymmetric modes in tokamaks using driven halo currents

    SciTech Connect

    Jardin, S.C.; Schmidt, J.A.

    1997-03-01

    The Tokamak Simulation Code (TSC) has been used to model a new method of feedback stabilization of the axisymmetric instability in tokamaks using driven halo (or scrapeoff layer) currents. The method appears to be feasible for a wide range of plasma edge parameters. It may offer significant advantages over the more conventional method of controlling this instability when applied in a reactor environment.

  7. Modeling of simultaneous denitrification--anaerobic digestion--organic matter aerobic oxidation and nitrification in an anoxic-anaerobic-aerobic compact filter reactor.

    PubMed

    Moya, Jaime; Huiliñir, César; Peredo, Karol; Aspé, Estrella; Roeckel, Marlene

    2012-08-31

    A mathematical model was developed for a compact anoxic-anaerobic-aerobic filter reactor with liquid recirculation for the treatment of fishing effluents. The model includes denitrification, anaerobic digestion, aerobic carbon oxidation and nitrification steps, as well as an evaluation of the liquid gas mass transfer and pH. The model was calibrated using one experimental condition at a recycling ratio (R)=10, and was validated with R equal to 2 and 0, with an organic concentration of 554±24 mg TOCL(-1), salinity of 24 g L(-1) and hydraulic retention time (HRT) of 2 d. Carbon total removal is higher than 98%, while maximum nitrogen removal is 62% using total nitrification in the aerobic zone, due to a higher quantity of NO(x) produced which were recirculated to the anoxic zone. In the aerobic zone, simultaneous nitrification and denitrification processes occur, because the diffusion limitations cause a low oxygen penetration in the biofilm. In the anoxic-anaerobic zone, denitrification or methanogenesis inhibition by DO (caused by the recycled oxygen) is not observed.

  8. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    NASA Astrophysics Data System (ADS)

    Azizov, E. A.

    2012-02-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.

  9. Environmental Assessment and Finding of No Significant Impact: The National Compact Stellarator Experiment at the Princeton Plasma Physics Laboratory

    SciTech Connect

    N /A

    2002-10-25

    If the United States is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels, the eventual depletion of fissionable uranium used in present-day nuclear reactors, and the limitations of solar, hydro and wind alternatives. Fusion energy, the power source of the sun and other stars, has the potential to become a major source of energy for the future. Power from fusion would provide substantially reduced environmental impacts as compared with current forms of energy generation. Thus, the United States and other countries around the world continue to pursue development of fusion energy as one of a number of potential power sources for the long term. Fusion research, using various machine configurations, has been proceeding since the early 1950's, and significant progress has been achieved in performance and in understanding of the underlying physics. For most of this period, fusion machines called stellarators and tokamaks, which are toroidal (doughnut-shaped) devices, have been most frequently used to conduct experiments for producing controlled nuclear fusion. It is now desirable to take a next step in the fusion development program, by providing an experimental device to investigate the attractiveness of a compact stellarator as the basis for a fusion power reactor. This concept has the potential to build upon advances in understanding of stellarators and tokamaks, and to combine the best features of both. The goal is to build a compact stellarator that would be smaller than conventional stellarators and operate more efficiently than previous tokamaks. Such a device would broaden our understanding of magnetic fusion science while contributing to the development of a potentially attractive fusion reactor solution that may have cost advantages over other fusion concepts.

  10. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  11. Improvement of tokamak performance by injection of electrons

    SciTech Connect

    Ono, Masayuki.

    1992-12-01

    Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas.

  12. Prospects for toroidal fusion reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.D.

    1994-06-01

    Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, {approximately}2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges.

  13. Computational methods in tokamak transport

    SciTech Connect

    Houlberg, W.A.; Attenberger, S.E.; Lao, L.L.

    1982-06-01

    A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements.

  14. Modular tokamak magnetic system

    DOEpatents

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  15. Bootstrap currents in radio-frequency-driven tokamak equilibria

    SciTech Connect

    Hsiao, Ming-Yuan; Ehst, D.A.; Evans, K. Jr.

    1988-05-01

    Interest in the bootstrap current arising from neoclassical transport in tokamaks has increased recently in view of certain experimental observations. In this study, the bootstrap current is calculated for a number of rf current-driven tokamaks. Two-dimensional, self-consistent, steady-state tokamak MHD equilibria are obtained by including both the transport-driven bootstrap current and the externally driven rf current. The self-consistency is acomplished by iterating between two-dimensional MHD equilibrium calculations and the current calculations (including bootstrap and rf ray-tracing). Calculations for other reactor parameters of interest are also carried out. It is found that for reactor-grade plasmas, the bootstrap current contribution to the toroidal current is, in general, important. An approxiamte scaling law for GAMMA, based on parametric survey performed, is also obtained. 16 refs., 8 figs., 1 tab.

  16. Designing tokamaks to withstand electromagnetic disruption loads

    NASA Astrophysics Data System (ADS)

    Crowell, Jeffrey Arnold

    1999-11-01

    Tokamaks, the toroidal plasma confinement devices used to study fusion energy, operate by driving a multi-MA current in the plasma while creating a strong confining magnetic field. In experimental tokamaks under some conditions, the plasma can become unstable, escape its magnetic confines and rapidly cool off. On a time scale of milliseconds, the plasma current decays away in the resulting cold and highly resistive plasma. In these events, called disruptions, the rapid change in plasma current induces large currents in the surrounding conducting structures. The induced currents, flowing in the presence of a strong magnetic field, can apply substantial electromagnetic forces. Some experimental devices, such as the JET facility, have experienced extensive damage from these events. In future power reactors, even greater loads must be absorbed by components also subject to neutron embrittlement. This study models the electromagnetic and structural behavior of conceptual designs of the first generation of power-producing tokamaks to identify the components that are at risk and illuminate design options which mitigate these loads. The problem is a coupled one: the geometry and resistivity of the structure affects the induced currents while the induced currents and resulting loads place demands on the structure. Several new analytical and computational tools for the evaluation of these systems are discussed including a dual-solution technique for taking advantage of the complex electromagnetic symmetries in a typical tokamak design. The finite element method with a differential formulation and an integral method using a Green's function have been applied to 2D and 3D electromagnetic models of tokamaks. The differential formulation was found to be superior in these highly symmetric systems. The most significant design issues arise with the components most proximate to the plasma. Despite toroidal segmentation, damaging electromagnetic loads threaten the first wall and

  17. A low aspect ratio tokamak transmutation system

    NASA Astrophysics Data System (ADS)

    Qiu, L. J.; Wu, Y. C.; Xiao, B. J.; Xu, Q.; Huang, Q. Y.; Wu, B.; Chen, Y. X.; Xu, W. N.; Chen, Y. P.; Liu, X. P.

    2000-03-01

    A low aspect ratio tokamak transmutation system is proposed as an alternative application of fusion energy on the basis of a review of previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a radioactivity-clean nuclear power system as blanket, and (3) a novel concept of liquid metal centre conductor post as part of the toroidal field coils. In the conceptual design, a driver of 100 MW fusion power under 1 MW/m2 neutron wall loading can transmute the amount of high level waste (including minor actinides and fission products) produced by ten standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium on a self-sustaining basis and an output of about 2 GW of electrical energy. After 30 years of operation, the biological hazard potential level of the whole system will decrease by two orders of magnitude.

  18. The physics of tokamak start-up

    SciTech Connect

    Mueller, D.

    2013-05-15

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  19. Tritium Retention and Removal in Tokamaks

    SciTech Connect

    Skinner, Charles H.

    2009-02-19

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy. Tritium is an important source term in safety assessments, it is expensive and in short supply. Tritium can be continuously retained in a tokamak by codeposition with eroded carbon or beryllium and JET and TFTR with carbon plasma facing components showed a tritium retention level that would be unacceptable in ITER or future fusion reactors. Asdex-U and Alcator C-mod have shown reduced hydrogenic retention with tungsten clad and molybdenum plasma facing components. Once the tritium inventory approaches the administrative limit, tritium must be removed to permit continued D-T plasma operations. Several candidate techniques are being considered and need to be proven at a relevant speed and efficiency in contemporary tokamaks. Projections for ITER are discussed.

  20. The physics of tokamak start-upa)

    NASA Astrophysics Data System (ADS)

    Mueller, D.

    2013-05-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  1. ADX: a high field, high power density, advanced divertor and RF tokamak

    NASA Astrophysics Data System (ADS)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  2. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  3. Resistive edge mode instability in stellarator and tokamak geometries

    NASA Astrophysics Data System (ADS)

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-01

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  4. Tokamak foundation in USSR/Russia 1950-1990

    NASA Astrophysics Data System (ADS)

    Smirnov, V. P.

    2010-01-01

    In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

  5. Resistive edge mode instability in stellarator and tokamak geometries

    SciTech Connect

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-15

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  6. Localized Ballooning Modes in Compact Quasiaxially Symmetric Stellarators

    SciTech Connect

    M.H. Redi; J. Canik; R.L. Dewar; E.D. Fredrickson; W.A. Cooper; J.L. Johnson; S. Klasky

    2001-06-14

    Understanding of ballooning mode stability boundaries may lead to performance improvement of toroidal devices through control of plasma disruptions. Toroidally localized ballooning modes have been found as precursors to high-beta plasma disruptions on the Tokamak Fusion Test Reactor (TFTR) arising in conditions of n=1 kink mode asymmetry. Recent optimization has shown that magnetohydrodynamic (MHD) stability as well as good particle confinement are likely to be achievable in the National Compact Stellarator Experiment (NCSX), a compact, quasiaxially symmetric stellarator (QAS) for values of the plasma near beta = 4%. The configuration, with a major radius of 1.42 m, an aspect ratio of 4.4, a toroidal magnetic field 1.2-1.7 T and 6 MW of neutral-beam heating, is stable to MHD instabilities, and is expected to be limited by high-n kink and ballooning modes. This paper describes the ballooning eigenvalue isosurfaces for NCSX, the first step in an examination of the kinetic stabilization of the ballooning beta limit using a hybrid WKB approach.

  7. Formation of a 100-kA tokamak discharge in the Princeton large torus by lower hybrid waves

    SciTech Connect

    Jobes, F.; Stevens, J.; Bell, R.; Bernabei, S.; Cavallo, A.; Chu, T.K.; Cohen, S.; Denne, B.; Efthimion, P.; Hinnov, E.

    1984-03-01

    The development of non-inductive current drive is of great importance in establishing the tokamak as a long-pulse or steady-state fusion reactor. Lower hybrid waves, carrying 200 kW of power at 800 MHz, have been launched into the PLT tokamak to initiate and drive the discharge current to a level in excess of 100 kA.

  8. Completely bootstrapped tokamak

    SciTech Connect

    Weening, R.H. ); Boozer, A.H. )

    1992-01-01

    Numerical simulations of the evolution of large-scale magnetic fields have been developed using a mean-field Ohm's law. The Ohm's law is coupled to a {Delta}{prime} stabilty analysis and a magnetic island growth equation in order to simulate the behavior of tokamak plasmas that are subject to tearing modes. In one set of calculations, the magnetohydrodynamic (MHD)-stable regime of the tokamak is examined via the construction of an {ital l}{sub {ital i}} -{ital q}{sub {ital a}} diagram. The results confirm previous calculations that show that tearing modes introduce a stability boundary into the {ital l}{sub {ital i}} -{ital q}{sub {ital a}} space. In another series of simulations, the interaction between tearing modes and the bootstrap current is investigated. The results indicate that a completely bootstrapped tokamak may be possible, even in the absence of any externally applied loop voltage or current drive.

  9. Texas Experimental Tokamak

    SciTech Connect

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported.

  10. A Midsize Tokamak As Fast Track To Burning Plasmas

    SciTech Connect

    E. Mazzucato

    2010-07-14

    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.

  11. Solenoid-free plasma start-up in spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

  12. Evaluation of the operational parameters for NBI-driven fusion in low-gain tokamaks with two-component plasma

    NASA Astrophysics Data System (ADS)

    Chirkov, A. Yu.

    2015-09-01

    Low gain (Q ~ 1) fusion plasma systems are of interest for concepts of fusion-fission hybrid reactors. Operational regimes of large modern tokamaks are close to Q  ≈  1. Therefore, they can be considered as prototypes of neutron sources for fusion-fission hybrids. Powerful neutral beam injection (NBI) can support the essential population of fast particles compared with the Maxwellial population. In such two-component plasma, fusion reaction rate is higher than for Maxwellian plasma. Increased reaction rate allows the development of relatively small-size and relatively inexpensive neutron sources. Possible operating regimes of the NBI-heated tokamak neutron source are discussed. In a relatively compact device, the predictions of physics of two-component fusion plasma have some volatility that causes taking into account variations of the operational parameters. Consequent parameter ranges are studied. The feasibility of regimes with Q  ≈  1 is shown for the relatively small and low-power system. The effect of NBI fraction in total heating power is analyzed.

  13. Tearing Modes in Tokamaks

    SciTech Connect

    White, R. B.

    2008-05-14

    This lecture gives a basic introduction to magnetic pound elds, magnetic surface destruction, toroidal equilibrium and tearing modes in a tokamak, including the linear and nonlinear development of these modes and their modi pound cation by current drive and bootstrap current, and sawtooth oscillations and disruptions.

  14. Approach to ignition of tokamak reactors

    SciTech Connect

    Sigmar, D.J.

    1981-02-01

    Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup.

  15. High Beta Tokamaks

    SciTech Connect

    Cowley, S.

    1998-11-14

    Perhaps the ideal tokamak would have high {beta} ({beta} {approx}> 1) and classical confinement. Such a tokamak has not been found, and we do not know if one does exist. We have searched for such a possibility, so far without success. In 1990, we obtained analytic equilibrium solutions for large aspect ratio tokamaks at {beta} {approx} {Omicron}(1) [1]. These solutions and the extension at high {beta} poloidal to finite aspect ratio [2] provided a basis for the study of high {beta} tokamaks. We have shown that these configurations can be stable to short scale MHD modes [3], and that they have reduced neoclassical transport [4]. Microinstabilities (such as the {del}T{sub i} mode) seem to be stabilized at high {beta} [5] - this is due to the large local shear [3] and the magnetic well. We have some concerns about modes associated with the compressional branch which may appear at high {beta}. Bill Dorland and Mike Kotschenreuther have studied this issue and our concerns may be unfounded. It is certainly tantalizing, especially given the lowered neoclassical transport values, that these configurations could have no microinstabilities and, one could assume, no anomalous transport. Unfortunately, while this work is encouraging, the key question for high {beta} tokamaks is the stability to large scale kink modes. The MHD {beta} limit (Troyon limit) for kink modes at large aspect ratio is problematically low. There is ample evidence from computations that the limit exists. However, it is not known if stable equilibria exist at much higher {beta}--none have been found. We have explored this question in the asymptotic high {beta} poloidal limit. Unfortunately, we are unable to find stable equilibrium and also unable to show that they don't exist. The results of these calculations will be published when a more definitive answer is found.

  16. 20 years of research on the Alcator C-Mod tokamak

    SciTech Connect

    Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; and others

    2014-11-15

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

  17. Considerations of the high magnetic field tokamak path on the approach to fusion energy

    NASA Astrophysics Data System (ADS)

    Marmar, Earl

    2015-11-01

    This tutorial will review the physics basis, and its applications, for high magnetic field, compact visions of steady-state pilot plants and fusion reactors. This includes: energy and particle confinement; transport barriers; heating and current drive; scrape-off layer and divertor physics including implications for power handling, and ash/impurity control. The development of new technologies, particularly high-temperature, high critical magnetic field superconducting materials opens a new opportunity to consider the leverage of on-axis magnetic fields of 10T or more, enabling the feasibility of smaller sized devices on the path to fusion energy, including a pilot plant which could produce hundreds of megawatts of net electricity in a 10T tokamak with major radius of order 3 meter. Incorporating jointed magnetic coils, also made feasible by the high temperature superconductors, can dramatically improve flexibility of experimental superconducting facilities, and ultimately maintainability for reactor systems. Steady-state requires high bootstrap fraction, combined with efficient off-axis current drive, and existing and new approaches for RF sustainment will be covered, including Lower Hybrid Current Drive (both from the low- and high-field side), ECCD, and fast-wave techniques. External torque drive from neutral beams, routinely used in most present-day experiments to enhance confinement and suppress instabilities, will be weak or absent in reactors. Alternative, RF-based flow drive, using mode-converted ICRF waves will be discussed. All reactor concepts have extraordinary power handling requirements, combined with stringent limits on PFC erosion and impurity sources; the current state of the art in divertor configurations will be compared with emerging and new concepts, including snowflake, x-point, x-divertor and liquid metals, to meet these challenges. Supported by USDOE.

  18. Tokamak plasma modelling and atomic processes

    NASA Astrophysics Data System (ADS)

    Kawamura, T.

    1986-06-01

    Topics addressed include: particle control in a tokomak device; ionizing and recombining plasmas; effects of data accuracy on tokamak impurity transport modeling; plasma modeling of tokamaks; and ultraviolet and X-ray spectroscopy of tokamak plasmas.

  19. Power supplies and quench protection for the Tokamak Physics Experiment

    SciTech Connect

    Neumeyer, C.L.

    1994-07-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems.

  20. VIBRATION COMPACTION

    DOEpatents

    Hauth, J.J.

    1962-07-01

    A method of compacting a powder in a metal container is described including the steps of vibrating the container at above and below the resonant frequency and also sweeping the frequency of vibration across the resonant frequency several times thereby following the change in resonant frequency caused by compaction of the powder. (AEC)

  1. Energy confinement in tokamaks

    SciTech Connect

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston.

  2. TPX tokamak construction management

    SciTech Connect

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-12-31

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly.

  3. Tokamak divertor maps

    NASA Astrophysics Data System (ADS)

    Punjabi, Alkesh; Verma, Arun; Boozer, Allen

    1994-08-01

    A mapping method is developed to investigate the problem of determination and control of heat-deposition patterns on the plates of a tokamak divertor. The deposition pattern is largely determined by the magnetic field lines, which are mathematically equivalent to the trajectories of a single-degree-of-freedom time-dependent Hamiltonian system. Maps are natural tools to study the generic features of such systems. The general theory of maps is presented, and methods for incorporating various features of the magnetic field and particle motion in divertor tokamaks are given. Features of the magnetic field include the profile of the rotational transform, single- versus double-null divertor, reverse map, the effects of naturally occurring low M and N, and externally imposed high-M, high-N perturbations. Particle motion includes radial diffusion, pitch angle and energy scattering, and the electric sheath at the plate. The method is illustrated by calculating the stochastic broadening in a single- null divertor tokamak. Maps provide an efficient, economic and elegant method to study the problem of motion of plasma particles in the stochastic scrape-off layer.

  4. Impurity transport in Tokamaks

    NASA Astrophysics Data System (ADS)

    Amano, T.

    1983-12-01

    Theoretical and experimental efforts directed towards gaining an understanding of impurity behavior in Tokamaks are reviewed. In the Alcator Tokamak experiments, a laser blow-off technique was used to introduce trace amounts of impurities into ohmically heated plasmas. After a series of experiments in which they injected Si, Al, Fe, Mo impurities, an equation representing empirical impurity confinement time was derived. The scaling of this equation was compared with the results of impurity injection experiments on other Tokamaks, FT-I, PDX, TFR, ISX-B. Impurity confinement times in all these cases agree remarkably well, except for the TFR confinement times, which were about a factor of two larger than predicted. In the presence of intense neutral beam injection impurity ions behave differently. Specifically, in the ISX-B experiments, a marked accumulation of impurity ions toward the center of the plasma was observed in the case of counter neutral beam injection. This was interpreted semi-quantitatively by the neoclassical effect of the rotation of the plasma driven by the neutral beam.

  5. Twenty Years of Research on the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin

    2013-10-01

    Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

  6. Development and investigations of compact heat-transfer equipment for a nuclear power station equipped with a high-temperature gas-cooled reactor

    NASA Astrophysics Data System (ADS)

    Golovko, V. F.; Dmitrieva, I. V.; Kodochigov, N. G.; Bykh, O. A.

    2013-07-01

    The project of a nuclear power station the reactor coolant system of which includes a high-temperature gas-cooled reactor combined with a gas-turbine energy conversion unit supposes the use of high-efficient gas-cycle-based heat-transfer equipment. An analysis aimed at selecting the optimal heat-transfer surfaces is presented together with the results from their calculated and experimental investigation. The design features of recuperators arranged integrally with end and intermediate coolers and placed in a vertical sealed high-pressure vessel of limited sizes are considered.

  7. Transmutation of nuclear waste with a low-aspect-ratio tokamak neutron source

    NASA Astrophysics Data System (ADS)

    Hong, Bong Guen; Moon, Se Youn

    2014-10-01

    The transmutation characteristics of transuranics (TRUs) in a transmutation reactor based on a LAR (Low-aspect-ratio) tokamak as a neutron source are investigated. The optimum radial build of a transmutation reactor is found by using a coupled analysis of the tokamak systems and the neutron transport. The dependences of the transmutation characteristics on the aspect ratio A in the range of 1.5 to 2.5 and on the fusion power in the range of 150 to 500 MW are investigated. An equilibrium fuel cycle is developed for effective transmutation, and show that with one unit of the transmutation reactor based on the LAR tokamak producing fusion power in the range of a few hundred MWs, up to 3 PWRs (1.0 GWe capacity) can be supported with a burn-up fraction larger than 50%.

  8. Magnetic confinement experiment -- 1: Tokamaks

    SciTech Connect

    Goldston, R.J.

    1994-12-31

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization.

  9. FEM (Free Electron Maser) for tokamak: Final report

    SciTech Connect

    Not Available

    1987-01-01

    This paper studies the feasibility of a microwave source for heating a tokamak reactor. The free electron maser (FEM) shows great promise for being this source. The topics covered in this paper are microwave generation with FEM, efficiency enhancement, parameter scaling, space charge scaling, beam energy spread and efficiency scaling, electron beam line with energy recovery, achromatic bend, multi-stage depressed voltage electron beam collector, and development plans. 12 refs., 10 figs., 5 tabs. (LSP)

  10. Thermally excited proton spin-flip laser emission in tokamaks

    SciTech Connect

    Arunasalam, V.; Greene, G.J.

    1993-07-01

    Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser.

  11. Mathematical modeling plasma transport in tokamaks

    SciTech Connect

    Quiang, Ji

    1995-12-31

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10{sup 20}/m{sup 3} with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

  12. Tritium Experience in Large Tokamaks: Application to ITER

    SciTech Connect

    Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

    1998-05-01

    Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R{ampersand}D avenues to address them. This paper summarizes the presentations, discussions, and recommendations.

  13. Alternative fusion concepts and the prospects for improved reactors

    NASA Astrophysics Data System (ADS)

    Krakowski, R. A.

    1985-05-01

    Past trends, present status, and future directions in the search for an improved fusion reactor are reviewed, and promising options available to both the principle tokamak and other supporting concept are summarized.

  14. Dust Measurements in Tokamaks

    SciTech Connect

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  15. Sawtooth oscillation in tokamaks

    SciTech Connect

    Park, W.; Monticello, D.A.

    1989-03-01

    A three-dimensional nonlinear toroidal full MHD code, MH3D, has been used to study sawtooth oscillations in tokamaks. The profile evolution during the sawtooth crash phase compares well with experiment, but only if neoclassical resistivity is used in the rise phase. (Classical resistivity has been used in most of the previous theoretical sawtooth studies.) With neoclassical resistivity, the q value at the axis drops from 1 to about 0.8 before the crash phase, and then resets to 1 through a Kadomtsev-type complete reconnection process. This ..delta..q/sub 0/ approx. = 0.2 is much larger than ..delta..q/sub o/ approx. = 0.01, which is obtained if classical resistivity is used. The current profile is strongly peaked at the axis with a flat region around the singular surface, and is similar to the Textor profile. To understand this behavior, approximate formulas for the time behavior of current and q values are derived. A functional dependence of sawtooth period scaling is also derived. A semi-empirical scaling is found which fits the experimental data from various tokamaks. Some evidence is presented which indicates that the fast crash time is due to enhanced effective resistivity inside the singular current sheet, generated by, e.g., microinstability and electron parallel viscosity with stochastic fields at the x-point. 16 refs., 5 figs.

  16. Ureilite compaction

    NASA Astrophysics Data System (ADS)

    Walker, D.; Agee, C. B.

    1988-03-01

    Ureilite meteorites show the simple mineralogy and compact recrystallized textures of adcumulate rock or melting residues. A certain amount of controversy exists about whether they are in fact adcumulate rocks or melting residues and about the nature of the precursor liquid or solid assemblage. The authors undertook a limited experimental study which made possible the evaluation of the potential of the thermal migration mechanism (diffusion on a saturation gradient) for forming ureilite-like aggregates from carbonaceous chondrite precursors. They find that the process can produce compact recrystallized aggregates of silicate crystals which do resemble the ureilities and other interstitial-liquid-free adcumulate rocks in texture.

  17. Plasma engineering analysis of Tennessee Tokamak

    SciTech Connect

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Shannon, T.E.; Uckan, N.A.

    1983-01-01

    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses dueterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. Detailed analyses are performed in the areas of (1) transport simulation using the 1-1/2-D WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control sytems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

  18. Observation of finite-. beta. MHD phenomena in tokamaks

    SciTech Connect

    McGuire, K.M.

    1984-09-01

    Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1 internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.

  19. Numerical simulation of fueling in tokamaks

    SciTech Connect

    Attenberger, S.E.; Houlberg, W.A.; Milora, S.L.

    1982-04-01

    We describe the numerical simulation of fueling and particle transport in both present and future tokamak plasmas. Models for pellet ablation and plasma density behavior after pellet injection are compared with experimental results in ISX and PDX plasmas and then extended to fusion reactor conditions. The role of fast ion ablation due to intense neutral beam injection and fusion alphas is examined along with pellet size and velocity considerations. In plasmas with high pumping efficiency (which may be obtained with divertor operation), pellet injection can significantly reduce fueling rates while maintaining more flexibility in control of the density profile than afforded by gas puffing. When fueling is dominated by gas puffing or high recycle from the walls or limiter, control of the fueling and density profiles is reduced and particle fluxes to the wall increase.

  20. Numerical simulation of fueling in tokamaks

    SciTech Connect

    Attenberger, S.E.; Houlberg, W.A.; Milora, S.L.

    1981-01-01

    We describe the numerical simulation of fueling and particle transport in both present and future tokamak plasmas. Models for pellet ablation and plasma density behavior after pellet injection are compared with experimental results in ISX and PDX plasmas and then extended to fusion reactor conditions. The role of fast ion ablation due to intense neutral beam injection and fusion alphas is examined along with pellet size and velocity considerations. In plasmas with high pumping efficiency (which may be obtained with divertor operation), pellet injection can significantly reduce fuel handling requirements and interaction of the plasma with the chamber walls while maintaining more flexibility in control of the density profile than afforded by gas puffing. When fueling is dominated by gas puffing or high recycle from the walls or limiter, control of the fueling and density profiles is reduced while plasma/wall interactions increase.

  1. Co3O4-based honeycombs as compact redox reactors/heat exchangers for thermochemical storage in the next generation CSP plants

    NASA Astrophysics Data System (ADS)

    Pagkoura, Chrysoula; Karagiannakis, George; Halevas, Eleftherios; Konstandopoulos, Athanasios G.

    2016-05-01

    Over the last years, several research groups have focused on developing efficient thermochemical heat storage (THS) systems, in-principle capable of being coupled with next generation high temperature Concentrated Solar Power plants. Among systems studied, the Co3O4/CoO redox system is a promising candidate. Currently, research efforts extend beyond basic level identification of promising materials to more application-oriented approaches aiming at validation of THS performance at pilot scale reactors. The present work focuses on the investigation of cobalt oxide based honeycomb structures as candidate reactors/heat exchangers to be employed for such purposes. In the evaluation conducted and presented here, cobalt oxide-based structures with different composition and geometrical characteristics were subjected to redox cycles in the temperature window between 800 and 1000°C under air flow. Basic aspects related to redox performance of each system are briefly discussed but the main focus lies on the evaluation of the segments structural stability after multi-cyclic operation. The latter is based on macroscopic visual observation and also supplemented by pre- (i.e. fresh samples) and post-characterization (i.e. after long term exposure) of extruded honeycombs via combined mercury porosimetry and SEM analysis.

  2. Diagnostics for hybrid reactors

    SciTech Connect

    Orsitto, Francesco Paolo

    2012-06-19

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  3. Compact accelerator

    DOEpatents

    Caporaso, George J.; Sampayan, Stephen E.; Kirbie, Hugh C.

    2007-02-06

    A compact linear accelerator having at least one strip-shaped Blumlein module which guides a propagating wavefront between first and second ends and controls the output pulse at the second end. Each Blumlein module has first, second, and third planar conductor strips, with a first dielectric strip between the first and second conductor strips, and a second dielectric strip between the second and third conductor strips. Additionally, the compact linear accelerator includes a high voltage power supply connected to charge the second conductor strip to a high potential, and a switch for switching the high potential in the second conductor strip to at least one of the first and third conductor strips so as to initiate a propagating reverse polarity wavefront(s) in the corresponding dielectric strip(s).

  4. Energy deposition in STARFIRE reactor components

    SciTech Connect

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  5. Experimental studies of compact toroids

    SciTech Connect

    Not Available

    1991-01-01

    The Berkeley Compact Toroid Experiment (BCTX) device is a plasma device with a Marshall-gun generated, low aspect ratio toroidal plasma. The device is capable of producing spheromak-type discharges and may, with some modification, produce low-aspect ratio tokamak configurations. A unique aspect of this experimenal devie is its large lower hybrid (LH) heating system, which consists of two 450MHz klystron tubes generating 20 megawatts each into a brambilla-type launching structure. Successful operation with one klystron at virtually full power (18 MW) has been accomplished with 110 {mu}s pulse length. A second klystron is currently installed in its socket and magnet but has not been added to the RF drive system. This report describes current activities and accomplishments and describes the anticipated results of next year's activity.

  6. Quick profile-reoriganization driven by helical field perturbation for suppressing tokamak major disruptions

    NASA Astrophysics Data System (ADS)

    Yamazaki, K.; Kawahata, K.; Ando, R.; Matsuoka, K.; Hirokura, S.; Kitagawa, S.; Mohri, A.; Tanahashi, S.; Taniguchi, Y.; Toi, K.

    1986-09-01

    Disruptive behavior of magnetic field configuration leading to tokamak major disruption is found to be controlled by a mild mini-disruption which is induced by the compact external modular multipole-field coils with m=3/n=2 dominant helical field component in the JIPP T-IIU tokamak. This mini-disruption ergodizes the m=2/n=1 magnetic island quickly but mildly and then prevents the profile of electron temperature from flattening. This quick profile-reorganization is effective to avoid the two-step disruption (pre- and major disruptions) responsible for the catastrophic current termination.

  7. LIDAR Thomson scattering for advanced tokamaks. Final report

    SciTech Connect

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-03-18

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured.

  8. Tooling concepts for ITER tokamak assembly and remote disassembly

    SciTech Connect

    Oikawa, A.; Puhn, F.; Helary, J.L.; Shaw, R.; Friend, M.; Piec, Z.; Tachikawa, N.; Acks, M.; Basile, A.

    1995-12-31

    Since ITER has many of the characteristics of a full-scale tokamak reactor, its provisions for assembly and replaceability are relevant to a future fusion power plant. The performance of ITER is dependent on tight tolerances, mainly for the magnets and plasma facing components. The magnetic field must be highly uniform in the toroidal direction to ensure good plasma energy and particle confinement. Alignment of the plasma facing surface of the first wall and divertor target plates is required to avoid large local heat loads on the plasma facing components and, as a consequence, their erosion and contamination of the plasma with impurities. Because of the large and heavy components the major challenge of the ITER tokamak assembly is to hold such tight tolerances using guide tools, adjustable interfaces, accurate measuring tools, and specific procedures to compensate for deformation and fabrication tolerances. The assembly tooling plan also includes verification of the essential remote handling operations.

  9. [alpha]-particle transport-driven current in tokamaks

    SciTech Connect

    Heikkinen, J.A. ); Sipilae, S.K. )

    1995-03-01

    It is shown that the radial transport of fusion-born energetic [alpha] particles, induced by electrostatic waves traveling in one poloidal direction, is directly connected to a net momentum of [alpha] particles in the toroidal direction in tokamaks. Because the momentum change is almost independent of toroidal velocity, the energy required for the momentum generation remains small on an [alpha]-particle population sustained by an isotropic time-independent source. By numerical toroidal Monte Carlo calculations it is shown that the current carried by [alpha] particles in the presence of intense well penetrated waves can reach several mega-amperes in reactor-sized tokamaks. The current obtained can greatly exceed the neoclassical bootstrap current of the [alpha] particles.

  10. Three-dimensional equilibria in axially symmetric tokamaks

    PubMed Central

    Garabedian, Paul R.

    2006-01-01

    The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of α particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158

  11. On the economic prospects of nuclear fusion with tokamaks

    NASA Astrophysics Data System (ADS)

    Pfirsch, D.; Schmitter, K. H.

    1987-12-01

    A method of cost and construction energy estimation for tokamak fusion power stations conforming to the present stage of fusion development is described. The method is based on first-wall heat load constraints rather than Beta limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled Environmental Impact and Economic Prospects of Nuclear Fusion (1986). It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is derived when it is properly applied for cost estimation of advanced gas-cooled and Magnox reactors, the two examples presented by the European study to disprove it.

  12. Limiter/vacuum system for plasma impurity control and exhaust in tokamaks

    SciTech Connect

    Abdou, M.; Brooks, J.; Mattas, R.

    1980-01-01

    A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification.

  13. Bibliography of fusion product physics in tokamaks

    SciTech Connect

    Hively, L. M.; Sigmar, D. J.

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category.

  14. ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

    NASA Astrophysics Data System (ADS)

    Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

    2014-10-01

    The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

  15. Moving Divertor Plates in a Tokamak

    SciTech Connect

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  16. Conceptual design study of the moderate size superconducting spherical tokamak power plant

    NASA Astrophysics Data System (ADS)

    Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

    2015-06-01

    A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

  17. Protection of tokamak plasma facing components by a capillary porous system with lithium

    NASA Astrophysics Data System (ADS)

    Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

    2015-08-01

    Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

  18. Compact magnetograph

    NASA Technical Reports Server (NTRS)

    Title, A. M.; Gillespie, B. A.; Mosher, J. W.

    1982-01-01

    A compact magnetograph system based on solid Fabry-Perot interferometers as the spectral isolation elements was studied. The theory of operation of several Fabry-Perot systems, the suitability of various magnetic lines, signal levels expected for different modes of operation, and the optimal detector systems were investigated. The requirements that the lack of a polarization modulator placed upon the electronic signal chain was emphasized. The PLZT modulator was chosen as a satisfactory component with both high reliability and elatively low voltage requirements. Thermal control, line centering and velocity offset problems were solved by a Fabry-Perot configuration.

  19. Advanced fueling system for steady-state operation of a fusion reactor

    SciTech Connect

    Raman, R.

    2008-07-15

    Steady-state Advanced Tokamak scenarios rely on optimized density and pressure profiles to maximize the bootstrap current fraction. Under this mode of operation, the fuelling system must deposit small amounts of fuel where it is needed, and as often as needed, so as to compensate for fuel losses, but not to adversely alter the established density and pressure profiles. A precision fuelling system has the capability for controlling the fusion burn by maintaining the required pressure profile to maximize the bootstrap current fraction. An advanced fuelling system based on Compact Toroid (CT) injection has the potential to meet these needs while simultaneously simplifying the requirements of the tritium handling systems. Simpler engineering systems would reduce reactor construction and maintenance cost through increased reliability. A CT fueling system is described together with the associated tritium handling requirements. (authors)

  20. Status of tokamak experiments

    SciTech Connect

    Wolf, G.H.

    1996-03-01

    Plasma-wall interaction, heat removal and ash exhaust have emerged as the dominant problems still to be solved in order to achieve ignition and - even more difficult - to maintain a state of self-sustained thermo-nuclear burn. This is of particular delicacy, since those operational regimes which yield the best energy confinement correspond to an even better particle confinement and confinement of impurities, which then tend to accumulate in the plasma core and to result in disruption or degradation of the tokamak discharge. Therefore, plasma-wall interaction, heat removal and particle exhaust will determine not only the structure and configuration of the plasma edge region, of the wall system and of the materials facing the plasma, but also the final choice of useful confinement regimes. Moreover, the potential effect of powerful {alpha}-particle heating on plasma stability and confinement has to be kept below critical values. For the latter requirement, a final answer can only be obtained in an ITER-type device where ignition and burn will become accessible. 72 refs., 12 figs.

  1. Resistive instabilities in tokamaks

    SciTech Connect

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed.

  2. Continuous tokamak operation with an internal transformer

    SciTech Connect

    Singer, C.E.; Mikkelsen, D.R.

    1982-10-01

    A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia timescales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors.

  3. Equilibrium system analysis in a tokamak ignition experiment

    SciTech Connect

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  4. Equilibrium system analysis in a tokamak ignition experiment. Final report

    SciTech Connect

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  5. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  6. Tokamak coordinate conventions: COCOS

    NASA Astrophysics Data System (ADS)

    Sauter, O.; Medvedev, S. Yu.

    2013-02-01

    Dealing with electromagnetic fields, in particular current and related magnetic fields, yields "natural" physical vector relations in 3-D. However, when it comes to choosing local coordinate systems, the "usual" right-handed systems are not necessarily the best choices, which means that there are several options being chosen. In the magnetic fusion community such a difficulty exists for the choices of the cylindrical and of the toroidal coordinate systems. In addition many codes depend on knowledge of an equilibrium. In particular, the Grad-Shafranov axisymmetric equilibrium solution for tokamak plasmas, ψ, does not depend on the sign of the plasma current Ip nor that of the magnetic field B0. This often results in ill-defined conventions. Moreover the sign, amplitude and offset of ψ are of less importance, since the free sources in the equation depend on the normalized radial coordinate. The signs of the free sources, dp/dψ and dF2/dψ (p being the pressure, ψ the poloidal magnetic flux and F=RBφ), must be consistent to generate the current density profile. For example, RF and CD calculations (Radio Frequency heating and Current Drive) require an exact sign convention in order to calculate a co- or counter-CD component. It is shown that there are over 16 different coordinate conventions. This paper proposes a unique identifier, the COCOS convention, to distinguish between the 16 most-commonly used options. Given the present worldwide efforts towards code integration, the proposed new index COCOS defining uniquely the COordinate COnventionS required as input by a given code or module is particularly useful. As codes use different conventions, it is useful to allow different sign conventions for equilibrium code input and output, equilibrium being at the core of any calculations in magnetic fusion. Additionally, given two different COCOS conventions, it becomes simple to transform between them. The relevant transformations are described in detail.

  7. Understanding disruptions in tokamaks

    NASA Astrophysics Data System (ADS)

    Zakharov, Leonid

    2011-10-01

    Disruptions in tokamaks are known since 1963 but even now some aspects of them remain a mystery. This talk describes progress made recently in understanding disruptions. A major step forward occurred in 2007 when the importance of galvanic contact of the plasma with the wall in plasma dynamics was pointed out. The toroidal asymmetry of plasma current, observed in JET vertical disruptions, was explained by the theory of the wall touching kink mode. The currents shared by the plasma with the wall and responsible for the asymmetry were identified as generated by the kink mode. Such currents are referred to as Hiro currents. They have shown exceptional consistency with the entire JET disruption data base (more than 5500 cases) and ruled out the long lasting interpretation based on ``halo currents,'' which contradict experiments even in the sign of the measured asymmetry. Accordingly, the sideways forces are understood and their scaling from JET to ITER was justified. Hiro currents provide also a plausible explanation of the current spike at the beginning of the disruptions. The important role of the plasma edge and its interaction with the wall was revealed. Based on this new understanding of disruptions, dedicated experiments on the current spike (J-TEXT, Wuhan, China) and runaway prevention by the repetitive triggering of kink modes (T-10, AUG, Tore Supra) were motivated and are in progress. Accordingly, the need for new, adaptive grid approaches to numerical simulations of disruptions became evident. In addition to the core MHD, simulations of realistic wall geometry, disruption specific plasma edge physics, plasma-wall interaction, and energetic particles need be developed. The first results of simulations of the fast MHD regime, Hiro current generation, and slower plasma decay due to a wall touching kink mode made with the new DSC code are presented. This work is supported by US DoE contract No. DE-AC02-09-CH11466.

  8. Gridded ionization chamber for detection of x-ray wave activity in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Goldman, M. A.; Hill, K. W.; Moshey, E. A.; Sauthoff, N. R.; von Goeler, S.

    1985-03-01

    In order to carry out x-ray observations of magnetohydrodynamic wave activity of the plasma during DD and DT (deuterium-beam-heated deuterium and tritium plasmas, respectively) operation of the Tokamak Fusion Test Reactor (TFTR), we will need detectors not susceptible to nuclear radiation damage. We have investigated the use of gridded ionization chambers as fast nondamageable x-ray detectors. A prototype chamber is described which was tested on the PDX tokamak. These tests and laboratory tests with a pulsed x-ray source suggest that the detector has sufficient sensitivity and speed for the required measurements.

  9. One-dimensional transport code modeling of the divertor-limiter region in tokamaks

    SciTech Connect

    Ogden, J.M.; Singer, C.E.; Post, D.E.; Jensen, R.V.; Seidl, F.G.P.

    1981-12-01

    A model of the diverter-limiter scrapeoff region has been incorporated into the BALDUR one-dimensional tokamak transport code. Simulations of the proposed Toroidal Fusion Test Reactor (TFTR), and Poloidal Diverter (PDX) experiments and existing Alcator-A tokamak experiments have been carried out for ohmic and neutral beam heated cases. In particular it is studied how the edge conditions and energy-loss mechanisms in PDX depend upon plasma density, and results are compared with analytic estimates. The sensitivity of the results to changes in the transport coefficients and scrapeoff model is discussed with particular reference to the power loading on the TFTR limiter. 13 refs.

  10. Gridded ionization chamber for detecion of x-ray wave activity in tokamak plasmas

    SciTech Connect

    Goldman, M.A.; Hill, K.W.; Moshey, E.A.; Sauthoff, N.R.; von Goeler, S.

    1982-11-01

    In order to carry out X-ray observations of magnetohydrodynamic wave activity of the plasma during DD and DT (deuterium-beam-heated deuterium and tritium plasmas, respectively) operation of the Tokamak Fusion Test Reactor (TFTR), we will need detectors not susceptible to nuclear radiation damage. We have investigated the use of gridded ionization chambers as fast nondamageable X-ray detectors. A prototype chamber is described, which was tested on the PDX tokamak. These tests and laboratory tests with a pulsed X-ray source suggest that the detector has sufficient sensitivity and speed for the required measurements.

  11. Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

    NASA Astrophysics Data System (ADS)

    Freidberg, J.; Mangiarotti, F.; Minervini, J.

    2014-10-01

    The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

  12. Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

    NASA Astrophysics Data System (ADS)

    Jassby, D. L.

    1987-03-01

    The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.

  13. Helium Refrigerator Design for Pulsed Heat Load in Tokamaks

    SciTech Connect

    Kuendig, A.; Schoenfeld, H.

    2006-04-27

    Nuclear fusion reactors of the Tokamak type will be operated in a pulsed mode requiring the helium refrigerator to remove periodically large heat loads in time steps of approximately one hour. What are the necessary steps for a refrigerator to cope with such load variations?A series of numerical simulations has been performed indicating the possibility of an active refrigerator control with low exergetic losses. A basic comparison is made between the largest existing refrigerator sizes and the size required to service for example the ITER requirements.

  14. Profile control of advanced tokamak plasmas in view of continuous operation

    NASA Astrophysics Data System (ADS)

    Mazon, D.

    2015-07-01

    The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named 'advanced scenarios' are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated 'bootstrap' current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.

  15. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    SciTech Connect

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs.

  16. Currents induced in tokamaks by electron cyclotron heating

    SciTech Connect

    Eldridge, O. C.

    1980-10-01

    Generation of a plasma current is predicted in association with strong electron cyclotron heating in tokamaks or in any plasma with transverse magnetic field gradients. The current predicted in present-day tokamaks is of the order of one-quarter ampere per watt, which is large enough to be detected in heating experiments in progress. The current scales linearly with electron temperature and heating power and inversely with density and major radius. The mechanism depends on the Doppler shift for electrons streaming along magnetic field lines. Electrons streaming toward the source of radiation are resonant at a larger magnetic field. When the interaction is strong, radiation incident from the high field side is absorbed before reaching the cold electron resonant surface, and, so, a unidirectional population of electrons is heated. The anisotropic electron distribution gains momentum by collisions with ions. For small tokamaks the extraordinary wave should be launched for current drive, but for reactors the ordinary wave produces a sufficiently strong interaction.

  17. Fusion product measurements in tokamaks

    SciTech Connect

    Strachan, J.D.

    1985-05-01

    Diagnostic methods and the applications of fusion product measurements in tokamaks are reviewed with emphasis on results from PLT, PDX, and TFTR. Measurements have been made using the 2.5-MeV neutron from the d(d, n)/sup 3/ He reaction, the 3-MeV proton from the d(d, p)t reaction, both the 3.7-MeV alpha and the 14.7-MeV proton from the d(/sup 3/He, p)..cap alpha.. reaction, and the 14-MeV neutron from the d(t, n)..cap alpha.. reaction. The common use of these measurements is the determination of the ion temperature from the magnitude of the d-d neutron emission. For tokamak plasmas, these results are usually in good agreement with the charge exchange ion temperature. Recently, the charged fusion products have been used for high-resolution spectroscopic purposes, and emission profile measurements. Pitch angle resolution of the escaping 3-MeV proton emission has been used to determine the poloidal magnetic field inside the tokamak. Major issues in this field include the expected tritium operation on TFTR where the neutron measurements will determine when tritium will be introduced into the TFTR vessel and provide a measurement of the fusion power multiplication value (Q). The TFTR Q approx. 1 experiments will also provide a chance to measure the confinement of 3.5-MeV alphas in a tokamak.

  18. Litization of FTU tokamak vacuum vessel by using a Li limiter

    NASA Astrophysics Data System (ADS)

    Mazzitelli, Giuseppe; Apicella, Maria Laura; Lazarev, Vladimir; Azizov, E. A.; Mirnov, S. V.; Petrov, Vladimir; Vertkov, Alexei; Evtikhin, V. A.; Lyublinski, I. E.

    2003-10-01

    The idea to use lithium, a low Z metallic material, for the first wall protection of the tokamak-reactor was known for a rather long time. However practical application of lithium on experimental installations began rather recently. Litization - the coating of tokamak vacuum vessel walls by a thin lithium film is one of methods for protection of plasma from high Z impurities fluxes. A possible method for litization of the vacuum walls is by sputtering and evaporation of a Li limiter . This method utilizes a new concept of limiter on the basis of capillary-porous structure (CPS), having a sufficient mechanical stability (Li confinement). Furthermore the proposed experiment aims to study the liquid Li erosion, Li accumulation and distribution in plasma, thermal load reduction on the limiter due to Li radiation losses in a medium size tokamak as FTU.

  19. Ceramic powder compaction

    SciTech Connect

    Glass, S.J.; Ewsuk, K.G.; Mahoney, F.M.

    1995-12-31

    With the objective of developing a predictive model for ceramic powder compaction we have investigated methods for characterizing density gradients in ceramic powder compacts, reviewed and compared existing compaction models, conducted compaction experiments on a spray dried alumina powder, and conducted mechanical tests and compaction experiments on model granular materials. Die filling and particle packing, and the behavior of individual granules play an important role in determining compaction behavior and should be incorporated into realistic compaction models. These results support the use of discrete element modeling techniques and statistical mechanics principals to develop a comprehensive model for compaction, something that should be achievable with computers with parallel processing capabilities.

  20. A charged fusion product diagnostic for a spherical tokamak

    NASA Astrophysics Data System (ADS)

    Perez, Ramona Leticia Valenzuela

    Designs for future nuclear fusion power reactors rely on the ability to create a stable plasma (hot ionized gas of hydrogen isotopes) as a medium with which to sustain nuclear fusion reactions. My dissertation work involves designing, constructing, testing, installing, operating, and validating a new diagnostic for spherical tokamaks, a type of reactor test facility. Through detecting charged particles emitted from the plasma, this instrument can be used to study fusion reaction rates within the plasma and how they are affected by plasma perturbations. Quantitatively assessing nuclear fusion reaction rates at specific locations inside the plasma and as a function of time can provide valuable data that can be used to evaluate theory-based simulations related to energy transport and plasma stability. The Proton Detector (PD), installed in the Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy (CCFE) in Abingdon, England, was the first instrument to experimentally detect 3 MeV Protons and 1 MeV Tritons created from deuterium- deuterium (hydrogen isotopes) nuclear fusion reactions inside a spherical tokamak's plasma. The PD consists of an array of particle detectors with a protective housing and the necessary signal conditioning electronics and readout. After several years of designing (which included simulations for detector orientations), fabricating, and testing the PD, it was installed in MAST and data were collected over a period of two months in the summer of 2013. Proton and triton rates as high as 200 kHz were measured and an initial radial profile of these fusion reaction rates inside the plasma was extracted. These results will be compared to a complementary instrument at MAST as well as theory-based simulations and form the knowledge basis for developing a larger future instrument. The design and performance of all instrument components (electrical, computational, mechanical), and subsequent data analysis methods and results are

  1. Simplified compact containment BWR plant

    SciTech Connect

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-07-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  2. STARFIRE: a commercial tokamak fusion power plant study

    SciTech Connect

    Not Available

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  3. Transport of Dust Particles in Tokamak Devices

    SciTech Connect

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  4. Microwave Tokamak Experiment: Overview and status

    SciTech Connect

    Not Available

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs.

  5. Compact reactor for onboard hydrogen generation

    NASA Technical Reports Server (NTRS)

    Brabbs, T. A.

    1980-01-01

    Hydrogen, chemically stored as methanol, is promising internal-combustion fuel. Methanol is readily obtainable from natural products such as wood, compost, or various organic wastes. Steam reformation of methanol as source for hydrogen is relatively simple operation.

  6. Tokamak power system studies at ANL

    SciTech Connect

    Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

    1986-06-01

    The following features, in particular, have been examined: (a) large aspect ratio (A approx. = 6), which may ease maintenance; (b) high beta (..beta.. greater than or equal to 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I approx. = 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors.

  7. Comprehensive numerical modelling of tokamaks

    SciTech Connect

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-03

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell.

  8. Chemical-vapor-deposition reactor

    NASA Technical Reports Server (NTRS)

    Chern, S.

    1979-01-01

    Reactor utilizes multiple stacked trays compactly arranged in paths of horizontally channeled reactant gas streams. Design allows faster and more efficient deposits of film on substrates, and reduces gas and energy consumption. Lack of dead spots that trap reactive gases reduces reactor purge time.

  9. Neoclassical magnetic microislands in tokamaks

    SciTech Connect

    Kovalishen, E.A.; Mikhailovskii, A.B.; Botov, P.V.; Shirokov, M.S.; Konovalov, S.V.; Tsypin, V.S.; Galvao, R.M.O.

    2005-09-15

    Possibility of existence of neoclassical magnetic microislands (island width smaller than the ion Larmor radius) in a tokamak in the banana regime is shown. The rotation frequency of such islands is found. It is shown that for the case of positive electron temperature gradient, the bootstrap current destabilizes the microislands while the polarization current leads to their stabilization. Maximally possible neoclassical microisland width is estimated.

  10. Gyrosheath near the tokamak edge

    SciTech Connect

    Hazeltine, R.D.; Xiao, H. . Inst. for Fusion Studies); Valanju, P.M. . Fusion Research Center)

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results.

  11. Transport Equations In Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Callen, J. D.

    2009-11-01

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for: neoclassical effects on the parallel Ohm's law (trapped particle effects on resistivity, bootstrap current); fluctuation-induced transport; heating, current-drive and flow sources and sinks; small B field non-axisymmetries; magnetic field transients etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed recently using a kinetic-based framework. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales (and constraints they impose) are considered sequentially: compressional Alfv'en waves (Grad-Shafranov equilibrium, ion radial force balance); sound waves (pressure constant along field lines, incompressible flows within a flux surface); and ion collisions (damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on the plasma fluid: 7 ambipolar collision-based ones (classical, neoclassical, etc.) and 8 non-ambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients etc.). The plasma toroidal rotation equation [1] results from setting to zero the net radial current induced by the non-ambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the non-ambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The resultant transport equations will be presented and contrasted with the usual ones. [4pt] [1] J.D. Callen, A.J. Cole, C.C. Hegna, ``Toroidal Rotation In

  12. Compact Process Development at Babcock & Wilcox

    SciTech Connect

    Eric Shaber; Jeffrey Phillips

    2012-03-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of compaction trials have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel at packing fractions exceeding 46% by volume. Results from these trials are included. The scale-up effort is nearing completion with the process installed and operable using nuclear fuel materials. Final process testing is in progress to certify the process for manufacture of qualification test fuel compacts in 2012.

  13. Can tokamaks PFC survive a single event of any plasma instabilities?

    NASA Astrophysics Data System (ADS)

    Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

    2013-07-01

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

  14. GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

    NASA Astrophysics Data System (ADS)

    Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

    2015-09-01

    An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

  15. Tokamak plasma position dynamics and feedback control

    SciTech Connect

    Burenko, L.; Bailey, J.M.

    1983-01-01

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form.

  16. Steady State Tokamak Equilibria without Current Drive

    SciTech Connect

    Shaing, K.C.; Aydemir, A.Y.; Lin-Liu, Y.R.; Miller, R.L.

    1997-11-01

    Steady state tokamak equilibria without current drive are found. This is made possible by including the potato bootstrap current close to the magnetic axis. Tokamaks with this class of equilibria do not need seed current or current drive, and are intrinsically steady state. {copyright} {ital 1997} {ital The American Physical Society}

  17. Behavior of Compact Toroid in the External Magnetic Fields

    NASA Astrophysics Data System (ADS)

    Fukumoto, N.; Ioroi, A.; Nagata, M.; Uyama, T.

    1999-11-01

    We have investigated the possibility of refueling and density control of tokamak plasmas by the spheromak-type Compact Toroid (CT) injection in the JFT-2M tokamak in collaboration with JAERI. We demonstrated the CT injection into OH plasmas and observed the core penetration at B_T=0.8 T. The tokamak electron density increased ~0.2× 10^19m-3 at a rate of 2× 10^21m-3/s. We also observed the decrease of the CT velocity by the external magnetic field of the tokamak, which is applied across the CT acceleration region. We have examined the behavior of the CT translated in the external fields B_ext using the magnetic probes and the fast framing camera at Himeji Inst. of tech.. CT plasma in the acceleration region is deformed by the Lorentz force of Jg × B_ext, where Jg is the gun current for CT acceleration. The magnetic field structures of a long CT in the drift region has been revealed to be the mixed relaxed state of m=0 and m=1. Results from CT acceleration and injection in a transverse field will be presented.

  18. The TITAN reversed-field-pinch fusion reactor study

    SciTech Connect

    Not Available

    1990-01-01

    This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors.

  19. Natural current profiles in a tokamak

    SciTech Connect

    Taylor, J.B.

    1990-08-01

    In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described.

  20. Compaction Scale Up and Optimization of Cylindrical Fuel Compacts for the Next Generation Nuclear Plant

    SciTech Connect

    Jeffrey J. Einerson; Jeffrey A. Phillips; Eric L. Shaber; Scott E. Niedzialek; W. Clay Richardson; Scott G. Nagley

    2012-10-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of designed experiments have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel. Results from these experiments are included. The scale-up effort is nearing completion with the process installed and operational using nuclear fuel materials. The process is being certified for manufacture of qualification test fuel compacts for the AGR-5/6/7 experiment at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL).

  1. Status of neutron diagnostics on the experimental advanced superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Zhong, G. Q.; Hu, L. Q.; Pu, N.; Zhou, R. J.; Xiao, M.; Cao, H. R.; Zhu, Y. B.; Li, K.; Fan, T. S.; Peng, X. Y.; Du, T. F.; Ge, L. J.; Huang, J.; Xu, G. S.; Wan, B. N.

    2016-11-01

    Neutron diagnostics have become a significant means to study energetic particles in high power auxiliary heating plasmas on the Experimental Advanced Superconducting Tokamak (EAST). Several kinds of neutron diagnostic systems have been implemented for time-resolved measurements of D-D neutron flux, fluctuation, emission profile, and spectrum. All detectors have been calibrated in laboratory, and in situ calibration using 252Cf neutron source in EAST is in preparation. A new technology of digitized pulse signal processing is adopted in a wide dynamic range neutron flux monitor, compact recoil proton spectrometer, and time of flight spectrometer. Improvements will be made continuously to the system to achieve better adaptation to the EAST's harsh γ-ray and electro-magnetic radiation environment.

  2. Transport equations in tokamak plasmas

    SciTech Connect

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-15

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfven waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The 'mean field' effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  3. Transport equations in tokamak plasmasa)

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-01

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfvén waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The "mean field" effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  4. Tokamak plasma interaction with limiters

    NASA Astrophysics Data System (ADS)

    Pitcher, Charles Spencer

    1988-08-01

    The importance of plasma purity is discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fueling/recycling and impurity production. The experiments, carried out on the DITE tokomak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behavior; new physical phenomena are presented in all three areas.

  5. Breakdown in the pretext tokamak

    SciTech Connect

    Benesch, J.F.

    1981-06-01

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges.

  6. Global Hybrid Simulations of Energetic Particle Effects on the n=1 Mode in Tokamaks: Internal Kink and Fishbone Instability

    SciTech Connect

    G.Y. Fu; W. Park; H.R. Strauss; J. Breslau; J. Chen; S. Jardin; L.E. Sugiyama

    2005-08-09

    Global hybrid simulations of energetic particle effects on the n=1 internal kink mode have been carried out for tokamaks. For the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis Editors et al., Nucl. Fusion 39:2137 (1999)], it is shown that alpha particle effects are stabilizing for the internal kink mode. However, the elongation of ITER reduces the stabilization effects significantly. Nonlinear simulations of the precessional drift fishbone instability for circular tokamak plasmas show that the mode saturates due to flattening of the particle distribution function near the resonance region. The mode frequency chirps down rapidly as the flattening region expands radially outward. Fluid nonlinearity reduces the saturation level.

  7. Nuclear design of a very-low-activation fusion reactor

    SciTech Connect

    Cheng, E.T.; Hopkins, G.R.

    1983-06-01

    An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design.

  8. Enhancement of confinement in tokamaks

    SciTech Connect

    Furth, H.P.

    1986-05-01

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime.

  9. Finite pressure effects on the tokamak sawtooth crash

    SciTech Connect

    Nishimura, Yasutaro

    1998-07-01

    The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed.

  10. Feasibility study of a fission-suppressed tokamak fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  11. Dynamic simulations of the cryogenic system of a tokamak

    NASA Astrophysics Data System (ADS)

    Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

    2015-12-01

    Power generation in the next decades could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to 4.4K. The plasma works cyclically and the coil system is subjected to pulsed heat load which has to be handled by the refrigerator. By smoothing the variable loads, the refrigerator capacity can be set close to the average power; optimizing investment and operational costs. Within the “Broader Approach agreement” related to ITER project, CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) is in charge of providing the cryogenic system for the Japanese tokamak (JT-60SA), that is currently under construction in Naka. The system has been designed to handle the pulsed heat loads. To prepare the acceptance tests of the cryogenic system foreseen in 2016, both dynamic modelling and experimental tests on a scaled down mock-up are of high interest for assessing pulsed load smoothing control. After explaining HELIOS (HElium Loop for hIgh lOad Smoothing) operating modes, a dynamic model is presented, with results on the pulsed heat load scenarios. All the simulations have been performed with EcosimPro® and the associated cryogenic library CRYOLIB.

  12. Negative hydrogen ion source for TOKAMAK neutral beam injector (invited)

    NASA Astrophysics Data System (ADS)

    Okumura, Y.; Fujiwara, Y.; Kashiwagi, M.; Kitagawa, T.; Miyamoto, K.; Morishita, T.; Hanada, M.; Takayanagi, T.; Taniguchi, M.; Watanabe, K.

    2000-02-01

    Intense negative ion source producing multimegawatt hydrogen/deuterium negative ion beams has been developed for the neutral beam injector (NBI) in TOKAMAK thermonuclear fusion machines. Negative ions are produced in a cesium seeded multi-cusp plasma generator via volume and surface processes, and accelerated with a multistage electrostatic accelerator. The negative ion source for JT-60U has produced 18.5 A/360 keV (6.7 MW) H- and 14.3 A/380 keV (5.4 MW) D- ion beams at average current densities of 11 mA/cm2 (H-) and 8.5 mA/cm2 (D-). A high energy negative ion source has been developed for the next generation TOKAMAK such as the International Thermonuclear Experimental Reactor (ITER). The source has demonstrated to accelerate negative ions up to 1 MeV, the energy required for ITER. Higher negative ion current density of more than 20 mA/cm2 was obtained in the ITER concept sources. It was confirmed that the consumption rate of cesium is small enough to operate the source for a half year in ITER-NBI without maintenance.

  13. A method of interpreting the Balmer-alpha high-resolution spectroscopy for tokamak edge plasmas with account of divertor stray light

    NASA Astrophysics Data System (ADS)

    Neverov, V. S.; Kukushkin, A. B.; Alekseev, A. G.

    2016-01-01

    A method is suggested for interpreting the data from the Balmer-alpha high- resolution spectroscopy diagnostics of the edge plasma in the tokamak main chamber, which additionally uses the data from direct observation of the divertor. Such an extension of the diagnostics is motivated by the fact that in a tokamak-reactor with the metal first wall, like ITER tokamak, a significant role of the divertor stray light (DSL), which is emitted by the plasma in the divertor in the same spectral line and reflected from the first wall of the vacuum chamber to a spectrometer in the main chamber, is expected. The results of the first applications of the developed model to interpret the data from the JET-ILW tokamak experiments, which simulate the conditions of occurrence of the DSL in ITER, are discussed.

  14. Optimization of out-vessel magnetic diagnostics for plasma boundary reconstruction in tokamaks

    NASA Astrophysics Data System (ADS)

    Romero, J. A.; Svensson, J.

    2013-03-01

    To improve the low-frequency spectrum of magnetic field measurements of future tokamak reactors such as ITER, several steady-state magnetic sensor technologies have been considered. For all the studied technologies it is always advantageous to place the sensors outside the vacuum vessel and as far away from the reactor core to minimize radiation damage and temperature effects, but not so far as to compromise the accuracy of the equilibrium reconstruction. We have studied to what extent increasing the distance between out-vessel sensors and plasma can be compensated for sensor accuracy and/or density before the limit imposed by the degeneracy of the problem is reached. The study is particularized for the Swiss TCV tokamak, due to the quality of its magnetic data and its ability to operate with a wide range of plasma shapes and divertor configurations. We have scanned the plasma boundary reconstruction error as a function of out-vessel sensor density, accuracy and distance to the plasma. The study is performed for both the transient and steady-state phases of the tokamak discharge. We find that, in general, there is a broad region in the parameter space where sensor accuracy, density and proximity to the plasma can be traded for one another to obtain a desired level of accuracy in the reconstructed boundary, up to some limit. Extrapolation of the results to a tokamak reactor suggests that a hybrid configuration with sensors inside and outside the vacuum vessel could be used to obtain a good boundary reconstruction during both the transient and the flat-top of the discharges, if out-vessel magnetic sensors of sufficient density and accuracy can be placed sufficiently far outside the vessel to minimize radiation damage.

  15. Suprathermal electron studies in the TCV tokamak: Design of a tomographic hard-x-ray spectrometer

    SciTech Connect

    Gnesin, S.; Coda, S.; Decker, J.; Peysson, Y.

    2008-10-15

    Electron cyclotron resonance heating and electron cyclotron current drive, disruptive events, and sawtooth activity are all known to produce suprathermal electrons in fusion devices, motivating increasingly detailed studies of the generation and dynamics of this suprathermal population. Measurements have been performed in the past years in the tokamak a configuration variable (TCV) tokamak using a single pinhole hard-x-ray (HXR) camera and electron-cyclotron-emission radiometers, leading, in particular, to the identification of the crucial role of spatial transport in the physics of ECCD. The observation of a poloidal asymmetry in the emitted suprathermal bremsstrahlung radiation motivates the design of a proposed new tomographic HXR spectrometer reported in this paper. The design, which is based on a compact modified Soller collimator concept, is being aided by simulations of tomographic reconstruction. Quantitative criteria have been developed to optimize the design for the greatly variable shapes and positions of TCV plasmas.

  16. Integrated modeling of temperature profiles in L-mode tokamak discharges

    SciTech Connect

    Rafiq, T.; Kritz, A. H.; Tangri, V.; Pankin, A. Y.; Voitsekhovitch, I.; Budny, R. V.

    2014-12-15

    Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

  17. Mouse Embryo Compaction.

    PubMed

    White, M D; Bissiere, S; Alvarez, Y D; Plachta, N

    2016-01-01

    Compaction is a critical first morphological event in the preimplantation development of the mammalian embryo. Characterized by the transformation of the embryo from a loose cluster of spherical cells into a tightly packed mass, compaction is a key step in the establishment of the first tissue-like structures of the embryo. Although early investigation of the mechanisms driving compaction implicated changes in cell-cell adhesion, recent work has identified essential roles for cortical tension and a compaction-specific class of filopodia. During the transition from 8 to 16 cells, as the embryo is compacting, it must also make fundamental decisions regarding cell position, polarity, and fate. Understanding how these and other processes are integrated with compaction requires further investigation. Emerging imaging-based techniques that enable quantitative analysis from the level of cell-cell interactions down to the level of individual regulatory molecules will provide a greater understanding of how compaction shapes the early mammalian embryo. PMID:27475854

  18. Results from deuterium-tritium tokamak confinement experiments

    SciTech Connect

    Hawryluk, R.J.

    1997-02-01

    Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues.

  19. Composition And Electrical Properties Of Dust From Tokamak Compass

    SciTech Connect

    Vaverka, J.; Beranek, M.; Pavlu, J.; Richterova, I.; Vysinka, M.; Safrankova, J.; Nemecek, Z.

    2011-11-29

    In spite of the fact that fusion is a subject of the study for many years, there are still a lot of open questions. One of the interesting topics in fusion research is a presence of dust grains in reactors. In the paper, dust grains born in tokamak Compass are studied and compared with samples of a spherical geometry and well known composition. A unique experimental setup was used for investigations of charging properties of such grains and the SEM and EDX spectroscopy was applied for a study of grain composition. We focus on the secondary emission because this process plays a prominent role when a portion of energetic electrons is present in surroundings of a particular grain. It was shown that depending on the grain size and material, energetic electrons charge the grains to positive potentials comparable with the energy of impinging electrons.

  20. Design and installation of a ferromagnetic wall in tokamak geometry.

    PubMed

    Hughes, P E; Levesque, J P; Rivera, N; Mauel, M E; Navratil, G A

    2015-10-01

    Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak-Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability.

  1. Turbulence-driven bootstrap current in low-collisionality tokamaks.

    PubMed

    McDevitt, C J; Tang, Xian-Zhu; Guo, Zehua

    2013-11-15

    Neoclassical bootstrap current is expected to provide a significant fraction of the equilibrium plasma current in tokamak reactors. Here we report a novel mechanism through which a bootstrap current may be driven even in a collisionless plasma. In analogy with the neoclassical mechanism, in which the collisional equilibrium established between trapped and passing electrons produces a steady state current, we show that resonant scattering of electrons by drift wave microturbulence provides an additional means of determining the equilibrium between trapped and passing electrons and thus driving a bootstrap current. Employing a linearized Fokker-Planck collision operator, the plasma current in the presence of both collisions and resonant electron scattering is computed, allowing for the relative strength of these two mechanisms to be quantified as a function of collisionality and fluctuation amplitude.

  2. Power and particle exhaust in tokamaks

    SciTech Connect

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER`s nominal design positions; important directions for further research are identified.

  3. Reversible DNA compaction.

    PubMed

    González-Pérez, Alfredo

    2014-01-01

    In this review we summarize and discuss the different methods we can use to achieve reversible DNA compaction in vitro. Reversible DNA compaction is a natural process that occurs in living cells and viruses. As a result these process long sequences of DNA can be concentrated in a small volume (compacted) to be decompacted only when the information carried by the DNA is needed. In the current work we review the main artificial compacting agents looking at their suitability for decompaction. The different approaches used for decompaction are strongly influenced by the nature of the compacting agent that determines the mechanism of compaction. We focus our discussion on two main artificial compacting agents: multivalent cations and cationic surfactants that are the best known compacting agents. The reversibility of the process can be achieved by adding chemicals like divalent cations, alcohols, anionic surfactants, cyclodextrins or by changing the chemical nature of the compacting agents via pH modifications, light induced conformation changes or by redox-reactions. We stress the relevance of electrostatic interactions and self-assembly as a main approach in order to tune up the DNA conformation in order to create an on-off switch allowing a transition between coil and compact states. The recent advances to control DNA conformation in vitro, by means of molecular self-assembly, result in a better understanding of the fundamental aspects involved in the DNA behavior in vivo and serve of invaluable inspiration for the development of potential biomedical applications. PMID:24444152

  4. Dust measurements in tokamaks (invited)

    SciTech Connect

    Rudakov, D. L.; Yu, J. H.; Boedo, J. A.; Hollmann, E. M.; Krasheninnikov, S. I.; Moyer, R. A.; Muller, S. H.; Pigarov, A. Yu.; Rosenberg, M.; Smirnov, R. D.; West, W. P.; Boivin, R. L.; Bray, B. D.; Brooks, N. H.; Hyatt, A. W.; Wong, C. P. C.; Roquemore, A. L.; Skinner, C. H.; Solomon, W. M.; Ratynskaia, S.

    2008-10-15

    Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 {mu}m in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C{sub 2} dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  5. Design study status of compact containment BWR

    SciTech Connect

    Heki, H.; Nakamaru, M.; Kuroki, M.; Kojima, Y.; Arai, K.; Tahara, M.; Hoshi, T.

    2006-07-01

    The reactor concept considered in this paper has a relatively mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Japan Atomic Power Company has been taking initiative in developing the concept of the Compact Containment Boiling Water Reactor (CCR). The CCR., which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's relatively mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, upper entry control rod drives (CRDs) and simplified safety system with high pressure resistible containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The upper entry CRDs enable a simplified safety system followed by in-vessel retention (IVR) capability with the compact primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration of RPV below the top of the core height, passive cooling system by isolation condenser (IC). The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. Further core design study has been carried out taking into account compact reactor size and reduction of fuel

  6. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    SciTech Connect

    Gentile, C.A.; LaMarche, P.H.; Anderson, J.L.

    1995-07-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.

  7. First-wall/blanket materials selection for STARFIRE tokamak reactor

    SciTech Connect

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed.

  8. Fusion energy calorimeter for the tokamak fusion test reactor

    SciTech Connect

    Jassby, D.L.; Imel, G.R.

    1981-04-01

    One and two-dimensional neutronic analyses treating the transport and scattering of neutrons and the production and transport of gamma rays in the TFTR demonstrate that the fusion energy production in a D-T pulse in the TFTR can be determined with an uncertainty of +- 15% or less, simply by integrating the measured profile of temperature increase along the central radial axis of a large hydrocarbon moderator that fills the bay between adjacent toroidal-field coils, just outside the vacuum vessel. Limitations in thermopile temperature measurements dictate a minimum fusion-neutron fluence at the vacuum vessel of the order of 10/sup 12/ n/cm/sup 2/ per pulse (a source strength of 10/sup 18/ n/pulse in TFTR), in order that this simple calorimeter can provide useful accuracy.

  9. Neutron spectroscopy on TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Nishitani, T.; Strachan, J.D.

    1988-05-01

    This paper describes the use of an /sup 3/He ionization chamber for neutron spectroscopy on TFTR during 1987. The ion temperature was measured using neutron spectroscopy for one set of ohmically heated plasmas. The deduced ion temperatures agreed to within 20% with those measured by other diagnostics. 11 refs., 11 figs., 1 tab.

  10. TFTR (Tokamak Fusion Test Reactor) neutral beam injected power measurement

    SciTech Connect

    Kamperschroer, J.H.; Grisham, L.R.; Dudek, L.E.; Gammel, G.M.; Johnson, G.A.; Kugel, H.W.; Lagin, L.; O'Connor, T.E.; Shah, P.A.; Sichta, P.

    1989-05-01

    Energy flow within TFTR neutral beamlines is measured with a waterfall calorimetry system capable of simultaneously measuring the energy deposited within four heating beamlines (three ion sources each), or of measuring the energy deposited in a separate neutral beam test stand. Of the energy extracted from the ion source in the well instrumented test stand, 99.5 +- 3.5% can be accounted for. When the ion deflection magnet is energized, however, 6.5% of the extracted energy is lost. This loss is attributed to a spray of devious particles onto unmonitored surfaces. A 30% discrepancy is also observed between energy measurements on the internal beamline calorimeter and energy measurements on a calorimeter located in the test stand target chamber. Particle reflection from the flat plate calorimeter in the target chamber, which the incident beam strikes at a near-grazing angle of 12/degree/, is the primary loss of this energy. A slight improvement in energy accountability is observed as the beam pulse length is increased. This improvement is attributed to systematic error in the sensitivity of the energy measurement to small fluctuations on the supply water temperature. An overall accuracy of 15% is estimated for the total power injected into TFTR. Contributions to this error are uncertainties in the beam neutralization efficiency, reionization and beam scrape-off in the drift duct, and fluctuations in the temperature of the supply water. 28 refs., 9 figs., 1 tab.

  11. Preliminary experiment of high-speed gas flow generation by a compact toroid injection into a gas neutralizer

    NASA Astrophysics Data System (ADS)

    Ito, Y.; Liu, D.; Shoji, T.; Nakanishi, R.; Fukumoto, N.; Sekioka, T.; Kikuchi, Y.; Nagata, M.

    2007-11-01

    A supersonic gas jet injection has been considered to be a new technique for future reactor fuelling and disruption mitigation in tokamak devices [1]. We have recently started to investigate a production of high-speed gas flow by using a compact toroid (CT) injection into a hydrogen gas neutralizer. The electron density of the CT plasma is 1˜4 x10^21 m-3, and the CT speed is 30˜70 km/s in the preliminary experiment. The kinetic-energy measurements of ions and neutrals after the neutralization were carried out by using an electrostatic ion energy analyzer and time-of-flight technique. An enhancement of the Hβ emission level, a significant decay of the CT plasma density and the magnetic field profile have been observed after the neutralization when the neutral pressure is about 10-3 Torr. It could be considered that high-energy neutral particles were generated by a charge exchange process from the CT plasma to the neutral particles. [1] V. Rozhansky, et al., Nucl. Fusion 46, 367 (2006).

  12. Do spherical tokamaks have a thermonuclear future?

    NASA Astrophysics Data System (ADS)

    Mirnov, S. V.

    2012-12-01

    This work has been initiated by the publication of a review by B.V.Kuteev et al., "Intense Fusion Neutron Sources" [Plasma Physics Reports 36, 281 (2010)]. It is stated that the key thesis of the above review that a spherical tokamak can be recommended for research neutron sources and for demonstration hybrid systems as an alternative to expensive "classical" tokamaks of the JET and ITER type is inconsistent. The analysis of the experimental material obtained during the last 10 years in the course of studies on the existing spherical tokamaks shows that the TIN-ST fusion neutron source spherical tokamak proposed by the authors of the review and intended, according to the authors' opinion, to replace "monsters" in view of its table-top dimensions (2 m3) and laboratory-level energetics cannot be transformed into any noticeable stationary megawatt-power neutron source competing with the existing classical tokamaks (in particular, with JET with its quasi-steady DT fusion power at a level of 5 MW). Namely, the maximum plasma current in the proposed tokamak will be not 3 MA, as the authors suppose erroneously, but, according to the present-day practice of spherical tokamaks, within 0.6-0.7 MA, which will lead to a reduction on the neutron flux by two to three orders of magnitude from the expected 5 MW. The possibility of the maintenance of the stationary process itself even in such a "weakened" spherical tokamak is very doubtful. The experience of the largest existing devices of this type (such as NSTX and MAST) has shown that they are incapable of operating even in a quasi-steady operating mode, because the discharge in them is spontaneously interrupted about 1 s after the beginning of the current pulse, although its expected duration is of up to 5 s. The nature of this phenomenon is the subject of further study of the physics of spherical tokamaks. This work deals with a critical analysis of the available experimental data concerning such tokamaks and a discussion of

  13. Mechanics of tissue compaction.

    PubMed

    Turlier, Hervé; Maître, Jean-Léon

    2015-12-01

    During embryonic development, tissues deform by a succession and combination of morphogenetic processes. Tissue compaction is the morphogenetic process by which a tissue adopts a tighter structure. Recent studies characterized the respective roles of cells' adhesive and contractile properties in tissue compaction. In this review, we formalize the mechanical and molecular principles of tissue compaction and we analyze through the prism of this framework several morphogenetic events: the compaction of the early mouse embryo, the formation of the fly retina, the segmentation of somites and the separation of germ layers during gastrulation.

  14. Influence of collisions on parametric instabilities induced by lower hybrid waves in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Castaldo, C.; Di Siena, A.; Fedele, R.; Napoli, F.; Amicucci, L.; Cesario, R.; Schettini, G.

    2016-01-01

    Parametric instabilities induced at the plasma edge by lower hybrid wave power externally coupled to tokamak plasmas have, via broadening of the antenna spectrum, strong influence on the power deposition and current drive in the core. For modeling the parametric instabilities at the tokamak plasma edge in lower hybrid current drive experiments, the effect of the collisions has been neglected so far. In the present work, a specific collisional parametric dispersion relation, useful to analyze these nonlinear phenomena near the lower hybrid antenna mouth, is derived for the first time, based on a kinetic model. Numerical solutions show that in such cold plasma regions the collisions prevent the onset of the parametric instabilities. This result is important for present lower hybrid current drive experiments, as well as in fusion reactor scenarios.

  15. Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

    PubMed

    Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

    2012-01-01

    Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas. PMID:22233634

  16. Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

    PubMed

    Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

    2012-01-10

    Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

  17. Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Graves, J. P.; Chapman, I. T.; Coda, S.; Lennholm, M.; Albergante, M.; Jucker, M.

    2012-01-01

    Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

  18. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    SciTech Connect

    Chektybayev, B. Shapovalov, G.; Kolodeshnikov, A.

    2015-05-15

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  19. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    NASA Astrophysics Data System (ADS)

    Chektybayev, B.; Shapovalov, G.; Kolodeshnikov, A.

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  20. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle.

    PubMed

    Chektybayev, B; Shapovalov, G; Kolodeshnikov, A

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  1. Observation of Energetic Particle Driven Modes Relevant to Advanced Tokamak Regimes

    SciTech Connect

    R. Nazikian; B. Alper; H.L. Berk; D. Borba; C. Boswell; R.V. Budny; K.H. Burrell; C.Z. Cheng; E.J. Doyle; E. Edlund; R.J. Fonck; A. Fukuyama; N.N. Gorelenkov; C.M. Greenfield; D.J. Gupta; M. Ishikawa; R.J. Jayakumar; G.J. Kramer; Y. Kusama; R.J. La Haye; G.R. McKee; W.A. Peebles; S.D. Pinches; M. Porkolab; J. Rapp; T.L. Rhodes; S.E. Sharapov; K. Shinohara; J.A. Snipes; W.M. Solomon; E.J. Strait; M. Takechi; M.A. Van Zeeland; W.P. West; K.L. Wong; S. Wukitch; L. Zeng

    2004-10-21

    Measurements of high-frequency oscillations in JET [Joint European Torus], JT-60U, Alcator C-Mod, DIII-D, and TFTR [Tokamak Fusion Test Reactor] plasmas are contributing to a new understanding of fast ion-driven instabilities relevant to Advanced Tokamak (AT) regimes. A model based on the transition from a cylindrical-like frequency-chirping mode to the Toroidal Alfven Eigenmode (TAE) has successfully encompassed many of the characteristics seen in experiments. In a surprising development, the use of internal density fluctuation diagnostics has revealed many more modes than has been detected on edge magnetic probes. A corollary discovery is the observation of modes excited by fast particles traveling well below the Alfven velocity. These observations open up new opportunities for investigating a ''sea of Alfven Eigenmodes'' in present-scale experiments, and highlight the need for core fluctuation and fast ion measurements in a future burning-plasma experiment.

  2. Selection of materials for tokamak plasma facing elements based on a liquid tin capillary pore system

    NASA Astrophysics Data System (ADS)

    Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu; Sevryukov, O. N.; Dzhumaev, P. S.; Shumskiy, V. A.; Ivannikov, A. A.

    2016-09-01

    Capillary-Pore Systems (CPS) filled by liquid metals are considered as an alternative solution of materials choice for plasma facing component of tokamak reactor. Tin is viewed as one of the candidates for CPS because it has lower corrosiveness than gallium and lower saturated vapour pressure compared to lithium. The corrosion resistance of Mo, Nb and W in pure liquid tin was investigated. The corrosion tests were carried out in the static isothermal conditions at a temperature up to 1050°C. As a result of the corrosion study, it was found that Mo does not corrode in liquid Sn, as opposed to Nb and is compatible with liquid tin in temperatures of up to approx. 1000°C. This allows considering Mo as an alloy base material of the in-vessel tokamak elements based on liquid tin capillary pore systems.

  3. Quasi-Isotropic Approximation of Geometrical Optics Method as Adequate Electrodynamical Basis for Tokamak Plasma Polarimetry

    NASA Astrophysics Data System (ADS)

    Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco

    Basic principles and recent findings of quasi-isotropic approximation (QIA) of a geometrical optics method are presented in a compact manner. QIA was developed in 1969 to describe electromagnetic waves in weakly anisotropic media. QIA represents the wave field as a power series in two small parameters, one of which is a traditional geometrical optics parameter, equal to wavelength ratio to plasma characteristic scale, and the other one is the largest component of anisotropy tensor. As a result, "" QIA ideally suits to tokamak polarimetry/interferometry systems in submillimeter range, where plasma manifests properties of weakly anisotropic medium.

  4. Theoretical design of a compact energy recovering divertor

    NASA Astrophysics Data System (ADS)

    Baver, D. A.

    2015-11-01

    An energy recovering divertor (ERD) is a type of plasma direct converter (PDC) designed to fit in the divertor channel of a tokamak. Such a device reduces the heat load to the divertor plate by converting a portion of it into electrical energy. This recovered energy can then be used for auxiliary heating and current drive, fundamentally altering the relationship between scientific and engineering breakeven and reducing dependence on bootstrap current. Previous work on the ERD concept focused on amplification of Alfven waves in a manner similar to a free-electron laser. While conceptually straightforward, this concept was also bulky, thus limiting its applicability to existing tokamak experiments. A design is presented for an ERD based on sheath-localized waves. This makes possible a device sufficiently compact to fit in the divertor channel of many existing tokamak experiments, and moreover requires no new shaping coils to achieve the desired magnetic geometry or topology. In addition, incidental advantages of this concept will be discussed.

  5. Tokamak Diagnostics Using Fusion Products.

    NASA Astrophysics Data System (ADS)

    Heidbrink, William Walter

    Measurements of neutrons and protons produced by the d(d,n)('3)He, d(t,n)(alpha), d(d,p)t, and d(('3)He,p)(alpha) fusion reactions are used to diagnose plasmas in the PLT and PDX tokamaks. An expression for the efficiency of proton detection is derived and confirmed experimentally. The time evolution of the ('3)He density indicates that a scoop limiter may pump ('3)He from the plasma faster than conventional limiters. The confinement of 1.0 MeV tritons and of 0.8 MeV ('3)He ions is studied by measuring the fraction of these fusion-produced ions that burn up in subsequent fusion reactions. In discharges with sawtooth activity and with B(,(phi)) > 2 T, the triton and ('3)He 'burnup' is consistent (within a factor of three) with predictions based on classical theories of ion confinement and slowing down. In discharges with large m = 2 or fishbone instabilities, the ('3)He burnup is less than classically predicted and, in PLT discharges at B(,(phi)) = 1.8 T, the triton burnup is over an order of magnitude smaller than predicted. Expressions for the energy spectrum of ions produced in beam-target fusion reactions are derived. Collimated measurements of the spectrum of 15 MeV protons produced by reactions between energetic ('3)He ions and relatively cold deuterons during fast wave minority heating indicate that the velocity distribution of fast ('3)He ions is peaked perpendicular to the tokamak magnetic field. The ion temperature profile and density of fast deuterons are measured with an array of collimated 3 MeV proton detectors. The fast ions produced by neutral beam injection and by launching lower hybrid waves are concentrated near the magnetic axis. Poloidal field measurements using 3 MeV protons also appear possible. In discharges in which the line radiation from central impurities does not decay, the plasma current profile is broader than in more typical discharges.

  6. Bifurcated helical core equilibrium states in tokamaks

    NASA Astrophysics Data System (ADS)

    Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

    2013-07-01

    Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

  7. Compaction properties of isomalt.

    PubMed

    Bolhuis, Gerad K; Engelhart, Jeffrey J P; Eissens, Anko C

    2009-08-01

    Although other polyols have been described extensively as filler-binders in direct compaction of tablets, the polyol isomalt is rather unknown as pharmaceutical excipient, in spite of its description in all the main pharmacopoeias. In this paper the compaction properties of different types of ispomalt were studied. The types used were the standard product sieved isomalt, milled isomalt and two types of agglomerated isomalt with a different ratio between 6-O-alpha-d-glucopyranosyl-d-sorbitol (GPS) and 1-O-alpha-d-glucopyranosyl-d-mannitol dihydrate (GPM). Powder flow properties, specific surface area and densities of the different types were investigated. Compactibility was investigated by compression of the tablets on a compaction simulator, simulating the compression on high-speed tabletting machines. Lubricant sensitivity was measured by compressing unlubricated tablets and tablets lubricated with 1% magnesium stearate on an instrumented hydraulic press. Sieved isomalt had excellent flow properties but the compactibility was found to be poor whereas the lubricant sensitivity was high. Milling resulted in both a strong increase in compactibility as an effect of the higher surface area for bonding and a decrease in lubricant sensitivity as an effect of the higher surface area to be coated with magnesium stearate. However, the flow properties of milled isomalt were too bad for use as filler-binder in direct compaction. Just as could be expected, agglomeration of milled isomalt by fluid bed agglomeration improved flowability. The good compaction properties and the low lubricant sensitivity were maintained. This effect is caused by an early fragmentation of the agglomerated material during the compaction process, producing clean, lubricant-free particles and a high surface for bonding. The different GPS/GPM ratios of the agglomerated isomalt types studied had no significant effect on the compaction properties. PMID:19327398

  8. Stabilization of compactible waste

    SciTech Connect

    Franz, E.M.; Heiser, J.H. III; Colombo, P.

    1990-09-01

    This report summarizes the results of series of experiments performed to determine the feasibility of stabilizing compacted or compactible waste with polymers. The need for this work arose from problems encountered at disposal sites attributed to the instability of this waste in disposal. These studies are part of an experimental program conducted at Brookhaven National Laboratory (BNL) investigating methods for the improved solidification/stabilization of DOE low-level wastes. The approach taken in this study was to perform a series of survey type experiments using various polymerization systems to find the most economical and practical method for further in-depth studies. Compactible dry bulk waste was stabilized with two different monomer systems: styrene-trimethylolpropane trimethacrylate (TMPTMA) and polyester-styrene, in laboratory-scale experiments. Stabilization was accomplished by wetting or soaking compactible waste (before or after compaction) with monomers, which were subsequently polymerized. Three stabilization methods are described. One involves the in-situ treatment of compacted waste with monomers in which a vacuum technique is used to introduce the binder into the waste. The second method involves the alternate placement and compaction of waste and binder into a disposal container. In the third method, the waste is treated before compaction by wetting the waste with the binder using a spraying technique. A series of samples stabilized at various binder-to-waste ratios were evaluated through water immersion and compression testing. Full-scale studies were conducted by stabilizing two 55-gallon drums of real compacted waste. The results of this preliminary study indicate that the integrity of compacted waste forms can be readily improved to ensure their long-term durability in disposal environments. 9 refs., 10 figs., 2 tabs.

  9. Microtearing modes in tokamak discharges

    NASA Astrophysics Data System (ADS)

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

  10. Upgrades for the TCV tokamak

    NASA Astrophysics Data System (ADS)

    Duval, Basil; TCV Team

    2013-10-01

    Major upgrades are being implemented on the TCV tokamak to extend its operational domain towards a burning plasma regime. The goals of obtaining high normalized plasma beta and comparable ion and electron temperatures will be achieved with the addition of a 1 MW neutral heating system and 2 MW additional third harmonic EC power. Spatial constraints together with beam occlusion required severe design optimization and the additional of a new large tangential port on the TCV vessel. For EC, the existing vertical launch mirror will be sufficient but new 1MW EC units will be employed with the legacy X3 systems modified for lateral launch. The modifications will not affect TCV's strong RT shaping and EC actuator ranges or the open divertor vacuum chamber that permits access to Snowflake divertor or doublet configurations although some wall protection enhancement is envisaged. TCV can then contribute to disentangling effects of electron-ion coupling, rotation, current and density profile control all as a function of shape in L and H-modes with ITER (or higher) values of plasma beta. Together with fast-ion physics, TCV will also be able to explore heat, particle and momentum transport and turbulence effects in electron-heat dominated discharges for Te/Ti in the (0.02 to 3) range.

  11. Tokamak x ray diagnostic instrumentation

    SciTech Connect

    Hill, K.W.; Beiersdorfer, P.; Bitter, M.; Fredrickson, E.; Von Goeler, S.; Hsuan, H.; Johnson, L.C.; Liew, S.L.; McGuire, K.; Pare, V.

    1987-01-01

    Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/..delta..E is greater than 10/sup 4/ and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally.

  12. Toroidal Flow in Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Cole, A. J.; Hegna, C. C.

    2007-11-01

    Many effects influence toroidal flow evolution in tokamak plasmas. Momentum sources and radial diffusion due to axisymmetric neoclassical, paleoclassical and anomalous transport are usually considered. In addition, the toroidal flow can be affected by field errors. Small, non-axisymmetric field errors arise from coil irregularities, active control coils and collective plasma magnetic distortions (e.g., NTMs, RWMs). Resonant field errors cause localized electromagnetic torques near rational surfaces in the plasma, which can lock the plasma to the wall leading to magnetic islands and reduced confinement or disruptions. Their penetration into the plasma is limited by flow-shielding effects; but they can be amplified by the plasma response at high beta. Non-resonant field errors cause magnetic pumping and radial banana drifts, and lead to toroidal flow damping over the entire plasma. Many of these processes can also produce momentum pinch and intrinsic flow effects. This poster will seek to present a coherent picture of all these effects and suggest ways they could be tested and distinguished experimentally.

  13. Polarimetric spectra analysis for tokamak pitch angle measurements

    NASA Astrophysics Data System (ADS)

    Ko, J.; Chung, J.; Lange, A. G. G.; de Bock, M. F. M.

    2013-10-01

    Measurements of the internal magnetic field structures using conventional polarimetric approaches are considered extremely challenging in fusion-reactor environments whereas the information on current density profiles is essential to establish steady-state and advance operation scenarios in such reactor-relevant devices. Therefore, on ITER a hybrid system is proposed for the current density measurements that uses both polarimetry and spectral measurements. The spectrum-based approaches have been tested in the Korea Superconducting Tokamak Advanced Research (KSTAR) during the past two plasma campaigns. As such, KSTAR is a test-bed for the proposed ITER hybrid system. Measurements in the plasma core are based on the motional Stark effect (MSE) spectrum of the neutral beam emission. For the edge profiles, the Zeeman effect (ZE) acting on the lithium emission spectrum of the newly installed (2013) Lithium-beam-diagnostic is exploited. The neutral beam emission spectra, complicated by the multi-ion-source beam injection, are successfully fitted making use of the data provided by the Atomic Data and Analysis Structure (ADAS) database package. This way pitch angle profiles could be retrieved from the beam emission spectra. With the same spectrometer/CCD hardware as on MSE, but with a different wavelength range and different lines of sight, the first ZE spectrum measurements have been made. The Zeeman splitting comparable to and greater than the instrumental broadening has been routinely detected at high toroidal field operations ( ~ 3 Tesla).

  14. REACTOR CONTROL DEVICE

    DOEpatents

    Graham, R.H.

    1962-09-01

    A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

  15. Merchant Marine Ship Reactor

    DOEpatents

    Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

    1961-05-01

    A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

  16. MERCHANT MARINE SHIP REACTOR

    DOEpatents

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  17. Helicity content and tokamak applications of helicity

    SciTech Connect

    Boozer, A.H.

    1986-05-01

    Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities.

  18. Physics of Tokamak Plasma Start-up

    NASA Astrophysics Data System (ADS)

    Mueller, Dennis

    2012-10-01

    This tutorial describes and reviews the state-of-art in tokamak plasma start-up and its importance to next step devices such as ITER, a Fusion Nuclear Science Facility and a Tokamak/ST demo. Tokamak plasma start-up includes breakdown of the initial gas, ramp-up of the plasma current to its final value and the control of plasma parameters during those phases. Tokamaks rely on an inductive component, typically a central solenoid, which has enabled attainment of high performance levels that has enabled the construction of the ITER device. Optimizing the inductive start-up phase continues to be an area of active research, especially in regards to achieving ITER scenarios. A new generation of superconducting tokamaks, EAST and KSTAR, experiments on DIII-D and operation with JET's ITER-like wall are contributing towards this effort. Inductive start-up relies on transformer action to generate a toroidal loop voltage and successful start-up is determined by gas breakdown, avalanche physics and plasma-wall interaction. The goal of achieving steady-sate tokamak operation has motivated interest in other methods for start-up that do not rely on the central solenoid. These include Coaxial Helicity Injection, outer poloidal field coil start-up, and point source helicity injection, which have achieved 200, 150 and 100 kA respectively of toroidal current on closed flux surfaces. Other methods including merging reconnection startup and Electron Bernstein Wave (EBW) plasma start-up are being studied on various devices. EBW start-up generates a directed electron channel due to wave particle interaction physics while the other methods mentioned rely on magnetic helicity injection and magnetic reconnection which are being modeled and understood using NIMROD code simulations.

  19. Beta limits in long-pulse tokamak discharges

    NASA Astrophysics Data System (ADS)

    Sauter, O.; La Haye, R. J.; Chang, Z.; Gates, D. A.; Kamada, Y.; Zohm, H.; Bondeson, A.; Boucher, D.; Callen, J. D.; Chu, M. S.; Gianakon, T. A.; Gruber, O.; Harvey, R. W.; Hegna, C. C.; Lao, L. L.; Monticello, D. A.; Perkins, F.; Pletzer, A.; Reiman, A. H.; Rosenbluth, M.; Strait, E. J.; Taylor, T. S.; Turnbull, A. D.; Waelbroeck, F.; Wesley, J. C.; Wilson, H. R.; Yoshino, R.

    1997-05-01

    The maximum normalized beta achieved in long-pulse tokamak discharges at low collisionality falls significantly below both that observed in short pulse discharges and that predicted by the ideal MHD theory. Recent long-pulse experiments, in particular those simulating the International Thermonuclear Experimental Reactor (ITER) [M. Rosenbluth et al., Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1995), Vol. 2, p. 517] scenarios with low collisionality νe*, are often limited by low-m/n nonideal magnetohydrodynamic (MHD) modes. The effect of saturated MHD modes is a reduction of the confinement time by 10%-20%, depending on the island size and location, and can lead to a disruption. Recent theories on neoclassical destabilization of tearing modes, including the effects of a perturbed helical bootstrap current, are successful in explaining the qualitative behavior of the resistive modes and recent results are consistent with the size of the saturated islands. Also, a strong correlation is observed between the onset of these low-m/n modes with sawteeth, edge localized modes (ELM), or fishbone events, consistent with the seed island required by the theory. We will focus on a quantitative comparison between both the conventional resistive and neoclassical theories, and the experimental results of several machines, which have all observed these low-m/n nonideal modes. This enables us to single out the key issues in projecting the long-pulse beta limits of ITER-size tokamaks and also to discuss possible plasma control methods that can increase the soft β limit, decrease the seed perturbations, and/or diminish the effects on confinement.

  20. Overview of physics research on the TCV tokamak

    NASA Astrophysics Data System (ADS)

    Fasoli, A.; TCV Team

    2009-10-01

    The Tokamak à Configuration Variable (TCV) tokamak is equipped with high-power (4.5 MW), real-time-controllable EC systems and flexible shaping, and plays an important role in fusion research by broadening the parameter range of reactor relevant regimes, by investigating tokamak physics questions and by developing new control tools. Steady-state discharges are achieved, in which the current is entirely self-generated through the bootstrap mechanism, a fundamental ingredient for ITER steady-state operation. The discharge remains quiescent over several current redistribution times, demonstrating that a self-consistent, 'bootstrap-aligned' equilibrium state is possible. Electron internal transport barrier regimes sustained by EC current drive have also been explored. MHD activity is shown to be crucial in scenarios characterized by large and slow oscillations in plasma confinement, which in turn can be modified by small Ohmic current perturbations altering the barrier strength. In studies of the relation between anomalous transport and plasma shape, the observed dependences of the electron thermal diffusivity on triangularity (direct) and collisionality (inverse) are qualitatively reproduced by non-linear gyro-kinetic simulations and shown to be governed by TEM turbulence. Parallel SOL flows are studied for their importance for material migration. Flow profiles are measured using a reciprocating Mach probe by changing from lower to upper single-null diverted equilibria and shifting the plasmas vertically. The dominant, field-direction-dependent Pfirsch-Schlüter component is found to be in good agreement with theoretical predictions. A field-direction-independent component is identified and is consistent with flows generated by transient over-pressure due to ballooning-like interchange turbulence. Initial high-resolution infrared images confirm that ELMs have a filamentary structure, while fast, localized radiation measurements reveal that ELM activity first appears

  1. Electron cyclotron emission diagnostics on KSTAR tokamak.

    PubMed

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  2. Electron cyclotron emission diagnostics on KSTAR tokamak

    SciTech Connect

    Jeong, S. H.; Lee, K. D.; Kwon, M.; Kogi, Y.; Kawahata, K.; Nagayama, Y.; Mase, A.

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  3. Origin of Tokamak Density Limit Scalings

    NASA Astrophysics Data System (ADS)

    Gates, D. A.; Delgado-Aparicio, L.

    2012-04-01

    The onset criterion for radiation driven islands [P. H. Rebut and M. Hugon, Plasma Physics and Controlled Nuclear Fusion Research 1984: Proc. 10th Int. Conf. London, 1984, (IAEA, Vienna, 1985), Vol. 2] in combination with a simple cylindrical model of tokamak current channel behavior is consistent with the empirical scaling of the tokamak density limit [M. Greenwald, Nucl. Fusion 28, 2199 (1988)NUFUAU0029-551510.1088/0029-5515/28/12/009]. Many other unexplained phenomena at the density limit are consistent with this novel physics mechanism.

  4. MHD stable regime of the tokamak

    SciTech Connect

    Cheng, C.Z.; Furth, H.P.; Boozer, A.H.

    1986-10-01

    A broad family of tokamak current profiles is found to be stable against ideal and resistive MHD kink modes for 1 less than or equal to q(0), with q(a) as low 2. For 0.5 less than or equal to q(0) < and q(a) > 1, current profiles can be found that are unstable only to the m = 1, n = 1 mode. A specific ''optimal'' tokamak profile can be selected from the range of stable solutions, by imposing a common upper limit on dj/dr - corresponding in ohmic equilibrium to a limitation of dT/sub e//dr by anomalous transport.

  5. Optimization of turn position of tokamak inductor

    NASA Astrophysics Data System (ADS)

    Aristov, Yu. A.; Vorobev, G. M.; Kuznetsov, A. V.

    Statement and methods of solution of the problem of optimizing turn position of tokamak induction are considered. Optimization is aimed at determination of inductor turn position, providing the minimal scattering of magnetic field in the region of chamber at any assigned value of volt-seconds. Algorithms of problem solution are described, and results of calculations for STX tokamak are presented. It is shown that development methods can be used for determining optimal position of turns of any coils of poloidal magnetic field, providing the assigned configuration and level of magnetic field.

  6. Tokamak D T fusion neutron source requirements for closing the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Stacey, W. M.

    2007-03-01

    This paper summarizes a series of conceptual design studies conducted with the purpose of determining if tokamak fusion neutron sources based on ITER physics and technology could meet the neutron source requirements for sub-critical fast-spectrum nuclear reactors that would help to close the nuclear fuel cycle by transmuting the transuranics in spent nuclear fuel. The studies were constrained to nuclear reactor and materials technologies under consideration in the US nuclear programme. Fuel cycle studies indicate that fusion neutron sources in the range ~200-500 MW would meet the needs of transmutation reactors, depending on other constraints such as materials damage to the nuclear fuel. A tokamak with R = 3.75 m, a = 1.1 m, B = 5.7-5.9 T, q95 = 3.00-4.0, I = 8.3-10 MA, βN = 2.0-2.85, HIPB98 = 1.0-1.06, γcd = 0.6 A Wm-2 would meet these requirements.

  7. Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

    SciTech Connect

    Majeski, R.

    2010-05-20

    At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700 deg. C. However, at a sufficiently high operating temperature (700 - 1000 deg. C), tungsten is self-annealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment.The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium.Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands and fusion experiments to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 deg. C), it has been now been used as a PFC in several confinement experiments (TFTR, T11-M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

  8. Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

    SciTech Connect

    R. Majeski

    2010-01-15

    At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700oC. However, at a sufficiently high operating temperature (700 - 1000 oC), tungsten is selfannealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment. The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium. Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands1 and fusion experiments2,3 to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 oC), it has been now been used as a PFC in several confinement experiments (TFTR, T11- M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

  9. Rotation of tokamak halo currents

    SciTech Connect

    Boozer, Allen H.

    2012-05-15

    During tokamak disruptions, halo currents, which can be tenths of the total plasma current, can flow at the plasma edge along the magnetic field lines that intercept the chamber walls. Non-axisymmetric halo currents are required to maintain force balance as the plasma kinks when the edge safety factor drops to about two in a vertical displacement event. The plasma quickly assumes a definite toroidal velocity v{sub a}(r) with respect to that of the magnetic kink, v{sub k}, where v{sub a}(r) is set by the radial electric field required for ambipolarity. The plasma velocity, v{sub pl}=v{sub a}+v{sub k}, near the edge is influenced by the interaction with neutrals and with the potential in the halo required for quasi-neutrality on open magnetic field lines, and the plasma velocity in the core is influenced by external error fields. When plasma effects dominate magnetic locking, the magnetic kink should rotate at a diamagnetic speed of either the edge or the core. If the magnetic field lines of the halo plasma intercept the wall at locations of very different electrical conductivity, the toroidal rotation of the halo currents can intermittently stall at wall locations of high conductivity. Such stalling is seen in experiments. The toroidal phase difference between the stalled halo currents and the kink, which is expected to rotate smoothly, must satisfy {delta}{phi}<{+-}{pi}/2. A concern cited by ITER engineers is that the time varying force of the rotating halo could substantially increase the disruption loads on in-vessel components.

  10. Compact microchannel system

    DOEpatents

    Griffiths, Stewart

    2003-09-30

    The present invention provides compact geometries for the layout of microchannel columns through the use of turns and straight channel segments. These compact geometries permit the use of long separation or reaction columns on a small microchannel substrate or, equivalently, permit columns of a fixed length to occupy a smaller substrate area. The new geometries are based in part on mathematical analyses that provide the minimum turn radius for which column performance in not degraded. In particular, we find that straight channel segments of sufficient length reduce the required minimum turn radius, enabling compact channel layout when turns and straight segments are combined. The compact geometries are obtained by using turns and straight segments in overlapped or nested arrangements to form pleated or coiled columns.

  11. Compact turbidity meter

    NASA Technical Reports Server (NTRS)

    Hirschberg, J. G.

    1979-01-01

    Proposed monitor that detects back-reflected infrared radiation makes in situ turbidity measurements of lakes, streams, and other bodies of water. Monitor is compact, works well in daylight as at night, and is easily operated in rough seas.

  12. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    SciTech Connect

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma.

  13. Preliminary Engineering Design of Toroidal Field Magnet System for Superconducting Tokamak HT-7U

    NASA Astrophysics Data System (ADS)

    Pan, Ying-nian; Weng, Pei-de; Chen, Zhuo-min; Li, Bao-zeng; Wu, Song-tao; Wu, Wei-yue; Gao, Bing-jun; Yu, Jie; Wu, Di; Wu, Xue-bing; Chen, Qiang; Chen, Weng-ge

    2000-04-01

    HT-7U is a large fusion experimental device. It will be built in the Institute of Plasma Physics of Chinese Academy of Sciences. The mission of HT-7U is to develop the scientific basis for a continuously operating tokamak fusion reactor. This paper describes only a toroidal field (TF) superconducting magnet system of HT-7U. In this paper, design criteria of conductor and stability analysis, coil winding and support structure design of magnet system, mechanical calculation and stress analysis, heat load evaluation are given.

  14. Physics Studies of a Proposed Small Aspect Ratio Torsatron-Tokamak Hybrid

    NASA Astrophysics Data System (ADS)

    Valanju, P. M.; Gentle, K. W.; McCool, S.; Miner, W. H., Jr.; Ross, D. W.; Wiley, J. C.; Wootton, A. J.

    1996-11-01

    Physics studies of a proposed small aspect ratio torsatron-tokamak hybrid (SMARTH) are presented. This small aspect ratio configuration attempts to remedy a major drawback of conventional stellarator reactors, viz. the large volume resulting from the large aspect ratio. Considering a small aspect ratio machine with 8 field periods (m=8) and with 8 or 16 modular coils, we investigate vacuum configurations and MHD equilibria to determine the optimum configuration which produces large rotational transform, large plasma volume, large beta, and small field ripple. Implications for particle confinement will also be discussed. Effects of radial electric fields on orbit losses and confinement will also be explored.

  15. A need for non-tokamak approaches to magnetic fusion energy

    NASA Astrophysics Data System (ADS)

    Bathke, C. G.; Krakowski, R. A.; Miller, R. L.

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested.

  16. Next-step-targeted experiments on the Mega-Amp Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Gryaznevich, M.; Akers, R. J.; Counsell, G. F.; Cunningham, G.; Dnestrovskij, A.; Field, A. R.; Hender, T. C.; Kirk, A.; Lloyd, B.; Meyer, H.; Morris, A. W.; Sykes, A.; Tabasso, A.; Valovic, M.; Voss, G. M.; Wilson, H. R.

    2003-05-01

    Since its first physics campaign, the principal parameters on MAST (Mega-Amp Spherical Tokamak) [A. Sykes et al., Nuclear Fusion 41, 1423 (2001)] have been brought up towards their design values. Considerable advances have been made in a range of physics areas of direct relevance to the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis, Nuclear Fusion 39, 2175 (1999)]. In this paper, results on H-mode access, global confinement and pedestal studies are presented and compared with conventional aspect ratio tokamak scalings. Physics and engineering requirements relevant to next step spherical tokamak devices are discussed, in particular the plasma formation, current ramp-up and sustainment, and plasma exhaust. Results of first experiments directly targeting these issues are presented: Plasma current up to 0.5 MA has been produced without use of the central solenoid flux, and current ramp-up and sustainment without use of the central solenoid flux has been demonstrated. Experiments on neutral beam heating and current drive (CD) demonstrate up to 50% bootstrap current fraction and good CD efficiency, and divertor power loading has been found to be tolerable and have a favorable outboard asymmetry.

  17. Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

    NASA Astrophysics Data System (ADS)

    Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; Holcomb, C. T.; Lao, L. L.; McKee, G. R.; Meneghini, O.; Staebler, G. M.; Grierson, B. A.; Qian, J. P.; Solomon, W. M.; Turnbull, A. D.; Holland, C.; Guo, W. F.; Ding, S. Y.; Pan, C. K.; Xu, G. S.; Wan, B. N.

    2016-06-01

    Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN , despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

  18. Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

    SciTech Connect

    Malik, M.A.

    1988-01-01

    There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such as INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism.

  19. Electron cyclotron heating at down-shifted frequencies in existing tokamak devices

    SciTech Connect

    Mazzucato, E.; Fidone, I.; Giruzzi, G.; Krivenski, V.

    1985-06-01

    Plasma heating in existing tokamak devices by electron cyclotron waves with frequency (f) significantly smaller than the electron gyrofrequency (f/sub c/) is investigated for the case of Maxwellian plasmas. It is shown that for central electron temperatures larger than 3 keV, strong absorption of extraordinary waves can occur at values of toroidal field for which the condition f = f/sub c/ is not satisfied in the plasma region. The cases of f = 60 GHz and f = 100 GHz are discussed for the PLT and TFTR devices as representative of medium (approx. =30 kG) and high (approx. =50 kG) magnetic field tokamaks, respectively. Numerical calculations with a ray tracing code indicate that most of the rf energy is absorbed in a central plasma region. These results are of practical interest because they significantly simplify the main technical problem of ECH in a tokamak reactor, i.e., the development of high frequency and high power microwave sources.

  20. Tokamak operation with safety factor q95 < 2 via control of MHD stability.

    PubMed

    Piovesan, P; Hanson, J M; Martin, P; Navratil, G A; Turco, F; Bialek, J; Ferraro, N M; La Haye, R J; Lanctot, M J; Okabayashi, M; Paz-Soldan, C; Strait, E J; Turnbull, A D; Zanca, P; Baruzzo, M; Bolzonella, T; Hyatt, A W; Jackson, G L; Marrelli, L; Piron, L; Shiraki, D

    2014-07-25

    Magnetic feedback control of the resistive-wall mode has enabled the DIII-D tokamak to access stable operation at safety factor q(95) = 1.9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at a given toroidal magnetic field. In tokamaks with a divertor, the limit occurs at q(95) = 2, as confirmed in DIII-D. Since the energy confinement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a whole new high-current regime not accessible before. This result brings significant possible benefits in terms of fusion performance, but it also extends resistive-wall mode physics and its control to conditions never explored before. In present experiments, the q(95) < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.

  1. Materials issues in fusion reactors

    NASA Astrophysics Data System (ADS)

    Suri, A. K.; Krishnamurthy, N.; Batra, I. S.

    2010-02-01

    The world scientific community is presently engaged in one of the toughest technological tasks of the current century, namely, exploitation of nuclear fusion in a controlled manner for the benefit of mankind. Scientific feasibility of controlled fusion of the light elements in plasma under magnetic confinement has already been proven. International efforts in a coordinated and co-operative manner are presently being made to build ITER - the International Thermonuclear Experimental Reactor - to test, in this first step, the concept of 'Tokamak' for net fusion energy production. To exploit this new developing option of making energy available through the route of fusion, India too embarked on a robust fusion programme under which we now have a working tokamak - the Aditya and a steady state tokamak (SST-1), which is on the verge of functioning. The programme envisages further development in terms of making SST-2 followed by a DEMO and finally the fusion power reactor. Further, with the participation of India in the ITER program in 2005, and recent allocation of half - a - port in ITER for placing our Lead - Lithium Ceramic Breeder (LLCB) based Test Blanket Module (TBM), meant basically for breeding tritium and extracting high grade heat, the need to understand and address issues related to materials for these complex systems has become all the more necessary. Also, it is obvious that with increasing power from the SST stages to DEMO and further to PROTOTYPE, the increasing demands on performance of materials would necessitate discovery and development of new materials. Because of the 14.1 MeV neutrons that are generated in the D+T reaction exploited in a tokamak, the materials, especially those employed for the construction of the first wall, the diverter and the blanket segments, suffer crippling damage due to the high He/dpa ratios that result due to the high energy of the neutrons. To meet this challenge, the materials that need to be developed for the tokamaks

  2. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  3. Elementary Processes Underlying Alpha Channeling in Tokamaks

    SciTech Connect

    NM.J. Fisch

    2012-06-15

    Alpha channeling in tokamaks is speculative, but also extraordinarily attractive. Waves that can accomplish this effect have been identified. Key aspects of the theory now enjoy experimental confirmation. This paper will review the elementary processes of wave-particle interactions in plasma that underlie the alpha channeling effect

  4. Banana drift transport in tokamaks with ripple

    SciTech Connect

    Linsker, R.; Boozer, A.H.

    1981-04-01

    Ripple transport in tokamaks is discussed for the banana drift collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and the resulting transport coefficients can consequently differ by several orders of magnitude.

  5. Banana drift transport in tokamaks with ripple

    SciTech Connect

    Linsker, R.; Boozer, A.H.

    1982-01-01

    Ripple transport in tokamaks is discussed for the ''banana drift'' collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and consequently the resulting transport coefficients can differ by several orders of magnitude.

  6. (High beta tokamak research and plasma theory)

    SciTech Connect

    Not Available

    1990-01-01

    Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report.

  7. UCLA Tokamak Program Close Out Report.

    SciTech Connect

    Taylor, Robert John

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  8. Microinstabilities in weak density gradient tokamak systems

    SciTech Connect

    Tang, W.M.; Rewoldt, G.; Chen, L.

    1986-04-01

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient.

  9. Toroidal Alfven wave stability in ignited tokamaks

    SciTech Connect

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

  10. Fusion product measurements in tokamaks (invited; abstract)

    NASA Astrophysics Data System (ADS)

    Strachan, J. D.

    1985-05-01

    Diagnostic methods and the applications of fusion product measurements in tokamaks are reviewed with emphasis on results from PLT, PDX, and TFTR. Measurements have been made using the 2.5-MeV neutron from the d(d, n)3He reaction, the 3-MeV proton from the d(d, p)t reaction, both the 3.7-MeV alpha and the 14.7-MeV proton from the d(3He, p)α reaction, and the 14-MeV neutron from the d(t, n)α reaction. The common use of these measurements is the determination of the ion temperature from the magnitude of the d-d neutron emission. For tokamak plasmas, these results are usually in good agreement with the charge exchange ion temperature. Recently, the charged fusion products have been used for high-resolution spectroscopic purposes, and emission profile measurements. Pitch angle resolution of the escaping 3-MeV proton emission has been used to determine the poloidal magnetic field inside the tokamak. Major issues in this field include the expected tritium operation on TFTR where the neutron measurements will determine when tritium will be introduced into the TFTR vessel and provide a measurement of the fusion power multiplication value (Q). The TFTR Q˜1 experiments will also provide a chance to measure the confinement of 3.5-MeV alphas in a tokamak.

  11. Analysis of sawtooth relaxation oscillations in tokamaks

    SciTech Connect

    Yamazaki, K.; McGuire, K.; Okabayashi, M.

    1982-07-01

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated.

  12. Spontaneous generation of rotation in tokamak plasmas

    SciTech Connect

    Parra Diaz, Felix

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  13. Diagnostics for neutral-beam-heated tokamaks

    SciTech Connect

    Goldston, R.J.

    1982-12-01

    Diagnostic techniques for neutral-beam-heated tokamak plasmas fall into three categories: (1) magnetic diagnostics for measurements of gross stored energy, (2) profile diagnostics for measurements of stored thermal and beam energy, impurity content and plasma rotation, and (3) fast time resolution diagnostics to study MHD fluctuations and micro-turbulence.

  14. Stabilization of tokamak plasma by lithium streams

    SciTech Connect

    L.E. Zakharov

    2000-08-07

    The stabilization theory of free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. While the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.

  15. Experimental test of the system of vertical and longitudinal lithium limiters on T-11M tokamak as a prototype of plasma facing components of a steady-state fusion neutron source

    NASA Astrophysics Data System (ADS)

    Mirnov, S. V.; Belov, A. M.; Djigailo, N. T.; Dzhurik, A. S.; Kravchuk, S. I.; Lazarev, V. B.; Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu.; Shcherbak, A. N.

    2015-11-01

    A new functional model of the prototype of closed Li circuit for protection of the chamber wall was tested in T-11M tokamak by simultaneous use of the vertical Li limiter as an emitter of Li and a new longitudinal Li limiter as its collector. Such technological scheme can be suggested for the steady-state fusion neutron source on the tokamak basis. During plasma shots the cryogenic target of T-11M collected Li flow emitted by the vertical capillary Li limiter almost completely (up to 80%). These Li and hydrogen isotopes were captured and extracted outside the tokamak vacuum chamber without venting of the vessel which is a key requirement for the use of Li in the steady-state tokamak reactor.

  16. Neutral particle analyzer diagnostics on the TCV tokamak

    SciTech Connect

    Karpushov, Alexander N.; Duval, Basil P.; Schlatter, Christian; Afanasyev, Valery I.; Chernyshev, Fedor V.

    2006-03-15

    Experimental apparatus and data analysis techniques used in neutral particle analyzer (NPA) diagnostics on the Tokamak a Configuration Variable (TCV) are described. Two NPAs are used on TCV to measure the energy spectrum of neutral particle fluxes from the plasma. The 'five-channel energy analyzer of atomic particles' used in double electrical analysis mode with fast voltage sweeping detect particles without atomic mass discrimination in the energy range of 0.6-8.0 keV with a time resolution of 0.5-2.0 ms and an energy resolution of 7%-20%. The 28-channel ''compact neutral particle analyzer'' (CNPA) is an EIIB spectrometer with mass and energy separations designed for medium sized fusion machines featuring a carbon neutral stripping foil, a permanent magnet for dispersion, and channel-electron multiplier detectors. The CNPA simultaneously detects two mass species [hydrogen (H) and deuterium (D) or D and helium (He)] in the 0.5-50 keV energy range with a resolution of 60%-10% and a time resolution of 0.5-4.0 ms. The CNPA views the plasma across the path of the diagnostic neutral beam and can perform active charge-exchange NPA measurement. Data analysis procedures and numerical algorithms developed for NPA measurement are routinely used on TCV to obtain information on the plasma ion temperature, ion energy distribution function, plasma isotope ratios, and other plasma characteristics.

  17. Design of multipulse Thomson scattering diagnostic for SST-1 tokamak

    NASA Astrophysics Data System (ADS)

    Kumar, Ajai; Chavda, Chhaya; Saxena, Y. C.; Singh, Ranjeet; Thakar, Aruna; Thomas, Jinto; Patel, Kiran; Pandya, Kaushal; Bedakihale, Vijay

    2007-04-01

    A multipulse Nd:YAG (Yttrium aluminum garnet) Thomson scattering (TS) system is designed and developed for measuring electron temperature (Te) and density (ne) profiles of SST-1 tokamak. The system operates at vertical, divertor, and horizontal (midplane) regions of plasma and measures the electron temperature of 20eVto1.5keV and density of 1018-1019m-3. Six Nd:YAG lasers synchronized with external control is used to get three different temporal resolutions (30Hz, 180Hz, and 1kHz). The entire system is laboratory tested for the stability of alignment and performance over a distance of 30m. Different imaging lens assemblies are designed to image the scattered photons from each of the scattering region to an array of optical fibers. A low cost and compact five-channel interference filter polychromator is designed, fabricated, and tested for its image quality and the filter transmission characteristics. Detection system with an avalanche photodiode and required signal conditioning electronics is developed for detecting the scattered photons. A data acquisition and control module operating on PXI bus is developed for the real time data acquisition and system control. A detailed description of design and testing of TS subsystems is presented in this article.

  18. SABR fusion-fission hybrid transmutation reactor design concept

    NASA Astrophysics Data System (ADS)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  19. Physically detached 'compact groups'

    NASA Technical Reports Server (NTRS)

    Hernquist, Lars; Katz, Neal; Weinberg, David H.

    1995-01-01

    A small fraction of galaxies appear to reside in dense compact groups, whose inferred crossing times are much shorter than a Hubble time. These short crossing times have led to considerable disagreement among researchers attempting to deduce the dynamical state of these systems. In this paper, we suggest that many of the observed groups are not physically bound but are chance projections of galaxies well separated along the line of sight. Unlike earlier similar proposals, ours does not require that the galaxies in the compact group be members of a more diffuse, but physically bound entity. The probability of physically separated galaxies projecting into an apparent compact group is nonnegligible if most galaxies are distributed in thin filaments. We illustrate this general point with a specific example: a simulation of a cold dark matter universe, in which hydrodynamic effects are included to identify galaxies. The simulated galaxy distribution is filamentary and end-on views of these filaments produce apparent galaxy associations that have sizes and velocity dispersions similar to those of observed compact groups. The frequency of such projections is sufficient, in principle, to explain the observed space density of groups in the Hickson catalog. We discuss the implications of our proposal for the formation and evolution of groups and elliptical galaxies. The proposal can be tested by using redshift-independent distance estimators to measure the line-of-sight spatial extent of nearby compact groups.

  20. Fuel provision for nonbreeding deuterium-tritium fusion reactors

    SciTech Connect

    Jassby, D.L.; Katsurai, M.

    1980-01-01

    Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

  1. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor.

    PubMed

    Leipold, F; Furtula, V; Salewski, M; Bindslev, H; Korsholm, S B; Meo, F; Michelsen, P K; Moseev, D; Nielsen, S K; Stejner, M

    2009-09-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic using gyrotrons operated at 60 GHz will meet the requirements for spatially and temporally resolved measurements of the velocity distributions of confined fast alphas in ITER by evaluating the scattered radiation (CTS signal). While a receiver antenna on the low field side of the tokamak, resolving near perpendicular (to the magnetic field) velocity components, has been enabled, an additional antenna on the high field side (HFS) would enable measurements of near parallel (to the magnetic field) velocity components. A compact design solution for the proposed mirror system on the HFS is presented. The HFS CTS antenna is located behind the blankets and views the plasma through the gap between two blanket modules. The viewing gap has been modified to dimensions 30x500 mm(2) to optimize the CTS signal. A 1:1 mock-up of the HFS mirror system was built. Measurements of the beam characteristics for millimeter-waves at 60 GHz used in the mock-up agree well with the modeling.

  2. Panel sees limited interest in compact nukes

    SciTech Connect

    Not Available

    1983-11-01

    Participants in the Joint Power Generation conference thought compact (200- to 300-MW) nuclear reactors would be useful to developing countries, but only the Canadians showed interest in becoming suppliers. Others said they would simply downsize existing designs. A 300-MW mini-Candu that can be built in 48 months will use proven components and have the same price tag as a full-sized unit. A market may develop in the future in the US and other industrialized countries for low-temperature heat sources. Another 5 to 10 developing countries would likely join the 7 now using nuclear power. (DCK)

  3. Processes with neutral hydrogen and deuterium molecules relevant to edge plasma in tokamaks

    NASA Astrophysics Data System (ADS)

    Cadez, I.; Markelj, S.; Rupnik, Z.; Pelicon, P.

    2008-07-01

    Detailed understanding and characterization of plasma-wall interaction and edge plasma in present tokamaks and future fusion reactors is becoming more and more important due to the ITER project. Involved processes determine the physical and chemical sputtering of the wall material, fuel retention in exposed material, edge plasma properties, disruption phenomena etc. Neutral hydrogen atoms and molecules are present in the edge plasma. They are continuously generated by ion recombination on the wall of the fusion reactor and on the other plasma facing components and subsequently reemitted in the plasma. Neutral molecules are especially important for plasma detachment in tokamak divertors. The interaction of excited neutral molecules with the walls and their importance for the edge plasma is still not well understood since there are not many experimental studies of relevant processes. Moreover, spectroscopic results from tokamak edge plasma are dominated by processes involving ions and electrons, so that direct evidence of the influence of neutrals is difficult to extract. Direct correlation of the observed phenomena to the processes with neutrals is mainly possible by numerical simulations. We have constructed a set-up for vibrational spectroscopy of hydrogen molecules (H_2 and D_2) that is based on the properties of the dissociative electron attachment in hydrogen in order to facilitate dedicated experimental studies of relevant processes with hydrogen molecules. For the same purpose we also developed a technique for in-situ hydrogen depth profiling on the samples exposed to the controlled hydrogen atmosphere. This is done by Ion Beam Analytical (IBA) method ERAD (Elastic Recoil Detection Analysis), utilizing 4.2 MeV probing beam of ^7Li^2+ ions. A short description of experimental techniques and results on chemical erosion of graphite layers, production of vibrationally excited hydrogen molecules on tungsten and isotope exchange on tungsten are to be presented in

  4. Theory and models of material erosion and lifetime during plasma instabilities in a tokamak environment.

    SciTech Connect

    Hassanein, A.; Konkashbaev, I.

    1999-11-08

    Surface and structural damage to plasma-facing components (PFCs) due to the frequent loss of plasma confinement remains a serious problem for the tokamak reactor concept. The deposited plasma energy causes significant surface erosion, possible structural failure, and frequent plasma contamination. Surface damage consists of vaporization, spallation, and liquid splatter of metallic materials. Structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. To evaluate the lifetimes of plasma-facing materials and nearby components and to predict the various forms of damage that they experience, comprehensive models (contained in the HEIGHTS computer simulation package) are developed, integrated self-consistently, and enhanced. Splashing mechanisms such as bubble boiling and various liquid magnetohydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials are being examined. The design requirements and implications of plasma-facing and nearby components are discussed, along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components.

  5. Banana orbits in elliptic tokamaks with hole currents

    NASA Astrophysics Data System (ADS)

    Martin, P.; Castro, E.; Puerta, J.

    2015-03-01

    Ware Pinch is a consequence of breaking of up-down symmetry due to the inductive electric field. This symmetry breaking happens, though up-down symmetry for magnetic surface is assumed. In previous work Ware Pinch and banana orbits were studied for tokamak magnetic surface with ellipticity and triangularity, but up-down symmetry. Hole currents appear in large tokamaks and their influence in Ware Pinch and banana orbits are now considered here for tokamaks magnetic surfaces with ellipticity and triangularity.

  6. Advanced Construction of Compact Containment BWR

    SciTech Connect

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-07-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  7. Sawtooth Instability in the Compact Toroidal Hybrid

    NASA Astrophysics Data System (ADS)

    Herfindal, J. L.; Maurer, D. A.; Hartwell, G. J.; Ennis, D. A.; Knowlton, S. F.

    2015-11-01

    Sawtooth instabilities have been observed in the Compact Toroidal Hybrid (CTH), a current-carrying stellarator/tokamak hybrid device. The sawtooth instability is driven by ohmic heating of the core plasma until the safety factor drops below unity resulting in the growth of an m = 1 kink-tearing mode. Experiments varying the vacuum rotational transform from 0.02 to 0.13 are being conducted to study sawtooth property dependance on vacuum flux surface structure. The frequency of the sawtooth oscillations increase from 2 kHz to 2.8 kHz solely due the decrease in rise time of the oscillation, the crash time is unchanged. CTH has three two-color SXR cameras, a three-channel 1mm interferometer, and a new bolometer system capable of detecting the signatures of sawtooth instabilities. The new bolometer system consists of two cameras, each containing a pair of diode arrays viewing the plasma directly or through a beryllium filter. Electron temperature measurements are found with the two-color SXR cameras through a ratio of the SXR intensities. Impurity radiation can drastically affect the electron temperature measurement, therefore new filters consisting of aluminum and carbon were selected to avoid problematic line radiation while maximizing the signal for a 100 eV plasma. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

  8. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  9. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  10. Compact Nanowire Sensors Probe Microdroplets.

    PubMed

    Schütt, Julian; Ibarlucea, Bergoi; Illing, Rico; Zörgiebel, Felix; Pregl, Sebastian; Nozaki, Daijiro; Weber, Walter M; Mikolajick, Thomas; Baraban, Larysa; Cuniberti, Gianaurelio

    2016-08-10

    The conjunction of miniature nanosensors and droplet-based microfluidic systems conceptually opens a new route toward sensitive, optics-less analysis of biochemical processes with high throughput, where a single device can be employed for probing of thousands of independent reactors. Here we combine droplet microfluidics with the compact silicon nanowire based field effect transistor (SiNW FET) for in-flow electrical detection of aqueous droplets one by one. We chemically probe the content of numerous (∼10(4)) droplets as independent events and resolve the pH values and ionic strengths of the encapsulated solution, resulting in a change of the source-drain current ISD through the nanowires. Further, we discuss the specificities of emulsion sensing using ion sensitive FETs and study the effect of droplet sizes with respect to the sensor area, as well as its role on the ability to sense the interior of the aqueous reservoir. Finally, we demonstrate the capability of the novel droplets based nanowire platform for bioassay applications and carry out a glucose oxidase (GOx) enzymatic test for glucose detection, providing also the reference readout with an integrated parallel optical detector.

  11. Compact Nanowire Sensors Probe Microdroplets.

    PubMed

    Schütt, Julian; Ibarlucea, Bergoi; Illing, Rico; Zörgiebel, Felix; Pregl, Sebastian; Nozaki, Daijiro; Weber, Walter M; Mikolajick, Thomas; Baraban, Larysa; Cuniberti, Gianaurelio

    2016-08-10

    The conjunction of miniature nanosensors and droplet-based microfluidic systems conceptually opens a new route toward sensitive, optics-less analysis of biochemical processes with high throughput, where a single device can be employed for probing of thousands of independent reactors. Here we combine droplet microfluidics with the compact silicon nanowire based field effect transistor (SiNW FET) for in-flow electrical detection of aqueous droplets one by one. We chemically probe the content of numerous (∼10(4)) droplets as independent events and resolve the pH values and ionic strengths of the encapsulated solution, resulting in a change of the source-drain current ISD through the nanowires. Further, we discuss the specificities of emulsion sensing using ion sensitive FETs and study the effect of droplet sizes with respect to the sensor area, as well as its role on the ability to sense the interior of the aqueous reservoir. Finally, we demonstrate the capability of the novel droplets based nanowire platform for bioassay applications and carry out a glucose oxidase (GOx) enzymatic test for glucose detection, providing also the reference readout with an integrated parallel optical detector. PMID:27417510

  12. New Technique of AC drive in Tokamak using Permanent Magnets

    NASA Astrophysics Data System (ADS)

    Matteucci, Jackson; Zolfaghari, Ali

    2013-10-01

    This study investigates a new technique of capturing the rotational energy of alternating permanent magnets in order to inductively drive an alternating current in tokamak devices. The use of rotational motion bypasses many of the pitfalls seen in typical inductive and non-inductive current drives. Three specific designs are presented and assessed in the following criteria: the profile of the current generated, the RMS loop voltage generated as compared to the RMS power required to maintain it, the system's feasibility from an engineering perspective. All of the analysis has been done under ideal E&M conditions using the Maxwell 3D program. Preliminary results indicate that it is possible to produce an over 99% purely toroidal current with a RMS d Φ/dt of over 150 Tm2/s, driven by 20 MW or less of rotational power. The proposed mechanism demonstrates several key advantages including an efficient mechanical drive system, the generation of pure toroidal currents, and the potential for a quasi-steady state fusion reactor. The following quantities are presented for various driving frequencies and magnet strengths: plasma current generated, loop voltage, torque and power required. This project has been supported by DOE Funding under the SULI program.

  13. Tritium Removal by Laser Heating and Its Application to Tokamaks

    SciTech Connect

    C.H. Skinner; C.A. Gentile; G. Guttadora; A. Carpe; S. Langish; K.M. Young; M. Nishi; W. Shu

    2001-11-16

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.

  14. EBT: an alternate concept to tokamaks and mirrors

    SciTech Connect

    Glowienka, J.C.

    1980-01-01

    The ELMO Bumpy Torus (EBT) is a hybrid magnetic trap formed by a series of toroidally connected simple mirrors. It differs from a tokamak, the present main-line approach, in that plasma stability and heating are obtained in a current-free geometry by the application of steady-state, high power, electron cyclotron resonance heating (ECH) producing a steady-state plasma. The primary motivation for EBT confinement research is the potential for a steady-state, highly accessible reactor with high ..beta... In the present EBT-I/S device, electron confinement has been observed to agree with the predictions of theory. The major emphasis of the experimental program is on the further scaling of plasma parameters in the EBT-I/S machine with ECH frequency (10.6, 18, and 28 GHz), resonant magnetic field (0.3, 0.6, and 1 T), and heating power (30, 60, and 200 kW). In addition, substantial efforts are under way or planned in the areas of ion cyclotron heating, neutral beam heating, plasma-wall interactions, impurity control, synchrotron radiation, and divertors. Recently, EBT has been selected as the first alternative concept to be advanced to the proof-of-principle stage; this entails a major device scale-up to allow a reasonable extrapolation to a DT-burning facility. The status and future plans of the EBT program, in particular the proof-of-principle experiment (EBT-P), are discussed.

  15. New reflectometer systems for the DIII-D tokamak (abstract)

    SciTech Connect

    Doyle, E.J.; Kim, K.W.; Burns, S.; Nguyen, X.; Peebles, W.A.; Rhodes, T.L. )

    1992-10-01

    During a machine vent in December 1991, two new reflectometer systems were successfully installed and tested on the DIII-D tokamak. The first is an {ital X}-mode broadband system primarily intended for density profile measurements, utilizing BWO sources and covering {ital Q} and {ital V} frequency bands (33--50 and 50--75 GHz). The second system is an adaptation of a pre-existing inside launch (high field side) ECRH waveguide to provide an inside launch reflectometer capability at the same frequencies and polarization as an outside launch fixed frequency {ital O}-mode system. The new systems will have a dual role in both directly supporting the DIII-D physics program, and also acting as flexible and adaptable test beds for the development of reactor relevant reflectometer systems, such as required for ITER. Specific examples of planned measurements include investigation of possible in/out plasma asymmetries at the {ital L}--{ital H} transition and ELMs, and demonstration of routine and reliable density profile measurements. It is expected that preliminary data from the inside launch system will be available by the time of the conference. This work is supported by the U. S. Department of Energy under Grant No. DE-FG03-86-ER53225 and General Atomics subcontract SC120536 under DOE Contract No. DE-AC03-89ER51114.

  16. Compact, Integrated Photoelectron Linacs

    NASA Astrophysics Data System (ADS)

    Yu, David

    2000-12-01

    The innovative compact high energy iniector which has been developed by DULY Research Inc., will have wide scientific industrial and medical applications. The new photoelectron injector integrates the photocathode directly into a multicell linear accelerator with no drift space between the injector and the linac. By focusing the beam with solenoid or permanent magnets, and producing high current with low emittance, extremely high brightness is achieved. In addition to providing a small footprint and improved beam quality in an integrated structure, the compact system considerably simplifies external subsystems required to operate the photoelectron linac, including rf power transport, beam focusing, vacuum and cooling. The photoelectron linac employs an innovative Plane-Wave-Transformer (PWT) design, which provides strong cell-to-cell coupling, relaxes manufacturing tolerance and facilitates the attachment of external ports to the compact structure with minimal field interference. DULY Research Inc. under the support of the DOE Small Business Innovation Research (SBIR) program, has developed, constructed and installed a 20-MeV, S-band compact electron source at UCLA. DULY Research is also presently engaged in the development of an X-band photoelectron linear accelerator in another SBIR project. The higher frequency structure when completed will be approximately three times smaller, and capable of a beam brightness ten times higher than the S-band structure.

  17. COMPACT SCHOOL AND $$ SAVINGS.

    ERIC Educational Resources Information Center

    BAIR, W.G.

    A REVIEW OF THE CRITERIA FOR CONSIDERING THE USE OF A TOTAL ENERGY SYSTEM WITHIN A SCHOOL BUILDING STATES THE WINDOWLESS, COMPACT SCHOOL OFFERS MORE EFFICIENT SPACE UTILIZATION WITH LESS AREA REQUIRED FOR GIVEN STUDENT POPULATION AND LOWER OPERATION COSTS. THE AUTHOR RECOMMENDS THAT THESE BUILDINGS BE WINDOWLESS TO REDUCE HEAT COSTS, HOWEVER, AT…

  18. Compact optical transconductance varistor

    SciTech Connect

    Sampayan, Stephen

    2015-09-22

    A compact radiation-modulated transconductance varistor device having both a radiation source and a photoconductive wide bandgap semiconductor material (PWBSM) integrally formed on a substrate so that a single interface is formed between the radiation source and PWBSM for transmitting PWBSM activation radiation directly from the radiation source to the PWBSM.

  19. Compact ultradense matter impactors.

    PubMed

    Rafelski, Johann; Labun, Lance; Birrell, Jeremiah

    2013-03-15

    We study interactions of meteorlike compact ultradense objects (CUDO), having nuclear or greater density, with Earth and other rocky bodies in the Solar System as a possible source of information about novel forms of matter. We study the energy loss in CUDO puncture of the body and discuss differences between regular matter and CUDO impacts.

  20. Unique features of space reactors

    SciTech Connect

    Buden, D.

    1990-01-01

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

  1. Models for impurity effects in tokamaks

    SciTech Connect

    Hogan, J.T.

    1980-03-01

    Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high ..beta.. and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high ..beta.. in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability.

  2. Boundary Plasma Turbulence Simulations for Tokamaks

    SciTech Connect

    Xu, X; Umansky, M; Dudson, B; Snyder, P

    2008-05-15

    The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.

  3. Properties of dc helicity injected tokamak plasmas

    SciTech Connect

    Darrow, D.S.; Ono, M.; Forest, C.B.; Greene, G.J.; Hwang, Y.S.; Park, H.K. ); Taylor, R.J.; Pribyl, P.A.; Evans, J.D.; Lai, K.F.; Liberati, J.R. )

    1990-06-01

    Several dc helicity injection experiments using an electron beam technique have been conducted on the Current Drive Experiment (CDX) (Phys. Rev. Lett. {bold 59}, 2165 (1987)) and the Continuous Current Tokamak (CCT) (Phys. Rev. Lett. {bold 63}, 2365 (1989)). The data strongly suggest that tokamak plasmas are being formed and maintained by this method. The largest currents driven to date are 1 kA in CDX ({ital q}{sub {ital a}} =5) and 6 kA in CCT ({ital q}{sub {ital a}} =3.5). An initial comparison of discharge properties with helicity theory indicates rough agreement. Current drive energy efficiencies are 9% and 23% of Ohmic efficiency in two cases analyzed. Strong radial electric fields are observed in these plasmas that cause poloidal rotation and, possibly, improved confinement.

  4. Energetics of runaway electrons during tokamak disruptions

    NASA Astrophysics Data System (ADS)

    Riemann, J.; Smith, H. M.; Helander, P.

    2012-01-01

    In a tokamak disruption, a substantial fraction of the plasma current can be converted into runaway electrons. Although these are usually highly relativistic, their total energy is initially much smaller than that of the pre-disruption plasma. However, following a suggestion by Putvinski et al. [Plasma Phys. Controlled Fusion 39, B157 (1997)], it is shown that as the post-disruption plasma drifts toward the first wall, a non-negligible part of the energy contained in the poloidal magnetic field can be converted into kinetic energy of the runaway electrons. This process is simulated numerically, and it is found that in an ITER-like tokamak runaway electrons can gain kinetic energies up to about 70 MJ by this mechanism.

  5. The Physics of Tokamak Start-up

    SciTech Connect

    D. Mueller

    2012-11-13

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. ITER, the National Spherical Torus eXperiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  6. Spherical Stellarator-Tokamak Hybrid Configurations

    NASA Astrophysics Data System (ADS)

    Hanson, James D.; Yuan, Ying; Gandy, Rex F.; Knowlton, Stephen F.; Doloc, Cristian; Carnevali, Antonino; Hartwell, Gregory

    1996-11-01

    We consider low-aspect ratio stellarator-tokamak hybrid configurations similar to the inclined coils configurations of Moroz(P. E. Moroz, Phys. Plasmas 2), 4269 (1995). and the Small-Aspect Ratio Toroidal Hybrid(D. B. Batchelor et al)., poster at this meeting. (SMARTH) configurations of Batchelor et al. The advantages of these configurations include a current-free q profile which increases with minor radius, (like a tokamak's), and a magnetic divertor structure which does not rotate about the magnetic axis. Our investigations center on configurations suitable to be built as a small, inexpensive exploratory device. Initial work has focused on planar coils (for ease of construction) and small numbers of toroidal coils (for ease of access). Results from field line tracing, equilibrium, and particle orbit studies will be shown.

  7. Rapidly Moving Divertor Plates In A Tokamak

    SciTech Connect

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  8. First Engineering Commissioning of EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Wan, Yuanxi; Li, Jiangang; Weng, Peide; EAST Team

    2006-05-01

    Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak. The first commissioning started on Feb. 1st of 2006 and finished on March 30th of 2006 at the Institute of Plasma Physics, Chinese Academy of Sciences. It consists of leakage testing at both room temperature and low temperature, pumping down, cooling down all coils, current leads, bus bar and the thermal shielding, exciting all the coils, measuring magnetic configuration and warming up the magnets. The electromagnetic, thermal hydraulic and mechanical performance of EAST Toroidal Field (TF) and Poloidal Field (PF) magnets have also been tested. All sub-systems, including pumping system, cryogenic system, PF& TF power supply systems, magnet instrumentation system, quench detection and protection system, water cooling system, data acquisition system, main control system, plasma control system (PCS), interlock and safety system have been successfully tested.

  9. Microtearing modes in spherical and conventional tokamaks

    NASA Astrophysics Data System (ADS)

    Moradi, S.; Pusztai, I.; Guttenfelder, W.; Fülöp, T.; Mollén, A.

    2013-06-01

    The onset and characteristics of microtearing modes (MTM) in the core of spherical (NSTX) and conventional tokamaks (ASDEX Upgrade and JET) are studied through local linear gyrokinetic simulations with GYRO (Candy and Belli 2011 General Atomics Report GA-A26818). For experimentally relevant core plasma parameters in the NSTX and ASDEX Upgrade tokamaks, in agreement with previous works, we find MTMs as the dominant linear instability. Also, for JET-like core parameters considered in our study an MTM is found as the most unstable mode. In all of these plasmas, finite collisionality is needed for MTMs to become unstable and the electron temperature gradient is found to be the fundamental drive. However, a significant difference is observed in the dependence of the linear growth rate of MTMs on electron temperature gradient. While it varies weakly and non-monotonically in JET and ASDEX Upgrade plasmas, in NSTX it increases with the electron temperature gradient.

  10. High beta plasmas in the PBX tokamak

    SciTech Connect

    Bol, K.; Buchenauer, D.; Chance, M.; Couture, P.; Fishman, H.; Fonck, R.; Gammel, G.; Grek, B.; Ida, K.; Itami, K.

    1986-04-01

    Bean-shaped configurations favorable for high ..beta.. discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present ..beta.. limit.

  11. Fast ion orbits in spherical tokamaks

    SciTech Connect

    Solano, E.R.

    1995-07-20

    In a spherical tokamak, the 1/R variation of the toroidal field is extreme, and for a given value of the safety factor a relatively low average toroidal field can be used, together with large plasma current and large plasma minor radius and elongation. The poloidal and toroidal fields are then of similar size. In consequence, the orbits of fast ions depart considerably from the guiding center orbits because of gyromotion in the small magnetic fields in the low field side.

  12. Self-Organized Stationary States of Tokamaks.

    PubMed

    Jardin, S C; Ferraro, N; Krebs, I

    2015-11-20

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  13. Neoclassical tearing modes in a tokamak

    SciTech Connect

    Hahm, T.S.

    1988-12-01

    Linear tearing instability is studied in the banana collisionality regime in tokamak geometry. Neoclassical effects produce significant modifications of Ohm's law and the vorticity equation, so that the growth rate of tearing modes driven by ..delta..' is dramatically reduced compared to the usual resistive magnetohydrodynamic values. Consequences of this result, regarding the presence of pressure-gradient-driven neoclassical resistive interchange instabilities and the evolution of magnetic islands in the Rutherford regime, are discussed.

  14. Neoclassical tearing modes in a tokamak

    SciTech Connect

    Hahm, T.S.

    1988-08-01

    Linear tearing instability is studied in the banana collisionality regime in tokamak geometry. Neoclassical effects produce significant modifications of Ohm's law and the vorticity equation so that the growth rate of tearing modes driven by ..delta..' is dramatically reduced compared to the usual resistive MHD value. Consequences of this result, regarding the presence of pressure-gradient-driven neoclassical resistive interchange instabilities and the evolution of magnetic islands in the Rutherford regime, are discussed. 10 refs.

  15. Self-Organized Stationary States of Tokamaks

    SciTech Connect

    Jardin, S. C.; Ferraro, N.; Krebs, I.

    2015-11-01

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  16. Neoclassical transport in high [beta] tokamaks

    SciTech Connect

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high [beta] large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high [beta] large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low [beta] values by a factor ([var epsilon]/q[sup 2][beta])[sup [1/2

  17. Confinement scaling and ignition in tokamaks

    SciTech Connect

    Perkins, F.W.; Sun, Y.C.

    1985-10-01

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10/sup 15/ cm/sup -3/, high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition.

  18. Tokamak with liquid metal toroidal field coil

    DOEpatents

    Ohkawa, Tihiro; Schaffer, Michael J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.

  19. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  20. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.