Science.gov

Sample records for compact tokamak reactors

  1. 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.

  2. Can high fields save the tokamak? The challenge of steady-state operation for low cost compact reactors

    NASA Astrophysics Data System (ADS)

    Freidberg, Jeffrey; Dogra, Akshunna; Redman, William; Cerfon, Antoine

    2016-10-01

    The development of high field, high temperature superconductors is thought to be a game changer for the development of fusion power based on the tokamak concept. We test the validity of this assertion for pilot plant scale reactors (Q 10) for two different but related missions: pulsed operation and steady-state operation. Specifically, we derive a set of analytic criteria that determines the basic design parameters of a given fusion reactor mission. As expected there are far more constraints than degrees of freedom in any given design application. However, by defining the mission of the reactor under consideration, we have been able to determine the subset of constraints that drive the design, and calculate the values for the key parameters characterizing the tokamak. Our conclusions are as follows: 1) for pulsed reactors, high field leads to more compact designs and thus cheaper reactors - high B is the way to go; 2) steady-state reactors with H-mode like transport are large, even with high fields. The steady-state constraint is hard to satisfy in compact designs - high B helps but is not enough; 3) I-mode like transport, when combined with high fields, yields relatively compact steady-state reactors - why is there not more research on this favorable transport regime?

  3. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.

  4. Compact Reactor

    NASA Astrophysics Data System (ADS)

    Williams, Pharis E.

    2007-01-01

    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.

  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. 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.

  8. 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.

  9. On the survivability of diagnostic windows in the CIT (Compact Ignition Tokamak) reactor

    SciTech Connect

    Taylor, A.

    1988-11-01

    The problem of radiation induced stresses in CIT diagnostic windows is discussed. Existing data indicate windows of existing design will probably survive if placed on the periphery of the reactor. There is a lack of adequate engineering data upon which the design and survivability of windows can be based. 22 refs., 5 figs., 2 tabs.

  10. Overview of the Compact Ignition tokamak

    SciTech Connect

    Flanagan, C. A.; Peng, Yueng Kay Martin

    1986-01-01

    The Compact Ignition Tokamak (CIT) mission is to achieve ignition and provide the capability to experimentally study burning plasma behavior. A national team has developed a baseline concept including definition of the necessary research and development. The baseline concept satisfies the physics performance objectives established for the project and complies with defined design specifications. To ensure that the mission is achieved, the design requires large magnetic fields on axis (10 T) and use of large plasma currents (10 MA). The design is capable of accommodating significant auxiliary heating to enter the ignited regime. The CIT is designed to operate in plasma parameter regimes that a are directly relevant to future fusion power reactors.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Vertical compact torus injection into the STOR-M tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Dazhi

    Central fuelling is a fundamental issue in the next generation tokamak-ITER (International Thermonuclear Experimental Reactor). It is essential for optimization of the bootstrap current which is proportional to the pressure gradient of trapped particles. The conventional fusion reactor fuelling techniques, such as gas puffing and cryogenic pellet injection, are considered inadequate to fulfill this goal due to premature ionization caused by high plasma temperature and density. Compact Torus (CT) injection is a promising fuelling technique for central fuelling a reactor-grade tokamak. An accelerated CT is expected to penetrate into the core region and deposit fuel there provided the CT kinetic energy density exceeds the magnetic energy density in a target plasma. This process is complicated and involves CT penetration into an external magnetic field, a CT stopping mechanism, magnetic reconnection, and excitation of plasma waves. CTs can be injected at different angles with respect to the tokamak toroidal magnetic field, either horizontally or vertically. Normally, CTs are injected radially in the mid-plane of a tokamak. In this configuration, CTs will undergo a decelerating force due to the gradient of the tokamak toroidal magnetic field. CTs will stop inside the tokamak chamber or bunce back depending on the relation between kinetic energy density of injected CTs and the tokamak toroidal magnetic field energy density. In the case of vertical injection, deeper penetration is expected due to the absence of the gradient of the tokamak toroidal field in that direction. Experimental investigations on vertical CT injection into a tokamak will be of great significance. The aim of this thesis is to experimentally investigate the feasibility of vertical CT injection into a tokamak and effects of CTs on tokamak plasma confinements. The Saskatchewan Torus-Modified (STOR-M) tokamak is currently the only tokamak equipped with a CT injector in the world. Vertical CT injection

  17. Fusion reactor design studies. [ARIES Tokamak

    SciTech Connect

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-10-12

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)

  18. 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.

  19. 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.

  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. The ARIES-I Tokamak Reactor Study

    SciTech Connect

    Najmabadi, F; Peng, Yueng Kay Martin

    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. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design si a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-{sup 3}He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

  2. 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.

  3. Physics aspects of the Compact Ignition Tokamak

    SciTech Connect

    Post, D.; Bateman, G.; Houlberg, W.; Bromberg, L.; Cohn, D.; Colestock, P.; Hughes, M.; Ignat, D.; Izzo, R.; Jardin, S.

    1986-11-01

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha-particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve the condition of ntau/sub E/ approx. 2 x 10/sup 20/ sec m/sup -3/ required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k approx. 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided.

  4. 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.

  5. 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.

  6. Tokamak reactor cost model based on STARFIRE/WILDCAT costing

    SciTech Connect

    Evans, K. Jr.

    1983-03-01

    A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately.

  7. Heating the Compact Ignition Tokamak (CIT)

    SciTech Connect

    Ignat, D.W.

    1989-11-01

    The proposed CIT starts operation in the late 1990's with 20 MW of rf heating power. The tokamak and facility are to be designed to accommodate 50 MW auxiliary heating. The heating methods new being considered are ion cyclotron heating (ICH) and electron cyclotron heating (ECH). Aspects of these systems are described, and the choice of power level and type is discussed. 18 refs.

  8. Summary discussion: An integrated advanced tokamak reactor

    SciTech Connect

    Sauthoff, N.R.

    1994-12-31

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ``figures of merit`` for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept.

  9. Scoping study for compact high-field superconducting net energy tokamaks

    NASA Astrophysics Data System (ADS)

    Mumgaard, R. T.; Greenwald, M.; Freidberg, J. P.; Wolfe, S. M.; Hartwig, Z. S.; Brunner, D.; Sorbom, B. N.; Whyte, D. G.

    2016-10-01

    The continued development and commercialization of high temperature superconductors (HTS) may enable the construction of compact, net-energy tokamaks. HTS, in contrast to present generation low temperature superconductors, offers improved performance in high magnetic fields, higher current density, stronger materials, higher temperature operation, and simplified assembly. Using HTS along with community-consensus confinement physics (H98 =1) may make it possible to achieve net-energy (Q>1) or burning plasma conditions (Q>5) in DIII-D or ASDEX-U sized, conventional aspect ratio tokamaks. It is shown that, by operating at high plasma current and density enabled by the high magnetic field (B>10T), the required triple products may be achieved at plasma volumes under 20m3, major radii under 2m, with external heating powers under 40MW. This is at the scale of existing devices operated by laboratories, universities and companies. The trade-offs in the core heating, divertor heat exhaust, sustainment, stability, and proximity to known plasma physics limits are discussed in the context of the present tokamak experience base and the requirements for future devices. The resulting HTS-based design space is compared and contrasted to previous studies on high-field copper experiments with similar missions. The physics exploration conducted with such HTS devices could decrease the real and perceived risks of ITER exploitation, and aid in quickly developing commercially-applicable tokamak pilot plants and reactors.

  10. 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.

  11. 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.

  12. Pulse length assessment of compact ignition tokamak designs

    SciTech Connect

    Stotler, D.P.; Pomphrey, N.

    1989-07-01

    A time-dependent zero-dimensional code has been developed to assess the pulse length and auxiliary heating requirements of Compact Ignition Tokamak (CIT) designs. By taking a global approach to the calculation, parametric studies can be easily performed. The accuracy of the procedure is tested by comparing with the Tokamak Simulation Code which uses theory-based thermal diffusivities. A series of runs is carried out at various levels of energy confinement for each of three possible CIT configurations. It is found that for cases of interest, ignition or an energy multiplication factor Q /approx gt/ 7 can be attained within the first half of the planned five-second flattop with 10--40 MW of auxiliary heating. These results are supported by analytic calculations. 18 refs., 7 figs., 2 tabs.

  13. 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.

  14. Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

    ScienceCinema

    None

    2016-07-12

    From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

  15. A design method of divertor in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Ueda, N.; Itoh, S.-I.; Tanaka, M.; Itoh, K.

    1990-08-01

    Computational method to design the efficient divertor configuration in tokamak reactor is presented. The two dimensional code was developed to analyze the distributions of the plasma and neutral particles for realistic configurations. Using this code, a method to design the efficient divertor configuration is developed. An example of new divertor, which consists of the baffle and fin plates, is analyzed.

  16. Compact Ignition Tokamak Program: R and D needs

    SciTech Connect

    Flanagan, C.A.

    1985-01-01

    This report on the Compact Ignition Tokamak Program supplies information concerning: segmented vacuum vessel joint development; first wall tile attachments; first wall/tile development - composite materials; vacuum leak detection; high frequency rf sources; Faraday shield development; design and testing of rf launchers for high power, ling pulse operation; radiation hardened, low loss, dielectric windows for rf, IR, visible, UV and X-rays, mirrors for changing direction and focusing IR, visible and UV radiation; radiation resistant optical dielectric wave guides; radiation resistant HV insulation for diagnostic magnetic pickup coils; compact radiation and/or magnetic shielding for in-vault diagnostics that need some attenuation to reduce S/N ratio; radiation hardened line-of-sight sensors such as bolometers, UV and soft X-ray detectors, neutral particle analyzers, torus pressure gauges; special maintenance fixtures and tools; material properties - design data base - all materials; and insulation - electrical/thermal and mechanical properties.

  17. Time-dependent simulations of a Compact Ignition Tokamak

    SciTech Connect

    Stotler, D.P.; Bateman, G.

    1988-05-01

    Detailed simulations of the Compact Ignition Tokamak are carried out using a 1-1/2-D transport code. The calculations include time-varying densities, fields, and plasma shape. It is shown that ignition can be achieved in this device if somewhat better than L-mode energy confinement time scaling is possible. We also conclude that the performance of such a compact, short-pulse device can depend greatly on how the plasma is evolved to its flat-top parameters. Furthermore, in cases such as the ones discussed here, where there is not a great deal of ignition margin and the electron density is held constant, ignition ends if the helium ash is not removed. In general, control of the deuterium--tritium density is equivalent to burn control. 48 refs., 15 figs.

  18. Tritium experience in the Tokamak Fusion Test Reactor

    SciTech Connect

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N.; Hogan, J.

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  19. In-vessel remote maintenance of the Compact Ignition Tokamak

    SciTech Connect

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref.

  20. Tokamak fusion reactors with less than full tritium breeding

    SciTech Connect

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed.

  1. Dust divertor for a tokamak fusion reactor

    SciTech Connect

    Tang, X Z; Delzanno, G L

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

  2. Deuterium and tritium separation in a tokamak reactor divertor layer

    NASA Astrophysics Data System (ADS)

    Tokar', M. Z.

    1989-04-01

    It's shown that the plasma isotope composition in a tokamak reactor divertor layer changes along the magnetic field and can notable differ from the gas composition in a pumping chamber. Heavier tritium must concentrate in the hot plasma far from the divertor plate due to thermal force stipulated by mutial collisions of deuterium and tritium ions. This circumstance is favourable from the point of view of tritium cycle optimization and must facilitate solution of the problem of tritium accumulation in the reactor construction elements.

  3. 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.

  4. EQUILIBRIUM AND STABILITY FOR THE ARIES COMPACT STELLERATOR REACTOR

    SciTech Connect

    TURNBULL AD; LAO LL; COOPER WA; FU GY; GARABEDIAN P; KU LP; ZARNSTORFF MC

    2004-06-01

    Equilibrium and ideal magnetohydrodynamic (MHD) stability studies are reported for Compact Stellarator (ARIES-CS) reactor design equilibria based on a scaled three-period NCSX configuration and a two-period quasi-axisymmetric variant, the MHH2 stellarator. With a stabilizing shell at about twice the minor radius, robustly stable equilibria up to {beta}=6% are achievable. Recent experiments raise questions as to the applicability of linear MHD stability in stellarators since the predicted stability limits appear to be significantly exceeded. A context for interpreting this question, consistent with tokamak experience, is discussed; both the equilibria and nonlinear consequences need to be more carefully considered. Nonlinear stability is analyzed by computing solutions with highly resolved discontinuities to effectively simulate current sheets and islands. This yields {beta} limits in better agreement with measured values.

  5. 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.

  6. 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.

  7. 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.

  8. DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR

    SciTech Connect

    Rule, Keith; Perry, Erik; Parsells, Robert

    2003-02-27

    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 techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

  9. 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.

  10. 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.

  11. 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.

  12. Delaying sawtooth oscillations in the Compact Ignition Tokamak

    SciTech Connect

    Bateman, G.

    1986-09-01

    A combination of pellets, off-axis heating, and current ramp is used to delay the onset of sawtooth oscillations for 3.4 seconds and achieve ignition with less than 0.2-second confinement time in a 1-1/2-D BALDUR simulation of the Compact Ignition Tokamak. Deuterium and tritium pellets are injected into an initially cold, relatively low density plasma, where they cool the center and produce a very centrally peaked density profile. A centrally peaked density profile (n/sub e0// = 4.0) is subsequently maintained by an inward particle pinch. Twenty megawatts of auxiliary heating is applied halfway between the magnetic axis and the edge of the plasma for 2 seconds after the pellets are injected. The plasma ignites and then burns from the time the auxiliary heating is turned off until the first large sawtooth crash occurs at 3.4 seconds. The burn would be expected to continue after that only if the sawtooth period is sufficiently long (roughly 0.3 seconds or longer).

  13. 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.).

  14. 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.

  15. 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.

  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. A New Method for Shear Stabilization of Advanced Tokamak Reactors via Mode Converted Ion Bernstein Waves*

    NASA Astrophysics Data System (ADS)

    Sund, Richard; Scharer, John

    2002-11-01

    We examine a new method for generating sheared flows in advanced tokamak D-T reactors with the goal of creating and controlling internal transport barriers. Ion-Bernstein waves (IBWs) have the recognized capacity to create internal transport barriers through sheared plasma flows resulting from ion absorption. Under reactor conditions, the IBW can be generated by mode conversion of a fast magnetosonic wave incident from the high-field side (HFS) on the second harmonic resonance of a minority hydrogen component, with near 100200 MHz) minimizes parasitic absorption and permits the converted IBW to approach the fifth tritium harmonic. It also facilitates compact antennas and feeds, and efficient fast wave launch. Placement of the 5T absorption layer on the HFS is advantageous for shear production. The scheme is applicable to reactors with aspect ratio < 3 such that the conversion and absorption layers are both on the high field side of the magnetic axis. Various factors (adequate separation of the mode conversion layer from the magnetic axis, concentration of the fast wave near the midplane, large machine size, and plasma elongation) minimize poloidal field effects in the conversion zone and permit a slab analysis. We use a 1-D full-wave code to analyze the conversion and absorption. A 2-D ray-tracing code incorporating poloidal magnetic fields is used to follow the IBW for various equilibria. Within this analysis a weak bean shape appears most favorable. This is an attractive scheme for future advanced tokamak reactors. *Research supported by the Univ. of Wisconsin, Madison

  18. 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 (> or = 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/--600/sup 0/C it is possible to achieve a quasisteady 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.

  19. 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.

  20. Thin blanket designs for the Elongated Tokamak Commercial Reactor

    SciTech Connect

    Wong, C.P.C.; Bourque, R.F.; Cheng, E.T.; Creedon, R.L.; Schultz, K.R.

    1986-11-01

    The Elongated Tokamak (ET)* is an innovative concept that uses as highly elongated plasma (plasma height-to-width ratio of 6-10) to allow high plasma current and high toroidal betas. ET has the potential for the development of small-size, high-power density, low-cost fusion reactors using normal conducting coils. The elongated plasma shape is achieved by use of a continuous stack of PF coils parallel to the plasma surface on both inbound and outbound sides. To achieve plasma stability, these coil stacks must be located no further than one plasma minor radius from the plasma edge, greatly restricting the space available for blankets. In order to assess the potential of a small reactor, the authors evaluated and designed blankets 30 to 40 cm thick. Three different thin blanket designs were found to be acceptable: FLiBe self-cooled, helium-cooled lithium, and helium-cooled 17Li83Pb blanket designs. A lithium-cooled integrated blanket-coil design (BLITZ-coil) was also found to be suitable for the ET commercial reactors.

  1. Winding insulation in electromagnetic systems for Tokamak reactor plants

    NASA Astrophysics Data System (ADS)

    Maslov, V. V.; Trubachev, S. G.

    1985-01-01

    Magnetic containment of the plasma in nuclear fusion reactors of the Tokamak type requires electromagnets with insulation which must withstand high temperatures and thermal shocks as well as ionizing radiation in various forms and electric fields, and mechanical loads. Insulation materials to ensure adequate thermophysical and mechanical properties are evaluated, followed by design of insulation systems with satisfactory performance characteristics. Data on neutron fluence energy characteristics and radiation absorption doses during neutron interactions are essential for such an evaluation. Materials considered for insulation in electromagnets with superconductor and cryoresistance windings are glass mica tape with epoxy compound impregnation, glass cloth with epoxy compound impregnation (STE), polyimide glass cloth with adhesive coating (LSNL), glass Textolite with epoxy phenolic binder (STEN), epoxy resin paste with mineral fillers (PE), and polyurethane compound modified by epoxy resin with mineral filler (KPU).

  2. 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.

  3. Tokamak power reactor ignition and time dependent fractional power operation

    SciTech Connect

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.

  4. 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.

  5. Ignition in a tokamak reactor with INTOR-like parameters

    SciTech Connect

    Singer, C.E.; Seidl, F.G.P.; Post, D.E.; Rutherford, P.H.

    1980-08-01

    The requirements for ignition in a tokamak reactor with INTOR-like parameters were studied using a one-dimensional transport code. With empirical electron energy diffusivity Chi/sub e/, ignition was obtained with 60 to 75 MW of neutral beam injection at a volume average pressure ratio <..beta..> = 4 to 5% under a variety of conditions. Charging Chi/sub e/ gave ignition at the same <..beta..> if the plasma minor radius varied as a proportional to Chi/sub e/sup 1/2/. The maximum impurity concentration which allows ignition was found to be comparable to that for the much simpler case of a homogeneous plasma with radiative losses only. In long pulse simulations with efficient helium pumping, the maximum toroidal field ripple which allowed ignition was 2.0% (peak to peak) at the plasma edge. Ignition was maintained with over 99% recycling of helium ash using 5% less than maximum ripple.

  6. Evaluation of polarized fuels in a commercial deuterium/tritium tokamak reactor

    SciTech Connect

    Finn, P.A.; Brooks, J.N.; Ehst, D.A.; Gohar, Y.; Mattas, R.F.; Baker, C.C.

    1985-12-01

    The use of polarized fuels in commercial deuterium-tritium tokamak fusion reactors has been assessed. Some of the advantages cited for this fueling modification have been evaluated. Although the advantages are real, their magnitude is such that polarized fuels do not appear to be a significant reactor in increasing the attractiveness of commercial reactor designs. 17 figs., 19 tabs.

  7. Compact formulas for guiding-center orbits in axisymmetric tokamak geometry

    SciTech Connect

    Brizard, Alain J.

    2011-02-15

    Compact formulas for trapped-particle and passing-particle guiding-center orbits in axisymmetric tokamak geometry are given in terms of the Jacobi elliptic functions and complete elliptic integrals. These formulas can find applications in bounce-center kinetic theory as well as guiding-center Fokker-Planck kinetic theory.

  8. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    SciTech Connect

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  9. Hollow current profile scenarios for advanced tokamak reactor operations

    SciTech Connect

    Gourdain, P.-A.; Leboeuf, J.-N.

    2009-11-15

    Advanced tokamak scenarios are a possible approach to boosting reactor performances. Such schemes usually trigger current holes, a particular magnetohydrodynamics equilibrium where no current or pressure gradients exist in the core of the plasma. While such equilibria have large bootstrap fractions, flat pressure profiles in the plasma core may not be optimal for a reactor. However, moderate modifications of the equilibrium current profile can lead to diamagnetism where most of the pressure gradient is now balanced by poloidal currents and the toroidal magnetic field. In this paper, we consider the properties of diamagnetic current holes, also called ''dual equilibria,'' and demonstrate that fusion throughput can be significantly increased in such scenarios. Their stability is investigated using the DCON code. Plasmas with a beta peak of 30% and an average beta of 6% are found stable to both fixed and free-boundary modes with toroidal mode numbers n=1-4, as well as Mercier and high-n ballooning modes. This is not surprising as these scenarios have a normal beta close to 3.

  10. Radioactivation characteristics for the tokamak fusion test reactor

    SciTech Connect

    Ku, L.; Kolibal, J.G.

    1983-11-01

    Activation analysis has been conducted for several primary fusion blanket materials based on a model of a commercial tokamak fusion reactor design, STARFIRE. The blanket materials studied include two solid tritium breeders, viz., Li/sub 2/O and ..cap alpha..-LiAlO/sub 2/, and four candidate structural materials, viz., PCA stainless steel, V15Cr5Ti, Ti6Al4V, and Al-6063 alloys. The importance of breeder material activation is identified in terms of its impurity contents such as potassium, iron, nickel, molybdenum, and zirconium trace elements. The breeder activation is also discussed with regard to its potential for recycling and its impact on the lithium resource requirements. The structural material activation is analyzed based on two measures, volumetric radioactivity concentration and contact biological dose due to decay gamma emission. Using the radioactivity concentration measure, it is revealed that a substantial advantage exists from a viewpoint of radwaste management, which is inherent in fusion reactor designs based on potential low-activation alloys such as V15Cr5Ti, Ti6Al4V, and Al-6063. On the other hand, from the dose standpoint, the V15Cr5Ti alloy is found to be the only alloy for which one could realize a significant dose reduction (below 2.5 mrem/h) within about 100 yr after shutdown, possibly by some extrapolation on alloy purification techniques.

  11. 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.

  12. 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.

  13. 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.

  14. Ion cyclotron transmission spectroscopy in the Tokamak Fusion Test Reactor

    SciTech Connect

    Greene, G.J.

    1993-09-01

    The propagation of waves in the ion cyclotron range of frequencies has been investigated experimentally in the Tokamak Fusion Test Reactor. A small, broadband, radiofrequency (rf) magnetic probe located outside the plasma limiter, at a major radius near that of the plasma center, was excited with a low power, frequency swept source (1--200 MHz). Waves propagating to a distant location were detected with a second, identical probe. The rf transmission spectrum revealed a region of attenuation over a band of frequencies for which the minority fundamental resonance was located between the outer plasma edge and the major radius of the probe location. Distinct, non-overlapping attenuation bands were observed from hydrogen and helium-3 minority species; a distinct tritium band should be observed in future DT experiments. Rapid spectrum acquisition during a helium-3 gas puff experiment showed that the wave attenuation involved the plasma core and was not a surface effect. A model in which the received power varied exponentially with the minority density, averaged over the resonance region, fit the time evolution of the probe signal relatively well. Estimation of a 1-d tunneling parameter from the experimental observations is discussed. Minority concentrations of less than 0.5 % can be resolved with this measurement.

  15. Sufficient condition for velocity-space stability of the alpha particle distribution in a tokamak reactor

    SciTech Connect

    Cordey, J.G.; Goldston, R.J.; Mikkelsen, D.R.

    1980-08-01

    A condition is derived for the velocity-space distribution of suprathermal alpha particles to be monotonically decreasing with energy - and hence stable to homogeneous plasma instabilities - during the heating phase of a tokamak reactor. This stability condition is easily satisfied for presently envisaged neutral injection heating of reactors, but may be violated in strong heating of smaller plasmas or during fast compressional heating.

  16. 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.

  17. Overview of the TIBER II (Tokamak Ignition/Burn Experimental Reactor) design

    SciTech Connect

    Henning, C.D.; Logan, B.G.

    1987-10-16

    The TIBER II Tokamak Ignition/Burn Experimental Reactor design is the result of efforts by numerous people and institutions, including many fusion laboratories, universities, and industries. While subsystems will be covered extensively in other reports, this overview will attempt to place the work in perspective. Major features of the design are compact size, low cost, and steady-state operation. These are achieved through plasma shaping and innovative features such as radiation tolerant magnets and optimized shielding. While TIBER II can operate in a pulsed mode, steady-state is preferred for nuclear testing. Current drive is achieved by a combination of lower hybrid and neutral beams. In addition, 10 MW of ECR is added for disruption control and current drive profiling. The TIBER II design has been the US option in preparation for the International Thermonuclear Experimental Reactor (ITER). Other equivalent national designs are the NET in Europe, the FER in Japan and the OTR in the USSR. These designs will help set the basis for the new international design effort. 9 refs.

  18. 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.

  19. High-energy tritium beams as current drivers in tokamak reactors

    SciTech Connect

    Mikkelsen, D.R.; Grisham, L.R.

    1983-04-01

    The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams.

  20. Tritium pellet injector for the Tokamak Fusion Test Reactor

    SciTech Connect

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability.

  1. Tritium pellet injector for the Tokamak Fusion Test Reactor

    SciTech Connect

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-11-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability.

  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. Alpha particle losses from Tokamak Fusion Test Reactor deuterium-tritium plasmas

    SciTech Connect

    Darrow, D.S.; Zweben, S.J.; Batha, S.

    1996-01-01

    Because alpha particle losses can have a significant influence on tokamak reactor viability, the loss of deuterium-tritium alpha particles from the Tokamak Fusion Test Reactor (TFTR) 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.

  4. 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.

  5. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    NASA Astrophysics Data System (ADS)

    Sager, G. T.; Wong, C. P. C.; Kapich, D. D.; McDonald, C. F.; Schleicher, R. W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  6. 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.

  7. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    SciTech Connect

    Sager, G.T.; Wong, C.P.C.; Kapich, D.D.; McDonald, C.F.; Schleicher, R.W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankie and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  8. 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.

  9. Internal transport barrier simulation with pellet injection in tokamak and helical reactor plasmas

    NASA Astrophysics Data System (ADS)

    Higashiyama, Y.; Yamazaki, K.; Garcia, J.; Arimoto, H.; Shoji, T.

    2008-07-01

    In the future fusion reactor, the plasma density peaking is important for the increase in the fusion power gain. The density control and the internal transport barrier (ITB) formation due to the pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. Firstly, the pellet injection simulation is carried out including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of the high field side (HFS) pellet injection is clarified. Secondly, the ITB simulation with the pellet injection is carried out with the confinement improvement model based on the E×B shear effects, and it is found that the deep pellet penetration is helpful for the ITB formation as well as the plasma core fuelling in the reversed shear tokamak reactor, but the deep pellet penetration is not effective in the helical reactor.

  10. 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.

  11. 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.

  12. 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.

  13. INTOR: a first-generation tokamak experimental reactor

    SciTech Connect

    Stacey, Jr, W M; Gilleland, J R; Kulcinski, G L; Rutherford, P H

    1980-02-01

    An intensive, year-long, international evaluation of the next major tokamak beyond the generation of large experiments currently under construction was carried out during 1979. This evaluation consisted of the definition of objectives, an assessment of the physics and technology base and R and D needs and the identification of a set of parameters that physically characterize the machine.

  14. 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.

  15. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    SciTech Connect

    Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

    1993-11-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility.

  16. 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.

  17. 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.

  18. Core Plasma Characteristics of a Spherical Tokamak D-3He Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Shi, Bingren

    2005-04-01

    The magnetic fusion reactor using the advanced D-3He fuels has the advantage of much less-neutron productions so that the consequent damages to the first wall are less serious. If the establishment of this kind of reactor becomes realistic, the exploration of 3He on the moon will be largely motivated. Based on recent progresses in the spherical torus (ST) research, we have physically designed a D-3He fusion reactor using the extrapolated results from the ST experiments and also the present-day tokamak scaling. It is found that the reactor size significantly depends on the wall reflection coefficient of the synchrotron radiation and of the impurity contaminations. The secondary reaction between D-D that promptly leads to the D-T reaction producing 14 MeV neutrons is also estimated. Comparison of this D-3He ST reactor with the D-T reactor is made.

  19. Fuel systems for compact fast space reactors

    SciTech Connect

    Cox, C.M.; Dutt, D.S.; Karnesky, R.A.

    1983-12-01

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO/sub 2/ and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO/sub 2/-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux.

  20. 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.

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

    SciTech Connect

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

    2016-11-15

    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.

  2. 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.

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

    PubMed

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

    2006-12-01

    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.

  5. Spectra of heliumlike krypton from Tokamak Fusion Test Reactor plasmas

    SciTech Connect

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M. ); Beiersdorfer, P.; Osterheld, A. ); Smith, A. ); Fraenkel, B. )

    1993-08-16

    Experiments were conducted on TFTR to study the radiation of krypton which will be important for future tokamaks, such as ITER, for the diagnostic of the central ion temperature and for the control of the energy release from the plasma by radiative cooling. The total krypton radiation was monitored, and satellite spectra of Kr XXXV were recorded with a high-resolution crystal spectrometer. Radiative cooling and reduced particle recycling at the plasma edge region were observed, in reasonable agreement with modeling calculations which included radial transport.

  6. Parametric analysis of the ARIES-III D-{sup 3}He tokamak reactor

    SciTech Connect

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

    1993-08-01

    The multi-institutional ARIES study has completed a series of conceptual designs of tokamak fusion reactors that varies the assumed advances in technology and physics. The ARIES-III design uses a D- {sup 3}He fuel cycle and requires significant advances in physics to enhance economic attractiveness. The optimal design was characterized through systems analysis for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the other ARIES designs is included.

  7. 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.

  8. 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.

  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. 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.

  11. A consistency analysis on the tokamak reactor plasmas

    NASA Astrophysics Data System (ADS)

    Fukuyama, A.; Itoh, Sanae-I.; Itoh, Kimitaka

    1990-12-01

    The parameter regime which simultaneously fulfills the various physics constraints are looked for in the case of ITER grade tokamaks. The consistency analysis code is applied. It is found that if the energy confinement time reaches 1.6 times the prediction of the L-mode scaling law, the Q-value of about 4 is possible for the full current drive operation at the input power P sub in of 100 MW (Q is the ratio of fusion output and P sub in). In the ignition mode where half of the current is inductively sustained, Q approaches to 15 for this circulating power. If only the L-mode is realized, Q is about 1.5 for P sub in approximately 100 MW.

  12. 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.

  13. Remote maintenance design activities and research and development accomplishments for the Compact Ignition Tokamak

    SciTech Connect

    Spampinato, P.T.

    1988-01-01

    The use of deuterium-tritium (D-T) fuel for the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations. The remote operations consist of removing and replacing such components as first wall armor protection tiles, radio-frequency (rf) heating modules, and diagnostic modules. The major pieces of equipment being developed for maintenance activities internal to the vacuum vessel include an articulated boom manipulator (ABM), an inspection manipulator, and special tooling. For activities external to the vessel, the equipment includes a bridge-mounted manipulator system, decontamination equipment, hot cell equipment, and solid radiation-waste (rad-waste) handling and packaging equipment. The CIT Project is completing the conceptual design phase; research and development (R and D) activities, which include demonstrations of remote maintenance operations on full-size partial mock-ups are under way. 5 figs.

  14. Impact of physics and technology innovations on compact tokamak fusion pilot plants

    NASA Astrophysics Data System (ADS)

    Menard, Jonathan

    2016-10-01

    missions including tritium breeding, high-thermal-efficiency liquid metal breeding blankets are attractive, and novel immersion blankets offer the potential for simplified fabrication and maintenance and reduced cost. Lastly, the optimal aspect ratio for a tokamak pilot plant is likely a function of the device mission and associated cost, with low aspect ratio favored for minimizing TF magnet mass and higher aspect ratio favored for minimizing blanket mass. The interplay between a range of physics and technology innovations for enabling compact pilot plants will be described. This work was supported by U.S. DOE Contract Number DE-AC02-09CH11466.

  15. Spectra of heliumlike krypton from tokamak fusion test reactor plasmas

    SciTech Connect

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Beiersdorfer, P.; Osterheld, A.; Smith, A.; Fraenkel, B.

    1993-04-01

    Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 {Angstrom} including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport.

  16. Spectra of heliumlike krypton from tokamak fusion test reactor plasmas

    SciTech Connect

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M. . Plasma Physics Lab.); Beiersdorfer, P.; Osterheld, A. ); Smith, A. ); Fraenkel, B. )

    1993-04-01

    Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 [Angstrom] including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport.

  17. 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.

  18. 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}.

  19. 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.

  20. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    SciTech Connect

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

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

  1. 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.

  2. 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.

  3. 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.

  4. Analysis of pumping requirement for exhausting duct in close vicinity of divertor in Tokamak Reactor

    SciTech Connect

    Saito, S.; Abe, T.; Fujisawa, N.; Sugihara, M.; Veda, K.

    1983-11-01

    An improved method for Monte Carlo simulation is described to calculate the neutral-particle transport in a divertor throat and to evaluate the helium removal efficiency from a burning plasma. The required pumping speed for the helium removal is discussed with special emphasis placed on the effects of long exhausting duct and of scrape-off plasma variables. The analysis for International Tokamak Reactor (INTOR) single null divertor suggests a possibility that the pumping requirement for INTOR could be drastically eased--e.g., <10/sup 4/ l/s, for the high scrape-off plasma density of the order of 10/sup 13/ cm/sup -3/.

  5. 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.

  6. A tritium-compatible piezoelectric valve for the Tokamak Fusion Test Reactor

    SciTech Connect

    Coffin, D.O.; Cole, S.P.; Wilhelm, R.C.

    1988-09-01

    A commercial piezoelectric valve has been modified to enhance its tritium compatibility, enabling it to provide about 125 tritium gas injections planned for Princeton's Tokamak Fusion Test Reactor (TFTR). The valve was modified at Los Alamos National Laboratory (LANL), exposed to progressively higher concentrations to tritium gas, and repeatedly tested for performance with tritium. The modified valve meets the basic TFTR flow requirements (50 torr L/s), and it survived 3400 torr hr exposure to tritium with neither decrease in performance nor significant leakage across the closed valve. A totally tritium-compatible piezo valve, containing no organic materials, is also proposed and described.

  7. 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.

  8. Combined electron cyclotron ray tracing and transport code studies in the Compact Ignition Tokamak

    SciTech Connect

    Bonoli, P.T.; Englade, R.C.; Porkolab, M. . Plasma Fusion Center); Kritz, A.H. )

    1990-01-01

    The Compact Ignition Tokamak (CIT) operating scenario is envisaged to consist of a start-up phase in which the toroidal magnetic field (B{sub T}), plasma current (I{sub P}) and central electron density (n{sub e}(0)) are simultaneously ramped, followed by a burn cycle and a ramp down cycle. Electron cyclotron radio-frequency (ECRF) power at fixed frequency is ideally suited to heating during the ramp up phase of CIT. The angle of injection of the incident microwave beam can be varied as the toroidal field is ramped, so as to maintain central rf-power deposition. Furthermore, since the EC wave is a propagating mode in vacuum, relatively high power densities can be easily coupled into a compact device. Finally, we note that recent advances in source technology makes ECR heating at (280--310) GHz a viable option. In order to realistically simulate this ramp up scenario, a combined code has been developed in which ECRF ray tracing and absorption, and MHD equilibrium calculation, and thermal and particle transport are treated self-consistently. Previous studies of electron cyclotron resonance heating in CIT were carried out using model profiles of rf absorption based on stand-alone ray tracing and absorption calculations. In addition, these studies held the plasma current, toroidal magnetic field, and density constant in time. In the present work, a time variation in these plasma quantities is imposed and ECRF power deposition is re-calculated during rf injection so as to more realistically and self-consistently simulate ECH-assisted start-up in CIT. The most recent CIT configuration calls for a toroidal field ramp to 11T. Thus, the present studies consider the use of an ECRF source frequency of 308 GHz, with EC waves launched in the ordinary mode of polarization.

  9. 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.

  10. 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.

  11. Implications of the DD fuel cycle on tokamak reactor technology considerations

    SciTech Connect

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

    1980-01-01

    The initial results of a study to develop a conceptual design of a DD tokamak reactor are reported in this paper. A comprehensive steady-state power and particle balance code has been used to survey a variety of design points. A reference set of parameters (R/sub 0/ = 8.6 m, B/sub max/ = 14 T, ..beta.. = 11%, P/sub th/ = 2500 to 2700 MW) has been selected for further study. The concepts of rf driven currents and a limiter-vacuum system have been examined and appear to offer promise. However, the penalty of rf current drive, in terms of recirculating power, is relatively larger for DD than for DT reactors. Blanket engineering studies have emphasized the optimization of energy multiplication and inboard blanket thickness.

  12. Alpha-particle losses in compact torsatron reactors

    SciTech Connect

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

    1989-01-01

    Loss of alpha particles in compact torsatron reactors is studied. For 6, 9, and 12 field period reactors, the direct loss is a relatively weak function of radius and energy and varies from approx. =33% for M = 6 to approx. =18% for M = 12. Loss of alpha particles through scattering into the loss region is calculated using the Fokker-Plank equation for fast ions and found to contribute an additional alpha-particle energy loss of approx. =15%. The consequences of these relatively large losses for torsatron reactor design are discussed. The relationship between the direct particle losses and the magnetic field structure is also studied. Orbit losses from a variety of stellarator configurations are calculated and a figure-of-merit that characterizes the orbit confinement of a magnetic configuration is deduced from these calculations. This figure-of-merit is used to show how the direct losses might be reduced at low aspect-ratio. Effects of finite beta on the direct particle losses are also addressed, and are shown to significantly increase the direct losses in some configurations. 15 refs., 8 figs.

  13. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    SciTech Connect

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  14. Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

    SciTech Connect

    Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

    1987-10-01

    Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared to conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA . cm/sup -2/ (at 77 K and approx.10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA . cm/sup -2/ most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of approx.7% are forecast for the overall capital cost of the power plant in the best case. An additional approx.3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs.

  15. Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

    SciTech Connect

    Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

    1988-08-01

    Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA /center dot/ cm/sup /minus/2/ (at 77 K and /approximately/10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA /center dot/ cm/sup /minus/2/ most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of /approximately/7% are forecast for the overall capital cost of the power plant in the best case. An additional /approximately/3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs.

  16. 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.

  17. Comparing linear ion-temperature-gradient-driven mode stability of the National Compact Stellarator Experiment and a shaped tokamak

    SciTech Connect

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

    2013-02-15

    One metric for comparing confinement properties of different magnetic fusion energy configurations 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 temperature gradient (ITG) mode in the National Compact Stellarator Experiment (NCSX) and a prototypical shaped tokamak, based on the profiles 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 effects of the stellarator configuration offset 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 adiabatic ITG mode critical gradients, although 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/L{sub T} Almost-Equal-To 1.5 a/L{sub T,crit}, when it surpassed the tokamak's. A discussion of the results presented with respect to a/L{sub T} vs. R/L{sub T} is included.

  18. 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.

  19. 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.

  20. Ion temperature from tangential charge exchange neutral analysis on the Tokamak Fusion Test Reactor

    SciTech Connect

    Fiore, C.L.; Medley, S.S.; Hammett, G.W.; Kaita, R.; Roquemore, A.L.; Scott, S.D.

    1987-09-01

    Fokker-Planck simulations of the Tokamak Fusion Test Reactor (TFTR) energetic ion mode discharges were performed to evaluate the utility of deriving the central ion temperature, T/sub i/, from deuterium neutral beam charge exchange spectra above the neutral beam injection energy. The T/sub i/ values obtained from fitting the calculated spectra obtained from sightlines nearly tangent to the neutral beam injection radius reproduce the central ion temperature within +-10% over the full range of TFTR energetic ion mode parameters. The code simulations demonstrate that the ion temperature obtained from the high energy tangential deuterium charge exchange spectrum is insensitive to variations in the plasma density, Z/sub eff/, plasma current, loop voltage, and injected neutral beam power and energy. Use of this method to reduce charge exchange data from TFTR energetic ion mode plasmas is demonstrated. 17 refs., 22 figs., 2 tabs.

  1. Initial confinement studies of ohmically heated plasmas in the Tokamak Fusion Test Reactor

    SciTech Connect

    Efthimion, P.C.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.L.; Coonrod, J.; Davis, S.; Dylla, H.F.; Fonck, R.; Furth, H.P.

    1984-06-01

    Initial operation of the Tokamak Fusion Test Reactor (TFTR) has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1 to 0.2 s for a line-average density range (anti n/sub e/) of 1 to 2.5 x 10/sup 19/ m/sup -3/ with electron temperatures of T/sub e/(o) approx. 1.2 to 2.2 keV, ion temperatures of T/sub i/(o) approx. 0.9 to 1.5 keV, and Z/sub eff/ approx. 3. A comparison of PLT, PDX, and TFTR plasma confinement supports a dimension-cubed scaling law.

  2. Mirror test for International Thermonuclear Experimental Reactor at the JET tokamak: An overview of the program

    NASA Astrophysics Data System (ADS)

    Rubel, M. J.; De Temmerman, G.; Coad, J. P.; Vince, J.; Drake, J. R.; Le Guern, F.; Murari, A.; Pitts, R. A.; Walker, C.; JET-EFDA Contributors

    2006-06-01

    Metallic mirrors will be essential components of all optical spectroscopy and imaging systems for plasma diagnosis that will be used at the next-step magnetic fusion experiment, International Thermonuclear Experimental Reactor (ITER). Any change of the mirror performance, in particular, reflectivity, will influence the quality and reliability of detected signals. At the instigation of the ITER Design Team, a dedicated technical and experimental activity aiming at the assessment of mirror surface degradation as a result of exposure to the plasma has been initiated on the JET tokamak. This article provides a comprehensive overview of the mirror test program, including design details of the mirror samples and their supports, their locations within JET, and the issue of optical characterization of the mirrors both before and after exposure. The postexposure characterization is particularly challenging in JET as a consequence of an environment in which both tritium and beryllium are present.

  3. Selected transport studies of a tokamak-based DEMO fusion reactor

    NASA Astrophysics Data System (ADS)

    Fable, E.; Wenninger, R.; Kemp, R.

    2017-02-01

    As a next-step in the tokamak-based fusion programme, the DEMO fusion reactor is foreseen to produce relevant output electricity, in the order of  ∼500 MW delivered to the network. The scenarios that are being presently investigated consist of a pulsed device, called DEMO1, and a steady-state device, called DEMO2. In this work, which is focused on the pulsed device DEMO1, scenarios are studied from the point of view of core transport, to assess plasma performance and limitations due to core microinstabilities. The role of radiated power, aspect ratio, and height of temperature pedestal are assessed as they impact both core energy and particle transport. Open issues in this framework are also discussed.

  4. 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).

  5. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    NASA Astrophysics Data System (ADS)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

  6. 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

  7. Application of a compact reactor to a submersible research vessel

    NASA Astrophysics Data System (ADS)

    Kusunoki, Tsuyoshi; Fujimoto, Hiromi; Nagata, Yutaka; Takahashi, Teruo; Ishida, Toshihisa

    Understanding of the global climate system is necessary to clarify the mechanism of the climate change and predict the fluctuations of the global climate and circumstance. Since the sign of these fluctuations can be predicted in the arctic zone where these fluctuations strongly affect the circumstance, the observation in this zone is very important. Observation data to be obtained there, however, are very few and limited due to difficulty caused by a thick ice covering on a wide-ranged area. A conceptual design of a submersible research vessel with a nuclear power source is studied to cope with this difficulty. The nuclear power is suitable for a submersible research vessel specialized to the undersea of the arctic zone, because it enables to operate for a long time without oxide or fuel supply. By taking account of working conditions in observation, the basic specifications of the vessel are decided; the total weight and the length of it are 500 t, 40 m, respectively, and the maximum ship speed is 12 knots. Two sets of nuclear compact reactor, SCR, light-weighted and of enhanced safety are installed in the vessel to supply the total electricity of 500 kW. This vessel is capable of providing a very rapid, fruitful activity in observation and research.

  8. Calculations of alpha particle loss for reversed magnetic shear in the Tokamak Fusion Test Reactor

    SciTech Connect

    Redi, M.H.; White, R.B.; Batha, S.H.; Levinton, F.M.; McCune, D.C.

    1997-03-01

    Hamiltonian coordinate, guiding center code calculations of the toroidal field ripple loss of alpha particles from a reversed shear plasma predict both total alpha losses and ripple diffusion losses to be greater than those from a comparable non-reversed magnetic shear plasma in the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. High central q is found to increase alpha ripple losses as well as first orbit losses of alphas in the reversed shear simulations. A simple ripple loss model, benchmarked against the guiding center code, is found to work satisfactorily in transport analysis modelling of reversed and monotonic shear scenarios. Alpha ripple transport on TFTR affects ions within r/a=0.5, not at the plasma edge. The entire plasma is above threshold for stochastic ripple loss of alpha particles at birth energy in the reversed shear case simulated, so that all trapped 3.5 MeV alphas are lost stochastically or through prompt losses. The 40% alpha particle loss predictions for TFTR suggest that reduction of toroidal field ripple will be a critical issue in the design of a reversed shear fusion reactor.

  9. Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

    SciTech Connect

    Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki; Itoh, Masami; Sekine, Tadashi

    2005-11-15

    An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

  10. 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.

  11. 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.

  12. 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.

  13. Ignition and burn criteria for D/sup 3/He tokamak and spherical torus reactors

    SciTech Connect

    Galambos, J.D.; Peng, Y-K. M.

    1989-01-01

    D-/sup 3/He ignition and burn criteria for tokamaks and spherical torus reactors are examined in a global analysis with profile corrections. Particle confinement and ash buildup effects are included with the power balance, which results in an increased sensitivity of the ignition criteria to losses via brehmsstrahlung and synchrotron radiation. Plasma beta scaling via an /epsilon//beta//sub p/ limit provides the needed aspect ratio (A) dependence, and permits an analysis in all A values of the first and second stability regimes. Energy confinement time (/tau//sub E/) associated with particle diffusion (/tau//sub p/) and energy conduction (/tau//sub c/) are used; parabolic profile are assumed with exponents /alpha//sub n/ = 1.0 and /alpha//sub T/ = 1.5; and we set /tau//sub p/ = 2/tau//sub c/. The ignition condition for minimum n/tau//sub E/ is found to be sensitive to /beta/ but not to the magnetic field. Steady state burn in second stability tokamaks (/epsilon//beta//sub p/ /ge/ 0.6) at high A (>4) with average synchrotron wall reflectivities below 95% requires n/tau//sub E/ above 5 /times/ 10/sup 21/ m/sup /minus/3/s or strong plasma elongation (/kappa/ > 3). Ignition in a spherical tori can be achieved with wall reflectivities below 80% and at n/tau//sub E/ /le/ 10/sup 21/ m/sup /minus/3/ s, without requiring strong plasma shaping or /epsilon//beta//sub p/ > 0.6. The need to minimize n/tau//sub E/ for ignition and burn strongly limits the synchrotron radiation loss to less than 20% of the fusion power for all values of A. Synchrotron power fractions can be increased to 40% by increasing n/tau//sub E/ to its upperbound of ignition. Further increases of this fraction can be obtained only by assuming preferential ash removal. 19 refs., 8 figs.

  14. Integrated Analysis on the Current Profile and the Operational Scenario of D-3He Spherical Tokamak Reactors

    NASA Astrophysics Data System (ADS)

    Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Mano, Junji

    We applied the TOTAL (toroidal transport analysis linkage) simulation code for the analysis of the operational scenario of D-3He spherical tokamak reactors with high beta values and high bootstrap current fractions. Several technical elements, such as the control of the fuel ratio or selective exhaust of the α particle, need to be developed to establish steady-state burning. Negative magnetic shear configuration is a candidate for the high bootstrap current fraction operation.

  15. 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

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

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

    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

  17. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    SciTech Connect

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

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) 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 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 (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

  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. The impact of thermal fatigue and other design limits on pulsed commercial tokamak reactor design

    NASA Astrophysics Data System (ADS)

    LeClaire, R. J.; Meyer, J. E.

    1985-10-01

    Methods of analysis for fusion first-wall design are developed. Several design limits have been evaluated and combined to present tradeoffs in the form of design windows. These considerations include limits related to thermal fatigue, primary membrane strength, displacement under loading, ratcheting, radiation damage, and plasma-wall interactions. Special emphasis is placed on the investigation of thermal fatigue using a two-dimensional treatment of a tubular first-wall configuration. The work is motivated by the proposal of the Ultra Long Pulse Commercial Reactor (ULTR), a machine capable of delivering plasma burn pulses of up to 24 hr in length. The present work looks in detail at the impact of pertinent characteristics of the first-wall design, such as pulse length, coolant pressure, first-wall thickness, and first-wall lifetime on the structural effects considered. Computer programs are developed and are used to consider several major structural effects on a cylindrical first-wall element for both 316 stainless steel and vanadium alloy. Results indicate that short pulse lengths (greater than a few minutes) can be tolerated in tokamak operation. For stainless steel this is true for heat depositions up to 1 MW/m2, while vanadium can tolerate heat depositions as high as 2 MW/m2. Long pulse operation can be used to increase modestly the allowable heat deposition or to increase useful wall thickness by 1-2 mm. It appears that irradiation swelling and embrittlement, not fatigue, ultimately limits the first-wall design.

  20. 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.

  1. 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.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.

    1994-04-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). 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 Fluorinet{sup {trademark}} 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 {approximately} 10 megawatts (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.

  2. Ion cyclotron range of frequency heating on the Tokamak Fusion Test Reactor

    SciTech Connect

    Taylor, G.; Bell, M.G.; Biglari, H.; Bitter, M.; Bretz, N.L.; Budny, R.; Chen, L.; Darrow, D.; Efthimion, P.C.; Ernst, D.; Fredrickson, E.; Fu, G.Y.; Grek, B.; Grisham, L.; Hammett, G.; Hosea, J.C.; Janos, A.; Jassby, D.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Majeski, R.; Mansfield, D.K.; Mazzucato, E.; Medley, S.S.; Mueller, D.; Nazikian, R.; Owens, D.K.; Paul, S.; Park, H.; Phillips, C.K.; Rogers, J.H.; Schilling, G.; Schivell, J.; Schmidt, G.L.; Stevens, J.E.; Stratton, B.C.; Strachan, J.D.; Synakowski, E.; Wilson, J.R.; Wong, K.L.; Zweben, S.J.; Baylor, L.; Bush, C.E.; Goldfinger, R.C.; Hoffman, D.J.; Murakami, M.; Qualls, A.L.; Rasmussen, D.; Machuzak, J.; Rimini, F.; Chang, Z.

    1993-06-01

    The complete ion cyclotron range of frequency (ICRF) heating system for the Tokamak Fusion Test Reactor (TFTR), consisting of four antennas and six generators designed to deliver 12.5 MW to the TFTR plasma, has now been installed. Recently a series of experiments has been conducted to explore the effect of ICRF heating on the performance of low recycling, Supershot plasmas in minority and non-resonant electron heating regimes. The addition of up to 7.4 MW of ICRF power to full size (R {approximately} 2.6 m, a {approximately} 0.95 m), helium-3 minority, deuterium Supershots heated with up to 30 MW of deuterium neutral beam injection has resulted in a significant increase in core electron temperature ({delta}T{sub e}=3--4 key). Simulations of equivalent deuterium-tritium (D-T) Supershots predict that such ICRF heating should result in an increase in {beta}{sub alpha}(O) {approximately} 30%. Direct electron heating has been observed and has been found to be in agreement with theory. ICRF heating has also been coupled to neutral beam heated plasmas fueled by frozen deuterium pellets. In addition ICRF heated energetic ion tails have been used to simulate fusion alpha particles in high recycling plasmas. Up to 11.4 MW of ICRF heating has been coupled into a hydrogen minority, high recycling helium plasma and the first observation of the toroidal Alfven eigenmode (TAE) instability driven by the energetic proton tail has been made in this regime.

  3. The effect of limiter conditioning on the Tokamak Fusion Test Reactor edge plasma

    SciTech Connect

    Kilpatrick, S.J.; Manos, D.M.; Nyberg, I.; Ramsey, A.T.; Stratton, B.C.; Timberlake, J.; Ulrickson, M.J.; Pitcher, C.S.; Dylla, H.F.

    1991-07-01

    Measurements by moveable Langmuir probes and edge spectroscopy diagnostics have documented the conditioning effect of low density helium-initiated discharge sequences on the Tokamak Fusion Test Reactor (TFTR) edge plasma. Langmuir probe measurements show in general that the edge electron density n{sub e} decreases by less than a factor of 2 while the edge electron temperature T{sub e} doubles. Radial profiles to the plasma boundary show that the density scrape-off length increases somewhat while the temperature scrape-off length decreases substantially. The particle flux density is unaffected. The spectral emission of C 2 decreases by a factor of 2, a much smaller change than that exhibited by the D{sub {alpha}} signal. These results complement previous accounts of the conditioning technique. Comparisons of these He conditioning measurements are made to edge measurements during a deuterium density scan experiment, showing many similarities, and to an existing edge model of the conditioning process, showing qualitative agreement. 20 refs., 5 figs.

  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. Equilibrium evolution on the resistive time scale in a tokamak reactor

    SciTech Connect

    Holmes, J.A.; Peng, Y.K.M.; Rothe, K.E.

    1981-04-01

    In the Engineering Test Facility (ETF) the plasma pulse duration is expected to be hundreds of seconds, which is comparable to the resistive time scale that governs the resistive diffusion of the equilibrium. The resistive evolution of the safety factor q profile may, for MHD stability reasons, limit the duration of the plasma burn in a tokamak reactor. It may be possible to control this evolution and extend the plasma burn time through proper profile tailoring. We study the evolution of the q profile on the resistive time scale numerically using a one-and-one-half-dimensional (1 1/2-D) single fluid transport code. Two high beta (anti ..beta../sub T/ approx. 7 to 16%) cases are considered: (1) a beam-driven hydrogen plasma with no nuclear alpha heating for which the beam energy is used as a device to control the tempature profile, and (2) an ignited D-T plasma in which the neutral injection has been turned off.

  6. Differential-phase reflectometry for edge profile measurements on Tokamak fusion test reactor

    SciTech Connect

    Hanson, G.R.; Wilgen, J.B.; Bigelow, T.S.; Collazo, I.; England, A.C.; Murakami, M.; Rasmussen, D.A.; Wilson, J.R. )

    1995-01-01

    Edge electron density profile measurements, including the scrape-off layer, have been made during ion cyclotron range of frequency (ICRF) heating with the two-frequency differential-phase reflectometer installed on an ICRF antenna on the Tokamak fusion test reactor (TFTR). This system probes the plasma using the extraordinary mode with two signals swept from 90 to 118 GHz, while maintaining a fixed-difference frequency of 125 MHz. The extraordinary mode is used to obtain density profiles in the range of 1[times]10[sup 11]--3[times]10[sup 13] cm[sup [minus]3] in high-field (4.5--4.9 T) full-size ([ital R][sub 0]=2.62 m, [ital a]=0.96 m) TFTR plasmas. The reflectometer launcher is located in an ICRF antenna and views the plasma through a small penetration in the center of the Faraday shield. A 26-m-long overmoded waveguide run connects the launcher to the reflectometer microwave electronics. Profile measurements made with this reflectometer system will be presented along with a discussion of the characteristics of this differential phase reflectometer and data analysis.

  7. The effect of limiter conditioning on the Tokamak Fusion Test Reactor edge plasma

    SciTech Connect

    Kilpatrick, S.J.; Pitcher, C.S.; Dylla, H.F.; Manos, D.M.; Nyberg, I.; Ramsey, A.T.; Stratton, B.C.; Timberlake, J.; Ulrickson, M.J. )

    1991-05-01

    Measurements by moveable Langmuir probes and edge spectroscopy diagnostics have documented the conditioning effect of low-density helium-initiated discharge sequences on the Tokamak Fusion Test Reactor edge plasma. Langmuir probe measurements shown in general that the edge electron density {ital n}{sub {ital e}} decreases by less than a factor of 2 while the edge electron temperature {ital T}{sub {ital e}} doubles. Radial profiles to the plasma boundary show that the density scrape-off length increases somewhat while the temperature scrape-off length decreases substantially. The particle flux density is unaffected. The spectral emission of C II decreases by a factor of 2, a much smaller change than that exhibited by the {ital D}{sub {alpha}} signal. These results complement previous accounts of the conditioning technique. Comparisons of these He conditioning measurements are made to edge measurements during a deuterium density scan experiment, showing many similarities, and to an existing edge model of the conditioning process, showing qualitative agreement.

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

    NASA Astrophysics Data System (ADS)

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

    2015-03-01

    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 βn ≈ 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.

  9. Conceptual physics design of a compact torus fusion reactor (CTOR)

    SciTech Connect

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

    1980-07-01

    The general approach to fusion power embodied by field-reversed plasmoid configurations is reviewed within the context of a power reactor. A simple analytic formulation is developed and applied to the field-reversed theta pinch as a core plasma for a thermonuclear reactor. These calculations and results are based on a minimum power constraint and will serve as a basis for more exact and detailed reactor modeling.

  10. 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.

  11. An evolutionary approach for a compact-split-core reactor

    NASA Technical Reports Server (NTRS)

    Breitwieser, R.; Lantz, E.

    1973-01-01

    An economical approach for advanced reactor power development is presented, and systems that result from the several stages of this plan are described. The development starts with a highly modularized heat pipe, radioisotopic design and evolves into a low specific weight high performance reactor system.

  12. A Compact Nuclear Fusion Reactor for Space Flights

    SciTech Connect

    Nastoyashchiy, Anatoly F.

    2006-05-02

    A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products -- charged particles. The energy balance considerably improves, as synchrotron radiation turn out 'captured' in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

  13. Spent fuel utilization in a compact traveling wave reactor

    NASA Astrophysics Data System (ADS)

    Hartanto, Donny; Kim, Yonghee

    2012-06-01

    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.

  14. A programmatic framework for the Tokamak Physics Experiment (TPX)

    SciTech Connect

    Thomassen, K.I. ); Goldston, R.J. ); Neilson, G.H. )

    1993-09-01

    Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the [open quotes]power breakeven[close quotes] level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program.

  15. Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

    SciTech Connect

    Budny, R.V.; Alper, B.; Borba, D.; Cordey, J.G.; Ernst, D.R.; Gowers, C.

    2001-02-02

    First results of gyrokinetic analysis of JET [Joint European Torus] ELMy [Edge Localized Modes] H-mode [high-confinement modes] plasmas are presented. ELMy H-mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER [International Thermonuclear Experimental Reactor]. Relatively high performance for long duration has been achieved and the scaling appears to be favorable. It will be necessary to sustain low Z(subscript eff) and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition between Type I to III ELMs at the heating power threshold. Linear gyrokinetic analysis gives the growth rate, gamma(subscript lin) of the fastest growing modes. The flow-shearing rate omega(subscript ExB) and gamma(subscript lin) are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high |gamma(subscript ExB)|/gamma(subscript lin) near the top of the pedestal for high confinement.

  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. New interpretation of alpha-particle-driven instabilities in deuterium-tritium experiments on the Tokamak Fusion Test Reactor.

    PubMed

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

    2003-09-19

    The original description of alpha particle driven instabilities in the Tokamak Fusion Test Reactor in terms of toroidal Alfvén 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 antiballooning 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. Alpha Particle-Driven Toroidal Alfven Eigenmodes in Tokamak Fusion Test Reactor Deuterium-Tritium Plasmas: Theory and Experiments

    SciTech Connect

    Budny, R.; Chang, Z.; Fu, G.Y.; Nazikian, R.

    1998-07-09

    The toroidal Alfvén eigenmodes (TAE) in the Tokamak Fusion Test Reactor [K. Young, et al., Plasma Phys. Controlled Fusion 26, 11 (1984)]deuterium-tritium plasmas are analyzed using the NOVA-K code [C.Z. Cheng, Phys. Reports 211, 1 (1992)]. The theoretical results are compared with the experimental measurements in detail. In most cases, the theory agrees with the observations in terms of mode frequency, mode structure, and mode stability. However, one mode with toroidal mode number n = 2 is observed to be poloidally localized on the high field side of the magnetic axis with a mode frequency substantially below the TAE frequency.

  20. PPPL tokamak program

    SciTech Connect

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT.

  1. Compact reversed-field pinch reactor (CRFPR): a high-density approach to magnetic fusion energy

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.; Byrne, R.N.; Dobrott, D.

    1982-01-01

    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 cost-optimized Compact RFP Reactor (CRFPR) design would operate with fusion-power-core power densities and mass utilizations that are comparable to fission power plants and are an order of magnitude more favorable than the conventional fusion approaches. A comprehensive system model predicts the CRFPR point design to be surprisingly resilient to changes in key, but relatively unknown, physics and systems parameters.

  2. 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)

  3. 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.

  4. Shielding Analysis of a Small Compact Space Nuclear Reactor

    DTIC Science & Technology

    1987-08-01

    87117-6008 I Dl’ _ -•, AFWL-TR-G7- 04 This final report was prepared by the Air Force Weapons Laboratory, Kirtland Air Force Base, New Mexico , under Job...Reactor Lennard W. Lee Jr. B.S. Nuclear Engineering, Mississippi State University, 1983 M.S. Nuclear Engineering, Univeisity of New Mexico , 1987 Early...6.15. FEMPID Reactor Model. 71 7.0 GeneraL.n4 Dese Response Functic-ns FFMP2D and FEMPID will generate gamma and neutron radiation fluxes which may be

  5. Collaboration on Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Text Reactor. Final report

    SciTech Connect

    Intrator, T.

    2000-06-01

    This proposal was peer reviewed and funded as a Collaboration on ''Low Phase Speed Radio Frequency Current Drive Experiments at the Tokamak Fusion Test Reactor''. The original plans we had were to carry out the collaboration proposal by including a post doctoral scientist stationed at PPPL. In response to a 60+% funding cut, all expenses were radically pruned. The post doctoral position was eliminated, and the Principal Investigator (T. Intrator) carried out the brunt of the collaboration. Visits to TFTR enabled T. Intrator to set up access to the TFTR computing network, database, and get familiar with the new antennas that were being installed in TFTR during an up to air. One unfortunate result of the budget squeeze that TFTR felt for its last year of operation was that the experiments that we specifically got funded to perform were not granted run time on TFTR., On the other hand we carried out some modeling of the electric field structure around the four strap direct launch Ion Bernstein Wave (IBW) antenna that was operated on TFTR. This turned out to be a useful exercise and shed some light on the operational characteristics of the IBW antenna and its coupling to the plasma. Because of this turn of events, the project was renamed ''Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Test Reactor''.

  6. 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.

  7. 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].

  8. 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.

  9. 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.

  10. Nuclear aspects of tokamak fusion test reactor (TFTR) diagnostics and instrumentation

    SciTech Connect

    Young, K.M.

    1982-01-01

    There are five principal aspects of the nuclear radiation from the high temperature plasmas of TFTR on its plasma diagnostic equipment. i) Important information about the plasma properties to be obtained from measurement of the neutrons, or other fusion reaction products. ii) Experimental studies to give design data for future tokamak devices and their instrumentation. iii) Transient noise or damage effects on the array of detectors for the collection of physics data about the plasma. iv) The effect of tritium on detectors that necessarily are in vacuum, directly connected to the tokamak vacuum vessel. v) Damage of diagnostic components mounted close to the vacuum vessel. Each of these topics will be addressed after a brief description of the TFTR tokamak and its radiation environment.

  11. 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.

  12. 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.

  13. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) T4B Experiment

    NASA Astrophysics Data System (ADS)

    McGuire, Thomas

    2016-10-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. The CFR uses a diamagnetic, high beta, magnetically encapsulated, linear ring cusp plasma confinement scheme. The goal of the T4B experiment is to demonstrate a suitable plasma target for heating experiments and to characterize the behavior of plasma sources in the CFR configuration. The design of the T4B experiment will be presented, including discussion of predicted behavior, plasma sources, heating mechanisms, diagnostics suite and relevant numerical modeling. ©2016 Lockheed Martin Corporation. All Rights Reserved.

  14. Evaluation of weldments in Type 21-6-9 stainless steel for Compact Ignition Tokamak structural applications: Phase 1

    SciTech Connect

    Alexander, D.J.; Goodwin, G.M.; Bloom, E.E.

    1991-06-01

    Primary design considerations for the Compact Ignition Tokamak toroidal field-coil cases are yield strength and toughness in the temperature range from 77 to 300 K. Type 21-6-9 stainless steel, also still known by its original Armco Steel Company trade name Nitronic 40, is the proposed alloy for this application. It has high yield strength and usually adequate base metal toughness, but weldments in thick sections have not been adequately characterized in terms of mechanical properties or hot-cracking propensity. In this study, weldability of the alloy in heavy sections and the mechanical properties of the resultant welds were investigated including tensile yield strength and Charpy V-notch toughness at 77 K and room temperature. Weldments were made in four different base metals using seven different filler metals. None of the weldments showed any indication of hot-cracking problems. All base metals, including weldment heat-affected zones, were found to have adequate strength and impact toughness at both test temperatures. Weld metals, on the other hand, except ERNiCr-3 and ENiCrFe-3 had impact toughnesses of less than 67 J at 77 K. Inconel 82 had an average weld metal impact toughness of over 135 J at 77 K, and although its strength at 77 K is less than that of type 21-6-9 base metal, at this point it is considered to be the first-choice filler metal. Phase 2 of this program will concentrate on composition refinement and process/procedure optimization for the generic ERNiCr-3 composition and will generate a design data base for base and weld metal, including tensile, fracture toughness, and crack growth rate data.

  15. 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.

  16. A Compact Torus Fusion Reactor Utilizing a Continuously Generated Strings of CT's. The CT String Reactor, CTSR.

    SciTech Connect

    Hartman, C W; Reisman, D B; McLean, H S; Thomas, J

    2007-05-30

    A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal field opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall.

  17. What we should do for transition from current tokamaks to fusion-fission reactor

    NASA Astrophysics Data System (ADS)

    Mirnov, S.

    2012-06-01

    The Russian fission community places several heavy demands to quality of fusion neutron source for the first step of investigation of minority transmutations ("burning") and breading of nuclear fuel. They are: the steady state regime of neutron production (not rare 80% of main operation time), the total power on neutron flux should be not lower than 20MW with surface neutron load not lower than 0.2MW/m2. Between the current fusion devices: mirror traps, reverse field pinches, stellarators, spherical torus and tokamaks only lasts have today the some probability to fulfill in the near future these hard demands. Two well known DT-tokamaks with neutron power production higher 10MW - TFTR and JET-had maximal neutron load approximately 0.1MW/m2 only in transient (with time scale lower 1s) regimes. The quasi steady state neutron emission regime (˜5MW, 5sec) was performed in JET with mean surface neutron load lower than 0.025MW/m2 only. In this communication it will be discussed the main needs of JET scale tokamak improvement for increase on neutron load up to 0.2MW/m2. They are: decrease of Zeff by ECRH and lithium use as plasma facing components, the increase of energy of steady state neutral injectors up to 150-170keV (tritium), the He removal and creation of closed loop of DT fuel circulation.

  18. 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.

  19. Spherical tokamaks with plasma centre-post

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2013-10-01

    The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

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

    SciTech Connect

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

    1981-03-01

    Early scoping studies based on approximate, analytic models have been 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.

  1. Compact-Toroid Fusion Reactor (CTOR) based on the Field-Reversed Theta Pinch

    SciTech Connect

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

    1981-01-01

    Scoping studies of a translating Compact Torus Reactor (CTOR) have been made on the basis of a dynamic plasma model and an overall systems approach. 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 plamoid 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 radition 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.

  2. The engineering design of the Tokamak Physics Experiment (TPX)

    SciTech Connect

    Reiersen, W.T.

    1994-09-01

    The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. TPX has a long pulse (1000s) capability, can accommodate high divertor heat loads, has a flexible poloidal field (PF) system, and auxiliary heating and current drive systems that make it an ideal test bed for development of attractive reactor concepts. The design incorporates superconducting magnets in both the toroidal field (TF) and poloidal field (PF) systems. Long pulse deuterium operation will produce 6 {times} 10{sup 21} neutrons per year requiring remote maintenance of the in-vessel hardware. This paper provides an overview of the TPX design with the emphasis on developments in the tokamak design since the Conceptual Design Review (CDR) in March, 1993.

  3. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    SciTech Connect

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio.

  4. Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

    SciTech Connect

    R.V. Budny

    2000-11-15

    ELMy H-mode plasmas form the basis of conservative performance predictions for tokalmak reactors of the size of ITER. Relatively high performace for long durations has been achieved and the scaling is favorable. It will be necessary to sustain low Zeff and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one in which the degradation occurs with an intense gas puff, and the other with a spontaneous transition at the heating power threshold from Type I to III ELMs. Linear gryokinetic analysis gives the growth rate, glin of the fastest growing mode. Our results indicate that the flow-shearing rate wExB and glin are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high *wExB*/glin near the top of the pedestal for high confinement.

  5. 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.

  6. Ballistic contributions to heat pulse propagation on TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Fredrickson, E.D.; McGuire, K.; Cavallo, A.; Budny, R.; Janos, A.; Monticello, D.; Nagayama, Y.; Park, W.; Taylor, G.; Zarnstorff, M.C.

    1990-07-01

    Measurements on the TFTR tokamak of the electron temperature profile evolution and soft x-ray emissivity on a fast (10 {mu}sec) time scale during a sawtooth crash, show that significant heat is deposited beyond the mixing (or reconnection) radius within 200 {mu}sec following a sawtooth crash. This extended region in which electron heat is redistributed during the sawtooth crash substantially complicates the determination of heat transport properties from the subsequent heat pulse propagation. It is shown that the relaxation of this extended perturbation is consistent with the power balance estimates of the local thermal diffusivity. 19 refs., 6 figs.

  7. 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.

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

    NASA Astrophysics Data System (ADS)

    Maise, George; Powell, James; Paniagua, John

    2006-01-01

    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 ~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 ~3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  9. TIBER II: Tokamak Ignition/Burn Experimental Reactor: 1986 status report

    SciTech Connect

    Henning, C.D.; Logan, B.G.

    1986-10-23

    Several chapters are presented that cover the following areas: (1) physics basis; (2) current drive; (3) compact divertors; (4) neutron shielding; (5) high-current density, radiation-tolerant magnets; and (6) costs. (MOW)

  10. 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.

  11. The Tokamak Physics Experiment

    SciTech Connect

    Davidson, R.C.; Goldston, R.J.; Neilson, G.H.; Thomassen, K.I.

    1995-06-01

    The mission of the Tokamak Physics Experiment (TPX) [Nevins {ital et} {ital al}., {ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Nuclear} {ital Fusion}, Wuerzburg (International Atomic Energy Agency, Vienna, 1992), Vol. 3, p. 279] is to develop the scientific basis for an economically competitive and continuously operating tokamak fusion power source. This complements the primary mission of the International Thermonuclear Experimental Reactor (ITER) [ITER Document Ser. No. 18 (International Atomic Energy Agency, Vienna, 1991)], the demonstration of ignition and long-pulse burn, and the integration of nuclear technologies. The TPX program is focused on making the demonstration power plant that follows ITER as compact and attractive as possible, and on permitting ITER to achieve its ultimate goal of steady-state operation. This mission of TPX requires the development of steady-state regimes with high beta, good confinement, and a high fraction of a self-driven bootstrap current. These regimes must be compatible with plasma stability, strong heat-flux dispersion in the divertor region, and effective particle control.

  12. The Tokamak Physics Experiment

    NASA Astrophysics Data System (ADS)

    Davidson, Ronald C.; Goldston, Robert J.; Neilson, George H.; Thomassen, Keith I.

    1995-06-01

    The mission of the Tokamak Physics Experiment (TPX) [Nevins et al., Plasma Physics and Controlled Nuclear Fusion, Würzburg (International Atomic Energy Agency, Vienna, 1992), Vol. 3, p. 279] is to develop the scientific basis for an economically competitive and continuously operating tokamak fusion power source. This complements the primary mission of the International Thermonuclear Experimental Reactor (ITER) [ITER Document Ser. No. 18 (International Atomic Energy Agency, Vienna, 1991)], the demonstration of ignition and long-pulse burn, and the integration of nuclear technologies. The TPX program is focused on making the demonstration power plant that follows ITER as compact and attractive as possible, and on permitting ITER to achieve its ultimate goal of steady-state operation. This mission of TPX requires the development of steady-state regimes with high beta, good confinement, and a high fraction of a self-driven bootstrap current. These regimes must be compatible with plasma stability, strong heat-flux dispersion in the divertor region, and effective particle control.

  13. 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.

  14. Atomic data and level populations of highly ionized Ti for tokamak plasmas. [reactor chamber alloy constituent

    NASA Technical Reports Server (NTRS)

    Bhatia, A. K.; Feldman, U.; Doschek, G. A.

    1980-01-01

    The paper presents calculations of electron impact collision strengths and spontaneous radiative decay rates for titanium ions of the LiI through FI isoelectronic sequences for transitions between levels of the 2S(2)2p(k), 2s2p(k+1), and 2p(k+2) configurations. From these atomic data, excitation-rate coefficients are calculated along with level populations for these three configurations. The calculations of level populations include the effects of proton excitation, and are carried out at electron temperatures and densities typical of tokamak plasmas. Wavelengths of forbidden and intersystem lines are given, and a synthetic spectrum is presented for a typical temperature and density.

  15. 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.

  16. Circular limiter H-mode plasmas in the Tokamak Fusion Test Reactor (TFTR)

    SciTech Connect

    Bush, C.E.

    1990-01-01

    Circular limiter H-modes with centrally peaked density profiles have been obtained in TFTR using a highly conditioned graphite limiter. The transition to these centrally peaked H-modes takes place from the supershot to the H-mode rather than the usual L- to H-mode transition observed in other tokamaks. Bidirectional beam heating is required and the threshold power needed to induce the transition increases linearly with plasma current. Density peaking factors, n{sub e}(0)/{l angle}n{sub e}{r angle}, greater than 2.3 are obtained and, at the same time, the H-mode characteristics are similar to those of limiter H-modes on other tokamaks and the global confinement, {tau}{sub E}, can be >2.5 times L-mode scaling. Microwave scattering data from the edge plasma shows broad spectra at k = 5.5 cm{sup {minus}1} which begin at the drop in D{sub {alpha}} radiation and are strongly shifted in the electron diamagnetic drift direction. This implies a poloidal rotation, which begins at the transition to the H-mode, of {approximately}10{sup 4} m/sec. During an edge localized mode instability (ELM), these apparent rotations cease and Mirnov fluctuations in the 50--500 kHz range increase in intensity. Electron cyclotron emission data shows the origin of the ELMs and probably the transition layer to be located a few centimeters inside the plasma surface. A short review of requirements for controlled thermonuclear reactions is given in the introduction. 16 refs., 7 figs.

  17. 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.

  18. 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)

  19. 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.

  20. Tests of local transport theory and reduced wall impurity influx with highly radiative plasmas in the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Hill, K. W.; Scott, S. D.; Bell, M.; Budny, R.; Bush, C. E.; Clark, R. E. H.; Denne-Hinnov, B.; Ernst, D. R.; Hammett, G. W.; Mikkelsen, D. R.; Mueller, D.; Ongena, J.; Park, H. K.; Ramsey, A. T.; Synakowski, E. J.; Taylor, G.; Zarnstorff, M. C.

    1999-03-01

    The electron temperature (Te) profile in neutral beam-heated supershot plasmas (Te0˜6-7 keV ion temperature Ti0˜15-20 keV, beam power Pb˜16 MW) was remarkably invariant when radiative losses were increased significantly through gas puffing of krypton and xenon in the Tokamak Fusion Test Reactor [McGuire et al., Phys. Plasmas 2, 2176 (1995)]. Trace impurity concentrations (nz/ne˜10-3) generated almost flat and centrally peaked radiation profiles, respectively, and increased the radiative losses to 45%-90% of the input power (from the normal ˜25%). Energy confinement was not degraded at radiated power fractions up to 80%. A 20%-30% increase in Ti, in spite of an increase in ion-electron power loss, implies a factor of ˜3 drop in the local ion thermal diffusivity. These experiments form the basis for a nearly ideal test of transport theory, since the change in the beam heating power profile is modest, while the distribution of power flow between (1) radiation and (2) conduction plus convection changes radically and is locally measurable. The decrease in Te was significantly less than predicted by two transport models and may provide important tests of more complete transport models. At input power levels of 30 MW, the increased radiation eliminated the catastrophic carbon influx (carbon "bloom") and performance (energy confinement and neutron production) was improved significantly relative to that of matched shots without impurity gas puffing.

  1. Microinstability properties of negative magnetic shear discharges in the Tokamak Fusion Test Reactor and DIII-D

    SciTech Connect

    Rewoldt, G.; Tang, W.M.; Lao, L.L.

    1997-03-01

    The microinstability properties of discharges with negative (reversed) magnetic shear in the Tokamak Fusion Test Reactor (TFTR) and DIII-D experiments with and without confinement transitions are investigated. A comprehensive kinetic linear eigenmode calculation employing the ballooning representation is employed with experimentally measured profile data, and using the corresponding numerically computed magnetohydrodynamic (MHD) equilibria. The instability considered is the toroidal drift mode (trapped-electron-{eta}{sub i} mode). A variety of physical effects associated with differing q-profiles are explained. In addition, different negative magnetic shear discharges at different times in the discharge for TFTR and DIII-D are analyzed. The effects of sheared toroidal rotation, using data from direct spectroscopic measurements for carbon, are analyzed using comparisons with results from a two-dimensional calculation. Comparisons are also made for nonlinear stabilization associated with shear in E{sub r}/RB{sub {theta}}. The relative importance of changes in different profiles (density, temperature, q, rotation, etc.) on the linear growth rates is considered.

  2. Local transport barrier formation and relaxation in reverse-shear plasmas on the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Synakowski, E. J.; Batha, S. H.; Beer, M. A.; Bell, M. G.; Bell, R. E.; Budny, R. V.; Bush, C. E.; Efthimion, P. C.; Hahm, T. S.; Hammett, G. W.; LeBlanc, B.; Levinton, F.; Mazzucato, E.; Park, H.; Ramsey, A. T.; Schmidt, G.; Rewoldt, G.; Scott, S. D.; Taylor, G.; Zarnstorff, M. C.

    1997-05-01

    The roles of turbulence stabilization by sheared E×B flow and Shafranov shift gradients are examined for Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] enhanced reverse-shear (ERS) plasmas. Both effects in combination provide the basis of a positive-feedback model that predicts reinforced turbulence suppression with increasing pressure gradient. Local fluctuation behavior at the onset of ERS confinement is consistent with this framework. The power required for transitions into the ERS regime are lower when high power neutral beams are applied earlier in the current profile evolution, consistent with the suggestion that both effects play a role. Separation of the roles of E×B and Shafranov shift effects was performed by varying the E×B shear through changes in the toroidal velocity with nearly steady-state pressure profiles. Transport and fluctuation levels increase only when E×B shearing rates are driven below a critical value that is comparable to the fastest linear growth rates of the dominant instabilities. While a turbulence suppression criterion that involves the ratio of shearing to linear growth rates is in accord with many of these results, the existence of hidden dependencies of the criterion is suggested in experiments where the toroidal field was varied. The forward transition into the ERS regime has also been examined in strongly rotating plasmas. The power threshold is higher with unidirectional injection than with balance injection.

  3. 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

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

    SciTech Connect

    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. {bold 50}, 891 (1983)) and they have now been seen in the TFTR tokamak (Phys. Fluids {bold 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.

  5. Soft-computing approach to plasma evolution tracking in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Morabito, Francesco C.

    1997-10-01

    Qualitative information about the structure of a mapping can surely be of help in learning a mapping by a collection of input-output pairs. However, there are conditions in which time and some other constraints make guessing the only plausible means for interpreting data. In this paper, the problem of the plasma boundary reconstruction in 'Tokamak' nuclear fusion rectors is assessed. The problem is formulated as an inverse 'identification' problem and the mapping is derived by a properly generated database of simulated experiments. Real data coming from experiments are also available to validate both numerically generated data and extracted model. The identification problem is solved for two different databases by using neural networks and more conventional models. The introduction of techniques derived from soft computing is shown to improve the performance in various respects. Dynamic identification systems appear to be rather demanding also for such systems, for the need of rapidly interpreting real time data for discharge control. Soft computing approaches may yet yield some low cost ways to take decisions during plasma evolution. The approximate analysis of experimental data could also improve the knowledge on the particular problem allowing an evolution of the knowledge base. Experimental data related to ASDEX-Upgrade machine are presented in this work and preliminary processed. Soft computing techniques also allow to simply get ideas about two other interesting problems in plasma engineering, namely, the fault tolerance and the minimization of the number of sensors.

  6. 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.

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

    SciTech Connect

    Herrmann, Hans 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 α-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 α-particle loss has led to a better understanding of α-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing α-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° lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an α-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized α-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.

  8. 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.

  9. Evaluation of current drive requirements and operating characteristics of a high bootstrap fraction advanced tokamak reactor

    NASA Astrophysics Data System (ADS)

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

    The reactor potential of some advanced physics operating modes proposed for the TPX physics program are examined. A moderate aspect ratio (A = 4.5 as in TPX), 2 GW reactor is analyzed because of its potential for steady-state, noninductive operation with high bootstrap current fraction. Particle, energy, and toroidal current equations are evolved to steady-state conditions using the 1 1/2-D time-dependent WHIST transport code. The solutions are therefore consistent with particle, energy, and current sources and assumed transport models. Fast wave current drive (FWCD) provides the axial seed current. The bootstrap current typically provides 80-90% of the current, while feedback on the lower hybrid current drive (LHCD) power maintains the total current. The sensitivity of the plasma power amplification factor, Q equivalent to P(sub fus)/P(sub aux), to variations in the plasma properties is examined. The auxiliary current drive power, P(sub aux) = P(sub LH) + P(sub FW); bootstrap current fraction; current drive efficiency; and other parameters are evaluated. The plasma is thermodynamically stable for the energy confinement model assumed (a multiple of ITER89P). The FWCD and LHCD sources provide attractive control possibilities, not only for the current profile, but also for the total fusion power since the gain on the incremental auxiliary power is typically 10-30 in these calculations when overall Q approximately equals 30.

  10. Evaluation of current drive requirements and operating characteristics of a high bootstrap fraction advanced tokamak reactor

    SciTech Connect

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

    1994-11-01

    The reactor potential of some advanced physics operating modes proposed for the TPX physics program are examined. A moderate aspect ratio (A = 4.5 as in TPX), 2 GW reactor is analyzed because of its potential for steady-state, non-inductive operation with high bootstrap current fraction. Particle, energy and toroidal current equations are evolved to steady-state conditions using the 1-1/2-D time-dependent WHIST transport code. The solutions are therefore consistent with particle, energy and current sources and assumed transport models. Fast wave current drive (FWCD) provides the axial seed current. The bootstrap current typically provides 80-90% of the current, while feedback on the lower hybrid current drive (LHCD) power maintains the total current. The sensitivity of the plasma power amplification factor, Q {triple_bond} P{sub fus}/P{sub aux}, to variations in the plasma properties is examined. The auxiliary current drive power, P{sub aux} = P{sub LH} + P{sub FW}; bootstrap current fraction; current drive efficiency; and other parameters are evaluated. The plasma is thermodynamically stable for the energy confinement model assumed (a multiple of ITER89P). The FWCD and LHCD sources provide attractive control possibilities, not only for the current profile, but also for the total fusion power since the gain on the incremental auxiliary power is typically 10-30 in these calculations when overall Q {approx} 30.

  11. Evaluation of current drive requirements and operating characteristics of a high bootstrap fraction advanced tokamak reactor

    SciTech Connect

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

    1995-02-01

    The reactor potential of some advanced physics operating modes proposed for the TPX physics program are examined. A moderate aspect ratio (A = 4.5 as in TPX), 2 GW reactor is analyzed because of its potential for steady-state, non-inductive operation with high bootstrap current fraction. Particle, energy and toroidal current equations are evolved to steady-state conditions using the 1-1/2-D time-dependent WHIST transport code. The solutions are therefore consistent with particle, energy and current sources and assumed transport models. Fast wave current drive (FWCD) provides the axial seed current. The bootstrap current typically provides 80-90% of the current, while feedback on the lower hybrid current drive (LHCD) power maintains the total current. The sensitivity of the plasma power amplification factor, Q {equivalent_to} P{sub fus}/P{sub aux}, to variations in the plasma properties is examined. The auxiliary current drive power, P{sub aux} = P{sub LH} + P{sub FW}; bootstrap current fraction: current drive efficiency; and other parameters are evaluated. The plasma is thermodynamically stable for the energy confinement model assumed (a multiple of ITER89P). The FWCD and LHCD sources provide attractive control possibilities, not only for the current profile, but also for the total fusion power since the gain on the incremental auxiliary power is typically 10-30 in these calculations when overall Q {approx} 30.

  12. Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

    SciTech Connect

    Remy, L.; Cheymol, G.; Morana, A.; Marin, E.; Girard, S.

    2015-07-01

    In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensor design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison of

  13. Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

    SciTech Connect

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

    1985-01-01

    Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.; ...

    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

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. Systems code assessment of innovations, major design drivers, and minimum sizes of INTOR (International Tokamak Reactor) and ETR-like designs

    SciTech Connect

    Galambos, J.D.; Peng, Y.K.M.; Strickler, D.J.; Reid, R.L.

    1987-10-01

    System studies of next-generation superconducting tokamaks are presented here. These studies include examining design changes suggested for the International Tokamak Reactor (INTOR) as a means of reducing the size and simplifying the device and assessing the impact of a series of more aggressive design assumptions suggested in recent Engineering Test Reactor (ETR) studies. Also, a set of candidate design points offering small machines (major radius = 4 m) with a relatively conservative mix of design assumptions is proposed. Some of the design assumptions found to have a major effect on the minimum size are TF coil current density, noninductive current drive, plasma elongation and edge q, plasma temperature for current drive, maximum allowable plasma beta, the minimum required wall load, and assumptions on fixed radial dimensions such as shield thickness, gaps, and plasma scrapeoff. Some design assumptions with less impact on the device size are the OH coil current density, PF configuration (limiter/divertor), and plasma current level. 22 refs., 5 figs., 6 tabs.

  4. 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.

  5. Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

    NASA Astrophysics Data System (ADS)

    Lell, R. M.; Hanan, N. A.

    Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors were examined. Homogeneous and space dependent multigroup cross section set were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design.

  6. Quantitative Homogeneity and In-Contact Particles of High Temperature Reactors (htr) Compacts Determination via X-Ray Tomography

    NASA Astrophysics Data System (ADS)

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

    2008-02-01

    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.

  7. 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.

  8. Advanced Catalysis Technologies: Lanthanum Cerium Manganese Hexaaluminate Combustion Catalysts for Flat Plate Reactor for Compact Steam Reformers

    DTIC Science & Technology

    2008-12-01

    REACTOR FOR COMPACT STEAM REFORMERS Aly H. Shaaban Applied Research Associates P.O. Box 40128 Tyndall Air Force Base, FL 32403...DRIVE, SUITE 2 TYNDALL AIR FORCE BASE, FL 32403-5323 NOTICE AND SIGNATURE PAGE Using Government drawings, specifications, or other data...been reported in the literature to give high light fractions yield with low selectivity to coke [24]. ZSM-5 type (MFI) zeolite (Zeolyst CBV5524G, SiO2

  9. High Power LaB6 Plasma Source Performance for the Lockheed Martin Compact Fusion Reactor Experiment

    NASA Astrophysics Data System (ADS)

    Heinrich, Jonathon

    2016-10-01

    Lockheed Martin's Compact Fusion Reactor (CFR) concept is a linear encapsulated ring cusp. Due to the complex field geometry, plasma injection into the device requires careful consideration. A high power thermionic plasma source (>0.25MW; >10A/cm2) has been developed with consideration to phase space for optimal coupling. We present the performance of the plasma source, comparison with alternative plasma sources, and plasma coupling with the CFR field configuration. ©2016 Lockheed Martin Corporation. All Rights Reserved.

  10. Status of the tokamak program

    SciTech Connect

    Sheffield, J.

    1981-06-01

    During the past two decades, the tokamak program has made significant advances. As a result of these advances, the program has expanded, and construction is nearly completed of large facilities, including the US Tokamak Fusion Test Reactor (TFTR), that are capable of achieving energy break-even in the plasma. Design work is under way on the next application phase, exemplified by the US Fusion Engineering Device (FED) and the IAEA INTOR programs. There are many excellent reviews that give a broad history of the evolution of the tokamak program. In addition, reviews are available that give detailed descriptions of individual tokamak elements. This paper reviews the status of the tokamak program, concentrating on recent advances and areas important to the further advancement of the tokamak concept.

  11. Modular tokamak configuration

    SciTech Connect

    Thomson, S.L.

    1985-01-01

    This report is concerned with the modular tokamak configuration, and presents information on the following topics: modularity; external vacuum boundary; vertical maintenance; combined reactor building/biological shield with totally remote maintenance; independent TF coils; minimum TF coil bore; saddle PF coils; and heat transport system in bore.

  12. Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

    SciTech Connect

    Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

    1985-12-01

    A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions.

  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. Tokamak Physics Experiment diagnostic plans (invited)

    NASA Astrophysics Data System (ADS)

    Medley, S. S.

    1995-01-01

    A superconducting Tokamak Physics Experiment (TPX) whose mission is to develop the scientific basis for a compact and continuously operating tokamak fusion reactor is being designed by an integrated U.S. national team. Key physics features such as strong shaping, a double-null poloidal divertor, full noninductive current drive, and current profile control capability will be used to explore improvements in energy confinement and beta limit scaling in high-aspect-ratio plasmas with a high bootstrap current fraction. Steady-state operation of TPX permits these studies to be extended to time scales significantly exceeding the global current-relaxation time and the plasma-wall equilibrium time. The diagnostic requirements are determined by the TPX mission and supporting objectives, such as optimization of plasma performance through active control of the current profile and of the plasma-wall interactions. Diagnostic measurements are needed to characterize the plasma behavior over the full range of conventional tokamak plasma parameters with appropriate spatial and temporal resolution as well as for control and monitoring of aspects of the machine operation such as the plasma position and shape, plasma current, vacuum vessel currents, electron density and temperature, and the divertor and limiter temperatures. In addition, several diagnostic capabilities that are especially critical for the TPX project will be discussed.

  15. Mission and physics design of the Tokamak Physics Experiment

    SciTech Connect

    Neilson, G.H.; Batchelor, D.B.; Mioduszewski, P.K.; Strickler, D.J.; Bonoli, P.T.; Porkolab, M.; Goldston, R.J.; Jardin, S.C.; Bialek, J.M.; Kessel, C.E.

    1994-11-01

    Improvements in the confinement, stability limits, current-drive efficiency and divertor performance, combined with steady-state operation, can lead to a more economical tokamak fusion reactor than one based on the present physics data base. The Tokamak Physics Experiment (TPX) is planned to extend the recent advances in these areas, achieved in pulsed tokamaks, to the steady-state regime. In so doing, it will develop a data base needed for the design of an economically attractive tokamak reactor.

  16. 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.

  17. 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.

  18. 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.

  19. An analysis of activation and the impact of tritium breeding media and structural materials for a commercial tokamak fusion reactor design

    SciTech Connect

    Jung, J.

    1983-11-01

    Activation analysis has been conducted for several primary fusion blanket materials based on a model of a commercial tokamak fusion reactor design, STARFIRE. The blanket materials studied include two solid tritium breeders, viz., Li/sub 2/O and ..cap alpha..-LiAlO/sub 2/, and four candidate structural materials, viz., PCA stainless steel, V15Cr5Ti, Ti6Al4V, and Al-6063 alloys. The importance of breeder material activation is identified in terms of its impurity contents such as potassium, iron, nickel, molybdenum, and zirconium trace elements. The breeder activation is also discussed with regard to its potential for recycling and its impact on the lithium resource requirements. The structural material activation is analyzed based on two measures, volumetric radioactivity concentration and contact biological dose due to decay gamma emission. Using the radioactivity concentration measure, it is revealed that a substantial advantage exists from a viewpoint of radwaste management, which is inherent in fusion reactor designs based on potential low-activation alloys such as V15Cr5Ti, Ti6Al4V, and Al-6063. On the other hand, from the dose standpoint, the V15Cr5Ti alloy is found to be the only alloy for which one could realize a significant dose reduction (below 2.5 mrem/h) within about100 yr after shutdown, possibly by some extrapolation on alloy purification techniques.

  20. Three-dimensional analysis of tokamaks and stellarators.

    PubMed

    Garabedian, Paul R

    2008-09-16

    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.

  1. 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

  2. Three-dimensional analysis of tokamaks and stellarators

    NASA Astrophysics Data System (ADS)

    Garabedian, Paul R.

    2008-09-01

    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.

  3. Stabilized Spheromak Fusion Reactors

    SciTech Connect

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  4. 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)

  5. 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.

  6. 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

  7. The Aneutronic Rodless Ultra Low Aspect Ratio Tokamak

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2016-10-01

    The replacement of the metal centre-post in spherical tokamaks (STs) by a plasma centre-post (PCP, the TF current carrier) is the ideal scenario for a ST reactor. A simple rodless ultra low aspect-ratio tokamak (RULART) using a screw-pinch PCP ECR-assisted with an external solenoid has been proposed in the most compact RULART [Ribeiro C, SOFE-15]. There the solenoid provided the stabilizing field for the PCP and the toroidal electrical field for the tokamak start-up, which will stabilize further the PCP, acting as stabilizing closed conducting surface. Relative low TF will be required. The compactness (high ratio of plasma-spherical vessel volume) may provide passive stabilization and easier access to L-H mode transition. It is presented here: 1) stability analysis of the PCP (initially MHD stable due to the hollow J profile); 2) tokamak equilibrium simulations, and 3) potential use for aneutronic reactions studies via pairs of proton p and boron 11B ion beams in He plasmas. The beams' line-of-sights sufficiently miss the sources of each other, thus allowing a near maximum relative velocities and reactivity. The reactions should occur close to the PCP mid-plane. Some born alphas should cross the PCP and be dragged by the ion flow (higher momentum exchange) towards the anode but escape directly to a direct electricity converter. Others will reach evenly the vessel directly or via thermal diffusion (favourable heating by the large excursion 2a), leading to the lowest power wall load possible. This might be a potential hybrid direct-steam cycle conversion reactor scheme, nearly aneutronic, and with no ash or particle retention problems, as opposed to the D-T thermal reaction proposals.

  8. 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.

  9. 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.

  10. Design and operation of the pellet charge exchange diagnostic for measurement of energetic confined α particles and tritons on the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Medley, S. S.; Mansfield, D. K.; Roquemore, A. L.; Fisher, R. K.; Duong, H. H.; McChesney, J. M.; Parks, P. B.; Petrov, M. P.; Khudoleev, A. V.; Gorelenkov, N. N.

    1996-09-01

    Radially resolved energy and density distributions of the confined α particles in D-T experiments on the Tokamak Fusion Test Reactor (TFTR) are being measured with the pellet charge exchange (PCX) diagnostic. Other energetic ion species can be detected as well, such as tritons produced in D-D plasmas and H, He3, or tritium rf-driven minority ion tails. The ablation cloud formed by injected low-Z impurity pellets provides the neutralization target for this active charge exchange technique. Because the cloud neutralization efficiency is uncertain, the PCX diagnostic is not absolutely calibrated so only relative density profiles are obtained. A mass and energy resolving E∥B neutral particle analyzer (NPA) is used which has eight energy channels covering the energy range of 0.3-3.7 MeV for α particles with energy resolution ranging from 5.8% to 11.3% and a spatial resolution of ˜5 cm. The PCX diagnostic views deeply trapped ions in a narrow pitch angle range around a mean value of v∥/v=-0.048±10-3. For D-T operation, the NPA was shielded by a polyethylene-lead enclosure providing 100× attenuation of ambient γ radiation and 14 MeV neutrons. The PCX diagnostic technique and its application on TFTR are described in detail.

  11. 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

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

    SciTech Connect

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

    2016-09-15

    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

  13. Pressing device for producing compacts from source material in powder form in particular pulverized nuclear reactor fuel

    SciTech Connect

    Heller, G.; Adelmann, M.; Konigs, W.; Wendorf, W.

    1984-04-17

    Pressing device for producing compacts from source material in powder form, in particular pulverized nuclear reactor fuel having a die-plate contained in platen and a bore associated with a ram, for receiving source material powder, a filling shoe, and a reservoir for powder connected by a hose to the filling shoe. The device is characterized by a passing wheel in the filling shoe as filling aid means; a tube containing a feedscrew disposed between the reservoir and hose as metering means; the reservoir having a bottom part with a can type place-on part with an opening eccentric to the axis; a coupling part and a cover part are placed on the open part of the can, these parts are also provided with a passageway to the feedscrew eccentric to the longitudinal axis.

  14. New Concepts for Compact Space Reactor Power Systems for Space Based Radar Applications: A Feasibility Study

    DTIC Science & Technology

    2007-11-02

    thermoelectric, thermionic, Rankine/Brayton/ Stirling engines and Alkali Metal Thermoelectric Convenors (AMTEC). Reactor types investigated include...Date: Apr 21, 1993 Document Type : HC Number of Copies In Library: 000001 Record ID: 26713 ram BDM INTERNATIONAL, INC. 7915 JONES BRANCH DRIVE...internationally recognized expert on heat engines . The help and contributions of the following individuals are also greatly appreciated: Dr. Rick Moats and

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

    SciTech Connect

    Singh, M. J.; De Esch, H. P. L.

    2010-01-15

    This paper describes the physics design of a 100 keV, 60 A H{sup -} 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.

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

    NASA Astrophysics Data System (ADS)

    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.

  17. 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.

  18. Construction and operation of parallel electric and magnetic field spectrometers for mass/energy resolved multi-ion charge exchange diagnostics on the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Medley, S. S.; Roquemore, A. L.

    1998-07-01

    A novel charge exchange spectrometer using a dee-shaped region of parallel electric and magnetic fields was developed at the Princeton Plasma Physics Laboratory for neutral particle diagnostics on the Tokamak Fusion Test Reactor (TFTR). The E∥B spectrometer has an energy range of 0.5⩽A (amu)E (keV)⩽600 and provides mass-resolved energy spectra of H+, D+, and T+ (or 3He+) ion species simultaneously during a single discharge. The detector plane exhibits parallel rows of analyzed ions, each row containing the energy dispersed ions of a given mass-to-charge ratio. The detector consists of a large area microchannel plate (MCP) which is provided with three rectangular, semicontinuous active area strips, one coinciding with each of the mass rows for detection of H+, D+, and T+ (or 3He+) and each mass row has 75 energy channels. To suppress spurious signals attending operation of the plate in the magnetic fringe field of the spectrometer, the MCP was housed in a double-walled iron shield with a wire mesh ion entrance window. Using an accelerator neutron generator, the MCP neutron detection efficiency was measured to be 1.7×10-3 and 6.4×10-3 counts/neutron/cm2 for 2.5 MeV-DD and 14 MeV-DT neutrons, respectively. The design and calibration of the spectrometer are described in detail, including the effect of MCP exposure to tritium, and results obtained during high performance D-D operation on TFTR are presented to illustrate the performance of the E∥B spectrometer. The spectrometers were not used during D-T plasma operation due to the cost of providing the required radiation shielding.

  19. 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.

  20. The design of the Tokamak Physics Experiment (TPX)

    NASA Astrophysics Data System (ADS)

    Schmidt, J. A.; Thomassen, K. I.; Goldston, R. J.; Neilson, G. H.; Nevins, W. M.; Sinnis, J. C.; Andersen, P.; Bair, W.; Barr, W. L.; Batchelor, D. B.; Baxi, C.; Berg, G.; Bernabei, S.; Bialek, J. M.; Bonoli, P. T.; Boozer, A.; Bowers, D.; Bronner, G.; Brooks, J. N.; Brown, T. G.; Bulmer, R.; Butner, D.; Campbell, R.; Casper, T.; Chaniotakis, E.; Chaplin, M.; Chen, S. J.; Chin, E.; Chrzanowski, J.; Citrolo, J.; Cole, M. J.; Dahlgren, F.; Davis, F. C.; Davis, J.; Davis, S.; Diatchenko, N.; Dinkevich, S.; Feldshteyn, Y.; Felker, B.; Feng, T.; Fenstermacher, M. E.; Fleming, R.; Fogarty, P. J.; Fragetta, W.; Fredd, E.; Gabler, M.; Galambos, J.; Gohar, Y.; Goranson, P. L.; Greenough, N.; Grisham, L. R.; Haines, J.; Haney, S.; Hassenzahl, W.; Heim, J.; Heitzenroeder, P. J.; Hill, D. N.; Hodapp, T.; Houlberg, W. A.; Hubbard, A.; Hyatt, A.; Jackson, M.; Jaeger, E. F.; Jardin, S. C.; Johnson, J.; Jones, G. H.; Juliano, D. R.; Junge, R.; Kalish, M.; Kessel, C. E.; Knutson, D.; LaHaye, R. J.; Lang, D. D.; Langley, R. A.; Liew, S.-L.; Lu, E.; Mantz, H.; Manickam, J.; Mau, T. K.; Medley, S.; Mikkelsen, D. R.; Miller, R.; Monticello, D.; Morgan, D.; Moroz, P.; Motloch, C.; Mueller, J.; Myatt, L.; Nelson, B. E.; Neumeyer, C. L.; Nilson, D.; O'Conner, T.; Pearlstein, L. D.; Peebles, W. A.; Pelovitz, M.; Perkins, F. W.; Perkins, L. J.; Petersen, D.; Pillsbury, R.; Politzer, P. A.; Pomphrey, N.; Porkolab, M.; Posey, A.; Radovinsky, A.; Raftopoulis, S.; Ramakrishnan, S.; Ramos, J.; Rauch, W.; Ravenscroft, D.; Redler, K.; Reiersen, W. T.; Reiman, A.; Reis, E.; Rewoldt, G.; Richards, D. J.; Rocco, R.; Rognlien, T. D.; Ruzic, D.; Sabbagh, S.; Sapp, J.; Sayer, R. O.; Scharer, J. E.; Schmitz, L.; Schnitz, J.; Sevier, L.; Shipley, S. E.; Simmons, R. T.; Slack, D.; Smith, G. R.; Stambaugh, R.; Steill, G.; Stevenson, T.; Stoenescu, S.; Onge, K. T. St.; Stotler, D. P.; Strait, T.; Strickler, D. J.; Swain, D. W.; Tang, W.; Tuszewski, M.; Ulrickson, M. A.; VonHalle, A.; Walker, M. S.; Wang, C.; Wang, P.; Warren, J.; Werley, K. A.; West, W. P.; Williams, F.; Wong, R.; Wright, K.; Wurden, G. A.; Yugo, J. J.; Zakharov, L.; Zbasnik, J.

    1993-09-01

    The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.

  1. The design of the Tokamak Physics Experiment (TPX)

    SciTech Connect

    Schmidt, J.A.; Goldston, R.J.; Sinnis, J.C.; Bernabei, S.; Bialek, J.M.; Bronner, G.; Chen, S.J.; Chrzanowski, J.; Citrolo, J.; Dahlgren, F.

    1993-09-01

    The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a low-activation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation of US industry. Operation will begin with first plasma in the year 2000.

  2. 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.; ...

    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

  3. 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

    SciTech Connect

    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 and 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.

  4. A compact CO selective oxidation reactor for solid polymer fuel cell powered vehicle application

    NASA Astrophysics Data System (ADS)

    Dudfield, C. D.; Chen, R.; Adcock, P. L.

    Solid polymer fuel cells (SPFCs) are attractive as electrical power plants for vehicle applications since they offer the advantages of high efficiency, zero emissions, and mechanical robustness. Hydrogen is the ideal fuel, but is currently disadvantaged for automotive applications by the lack of refuelling infrastructure, bulky on-board storage, and safety concerns. On-board methanol reforming offers an attractive alternative due to its increased energy storage density. Since CO is always present as a by-product during the reforming reaction, it must be reduced to a level less than 20 ppm in order to avoid rapid deactivation of the platinum electro-catalyst in the fuel cells. In this paper, a compact CO selective oxidation unit based upon two coated aluminium heat exchangers, developed at Loughborough University, is reported. The geometric size of the whole unit is 4 litre and experimental results show that the selective oxidation unit can reduce the CO from up to 2% to less than 15 ppm and is suitable for a vehicle fuel cell power plant of 20 kW e.

  5. Hexagonal arrayed patterns with bright and dark spots observed in a compact plasma reactor based on a piezoelectric transformer

    NASA Astrophysics Data System (ADS)

    Itoh, H.; Suzuki, S.

    2014-10-01

    In this paper, we describe the formation of hexagonal arrayed patterns in a compact plasma reactor that consists of a piezoelectric transformer (PT) and a dielectric electrode. A high voltage induced on the PT surface by the piezoelectric effect generated diffuse glow-like and patterned dielectric barrier discharges (DBDs) in a gap. Vacuum ultraviolet (VUV) spectroscopy was applied to observe light emitted from Xe2\\ast and Ar2\\ast excimers, which were formed in a He-Xe mixture and pure Ar, respectively. Peaks at 172 and 126 nm were observed in the emission spectra. Hexagonal arrayed patterns were observed in both He and Ar by adding a small amount of air. Then, static and time-resolved images of the hexagonal arrayed patterns were captured using a single-lens reflex camera and an intensified charge-coupled device camera, respectively. The hexagonal arrayed patterns alternately exhibited bright and dark spots in weak uniform emission on the PT surface depending on the applied sinusoidal voltage. Both patterns were similar in the observed time-resolved images. Thus, it was confirmed that the observed dark spots were the traces of accumulated charges on the PT. Here, we discuss the necessary conditions for the generation of such patterns in terms of the charges accumulated on the PT and compare our results with those reported elsewhere.

  6. Simultaneous ammonium-nitrogen and copper removal, and copper recovery using nitrifying biofilm from the ultra-compact biofilm reactor.

    PubMed

    Lee, L Y; Ong, S L; Ng, H Y; Hu, J Y; Koh, Y N

    2008-09-01

    Simultaneous ammonium-nitrogen (NH(4)(+)-N) and copper removal, and copper recovery in synthetic wastewater using nitrifying biofilm from an ultra-compact biofilm reactor (UCBR) was demonstrated in batch studies, which consisted of three phases: Phase 1 for NH(4)(+)-N and copper removals, Phase 2 for copper recovery, and Phase 3 for NH(4)(+)-N removal. The results showed that more than 96.3% of copper was removed within 60min, while 60.1% of the adsorbed copper was recovered through rinsing the biofilms with 0.1mM of ethylenediaminetetraacetic acid (EDTA). The nitrifying biofilm was able to adsorb 0.245mg of copper/g of biofilms. After recovery treatment, 29.4% of copper remained bound within the nitrifying biofilms. No significant inhibitory effects towards NH(4)(+)-N removal in the presence of 0.92mg copper/L was noted in Phase 1 compared with the control test. However, lower initial pH condition in the recovery process and the accumulation of copper on the biofilm led to 50% inhibition on NH(4)(+)-N removal efficiency in the subsequent phase.

  7. 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.

  8. Tokamak burn control

    SciTech Connect

    Sager, G.T.

    1988-06-01

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs.

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

    SciTech Connect

    Kenneth M. Young

    2010-02-22

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

  10. Linear optimal control of tokamak fusion devices

    SciTech Connect

    Kessel, C.E.; Firestone, M.A.; Conn, R.W.

    1989-05-01

    The control of plasma position, shape and current in a tokamak fusion reactor is examined using linear optimal control. These advanced tokamaks are characterized by non up-down symmetric coils and structure, thick structure surrounding the plasma, eddy currents, shaped plasmas, superconducting coils, vertically unstable plasmas, and hybrid function coils providing ohmic heating, vertical field, radial field, and shaping field. Models of the electromagnetic environment in a tokamak are derived and used to construct control gains that are tested in nonlinear simulations with initial perturbations. The issues of applying linear optimal control to advanced tokamaks are addressed, including complex equilibrium control, choice of cost functional weights, the coil voltage limit, discrete control, and order reduction. Results indicate that the linear optimal control is a feasible technique for controlling advanced tokamaks where the more common classical control will be severely strained or will not work. 28 refs., 13 figs.

  11. A Fusion Breeder Reactor Based on a Catalyzed D-D Spherical Torus.

    DTIC Science & Technology

    1986-08-08

    cooling and fissile breeding. The need for tritium breeding is eliminated by the use of’ a catalyzed 0-0 fuel cycle. Analysis of this novel reactor...in heavy water which flows through the first wall and blanket providing both cooling and fissile breeding. The need I for tritium breeding is...studies in that: (1) a deuterium fuel cycle is used to eliminate the : need to breed tritium ; (2) a compact tokamak (spherical torus) is used as a

  12. Superconducting magnet system for the TPX Tokamak

    NASA Astrophysics Data System (ADS)

    Hassenzahl, W. V.; Chaplin, M. R.; Heim, J. R.; Lang, D. D.; O'Connor, T. G.; Slack, D. S.; Wong, R. L.; Zbasnik, J. P.; Brown, T. G.; Citrolo, J. C.

    1994-07-01

    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.

  13. 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.

  14. Superconducting magnet system for the TPX Tokamak

    SciTech Connect

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

    1994-07-01

    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 TPC 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.

  15. Tokamak Physics Experiment (TPX) design

    SciTech Connect

    Schmidt, J.A.

    1995-12-31

    TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995.

  16. Tokamak and RFP ignition requirements

    SciTech Connect

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (n{tau}{sub E}) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f{sub RAD} {approximately} 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the n{tau}{sub E} transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab.

  17. 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)

  18. Advanced tokamak operating modes in TPX and ITER

    SciTech Connect

    Nevins, W.M.

    1994-12-31

    A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER.

  19. Advanced tokamak operating modes in TPX and ITER

    NASA Astrophysics Data System (ADS)

    Nevins, W. M.

    1994-09-01

    A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER.

  20. 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.

  1. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  2. 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.

  3. 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.

  4. 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.

  5. Burn Control Mechanisms in Tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, M. A.; Stacey, W. M.

    2015-11-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Overview of spherical tokamak research in Japan

    NASA Astrophysics Data System (ADS)

    Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.

    2017-10-01

    Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

  11. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Aug 15, 2002 to Nov. 15, 2002) - DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Lawrence Townsend; Martin Williamson; Rupy Sawhney; Jacob Fife

    2002-12-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. This report covers the ninth quarter of the project. The three reactor concept teams have completed initial plant concept development, evaluation and layout. A significant design effort has proceeded with substantial change and evolution from original ideas. The concepts have been reviewed by the industry participants and improvements have been implemented. The third phase, industrial engineering simulation of reactor fabrication has begun.

  12. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

    2002-03-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Compact stellarators with modular coils.

    PubMed

    Garabedian, P R

    2000-07-18

    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.

  20. 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

  1. 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.

  2. Compact light-emitting diode optical fiber immobilized TiO2 reactor for photocatalytic water treatment.

    PubMed

    O'Neal Tugaoen, Heather; Garcia-Segura, Sergi; Hristovski, Kiril; Westerhoff, Paul

    2017-09-27

    A key barrier to implementing photocatalysis is delivering light to photocatalysts that are in contact with aqueous pollutants. Slurry photocatalyst systems suffer from poor light penetration and require post-treatment to separate the catalyst. The alternative is to deposit photocatalysts on fixed films and deliver light onto the surface or the backside of the attached catalysts. In this study, TiO2-coated quartz optical fibers were coupled to light emitting diodes (OF/LED) to improve in situ light delivery. Design factors and mechanisms studied for OF/LEDs in a flow-through reactor included: (i) the influence of number of LED sources coupled to fibers and (ii) the use of multiple optical fibers bundled to a single LED. The light delivery mechanism from the optical fibers into the TiO2 coatings is thoroughly discussed. To demonstrate influence of design variables, experiments were conducted in the reactor using the chlorinated pollutant para-chlorobenzoic acid (pCBA). From the degradation kinetics of pCBA, the quantum efficiencies (Φ) of oxidation and electrical energies per order (EEO) were determined. The use of TiO2 coated optical fiber bundles reduced the energy requirements to deliver photons and increased available surface area, which improved Φ and enhanced oxidative pollutant removal performance (EEO). Copyright © 2017 Elsevier B.V. All rights reserved.

  3. Acceleration of compact toroid plasma rings for fusion applications

    SciTech Connect

    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.

    1988-08-26

    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 to30% 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. 14 refs., 4 figs.

  4. Nonlinear stabilization of tokamak microturbulence by fast ions.

    PubMed

    Citrin, J; Jenko, F; Mantica, P; Told, D; Bourdelle, C; Garcia, J; Haverkort, J W; Hogeweij, G M D; Johnson, T; Pueschel, M J

    2013-10-11

    Nonlinear electromagnetic stabilization by suprathermal pressure gradients found in specific regimes is shown to be a key factor in reducing tokamak microturbulence, augmenting significantly the thermal pressure electromagnetic stabilization. Based on nonlinear gyrokinetic simulations investigating a set of ion heat transport experiments on the JET tokamak, described by Mantica et al. [Phys. Rev. Lett. 107, 135004 (2011)], this result explains the experimentally observed ion heat flux and stiffness reduction. These findings are expected to improve the extrapolation of advanced tokamak scenarios to reactor relevant regimes.

  5. 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.

  6. Prospects for toroidal fusion reactors

    NASA Astrophysics Data System (ADS)

    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.

  7. Hydrogen Retention and Helium Pumping by Flowing Liquid Lithium: A Path to a Compact (and therefore less costly) Magnetic Fusion Power Reactor

    NASA Astrophysics Data System (ADS)

    Ruzic, David

    2004-11-01

    The cost of a fusion power plant scales roughly as the volume of the torus. Scaling to a larger size typically allows almost any magnetic confinement concept to reach ignition, but also leads to a higher cost. The need for a larger size plasma results from the need to support a high core temperature while still preserving an edge temperature consistent with the requirements of the confinement scheme and the limitations of the wall material. Recent work has shown that a wall that has near-zero recycling will have very high edge plasma temperatures and can therefore support a high core temperature in a smaller volume.[1] The perceived difficulty with such concepts is that the only known low-Z, low-recycling surface is one composed of lithium, and lithium melts at 180C, has a very high vapor pressure at elevated temperatures, and is highly corrosive. The very fact that it is a low-recycling surface could be its downfall: what if it attracts so much tritium as to make the inventories untenable? In addition, a wall completely covered with lithium and a divertor protected by flowing lithium leaves no exit for the helium in a reactor concept. This talk will show results from the Flowing Illinois Lithium Retention Experiment[2] (FLIRE) which show that (1) the lithium adsorption of D (and by extension, T) is manageable, (2) the helium retention coefficient in flowing lithium may be high enough for He to be pumped, and (3) the corrosiveness of molten lithium is not beyond the technological capabilities of the fusion program. [1] L.E. Zakharov, N.N. Gorelenkov, R.B. White, S.I. Krasheninnikov, G.V. Pereverzev, Ignited spherical tokamaks and plasma regimes with Li walls, Fusion Engineering and Design, Sept. 2004. [2] J.P. Allain, M. Nieto, M.D. Coventry, R. Stubbers, D.N. Ruzic, "Studies of liquid-metal erosion and free surface flowing liquid-lithium retention of helium at the University of Illinois", Fusion Engineering and Design, Sept. 2004.

  8. Compact secondary treatment train combining a lab-scale moving bed biofilm reactor and enhanced flotation processes.

    PubMed

    Brosseau, Catherine; Émile, Bettina; Labelle, Marc-André; Laflamme, Édith; Dold, Peter L; Comeau, Yves

    2016-12-01

    High-rate wastewater processes are receiving a renewed interest to obtain energy positive/efficient water resource recovery facilities. An innovative treatment train combining a high-rate moving bed biofilm reactor (HR-MBBR) with an enhanced flotation process was studied. The two objectives of this work were 1) to maximize the conversion of soluble organics to particulate matter in an HR-MBBR and 2) to maximize the particulate matter recovery from the HR-MBBR effluent by green chemicals to enhance biogas production by anaerobic digestion. To achieve these objectives, lab-scale MBBRs fed with synthetic soluble wastewater were operated at organic loading rates (OLRs) between 4 and 34 kg COD m(-3) reactor d(-1) corresponding to hydraulic retention times (HRTs) between 6 and 54 min. Colloidal and soluble chemical oxygen demand (COD) removal efficiency in the HR-MBBR increased with HRT to reach a plateau of 85% at an HRT longer than 27 min. Carrier clogging observed at an OLR higher than 16 kg COD m(-3) d(-1) (HRT < 13 min) resulted in about 23% loss in colloidal and soluble COD removal efficiency. Thus, the recommended parameters were between 22 and 37 min and between 6 and 10 kg COD m(-3) d(-1) for the HRT and the OLR, respectively, to maximize the conversion of soluble organics to particulate matter. Total suspended solids (TSS) recovery of 58-85% and 90-97% were achieved by enhanced flotation using green and unbiodegradable chemicals, respectively, corresponding to a TSS effluent concentration below 14 and 7 mg TSS/L. Among the synthetic polymers tested, a high molecular weight and low charge density cationic polyacrylamide was found to give the best results with less than 2 mg TSS/L in the clarified effluent (97% TSS recovery). Green chemicals, although performing slightly less for solids separation than unbiodegradable chemicals, achieved a mean TSS concentration of 10 ± 3 mg/L in the clarified effluent.

  9. 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.

  10. 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.

  11. Millimeter wave tokamak heating and current drive with a high power free electron laser

    SciTech Connect

    Thomassen, K.I.

    1987-01-01

    Experiments on microwave generation using a free electron laser (FEL) have shown this to be an efficient way to generate millimeter wave power in short, intense pulses. Short pulse FEL's have several advantages that make them attractive for application to ECR heating of tokamak fusion reactors. This paper reports on plans made to demonstrate the technology at the Microwave Tokamak Experiment (MTX) Facility.

  12. Operating points and feedback control of plasma characteristics in tokamaks with full current drive

    SciTech Connect

    Swain, D.W.; Attenberger, S.E.; Houlberg, W.A.; Bonoli, P.T.; Nevins, W.M.

    1994-12-31

    Tokamak operation using advanced tokamak modes with steady-state current drive may provide better performance, thereby improving the prospects of power-producing fusion reactors. Scenarios for such operation in TPX and in an ITER-like scale-up of TPX are modeled. Results indicate that the heating and current drive power and feedback control requirements may be achievable.

  13. Neutral-beam current drive in tokamaks

    SciTech Connect

    Devoto, R.S.

    1986-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500 to 700 keV are needed for this device.

  14. 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.

  15. Energetic Particle Transport in Compact Quasi-axisymmetric Stellarators

    SciTech Connect

    Redi, M.H.; Mynick, H.E.; Suewattana, M.; White, R.B.; Zarnstorff, M.C.; Isaev, M.Yu.; Mikhailov, M.I.; Subbotin, A.A.

    1999-03-01

    Hamiltonian coordinate, guiding-center code calculations of the confinement of suprathermal ions in quasi-axisymmetric stellarator (QAS) designs have been carried out to evaluate the attractiveness of compact configurations which are optimized for ballooning stability. A new stellarator particle-following code is used to predict ion loss rates and particle confinement for thermal and neutral beam ions in a small experiment with R = 145 cm, B = 1-2 T and for alpha particles in a reactor-size device. In contrast to tokamaks, it is found that high edge poloidal flux has limited value in improving ion confinement in QAS, since collisional pitch-angle scattering drives ions into ripple wells and stochastic field regions, where they are quickly lost. The necessity for reduced stellarator ripple fields is emphasized. The high neutral beam ion loss predicted for these configurations suggests that more interesting physics could be explored with an experiment of less constrained size and magnetic field geometry.

  16. 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.

  17. Destabilization of fast-ion-induced long sawteeth by localized current drive in the JET tokamak.

    PubMed

    Eriksson, L-G; Mueck, A; Sauter, O; Coda, S; Mantsinen, M J; Mayoral, M-L; Westerhof, E; Buttery, R J; McDonald, D; Johnson, T; Noterdaeme, J-M; De Vries, P

    2004-06-11

    In a tokamak fusion reactor the energetic alpha particles will transiently stabilize the magnetohydrodynamic activity causing sawtooth oscillations. The crash events terminating long sawtooth free periods can provide seed islands for neoclassical tearing modes [Phys. Rev. Lett. 88, 105001 (2002)

  18. 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.

  19. Enhanced confinement regimes and control technology in the DIII-D tokamak

    SciTech Connect

    Lohr, J.; Burrell, K.H.; Coda, S.

    1993-07-01

    Advanced tokamak performance has been demonstrated in the DIII-D tokamak in a series of experiments which brought together developments in technology and improved understanding of the physical principles underlying tokamak operation. The achievement of greatly improved confinement coupled with development of new systems for real time plasma control have permitted investigation of the heretofore hidden or poorly controlled variables which together determine global confinement. These experiments, which included work in transport and control of the plasma boundary, point toward development of operationally and economically attractive reactors based on the tokamak. Some of these experiments are described.

  20. 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.

  1. 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.

  2. An overview of the Tokamak Physics Experiment vacuum vessel preliminary design

    SciTech Connect

    Rocco, R.E.

    1995-12-31

    The mission of the Tokamak Physics Experiment (TPX) Project is to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. The vacuum vessel, which consists of a double walled torus, ports and supports, is a major element of the TPX machine. This paper provides an overview of the vacuum vessel preliminary design work. The design of the vacuum vessel is being carried out by an industrial team under subcontract to the Princeton Plasma Physics Laboratory. The respective work scopes of this team are discussed. The role of concurrent engineering is presented in the context of this design-build subcontract. A discussion of the engineering requirements, material selection rationale and vacuum vessel configuration is provided. Titanium 6Al-4V will be used to fabricate the vacuum vessel. Significant material concerns were identified with the use of titanium; hydrogen embrittlement and the effects of borated water were the major issues. A research and development (R and D) program was established to resolve these material issues as well as to develop the vessel weld details. A comprehensive analytical effort was established to perform the structural and thermal analysis of the vessel. Design details of the vessel, supports, ports, and flanges are presented.

  3. 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.

  4. Ripple-induced energetic particle loss in tokamaks

    NASA Astrophysics Data System (ADS)

    White, R. B.; Goldston, R. J.; Redi, M. H.; Budny, R. V.

    1996-08-01

    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 axisymmetic orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor [R. Hawryluk, Plasma Phys. Controlled Fusion 33, 1509 (1991)] and International Thermonuclear Experimental Reactor [K. Tomabechi, Proceedings of the 12th International Conference on Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1989), Vol. 3, p. 214] equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks.

  5. Options for commercial tokamaks

    SciTech Connect

    Dabiri, A.E.; Keeton, D.C.; Thomson, S.L.

    1986-07-01

    Systems studies have been performed at the Fusion Engineering Design Center (FEDC) to assess commercial tokamak options. One study investigates the economics of high-beta operation and determines an optimum operating range of 10 to 20% beta, with a corresponding neutron wall loading of 6 to 8 MW/m/sup 2/. A second study determines conditions under which small, low-power tokamaks can be economically combined into a 1200-MW(electric) multiplex power plant. The results of these studies have directed future efforts at the FEDC toward a high-beta, tokamak design using a modular maintenance configuration.

  6. ITER (International Thermonuclear Experimental Reactor) reactor building design study

    SciTech Connect

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.; Canadian Fusion Fuels Technology Project, Mississauga, ON )

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. A self-consistent model of an isothermal tokamak

    NASA Astrophysics Data System (ADS)

    McNamara, Steven; Lilley, Matthew

    2014-10-01

    Continued progress in liquid lithium coating technologies have made the development of a beam driven tokamak with minimal edge recycling a feasibly possibility. Such devices are characterised by improved confinement due to their inherent stability and the suppression of thermal conduction. Particle and energy confinement become intrinsically linked and the plasma thermal energy content is governed by the injected beam. A self-consistent model of a purely beam fuelled isothermal tokamak is presented, including calculations of the density profile, bulk species temperature ratios and the fusion output. Stability considerations constrain the operating parameters and regions of stable operation are identified and their suitability to potential reactor applications discussed.

  16. Resonant-cavity ICRF coupler for large tokamaks

    SciTech Connect

    Perkins, F.W.; Kluge, R.F.

    1983-04-01

    A new resonant-cavity ICRF coupler is proposed for large tokamaks. The design features a novel resonant cavity, an rf magnetic-field orientation that effectively radiates fast Alfven waves, matching to 40 ..cap omega.. transmission lines, and an electric-field orientation so that the strongest rf electric fields are orthogonal to the main toroidal magnetic field thereby benefitting from magnetic insulation. As a result, the power handling capability is excellent. For the case of the Big-Dee Doublet III tokamak, a single 35 cm x 50 cm coupler can launch 20 MW of fast Alfven waves. Extrapolation to fusion reactor parameters is straightforward.

  17. 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.

  18. 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.

  19. Nondiffusive plasma transport at tokamak edge

    NASA Astrophysics Data System (ADS)

    Krasheninnikov, S. I.

    2000-10-01

    Recent findings show that cross field edge plasma transport at tokamak edge does not necessarily obey a simple diffusive law [1], the only type of a transport model applied so far in the macroscopic modeling of edge plasma transport. Cross field edge transport is more likely due to plasma filamentation with a ballistic motion of the filaments towards the first wall. Moreover, it so fast that plasma recycles on the main chamber first wall rather than to flow into divertor as conventional picture of edge plasma fluxes suggests. Crudely speaking particle recycling wise diverted tokamak operates in a limiter regime due to fast anomalous non-diffusive cross field plasma transport. Obviously that this newly found feature of edge plasma anomalous transport can significantly alter a design of any future reactor relevant tokamaks. Here we present a simple model describing the motion of the filaments in the scrape off layer and discuss it implications for experimental observations. [1] M. Umansky, S. I. Krasheninnikov, B. LaBombard, B. Lipschultz, and J. L. Terry, Phys. Plasmas 6 (1999) 2791; M. Umansky, S. I. Krasheninnikov, B. LaBombard and J. L. Terry, Phys. Plasmas 5 (1998) 3373.

  20. LHCD Scenarios for Spherical Tokamak Plasmas

    SciTech Connect

    Takase, Y.; Ejiri, A.; Oosako, T.; Bonoli, P. T.; Wright, J. C.

    2007-09-28

    Noninductive plasma current start-up and sustainment are crucial issues for spherical tokamak reactors and other applications such as component test facility. It is widely recognized that the lower hybrid wave (the slow wave), which is most efficient in driving current, is not accessible to the core of a fully developed spherical tokamak plasma with very high dielectric constant. However, it may be useful in the initial plasma current ramp-up phase while the density is still low, where it is not practical to use other methods of noninductive current drive. Such a possibility is investigated theoretically for planned experiments on the TST-2 spherical tokamak at the University of Tokyo. The transmitters previously used for FWCD experiments on JFT-2M (200 MHz) are being prepared for this experiment. The combline antenna used for JFT-2M has been modified for use in TST-2. This antenna will be used to excite a unidirectional fast wave traveling in the toroidal direction with a toroidal mode number of 12 (corresponding to an initial parallel index of refraction of about 5). The fast wave can mode convert to the lower hybrid wave and drive current under some conditions. Examination of the dispersion relation indicates that there may be a suitable regime at relatively high field (0.3 T) and low density (<1x10{sup 19} m{sup -3})

  1. 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.

  2. 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.

  3. 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.

  4. Prospects and status of low-aspect-ratio tokamaks

    SciTech Connect

    Peng, Y.K.M.

    1994-12-31

    The prospects for the low-aspect-ratio (A) tokamak to fulfill the requirements of viable fusion power plants are considered relative to the present status in data and modeling. Desirable physics and design features for an attractive Blanket Test Facility and power reactors are estimated for low-A tokamaks based on calculations improved with the latest data from small pioneering experiments. While these experiments have confirmed some of the recent predictions for low-A, they also identify the remaining issues that require verification before reliable projections can be made for these deuterium-tritium applications. The results show that the low-A regime of small size, modest field, and high current offers a path complementary to the standard and high A tokamaks in developing the full potential of fusion power.

  5. 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

  6. Mathematical modeling plasma transport in tokamaks

    NASA Astrophysics Data System (ADS)

    Qiang, Ji

    1998-11-01

    In this work, we have 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 the next generation machine, ITER. The ignition probability of ITER for engineering design activity (EDA) parameters can be formally as high as 99.9% in the present context. The same probability for conceptual design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%. This suggests that EDA parameters for ITER tokamak are very likely to achieve the self- sustained thermonuclear reaction, but CDA parameters are risky for the realization of ignition.

  7. Magnet technology for the Engineering Test Reactor

    SciTech Connect

    Henning, C.D.; Miller, J.R.

    1987-09-17

    The consideration for building an international Engineering Test Reactor emerged from the November 1985 Geneva Summit, in which President Reagan and Secretary Gorbachev called for the ''widest practical development of international cooperation'' in fusion. In parallel with the OTR design in the USSR, the FER in Japan, and the NET in Europe, the US has pursued the TIBER (Tokamak Ignition/Burn Experimental Reactor). This compact design of 3-m major radius is achievable because of high-current-density, radiation-tolerant magnets with nuclear heating rates up to 10 mW . cm/sup -3/. Full development of cable-in-conduit conductors (CICC) is seen as a credible path to achieve 40 A . mm/sup -2/ at fields of 12 T in the toroidal field (TF) coils and 14 T in the central poloidal field (PF) coils. Since neutron fluences of up to 10/sup 19/ n . cm/sup -2/ are expected in the TF coils, the unalloyed niobium-tin would be superior at 12 T. However, the central PF coil at 14 T is better shielded, so modified niobium tin would be advantageous. Polyimide insulation in the TF coils would withstand the equivalent 10/sup 10/) rads if loads in the winding pack are taken in compression. 13 refs., 7 figs., 2 tabs.

  8. 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.

  9. Proposals for an influential role of small tokamaks in mainstream fusion physics and technology research

    SciTech Connect

    Oost, G. van; Bosco, E. del; Gryaznevich, M. P.

    2006-12-04

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confinement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies, which because of the risky nature cannot be done in large machines without preliminary studies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. All these tasks are well recognised and reflected in documents and understood by the large tokamak teams. Recent experimental results will be presented of contributions to mainstream fusion physics and technology research on small tokamaks involved in the IAEA Coordinated Research Project 'Joint Research using small tokamaks', started in 2004.

  10. 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.

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

    SciTech Connect

    Neumeyer, C.L.

    1994-11-01

    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.

  12. 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.

  13. Diagnostics for hybrid reactors

    NASA Astrophysics Data System (ADS)

    Orsitto, Francesco Paolo

    2012-06-01

    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.

  14. 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.

  15. Physics of Compact Advanced Stellarators

    SciTech Connect

    M.C. Zarnstorff; L.A. Berry; A. Brooks; E. Fredrickson; G.-Y. Fu; S. Hirshman; S. Hudson; L.-P. Ku; E. Lazarus; D. Mikkelsen; D. Monticello; G.H. Neilson; N. Pomphrey; A. Reiman; D. Spong; D. Strickler; A. Boozer; W.A. Cooper; R. Goldston; R. Hatcher; M. Isaev; C. Kessel; J. Lewandowski; J. Lyon; P. Merkel; H. Mynick; B.E. Nelson; C. Nuehrenberg; M. Redi; W. Reiersen; P. Rutherford; R. Sanchez; J. Schmidt; R.B. White

    2001-08-14

    Compact optimized stellarators offer novel solutions for confining high-beta plasmas and developing magnetic confinement fusion. The 3-D plasma shape can be designed to enhance the MHD stability without feedback or nearby conducting structures and provide drift-orbit confinement similar to tokamaks. These configurations offer the possibility of combining the steady-state low-recirculating power, external control, and disruption resilience of previous stellarators with the low-aspect ratio, high beta-limit, and good confinement of advanced tokamaks. Quasi-axisymmetric equilibria have been developed for the proposed National Compact Stellarator Experiment (NCSX) with average aspect ratio 4-4.4 and average elongation of approximately 1.8. Even with bootstrap-current consistent profiles, they are passively stable to the ballooning, kink, vertical, Mercier, and neoclassical-tearing modes for beta > 4%, without the need for external feedback or conducting walls. The bootstrap current generates only 1/4 of the magnetic rotational transform at beta = 4% (the rest is from the coils), thus the equilibrium is much less nonlinear and is more controllable than similar advanced tokamaks. The enhanced stability is a result of ''reversed'' global shear, the spatial distribution of local shear, and the large fraction of externally generated transform. Transport simulations show adequate fast-ion confinement and thermal neoclassical transport similar to equivalent tokamaks. Modular coils have been designed which reproduce the physics properties, provide good flux surfaces, and allow flexible variation of the plasma shape to control the predicted MHD stability and transport properties.

  16. 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.

  17. Experimental modelling of eddy currents and deflection for tokamak limiters

    SciTech Connect

    Hua, T.Q.; Knott, M.J.; Turner, L.R.; Wehrle, R.B.

    1986-11-01

    During plasma disruptions in a tokamak fusion reactor, eddy currents are induced in the limiters and other conducting structures surrounding the plasma. Interactions between these currents with the toroidal field causes deflection and stress in the structural components. The structural motion in the strong magnetic field induces additional eddy current opposing the initial eddy current and modifying subsequent structural dynamics. Therefore, the motion and current are coupled and must be solved simultaneously. The coupling between current and deflection in cantilevered beams was investigated experimentally. The beams provide a simple model for the limiter blade of a tokamak fusion reactor. Several test pieces and various magnetic field conditions were employed to study the extend of the coupling effect from weak to strong coupling. Experimental results are compared with analytical predictions.

  18. 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

  19. 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.

  20. Advanced tokamak operations with ICRF and lower-hybrid power

    NASA Astrophysics Data System (ADS)

    Mau, T. K.; Lee, B. J.; Ehst, D. A.

    1994-10-01

    Advanced tokamak operating modes based on high bootstrap current, first- and second-stability regime plasma are examined in the context of the TPX experiment and the ARIES reactors, using a combination of ICRF fast wave and lower hybrid power. The main method of analysis entails the alignment of driven current density profiles with those required for stability. In most of the cases studied, the required power levels and launched spectra are found to be reasonable.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. JET, the largest tokamak on the eve of DT operation

    NASA Astrophysics Data System (ADS)

    Horton, L. D.

    2016-11-01

    The Joint European Torus (JET) is the world's largest operating tokamak and the only such machine capable of operating with the fuel mixture (deuterium-tritium) foreseen for a fusion reactor. Since it came into operation in 1983, JET has explored fusion plasmas "in conditions and dimensions approaching those of a fusion reactor" [1]. JET has demonstrated world-record levels of fusion power and energy production, in conditions where the ratio of the fusion power generated to the input power to the plasma, Q, approaches unity.

  6. Joint Experiments on the Tokamaks CASTOR and T-10

    SciTech Connect

    Oost, G. van; Peleman, P.; Gryaznevich, M.; Malaquias, A.; Mank, G.; Berta, M.; Brotankova, J.; Dejarnac, R.; Dufkova, E.; Duran, I.; Hron, M.; Sentkerestiova, J.; Stoeckel, J.; Weinzettl, V.; Zajac, J.; Budaev, V.

    2008-04-07

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confiement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. The first Joint (Host Laboratory) Experiment (JE1) has been carried out in 2005 on the CASTOR tokamak at the IPP Prague, Czech Republic. It was jointly organized by the IPP-ASCR and KFKI HAC, Budapest, involved 20 scientists from 7 countries and was supported through the IAEA and the ICTP, Trieste. The objective of JE1 was to perform studies of plasma edge turbulence and plasma confinement. Following the success of JE1, JE2 has been performed on T-10 at RRC 'Kurchatov Institute' in Moscow; 30 scientists from 13 countries participated in this experiment. This experiment aimed to continue JE1 turbulence studies, now extending them to the plasma core. Results of JE1 and JE2 will be overviewed and compared.

  7. 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.

  8. 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.

  9. Mathematical modeling plasma transport in tokamaks

    SciTech Connect

    Quiang, Ji

    1997-01-01

    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 1020/m3 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%.

  10. First neutron spectrometry measurement at the HL-2A Tokamak

    NASA Astrophysics Data System (ADS)

    Yuan, Xi; Zhang, Xing; Xie, Xu-Fei; Chen, Zhong-Jing; Peng, Xing-Yu; Fan, Tie-Shuan; Chen, Jin-Xiang; Li, Xiang-Qing; Yuan, Guo-Liang; Yang, Qing-Wei; Yang, Jin-Wei

    2013-12-01

    A compact neutron spectrometer based on the liquid scintillator is presented for neutron energy spectrum measurements at the HL-2A Tokamak. The spectrometer was well characterized and a fast digital pulse shape discrimination software was developed using the charge comparison method. A digitizer data acquisition system with a maximum frequency of 1 MHz can work under an environment with a high count rate at HL-2A Tokamak. Specific radiation and magnetic shielding for the spectrometer were designed for the neutron spectrum measurement at the HL-2A Tokamak. For pulse height spectrum analysis, dedicated numerical simulation utilizing NUBEAM combined with GENESIS was performed to obtain the neutron energy spectrum. Subsequently, the transportation process from the plasma to the detector was evaluated with Monte Carlo calculations. The distorted neutron energy spectrum was folded with the response matrix of the liquid scintillation spectrometer, and good consistency was found between the simulated and measured pulse height spectra. This neutron spectrometer based on a digital acquisition system could be well adopted for the investigation of the auxiliary heating behavior and the fast-ion related phenomenon on different tokamak devices.

  11. 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.

  12. The external kink mode in diverted tokamaks

    DOE PAGES

    Turnbull, Alan D.; Hanson, Jeremy M.; Turco, Francesca; ...

    2016-06-16

    Here, an explanation is provided for the disruptive instability in diverted tokamaks when the safety factor at the 95% poloidal flux surface, q95, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections when qedge, the safety factor at the outermost closed flux surface, lies just below a rational value. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and qedge decreases through successive rational surfaces. For qedge < 2, the instability is always encountered andmore » is highly disruptive. However, diverted plasmas, in which qedge is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when q95, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when q95 < 2, but qedge >> 2. The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the q = 2 surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.« less

  13. Gyrokinetic simulation of microturbulence in EAST tokamak

    NASA Astrophysics Data System (ADS)

    Xiao, Yong; Zhang, Taige; Zhao, Chen

    2014-10-01

    A complete understanding of anomalous transport is critical for designing future magnetic fusion reactors. It is generally accepted that the micro-scale turbulence leads to anomalous transport. For low beta toroidal plasmas, the electrostatic modes may dominate and ion temperature gradient (ITG) mode and trapped electron mode (TEM) are two very important candidates accounting for ion and electron turbulent transport respectively. Recently the massively parallel gyrokinetic simulation has emerged as a major tool to investigate the nonlinear physics of the turbulent transport. The newly-developed capabilities enable the gyrokinetic code GTC to simulate the turbulent transport for real tokamak plasma shape and profiles. These capabilities include a new gyrokinetic Poisson solver and zonal flow solver suitable for general plasma shape and profiles, improvements on the conventional four-point gyroaverage and newly-developed nonuniform initial marker loading. The GTC code is now able to import experimental plasma profiles and equilibrium magnetic field that come from the EFIT or TRANSP equilibrium reconstruction. Linear and nonlinear gyrokinetic simulations are carried out with the new capabilities in GTC for the electron coherent mode (ECM) recently observed in the EAST tokamak (EAST shot # 38300). We found that in the pedestal region with strong electron temperature gradient, the unstable waves propagate in the electron diamagnetic direction, showing a trapped electron mode (TEM) feature. It is also found in the collisionless limit, the linear mode frequency is higher than that from the experiment.

  14. Predicting temperature and density profiles in tokamaks

    SciTech Connect

    Bateman, G.; Kritz, A.H.; Kinsey, J.E.; Redd, A.J.; Weiland, J.

    1998-05-01

    A fixed combination of theory-based transport models, called the Multi-Mode Model, is used in the BALDUR [C. E. Singer {ital et al.}, Comput. Phys. Commun. {bold 49}, 275 (1988)] transport simulation code to predict the temperature and density profiles in tokamaks. The choice of the Multi-Mode Model has been guided by the philosophy of using the best transport theories available for the various modes of turbulence that dominate in different parts of the plasma. The Multi-Mode model has been found to provide a better match to temperature and density profiles than any of the other theory-based models currently available. A description and partial derivation of the Multi-Mode Model is presented, together with three new examples of simulations of the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire {ital et al.}, Phys. Plasmas {bold 2}, 2176 (1995)]. The first simulation shows the strong effect of recycling on the ion temperature profile in TFTR supershot simulations. The second simulation explores the effect of a plasma current ramp{emdash}where the plasma energy content changes slowly on the energy confinement time scale. The third simulation shows that the Multi-Mode Model reproduces the experimentally measured profiles when tritium is used as the hydrogenic isotope in L-mode (low confinement mode) plasmas. {copyright} {ital 1998 American Institute of Physics.}

  15. High-beta equilibria of drift-optimized compact stellarators.

    PubMed

    Ware, A S; Hirshman, S P; Spong, D A; Berry, L A; Deisher, A J; Fu, G Y; Lyon, J F; Sanchez, R

    2002-09-16

    Compact stellarator configurations have been obtained with good neoclassical confinement that are stable to both pressure- and current-driven modes for high values of beta. These configurations are drift-optimized tokamak-stellarator hybrids with a high-shear tokamak-like rotational transform profile and /B/ that is approximately poloidally symmetric. The bootstrap current is consistent with the required equilibrium current and, while larger than that in existing stellarators, is typically only a small fraction (1/3-1/5) of that in an equivalent tokamak. These configurations have strong magnetic wells and consequently high interchange stability beta limits up to beta=23%. Because of the reduced bootstrap current, these configurations are stable to low-n ideal MHD kink modes with no wall stabilization for values of beta ( approximately 7%-11%) significantly larger than in an equivalent advanced tokamak.

  16. 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.

  17. Midplane Faraday rotation: A densitometer for large tokamaks

    NASA Astrophysics Data System (ADS)

    Jobes, F. C.; Mansfield, D. K.

    1992-10-01

    The density in a large tokamak such as International Thermonuclear Experimental Reactor (ITER), or any of the proposed future US machines, can be determined by measuring the Faraday rotation of a 10.6 μm laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then ne(R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, operated at a full field of 30 T m and a density of 2×1020/m3, the rotation angle would be quite large-about 60° for two passes. A layout in which a single laser beam is fanned out in the horizontal midplane of the tokamak, with a set of retroreflectors on the far side of the vacuum vessel, would provide good spatial resolution, depending only upon the number of reflectors. With this proposed layout, only one window would be needed. Because the rotation angle is never more than 1 ``fringe,'' the data is always good, and it is also a continuous measurement in time. Faraday rotation is dependent only upon the plasma itself, and thus is not sensitive to vibration of the optical components. Simulations of the expected results show that ITER, or any large tokamak, existing or proposed, would be well served even at low densities by a midplane Faraday rotation densitometer of ˜64 channels.

  18. Argonne Plasma Engineering Experiment (APEX) Tokamak

    SciTech Connect

    Norem, J.H.; Balka, L.J.; Kulovitz, E.E.; Magill, S.R.; McGhee, D.G.; Moretti, A.; Praeg, W.F.

    1981-03-01

    The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials.

  19. Edge turbulence in tokamaks

    NASA Astrophysics Data System (ADS)

    Nedospasov, A. V.

    1992-12-01

    Edge turbulence is of decisive importance for the distribution of particle and energy fluxes to the walls of tokamaks. Despite the availability of extensive experimental data on the turbulence properties, its nature still remains a subject for discussion. This paper contains a review of the most recent theoretical and experimental studies in the field, including mainly the studies to which Wootton (A.J. Wooton, J. Nucl. Mater. 176 & 177 (1990) 77) referred to most in his review at PSI-9 and those published later. The available theoretical models of edge turbulence with volume dissipation due to collisions fail to fully interpret the entire combination of experimental facts. In the scrape-off layer of a tokamak the dissipation prevails due to the flow of current through potential shifts near the surface of limiters of divertor plates. The different origins of turbulence at the edge and in the core plasma due to such dissipation are discussed in this paper. Recent data on the electron temperature fluctuations enabled one to evaluate the electric probe measurements of turbulent flows of particles and heat critically. The latest data on the suppression of turbulence in the case of L-H transitions are given. In doing so, the possibility of exciting current instabilities in biasing experiments (rather than only to the suppression of existing turbulence) is given some attention. Possible objectives of further studies are also discussed.

  20. The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    SciTech Connect

    Sokolov, Mikhail A.; Nanstad, Randy K.

    2016-09-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

  1. The theory of the quasi-optical grill: A lower hybrid wave launcher in the 4 - 10 GHz range for high field tokamaks

    SciTech Connect

    Preinhaelter, J.; Vahala, L.; Vahala, G.

    1996-12-31

    Lower hybrid (LH) waves have been utilized for plasma heating and current drive in tokamaks. LH current drive has good efficiency in low to moderate plasma temperatures and is an excellent tool for attaining the reversed shear regions of much interest in advanced steady state tokamak scenarios. For high field tokamaks, the waveguides of the standard multifunction grills would become very narrow and the walls separating the waveguides would need to be very thin. As a result, the cooling of such structures becomes very difficult. Moreover, there are concerns that the classical grill launcher could not withstand the conditions at the reactor first wall. The Quasi-Optical Grill (QOG) was first proposed by Petelin & Suvorov to overcome some of these difficulties. QOG attempts to couple the RF power to the plasma slow wave by means of the diffraction of the incident wave on an array of rods. However, these original calculations are based on certain idealized assumptions and lead to poor coupling to the plasma. Preinhaelter has suggested a new QOG in which the rods are placed in one oversized waveguide ({open_quotes}hyperguide{close_quotes}) and irradiated obliquely by the wave emerging as a higher order mode from an auxiliary oversized waveguide. The confining walls are now an intrinsic part of the structure and thus one avoids the need for mirrors and the introduction of {open_quote}point-like{close_quote} structures. This new QOG is compact - with several orders of magnitude less construction elements than the classical LH launcher - and the problem of wave diffraction can be readily solved using the full wave method. Here we consider the optimization of a large scale QOG at a given frequency. The irradiation of either a single row or double set of rows of rods are considered as well as their optimal separation. One can achieve transmissivity and directivity comparable to those of the multifunction grill. Design of a QOG for TORE-SUPRA will also be discussed.

  2. 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)

  3. The physics design of the Tokamak Physics Experiment

    NASA Astrophysics Data System (ADS)

    Thomassen, K. I.; Batchelor, D. B.; Bialek, J.

    1994-08-01

    The physics approaches to improved, steady-state tokamak reactors, as evolved through reactor design studies, ideas based on experimental results, and better theoretical understanding, are the foundation for the mission and physics design of the Tokamak Physics Experiment (TPX). The mission of TPX is to develop the scientific basis for cost-competitive, continuously operating tokamak power plants. We report here the design status of TPX, a device optimized to achieve improved performance through strong plasma shaping, recycling control, and current profile shaping, while operating continuously. The design incorporates poloidal field flexibility for a wide range of operation in normalized beta and internal inductance, a double-null 'Vee' divertor configuration for power and particle control, internal and external n (ne) 0 coils, as well as passive stabilizers, for control of MHD activity, and remote maintenance for continuous high-power operation in deuterium. Having superconducting poloidal and toroidal coils, the TPX device itself is capable of continuous operation, although initially auxiliary equipment limits the pulse length to 1000 sec.

  4. The physics design of the Tokamak Physics Experiment

    SciTech Connect

    Thomassen, K.I.; Batchelor, D.B.; Bialek, J.

    1994-08-08

    The physics approaches to improved, steady-state tokamak reactors, as evolved through reactor design studies, ideas based on experimental results, and better theoretical understanding, are the foundation for the mission and physics design of the Tokamak Physics Experiment (TPX). The mission of TPX is to develop the scientific basis for cost-competitive, continuously operating tokamak power plants. We report here the design status of TPX, a device optimized to achieve improved performance through strong plasma shaping, recycling control, and current profile shaping, while operating continuously. The design incorporates poloidal field flexibility for a wide range of operation in normalized beta and internal inductance, a double-null ``Vee`` divertor configuration for power and particle control, internal and external n {ne} 0 coils, as well as passive stabilizers, for control of MHD activity, and remote maintenance for continuous high-power operation in deuterium. Having superconducting poloidal and toroidal coils, the TPX device itself is capable of continuous operation, although initially auxiliary equipment limits the pulse length to 1000 sec.

  5. 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)

  6. Tokamak building-design considerations for a large tokamak device

    SciTech Connect

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release.

  7. 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.

  8. Cooling a birth distribution of {alpha}-particles in a tokamak with waves

    SciTech Connect

    Herrmann, M C; Fisch, N J

    1996-01-01

    Alpha particles, the byproducts of the DT reaction in tokamak fusion reactor, might be cooled through interactions with waves. Numerical simulations employing two waves,one with frequency about the alpha cyclotron frequency, and one at much lower frequency, show the existence of parameter regimes where more than half of the {alpha}-particle power can be diverted to the waves.

  9. Causes of major tokamak disruptions

    SciTech Connect

    White, R.B.; Monticello, D.A.

    1980-07-01

    The nonlinear saturation theory of the tearing mode is used to examine the necessary conditions for the occurrence of a major tokamak disruption. The results are compared with full three-dimensional numerical simulations, and with experimental data.

  10. Connections between physics and economics for Tokamak fusion power plants

    NASA Astrophysics Data System (ADS)

    Krakowski, R. A.; Delene, J. G.

    1988-03-01

    A simplified physics, engineering, and costing model of a tokamak fusion reactor is used to examine quantitatively the connection between physics performance and power-plant economics. The material contained herein was generated as part of a broader study of the economic, safety, and environmental impact of fusion based on a range of confinement schemes, fusion fuels, blanket/shield configurations, power-conversion schemes, and commercial end products. Only a DT-fuelled tokamak reactor that produces electricity through an intermediate heat exchange and a conventional thermal-electric conversion cycle is considered; a self-cooled lithium-metal blanket with vanadium-alloy structure, steel shield, and superconducting magnets is used for all cases studied. An optimistic extension of Troyon scaling is applied to a high-elongation ( κ = 2.5) and low-safety-factor ( q ψ =2.3) plasma with β=0.1 and efficient ( I φ P CD =0.2 A/W) current drive. This 1200-MWe (net) power plant provides an economically competitive base case with which to compare other approaches to tokamak fusion power. The base case chosen for comparisons represents an optimistic extrapolation of present tokamak physics and technology. Troyon scaling with a coefficient β B φ a/ I φ equal to 0.04 is applied; the impact of an ad hoc but pessimistic scaling that diminished the Troyon coefficient with plasma elongation was also examined. Additionally, a constant current-drive efficiency, ϒ= nI φ R T / P CD =0.2 A/W, at T=10 keV plasma temperature is assumed; although representing an aggressive R&D target relative to present experience, the realization of bootstrap currents for the basecase, and especially for the second-stability-region tokamak, can significantly reduce this problem. The impact and reoptimization for a constant normalized current-drive efficiency, ϒ= nI φ R T/ P CD, was also examined. Although the focus of this study has been the optimistic basecase tokamak, comparisons are made with

  11. 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.

  12. Bootstrap current in a tokamak

    SciTech Connect

    Kessel, C.E.

    1994-03-01

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.

  13. 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

  14. Tokamak Plasma Flows Induced by Local RF Forces

    NASA Astrophysics Data System (ADS)

    Chen, Jiale; Gao, Zhe

    2015-10-01

    The tokamak plasma flows induced by the local radio frequency (RF) forces in the core region are analyzed. The effective components of local RF forces are composed of the momentum absorption term and the resonant parallel momentum transport term (i.e. the parallel component of the resonant ponderomotive forces). Different momentum balance relations are employed to calculate the plasma flows depending on different assumptions of momentum transport. With the RF fields solved from RF simulation codes, the toroidal and poloidal flows by these forces under the lower hybrid current drive and the mode conversion ion cyclotron resonance heating on EAST-like plasmas are evaluated. supported by National Natural Science Foundation of China (Nos. 11405218, 11325524, 11375235 and 11261140327), in part by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB111002, 2013GB112001 and 2013GB112010), and the Program of Fusion Reactor Physics and Digital Tokamak with the CAS “One-Three-Five” Strategic Planning

  15. [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.

  16. The Thor tokamak experiment

    NASA Astrophysics Data System (ADS)

    Argenti, L.; Bonizzoni, G.; Cirant, S.; Corti, S.; Grosso, G.; Lampis, G.; Rossi, L.; Carretta, U.; Jacchia, A.; de Luca, F.

    1981-06-01

    The principle characteristics of plasma discharges produced in Thor tokamak experiments are discussed. The equilibrium and stability characteristics of the plasma produced are considered, with attention given to the density limits and critical streaming parameter for stable operation. The temporal evolution of the main plasma parameters, including electron density, electron temperature distribution, hard X-ray emission from suprathermal electrons, neutral gas influx, plasma density and Ohmic heating efficiency, is then examined, with particular emphasis on means used to control the electron runaway. The results achieved are noted to have demonstrated the possibility of controlling both plasma equilibrium and discharge regime, and further improvements expected by the use of more efficient preionization, gas puffing and feedback poloidal control of column position are indicated.

  17. Filamentation in tokamaks

    SciTech Connect

    Cardozo, N.J.; Barth, C.J.; Chu, C.C.; Lok, J.; Montvai, A.; Oomens, A.A.; Peters, M.; Pijper, F.J.; de Rover, M.; Schueller, F.C.; Steenbakkers, M.F.; RTP team

    1995-09-01

    The relevance of a nest of toroidal flux surfaces as a paradigm of the magnetic topology of a tokamak plasma is challenged. High resolution Thomson scattering measurements of electron temperature and density in RTP show several hot filaments in the plasma center and sharp gradients near the sawtooth inversion radius and structures outside the sawtooth region under central ECH. In ohmic plasmas, too, the pressure and temperature profiles show significant bumps. These measurements give evidence of a complex magnetic topology. Transport in a medium with spatially strongly varying diffusivity is considered. It is shown that macroscopic transport is determined by the microscopic structure: a transport theory must predict this structure and the diffusivity in the insulating regions, while the {open_quote}turbulent{close_quote} diffusivity is irrelevant. A numerical approach to equilibria with broken surfaces is presented. {copyright} {ital 1995 American Institute of Physics.}

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Ion Bernstein wave heating on the Compact Ignition Tokamak (CIT)

    SciTech Connect

    Ignat, D.W.; Ono, M.

    1989-02-01

    In the present plan, CIT is to be heated by power in the ion cyclotron range of frequencies (ICRF), and electron cyclotron heating (ECH) may be used if suitable rf sources can be developed. We consider the option of ion Bernstein wave heating (IBWH). The key points are that a simple vacuum waveguide launcher can be well- removed from high fluxes of heat and particles and that the development of a suitable source is straightforward. A practical point is that an IBWH waveguide launcher, including transition from coaxial power feeds, fits inside the shield wall surrounding CIT. To confirm IBWH as an option for CIT, experiments are needed on a shaped, H-mode plasma at high power. Successful experiments should be followed by a tube development program to allow CIT heating at 200 - 275 MHz. 2 refs., 3 figs.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. The external kink mode in diverted tokamaks

    SciTech Connect

    Turnbull, Alan D.; Hanson, Jeremy M.; Turco, Francesca; Ferraro, Nathaniel M.; Lanctot, Matthew J.; Lao, Lang L.; Strait, Edward J.; Piovesan, Paolo; Martin, Piero

    2016-06-16

    Here, an explanation is provided for the disruptive instability in diverted tokamaks when the safety factor at the 95% poloidal flux surface, q95, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections when qedge, the safety factor at the outermost closed flux surface, lies just below a rational value. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and qedge decreases through successive rational surfaces. For qedge < 2, the instability is always encountered and is highly disruptive. However, diverted plasmas, in which qedge is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when q95, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when q95 < 2, but qedge >> 2. The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the q = 2 surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.

  9. 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.

  10. 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.

  11. The engineering design of the Tokamak Physics Experiment

    SciTech Connect

    Schmidt, J.A.

    1994-07-01

    A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design. If adequately funded the construction project should be completed in the year 2000.

  12. 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.

  13. Prospects for a dominantly microwave-diagnosed magnetically confined fusion reactor

    NASA Astrophysics Data System (ADS)

    Volpe, F. A.

    2017-01-01

    Compared to present experiments, tokamak and stellarator reactors will be subject to higher heat loads, sputtering, erosion and subsequent coating, tritium retention, higher neutron fluxes, and a number of radiation effects. Additionally, neutral beam penetration in tokamak reactors will only be limited to the plasma edge. As a result, several optical, beam-based and magnetic diagnostics of today's plasmas might not be applicable to tomorrow's reactors, but the present discussion suggests that reactors could largely rely on microwave diagnostics, including techniques based on mode conversions and Collective Thomson Scattering.

  14. Engineering, safety, and economic evaluations of ASPIRE (Advanced Safe Pool Immersed REactor)

    SciTech Connect

    Sze, D.K.; Gordon, J.; Piet, S.; Cheng, E.T.; Klein, A.

    1988-02-01

    A preconceptual design of a tokamak fusion reactor concept called ASPIRE (Advanced Safe Pool Immersed REactor) has been developed. This concept provides many of the attractive features that are needed to enhance the capability of fusion to become the power generation technology for the 21st century. Specifically, these features are: inherent safety, low pressure, environmental compatibility, moderate unit size, high availability, high thermal efficiency, simplicity, low radioactive inventory, Class C radioactive waste disposal, and low cost of electricity. We have based ASPIRE on a second stability tokamak. However, the concept is equally applicable to a first stability tokamak or to most other magnetic fusion systems.

  15. Optimization study of normal conductor tokamak for commercial neutron source

    NASA Astrophysics Data System (ADS)

    Fujita, T.; Sakai, R.; Okamoto, A.

    2017-05-01

    The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

  16. Intrinsic momentum transport in up-down asymmetric tokamaks

    NASA Astrophysics Data System (ADS)

    Ball, Justin; Parra, Felix I.; Barnes, Michael; Dorland, William; Hammett, Gregory W.; Rodrigues, Paulo; Loureiro, Nuno F.

    2014-09-01

    Recent work has demonstrated that breaking the up-down symmetry of tokamak flux surfaces removes a constraint that limits intrinsic momentum transport, and hence toroidal rotation, to be small. We show, through MHD analysis, that ellipticity is most effective at introducing up-down asymmetry throughout the plasma. We detail an extension to GS2, a local δf gyrokinetic code that self-consistently calculates momentum transport, to permit up-down asymmetric configurations. Tokamaks with tilted elliptical poloidal cross-sections were simulated to determine nonlinear momentum transport. The results, which are consistent with the experiment in magnitude, suggest that a toroidal velocity gradient, (∂uζi/∂ρ)/vthi, of 5% of the temperature gradient, (∂Ti/∂ρ)/Ti, is sustainable. Here vthi is the ion thermal speed, uζi is the ion toroidal mean flow, ρ is the minor radial coordinate normalized to the tokamak minor radius, and Ti is the ion temperature. Though other known core intrinsic momentum transport mechanisms scale poorly to larger machines, these results indicate that up-down asymmetry may be a feasible method to generate the current experimentally measured rotation levels in reactor-sized devices.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Compact vortices

    NASA Astrophysics Data System (ADS)

    Bazeia, D.; Losano, L.; Marques, M. A.; Menezes, R.; Zafalan, I.

    2017-02-01

    We study a family of Maxwell-Higgs models, described by the inclusion of a function of the scalar field that represent generalized magnetic permeability. We search for vortex configurations which obey first-order differential equations that solve the equations of motion. We first deal with the asymptotic behavior of the field configurations, and then implement a numerical study of the solutions, the energy density and the magnetic field. We work with the generalized permeability having distinct profiles, giving rise to new models, and we investigate how the vortices behave, compared with the solutions of the corresponding standard models. In particular, we show how to build compact vortices, that is, vortex solutions with the energy density and magnetic field vanishing outside a compact region of the plane.

  3. Compact HPD

    SciTech Connect

    Suyama, M.; Kawai, Y.; Kimura, S.

    1996-12-31

    In order to be utilized in such application fields as high energy physics or medical imaging, where a huge number of photodetectors are assembled in designated small area, the world`s smallest HPD, the compact BFD, has been developed. The overall diameter and the length of the tube are 16mm and 15mm, respectively. The effective photocathode area is 8mm in diameter. At applied voltage of -8kV to the photocathode, the electron multiplication gain of a PD incorporated HPD (PD-BPD) is 1,600, and that of an APD (APD-BPD) is 65,000. In the pulse height distribution measurement, photoelectron peaks up to 6 photoelectrons are clearly distinguishable with the APD-BPD. Experiments established that there was no degradation of gain in magnetic fields up to 1.5T, an important performance characteristic of the compact BPD for application in high energy physics.

  4. 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).

  5. Comment on ‘On the fusion triple product and fusion power gain of tokamak pilot plants and reactors’, by A. Costley

    NASA Astrophysics Data System (ADS)

    Biel, W.; Lackner, K.; Sauter, O.; Wenninger, R.; Zohm, H.

    2017-03-01

    In this comment, we discuss the arguments raised in two recent papers (Costley 2016 Nucl. Fusion 56 066003, Costley et al 2015 Nucl. Fusion 55 033001) on the claimed size independence of fusion power, triple product and fusion gain in tokamak reactors, and we show that all these three quantities actually do depend on the size of the tokamak, when distinguishing between independent input parameters (design parameters) and output quantities, and when taking into account technological limitations.

  6. An enhanced tokamak startup model

    NASA Astrophysics Data System (ADS)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  7. High Energy Particles in Tokamaks

    SciTech Connect

    White, R. B.

    2008-05-14

    This lecture covers the derivation of guiding center equations in a tokamak, orbit classification, the effect of magnetic perturbations and ripple, the interaction of particles with magnetohydrodynamic modes, including passing particle resonance, toroidal Alfven mode drive and saturation, the fishbone mode, and sawtooth stabilization.

  8. 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.

  9. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Hughes, Paul E.

    It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

  10. 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

  11. Experimental modeling of eddy currents and defections for tokamak limiters

    SciTech Connect

    Hua, T.Q.; Knott, M.J.; Turner, L.R.; Wehrle, R.B.

    1986-01-01

    In a tokamak fusion reactor or other magnetically confined fusion device, a rapid decay of the magnetic field due to a disrupting plasma current induces eddy currents in the surrounding structures. These include the limiters, blanket, first wall, and vacuum vessel. The eddy currents, through interaction with the applied toroidal and poloidal magnetic fields, produce large mechanical torques and forces that deflect the structural components. An important coupling effect exists between the dynamic behavior of the structure and the transient eddy current. This coupling occurs when, as it deflects, the component intercepts additional magnetic flux. The coupling between deflection and eddy current could reduce the peak current, deflection, and other electromagnetic effects to a level far less severe than would be predicted if coupling is disregarded. In this study, experiments were performed to investigate deflection, current, and material stress in cantilever beams with the Fusion ELectromagnetic Induction eXperiment (FELIX) at the Argonne National Laboratory. Since structures near the plasma are typically cantilevered, the beams provide a good model for the limiter blades of a tokamak fusion reactor. 2 figs.

  12. 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.

  13. Neoclassical transport of impurtities in tokamak plasmas

    SciTech Connect

    Hirshman, S.P.; Sigmar, D.J.

    1981-05-01

    Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I//sup 2//n/sub H/e/sup 2/ greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Cryogenic needs for future tokamaks

    NASA Astrophysics Data System (ADS)

    Katheder, H.

    The ITER tokamak is a machine using superconducting magnets. The windings of these magnets will be subjected to high heat loads resulting from a combination of nuclear energy absorption and AC-losses. It is estimated that about 100 kW at 4.5 K are needed. The total cooling mass flow rate will be around 10 - 15 kg/s. In addition to the large cryogenic power required for the superconducting magnets cryogenic power is also needed for refrigerated radiation shield, various cryopumps, fuel processing and test beds. A general description of the overall layout and the envisaged refrigerator cycle, necessary cold pumps and ancillary equipment is given. The basic cryogenic layout for the ITER tokakmak design, as developed during the conceptual design phase and a short overview about existing tokamak designs using superconducting magnets is given.

  19. Options for an ignited tokamak

    SciTech Connect

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon ..beta../sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed.

  20. Folded waveguide designs for tokamaks

    NASA Astrophysics Data System (ADS)

    Hoffman, D. J.; Bigelow, T. S.; Fogelman, C. H.; Yugo, J. J.; Caughman, J. B. O.; Gardner, W. L.; Carter, M. D.; Probert, P. H.; Barbato, E.

    The folded waveguide (FWG) has been tested to the megawatt level in RFTF and shows great promise for tokamak use. It has three primary advantages: low electric field (anywhere) per unit power coupled to the plasma, strong structural capabilities, and better spectral content than loops. A tokamak test is now needed. Potential candidates include C-Mod at 80 MHz and FTU at 433 MHz. The waveguide test on the first machine will be directed at conventional ion cyclotron heating, while the test on the latter will be directed at direct electron heating. In addition, a variation of the folded waveguide is proposed to be tested on Phaedrus-T. In this paper, we discuss the advantages of the waveguide, the design layout, some of the potential physics programs, and how these programs may have an impact on its potential use in ITER.

  1. 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.

  2. 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

  3. Magnetic island formation in tokamaks

    SciTech Connect

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs.

  4. Equilibrium Reconstruction in EAST Tokamak

    NASA Astrophysics Data System (ADS)

    Qian, Jinping; Wan, Baonian; L. Lao, L.; Shen, Biao; A. Sabbagh, S.; Sun, Youwen; Liu, Dongmei; Xiao, Bingjia; Ren, Qilong; Gong, Xianzu; Li, Jiangang

    2009-04-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.

  5. Compact torus

    SciTech Connect

    Furth, H.P.

    1980-10-01

    The objective of the compact torus approach is to provide toroidal magnetic-field configurations that are based primarily on plasma currents and can be freed from closely surrounding mechanical structures. Some familiar examples are the current-carrying plasma rings of reversed-field theta pinches and relativistic-electron smoke ring experiments. The spheromak concept adds an internal toroidal magnetic field component, in order to enhance MHD stability. In recent experiments, three different approaches have been used to generate spheromak plasmas: (1) the reversed-field theta pinch; (2) the coaxial plasma gun; (3) a new quasi-static method, based on the initial formation of a toroidal plasma sleeve around a mechanical ring that generates poloidal and toroidal fluxes, followed by field-line reconnection to form a detached spheromak plasma. The theoretical and experimental MHD stability results for the spheromak configuration are found to have common features.

  6. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  7. Global hybrid simulations of energetic particle effects on the n=1 mode in tokamaks: Internal kink and fishbone instability

    SciTech Connect

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

    2006-05-15

    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.

  8. 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.

  9. Leakage of runaway electrons from tokamaks

    SciTech Connect

    Wong, K.L.

    1982-02-01

    Runaway electron orbits are calculated in a tokamak magnetic field. It is shown that these electrons tend to drift towards a larger major radius with a velocity v Vector/sub R/ = qcE/B/sub 0/ R. This effect may be relevant to some recent experimental observations in tokamaks.

  10. Numerical tokamak turbulence project (OFES grand challenge)

    SciTech Connect

    Beer, M; Cohen, B I; Crotinger, J; Dawson, J; Decyk, V; Dimits, A M; Dorland, W D; Hammett, G W; Kerbel, G D; Leboeuf, J N; Lee, W W; Lin, Z; Nevins, W M; Reynders, J; Shumaker, D E; Smith, S; Sydora, R; Waltz, R E; Williams, T

    1999-08-27

    The primary research objective of the Numerical Tokamak Turbulence Project (NTTP) is to develop a predictive ability in modeling turbulent transport due to drift-type instabilities in the core of tokamak fusion experiments, through the use of three-dimensional kinetic and fluid simulations and the derivation of reduced models.

  11. 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.

  12. 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.

  13. 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)

  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. 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.

  16. Recent designs for advanced fusion reactor blankets

    SciTech Connect

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies.

  17. Compaction behavior of roller compacted ibuprofen.

    PubMed

    Patel, Sarsvatkumar; Kaushal, Aditya Mohan; Bansal, Arvind Kumar

    2008-06-01

    The effect of roller compaction pressure on the bulk compaction of roller compacted ibuprofen was investigated using instrumented rotary tablet press. Three different roller pressures were utilized to prepare granules and Heckel analysis, Walker analysis, compressibility, and tabletability were performed to derive densification, deformation, course of volume reduction and bonding phenomenon of different pressure roller compacted granules. Nominal single granule fracture strength was obtained by micro tensile testing. Heckel analysis indicated that granules prepared using lower pressure during roller compaction showed lower yield strength. The reduction in tabletability was observed for higher pressure roller compacted granules. The reduction in tabletability supports the results of granule size enlargement theory. Apart from the granule size enlargement theory, the available fines and relative fragmentation during compaction is responsible for higher bonding strength and provide larger areas for true particle contact at constant porosity for lower pressure roller compacted granules. Overall bulk compaction parameters indicated that granules prepared by lower roller compaction pressure were advantageous in terms of tabletability and densification. Overall results suggested that densification during roller compaction affects the particle level properties of specific surface area, nominal fracture strength, and compaction behavior.

  18. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Pellet injectors for the tokamak fusion test reactor

    SciTech Connect

    Combs, S.K.

    1986-01-01

    The repeating pneumatic injector is a device from the ORNL development program. A new eight-shot deuterium pellet injector has been designed and constructed specifically for the TFTR application and is scheduled to replace the repeating injector this year. The new device combines a cryogenic extruder and a cold wheel rotary mechanism to form and chamber eight pellets in a batch operation; the eight pellets can then be delivered in any time sequence. Another unique feature of the device is the variable pellet size with three pellets each of 3.0 and 3.5 mm diam and two each of 4.0 mm diam. The experience and technology that have been developed on previous injectors at ORNL have been utilized in the design of this latest pellet injection system.

  5. Transient getter scheme for the tokamak fusion test reactor

    NASA Astrophysics Data System (ADS)

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

    1980-02-01

    A method has been developed of utilizing SAES Zr/Al getter modules, which obviates the need for frequent interruptions of machine operation by using the pulsed operation of TFTR. These interruptions are required for regeneration of the absorbed tritium after as few as 50 machine pulses. With the Zr/Al getter at 500-600 C it is possible to achieve a quasisteady state in 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.

  6. 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.

  7. 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.

  8. The Microwave Tokamak Experiment (MTX)

    SciTech Connect

    Thomassen, K.I.; Cohen, B.I.; Hooper, E.B.; Lang, D.D.; Nevins, W.M.

    1987-10-02

    A new experimental facility is being assembled at the Lawrence Livermore National Laboratory (LLNL) for studying microwave propagation and absorption in high density plasmas. A unique feature of the facility is the free electron laser (FEL) used to generate high peak power microwaves at 250 GHz, at a repetition rate so as to produce up to 2 MW of average power for up to 30 s. Called the Microwave Tokamak Experiment (MTX), the facility will be used for studies of plasma heating, current drive, and confinement.

  9. Alpha particle confinement in tokamaks

    SciTech Connect

    White, R.B.; Mynick, H.E.

    1988-11-01

    An assessment of diffusive tokamak transport mechanisms of concern for alpha particles indicates that the ''stochastic regime'' is the only one which appears to pose a real danger for adequate alpha confinement. This fact, in conjunction with the threshold character of that mechanism, allows one to decide whether an alpha born at a given location will be lost or confined, according to a very simple criterion. Implementing this criterion numerically results in a new code for the assessment of alpha confinement, which is orders of magnitude faster than earlier codes used for this purpose. 13 refs., 3 figs., 1 tab.

  10. 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.

  11. Noniterative reconstruction of tokamak equilibria

    NASA Astrophysics Data System (ADS)

    Rodrigues, Paulo; Bizarro, João P. S.

    2009-02-01

    Unlike iterative approaches, noniterative equilibria reconstruction schemes are designed to keep two measured internal profiles fixed along a given chord while solving a sequence of linear differential equations, providing a unique and asymptotic solution to the Grad-Shafranov (GS) equation directly in laboratory coordinates. A noniterative algorithm is extended to handle plasma configurations that are not symmetric with respect to the tokamak midplane and then used to compute an equilibrium solution from an actual experimental data set. A number of issues concerning how available experimental data can be handled and provided as input to the GS solver in practical situations are also discussed.

  12. Global modelling of non-axisymmetric disruptions and halo currents in tokamaks

    NASA Astrophysics Data System (ADS)

    McCarrick, James F.

    1997-12-01

    As tokamak plasmas become more robust with the development of increasingly advanced operating regimes, the occurrence of plasma disruptions places a greater demand on the tokamak structure. In particular, the flow of halo currents, large currents which appear in tokamak vacuum vessels as a result of direct contact with bulk plasma, has become a matter of increasing concern. Experimental measurements have confirmed the existence of large, toroidally asymmetric currents which flow poloidally in the wall, exerting strong localized forces on the wall as they interact with the toroidal magnetic field. A new model has been developed to study this phenomenon, based on the use of nested sheet currents to represent a disrupting plasma. This model contains the minimum number of degrees of freedom which permit the flow of continuous, non-axisymmetric poloidal and toroidal currents; furthermore, the model can be put into a compact integral formulation which allows rapid numerical solution even in the presence of complicated tokamak geometries. A fast code called TSPS-3D has been written to solve the sheet current model; the code has been matched against experimental data and used to examine basic scaling relationships of halo currents and the resulting J x B loads with plasma parameters. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253- 1690.)

  13. Theoretical Predictions for Kr Radiation in a Tokamak

    NASA Astrophysics Data System (ADS)

    Lippmann, S. I.; Fournier, K. B.; Goldstein, W. H.

    1996-11-01

    There is current interest in the idea of using impurity radiation to control the energy exhaust from a future tokamak reactor. A suitable radiator would be one which radiates most strongly in the edge and does not cause a serious fuel dilution problem. Recently, the Hebrew University Lawrence Livermore Atomic Code (HULLAC) was adapted to this problem by making a prediction of the Kr radiation in a TFTR-like background plasma^1. Kr was considered an interesting candidate since the standard Post-Jensen cooling curves indicate a relatively large amount of radiation per ion in the edge relative to the center. HULLAC predicts even much more radiation (up to ~ 10^2× in the tokamak edge) from Kr in the edge than does the average-ion model. A similar result has is found for molybdenum^2. In this poster we will try to determine the nature of this descrepancy; whether it is do to a different ionization-recombination rates being used, an effect of the impurity transport assumptions used, or to some other theoretical problem. This work Supported by DOE subcontract B310463 to LLNL. ^1S. Lippmann, et al. Procedings of the 11th APS Topical Conference on Atomic Processes in Plasmas, 1996, p B20. ^2K. Fournier, these proceedings.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

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

    SciTech Connect

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

    2016-11-15

    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 {sup 252}Cf 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.

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

    PubMed

    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 (252)Cf 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.