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Sample records for fusion reactor applications

  1. Demountable vacuum seals for fusion reactor applications

    SciTech Connect

    Batzer, T.H.; Call, W.R.

    1987-10-16

    Demountable vacuum seals for fusion reactor applications must be compatible with the reactor environment, easily scalable, very reliable and readily maintained by remote handling methods. We are investigating gate valves as well as flanges in our efforts to provide such seals. They are all metal and scalable without becoming massive and require no axial fasteners. Preliminary tests on an initial 30 cm aluminum flange using no soft metal coatings or gaskets have given several vacuum tight closures. Weld fatigue of this preliminary design caused degradation of the seal with further cycling to leakage levels of 10/sup -6/ Tl/sec, which is acceptable with differential pumping for either valves or flanges. Additional flange pairs using slightly altered geometry, fabrication techniques, and seal plating materials will be tested and reported on.

  2. Chromium-molybdenum steels for fusion-reactor applications

    SciTech Connect

    Klueh, R.L.

    1981-08-01

    Because ferritic steels have been found to have excellent resistance to swelling when irradiated in a fast-breeder reactor, Cr-Mo steels have recently become of interest for nuclear applications, both as cladding and duct material for fast-breeder reactors and as a first-wall and blanket structural material for fusion reactors. In this paper we will assess the Cr-Mo steels for fusion reactor applications. Possible approaches on how Cr-Mo steels may be further developed for this application will be proposed.

  3. Organic materials for fusion-reactor applications

    SciTech Connect

    Hurley, G.F.; Coltman, R.R. Jr.

    1983-09-01

    Organic materials requirements for fusion-reactor magnets are described with reference to the temperature, radiation, and electrical and mechanical stress environment expected in these magnets. A review is presented of the response to gamma-ray and neutron irradiation at low temperatures of candidate organic materials; i.e. laminates, thin films, and potting compounds. Lifetime-limiting features of this response as well as needed testing under magnet operating conditions not yet adequately investigated are identified and recomendations for future work are made.

  4. Vanadium-base alloys for fusion reactor applications

    SciTech Connect

    Smith, D.L.; Loomis, B.A.; Diercks, D.R.

    1984-10-01

    Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.

  5. The properties and weldability of materials for fusion reactor applications

    SciTech Connect

    Chin, B.A.; Kee, C.K.; Wilcox, R.C.; Zinkle, S.J.

    1991-11-15

    Low-activation austenitic stainless steels have been suggested for applications within fusion reactors. The use of these nickel-free steels will help to reduce the radioactive waste management problem after service. one requirement for such steels is the ability to obtain sound welds for fabrication purposes. Thus, two austenitic Fe-Cr-Mn alloys were studied to characterize the welded microstructure and mechanical properties. The two steels investigated were a Russian steel (Fe-11.6Cr19.3Mn-0.181C) and an US steel (Fe-12.lCr-19.4Mn-0.24C). Welding was performed using a gas tungsten arc welding (GTAW) process. Microscopic examinations of the structure of both steels were conducted. The as-received Russian steel was found to be in the annealed state. Only the fusion zone and the base metal were observed in the welded Russian steel. No visible heat affected zone was observed. Examination revealed that the as-received US steel was in the cold rolled condition. After welding, a fusion zone and a heat affected zone along with the base metal region were found.

  6. 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.

  7. Aerosol Resuspension Model for MELCOR for Fusion and Very High Temperature Reactor Applications

    SciTech Connect

    B.J. Merrill

    2011-01-01

    Dust is generated in fusion reactors from plasma erosion of plasma facing components within the reactor’s vacuum vessel (VV) during reactor operation. This dust collects in cooler regions on interior surfaces of the VV. Because this dust can be radioactive, toxic, and/or chemically reactive, it poses a safety concern, especially if mobilized by the process of resuspension during an accident and then transported as an aerosol though out the reactor confinement building, and possibly released to the environment. A computer code used at the Idaho National Laboratory (INL) to model aerosol transport for safety consequence analysis is the MELCOR code. A primary reason for selecting MELCOR for this application is its aerosol transport capabilities. The INL Fusion Safety Program (FSP) organization has made fusion specific modifications to MELCOR. Recent modifications include the implementation of aerosol resuspension models in MELCOR 1.8.5 for Fusion. This paper presents the resuspension models adopted and the initial benchmarking of these models.

  8. Development of ferritic steels for fusion reactor applications

    SciTech Connect

    Klueh, R.L.; Maziasz, P.J.; Corwin, W.R.

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs.

  9. Proceedings of the Office of Fusion Energy/DOE workshop on ceramic matrix composites for structural applications in fusion reactors

    SciTech Connect

    Jones, R.H. ); Lucas, G.E. )

    1990-11-01

    A workshop to assess the potential application of ceramic matrix composites (CMCs) for structural applications in fusion reactors was held on May 21--22, 1990, at University of California, Santa Barbara. Participants included individuals familiar with materials and design requirements in fusion reactors, ceramic composite processing and properties and radiation effects. The primary focus was to list the feasibility issues that might limit the application of these materials in fusion reactors. Clear advantages for the use of CMCs are high-temperature operation, which would allow a high-efficiency Rankine cycle, and low activation. Limitations to their use are material costs, fabrication complexity and costs, lack of familiarity with these materials in design, and the lack of data on radiation stability at relevant temperatures and fluences. Fusion-relevant feasibility issues identified at this workshop include: hermetic and vacuum properties related to effects of matrix porosity and matrix microcracking; chemical compatibility with coolant, tritium, and breeder and multiplier materials, radiation effects on compatibility; radiation stability and integrity; and ability to join CMCs in the shop and at the reactor site, radiation stability and integrity of joints. A summary of ongoing CMC radiation programs is also given. It was suggested that a true feasibility assessment of CMCs for fusion structural applications could not be completed without evaluation of a material tailored'' to fusion conditions or at least to radiation stability. It was suggested that a follow-up workshop be held to design a tailored composite after the results of CMC radiation studies are available and the critical feasibility issues are addressed.

  10. Proton Collimators for Fusion Reactors

    NASA Technical Reports Server (NTRS)

    Miley, George H.; Momota, Hiromu

    2003-01-01

    Proton collimators have been proposed for incorporation into inertial-electrostatic-confinement (IEC) fusion reactors. Such reactors have been envisioned as thrusters and sources of electric power for spacecraft and as sources of energetic protons in commercial ion-beam applications.

  11. 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.

  12. Volatility from copper and tungsten alloys for fusion reactor applications

    SciTech Connect

    Smolik, G.R.; Neilson, R.M. Jr.; Piet, S.J. )

    1989-01-01

    Accident scenarios for fusion power plants present the potential for release and transport of activated constituents volatilized from first wall and structural materials. The extent of possible mobilization and transport of these activated species, many of which are oxidation driven'', is being addressed by the Fusion Safety Program at the Idaho National Engineering Laboratory (INEL). This report presents experimental measurements of volatilization from a copper alloy in air and steam and from a tungsten alloy in air. The major elements released included zinc from the copper alloy and rhenium and tungsten from the tungsten alloy. Volatilization rates of several constituents of these alloys over temperatures ranging from 400 to 1200{degree}C are presented. These values represent release rates recommended for use in accident assessment calculations. 8 refs., 3 figs., 5 tabs.

  13. Ion cyclotron and lower hybrid arrays applicable to current drive in fusion reactors

    NASA Astrophysics Data System (ADS)

    Bosia, G.; Helou, W.; Goniche, M.; Hillaret, J.; Ragona, R.

    2014-02-01

    This paper presents concepts for Ion Cyclotron and Lower Hybrid Current Drive arrays applicable to fusion reactors and based on periodically loaded line power division. It is shown that, in large arrays, such as the ones proposed for fusion reactor applications, these schemes can offer, in principle, a number of practical advantages, compared with currently adopted ones, such as in-blanket operation at significantly reduced power density, lay out suitable for water cooling, single ended or balanced power feed, simple and load independent impedance matching In addition, a remote and accurate real time measurement of the complex impedance of all array elements as well as detection, location, and measurement of the complex admittance of a single arc occurring anywhere in the structure is possible.

  14. Properties of V-4Cr-4Ti for application as fusion reactor structural components

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1994-08-01

    Vanadium-base alloys are promising candidate materials for application in fusion reactor first-wall and blanket structures because they offer several important advantages, i.e., inherently low irradiation-induced activity, good mechanical properties, good compatibility with lithium, high thermal conductivity, and good resistance to irradiation-induced swelling and damage. As part of a program to screen candidate alloys and develop an optimized vanadium-base alloy, extensive investigations of various V-Ti, V-Cr-Ti, and V-Ti-Si alloys have been conducted after irradiation in lithium in fission reactors. From these investigations, V-4 wt.% Cr-4 wt.% Ti was identified as the most promising alloy. The alloy exhibited attractive mechanical and physical properties that are prerequisites for first-wall and blanket structures, i.e., high tensile strength, high ductility, good creep properties, high impact energy, low ductile-brittle transition temperature before and after irradiation, excellent resistance to irradiation-induced swelling and microstructural instability, and good resistance to corrosion in lithium. In particular, the alloy is virtually immune to irradiation-induced embrittlement, a remarkable property compared to other candidate materials being investigated in the fusion-reactor-materials community. Effects of helium, charged dynamically in simulation of realistic fusion reactor conditions, on tensile, ductile-brittle transition, and swelling properties were insignificant.

  15. Colliding Beam Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  16. Fusion reactor materials

    SciTech Connect

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  17. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  18. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  19. 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.

  20. Fusion reactor pumped laser

    DOEpatents

    Jassby, Daniel L.

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  1. Fusion reactor pumped laser

    DOEpatents

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  2. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    NASA Astrophysics Data System (ADS)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  3. Physical and mechanical characteristics and chemical compatibility of aluminum nitride insulator coatings for fusion reactor applications

    SciTech Connect

    Natesan, K.; Rink, D.L.

    1996-04-01

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of various concepts, including liquid metal, molten salt, water, and helium. Based on the requirements for an electrically insulating coating on the first-wall structural material to minimize the MHD pressure drop during the flow of liquid metal in a magnetic field, AlN was selected as a candidate coating material for the Li self-cooled blanket concept. This report discusses the results from an ongoing study of physical and mechanical characteristics and chemical compatibility of AlN electrical insulator coatings in a liquid Li environment. Details are presented on the AlN coating fabrication methods, and experimental data are reported for microstructures, chemistry of coatings, pretreatment of substrate, heat treatment of coatings, hardness data for coatings, coating/lithium interactions, and electrical resistance before and after exposure to lithium. Thermodynamic calculations are presented to establish regions of stability for AlN coatings in an Li environment as a function of O concentration and temperature, which can aid in-situ development of AlN coatings in Li.

  4. Radiation hardening of diagnostics for fusion reactors

    SciTech Connect

    Baur, J.F.; Engholm, B.A.; Hacker, M.P.; Maya, I.; Miller, P.H.; Toffolo, W.E.; Wojtowicz, S.S.

    1981-12-01

    A list of the diagnostic systems presently used in magnetic confinement fusion experiments is compiled herein. The radiation-sensitive components are identified, and their locations in zones around the machine are indicated. A table of radiation sensitivities of components is included to indicate the data available from previous work in fission reactor, space probe, and defense-related programs. Extrapolation and application to hardening of fusion diagnostic systems requires additional data that are more specific to the fusion radiation environment and fusion components. A list is also given of present radiation-producing facilities where near-term screening tests of materials and components can be performed.

  5. Optimization of the chemical composition and manufacturing route for ODS RAF steels for fusion reactor application

    NASA Astrophysics Data System (ADS)

    Oksiuta, Z.; Baluc, N.

    2009-05-01

    As the upper temperature for use of reduced activation ferritic/martensitic steels is presently limited by a drop in mechanical strength at about 550 °C, Europe, Japan and the US are actively researching steels with high strength at higher operating temperatures, mainly using stable oxide dispersion. In addition, the numerous interfaces between matrix and oxide particles are expected to act as sinks for the irradiation-induced defects. The main R&D activities aim at finding a compromise between good tensile and creep strength and sufficient ductility, especially in terms of fracture toughness. Oxide dispersion strengthened (ODS) reduced activation ferritic (RAF) steels appear as promising materials for application in fusion power reactors up to about 750 °C. Six different ODS RAF steels, with compositions of Fe-(12-14)Cr-2W-(0.1-0.3-0.5)Ti-0.3Y2O3 (in wt%), were produced by powder metallurgy techniques, including mechanical alloying, canning and degassing of the milled powders and compaction of the powders by hot isostatic pressing, using various devices and conditions. The materials have been characterized in terms of microstructure and mechanical properties. The results have been analysed in terms of optimal chemical composition and manufacturing conditions. In particular, it was found that the composition of the materials should lie in the range Fe-14Cr-2W-(0.3-0.4)Ti-(0.25-0.3)Y2O3, as 14Cr ODS RAF steels exhibit higher tensile strength and better Charpy impact properties and are more stable than 12Cr materials (no risk of martensitic transformation), while materials with 0.5% Ti or more should not be further investigated, due to potential embrittlement by large TiO2 particles.

  6. Laser surface inspections: fundamentals and applications to monitor inner surface conditions of nuclear fusion reactor chambers

    NASA Astrophysics Data System (ADS)

    Kasuya, Koichi; Ozawa, S.; Norimatsu, T.; Azechi, H.; Mima, K.; Nakai, S.; Suzuki, S.; Budner, B.; Mroz, W.; Kasuya, N.; Kasuya, W.; Kasuya, Kei.; Izawa, Y.; Furukawa, H.; Shimada, Y.; Yamanaka, T.; Nakai, M.; Nagai, K.; Yokoyama, K.; Ezato, K.; Enoeda, M.; Akiba, M.; Prokopiuk, A.

    2010-09-01

    The most recent fundamental research results to investigate surface erosions of nuclear fusion candidate chamber materials are described in short. We used a commercial surface profiler with a red semiconductor laser. Various material surfaces ablated and eroded by a rather short pulse electron beam and a short pulse ArF laser light were measured with this surface profiler and the associated three-dimensional analysis software. Threshold input levels for various sample surface erosions with electron and laser beams were clearly decided for the first time with our new method in this article. After the above fundamental results were gathered, the methods to inspect inner surface conditions of nuclear fusion reactor chambers were newly proposed with various kinds of laser displacement sensors. The first one is the erosion monitor with the above profiler, and the second one is the laser induced ultrasonic wave detection method to inspect deeper surface layers than the first one.

  7. 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.

  8. 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.

  9. Outlook for the fusion hybrid and tritium-breeding fusion reactors

    NASA Astrophysics Data System (ADS)

    Richardson, J. M.; Cohen, R.; Simpson, J. W.

    The study examines the outlook for fusion hybrid reactors. The study evaluates the status of fusion hybrid technology in the United States and analyzes the circumstances under which such reactors might be deployed. The study also examines a related concept, the tritium-breeding fusion reactor. The study examined two potential applications for fusion hybrid technology: (1) the production of fissile material to fuel light-water reactors, and (2) the direct production of baseload electricity. For both applications, markets were sufficiently problematical or remote (mid-century or later) to warrant only modest current research and development emphasis on technology specific to the fusion hybrid reactor. For the tritium-breeding fusion reactor, a need for tritium for use in nuclear weapons might arise well before the middle of the next century, so that a program of design studies, experimentation, and evaluation should be undertaken.

  10. The characterization of copper alloys for the application of fusion reactors

    SciTech Connect

    Ishiyama, S.; Fukaya, K.; Eto, M.; Akiba, M.

    1995-12-31

    Three kinds of candidate copper alloys for divertor structural materials of fusion experimental reactors, that is, Oxygen Free High thermal conductivity Copper (OFHC), alumina disperse reinforced copper (DSC) and the composite of W and Cu (W/Cu), were prepared for strength and fatigue tests at temperatures ranging from R.T. to 500 C in a vacuum. High temperature strength of DSC and W/Cu with rapid fracture after peak loading at the temperatures is higher than that of OFHC by factor of 2, but fracture strains of DFC and W/Cu are smaller than that of OFHC. Fatigue life of DSC, which shows the same fatigue behavior of OFHC at room temperature, is longer than other materials at 400 C. Remarkable fatigue life reduction of OFHC found in this experiment is to be due to recrystallization of OFHC yielded above 400 C.

  11. Fatigue behavior of irradiated helium-containing ferritic steels for fusion reactor applications*1

    NASA Astrophysics Data System (ADS)

    Grossbeck, M. L.; Vitek, J. M.; Liu, K. C.

    1986-11-01

    The martensitic alloys 12Cr-1MoVW and 9Cr-1MoVNb have been irradiated in the High Flux Isotope Reactor (HFIR) and subsequently tested in fatigue. In order to achieve helium levels characteristic of fusion reactors, the 12Cr-1MoVW was doped with 1 and 2% Ni, resulting in helium levels of 210 and 410 appm at damage levels of 25 dpa. The 9Cr-1MoVNb was irradiated to a damage level of 3 dpa and contained less than 5 appm He. Irradiations were carried out at 55°C and testing at 22°C. No significant changes were found in 9Cr-1MoVNb upon irradiation at this damage level, but effects that could possibly be attributed to helium were found in 12Cr-1MoVW. Levels of 210 and 410 appm He produced cyclic strengthening of 29 and 34% over unirradiated nickel-doped materials, respectively. This cyclic hardening, attributable largely to helium, resulted in degradation of the cyclic life. However, the fatigue life remained comparable to or better than unirradiated 20%-cold-worked 316 stainless steel.

  12. Prospects for toroidal fusion reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.D.

    1994-06-01

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

  13. An evaluation of potential material coolant compatibility for applications in advanced fusion reactors

    NASA Astrophysics Data System (ADS)

    Kondo, T.; Watanabe, Y.; Yi, Y. S.; Hishinuma, A.

    1998-10-01

    In assessing possible potential issues for fusion applications, the compatibility of several metallic structural materials was examined using high temperature/pressure steam as test environment. High corrosion resistance associated with protective oxide film formation was regarded as essential for the function of protecting from tritium permeation and corrosion damage. A Ti-Al-based intermetallic compound with V addition, recently developed, showed excellent performance. A low-activation ferritic/martensitic steel, F82-H, was comparable with the current advanced materials for modern supercritical fossil boilers, while some potential vanadium alloys, although not intended for use in steam, were found less compatible.

  14. (Meeting on fusion reactor materials)

    SciTech Connect

    Jones, R.H. ); Klueh, R.L.; Rowcliffe, A.F.; Wiffen, F.W. ); Loomis, B.A. )

    1990-11-01

    During his visit to the KfK, Karlsruhe, F. W. Wiffen attended the IEA 12th Working Group Meeting on Fusion Reactor Materials. Plans were made for a low-activation materials workshop at Culham, UK, for April 1991, a data base workshop in Europe for June 1991, and a molecular dynamics workshop in the United States in 1991. At the 11th IEA Executive Committee on Fusion Materials, discussions centered on the recent FPAC and Colombo panel review in the United States and EC, respectively. The Committee also reviewed recent progress toward a neutron source in the United States (CWDD) and in Japan (ESNIT). A meeting with D. R. Harries (consultant to J. Darvas) yielded a useful overview of the EC technology program for fusion. Of particular interest to the US program is a strong effort on a conventional ferritic/martensitic steel for fist wall/blanket operation beyond NET/ITER.

  15. Advances in Tandem Mirror fusion power reactors

    SciTech Connect

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  16. Application of carbon-aluminum nanostructures in divertor coatings from fusion reactor

    NASA Astrophysics Data System (ADS)

    Ciupina, V.; Lungu, C. P.; Vladoiu, R.; Epure, T. D.; Prodan, G.; Porosnicu, C.; Prodan, M.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Zarovschi, V.

    2012-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Carbon-Aluminum composites are the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-aluminum nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. Moreover, the energy of ions can be controlled. Thermo-electrons emitted by an externally heated cathode and focused by a Wehnelt focusing cylinder are strongly accelerated towards the anode whose material is evaporated and bright plasma is ignited by a high voltage DC supply. The nanostructured C-Al films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM). Tribological properties in dry sliding were evaluated using a CSM ball-on-disc tribometer. The carbon - aluminum films were identified as a nanocrystals complex (from 2nm to 50 nm diameters) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The friction coefficients (0.1 - 0.2, 0.5) of the C-Al coatings was decreased more than 2-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-Al nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures

  17. Vanadium recycling for fusion reactors

    SciTech Connect

    Dolan, T.J.; Butterworth, G.J.

    1994-04-01

    Very stringent purity specifications must be applied to low activation vanadium alloys, in order to meet recycling goals requiring low residual dose rates after 50--100 years. Methods of vanadium production and purification which might meet these limits are described. Following a suitable cooling period after their use, the vanadium alloy components can be melted in a controlled atmosphere to remove volatile radioisotopes. The aim of the melting and decontamination process will be the achievement of dose rates low enough for ``hands-on`` refabrication of new reactor components from the reclaimed metal. The processes required to permit hands-on recycling appear to be technically feasible, and demonstration experiments are recommended. Background information relevant to the use of vanadium alloys in fusion reactors, including health hazards, resources, and economics, is provided.

  18. Modular Stellarator Fusion Reactor concept

    SciTech Connect

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

    1981-08-01

    A preliminary conceptual study is made of the Modular Stellarator Reactor (MSR). A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. The physics basis of the design point is described together with supporting magnetics, coil-force, and stress computations. The approach and results presented herein will be modified in the course of ongoing work to form a firmer basis for a detailed conceptual design of the MSR.

  19. Nonlinear control in fusion reactors

    NASA Astrophysics Data System (ADS)

    Schuster, Eugenio

    There is consensus in the fusion reactor community that active control will be one of the key enabling technologies. With further advancements in reduced-order fusion modeling, advances in control systems for fusion will continue, including vertical and shape control, kinetic and current profile control, MHD (magnetohydrodynamic) stabilization and plasma transport reduction. This dissertation addresses different control problems in tokamaks using as common denominator a nonlinear control approach. Contributions are made in the areas of kinetic control, magnetic control, and MHD flow control. In the area of kinetic control, we approach the problem of nonlinear control of burn instability in fission reactors, where a lumped-parameter nonlinear model involving approximate conservation equations for the energy and the densities of the species is used to synthesize a nonlinear feedback controller (backstepping, feedback linearization, passivity and input to state stability) for stabilizing the thermally unstable burn condition of a fusion reactor. In addition, the problem of control of kinetic profiles in non-burning plasmas, where a set of nonlinear partial differential equations (PDE's) describing approximately the dynamics of the density and energy was considered as the plant model used to synthesize a boundary controller (infinite-dimensional nonlinear backstepping) whose goal was the control of the density and energy spatial distributions, is also considered. In the area of magnetic control, the problem of plasma vertical position stabilization and shape control under actuation saturation in the DIII-D Tokamak at General Atomics is approached. In this case, modifications of the nominal control loops (nonlinear anti-windup augmentation) are proposed to ensure stability of the plant and good behavior of the nominal controller under the presence of voltage saturation in the coils that are used to vertically position and shape the plasma inside the tokamak. In the area

  20. Materials issues in fusion reactors

    NASA Astrophysics Data System (ADS)

    Suri, A. K.; Krishnamurthy, N.; Batra, I. S.

    2010-02-01

    The world scientific community is presently engaged in one of the toughest technological tasks of the current century, namely, exploitation of nuclear fusion in a controlled manner for the benefit of mankind. Scientific feasibility of controlled fusion of the light elements in plasma under magnetic confinement has already been proven. International efforts in a coordinated and co-operative manner are presently being made to build ITER - the International Thermonuclear Experimental Reactor - to test, in this first step, the concept of 'Tokamak' for net fusion energy production. To exploit this new developing option of making energy available through the route of fusion, India too embarked on a robust fusion programme under which we now have a working tokamak - the Aditya and a steady state tokamak (SST-1), which is on the verge of functioning. The programme envisages further development in terms of making SST-2 followed by a DEMO and finally the fusion power reactor. Further, with the participation of India in the ITER program in 2005, and recent allocation of half - a - port in ITER for placing our Lead - Lithium Ceramic Breeder (LLCB) based Test Blanket Module (TBM), meant basically for breeding tritium and extracting high grade heat, the need to understand and address issues related to materials for these complex systems has become all the more necessary. Also, it is obvious that with increasing power from the SST stages to DEMO and further to PROTOTYPE, the increasing demands on performance of materials would necessitate discovery and development of new materials. Because of the 14.1 MeV neutrons that are generated in the D+T reaction exploited in a tokamak, the materials, especially those employed for the construction of the first wall, the diverter and the blanket segments, suffer crippling damage due to the high He/dpa ratios that result due to the high energy of the neutrons. To meet this challenge, the materials that need to be developed for the tokamaks

  1. Advanced Real-Time Feedback Control in JT-60U High Performance Discharges for Application to Fusion Reactor Plasmas

    SciTech Connect

    Fukuda, T.; Oikawa, T.; Takeji, S.; Isayama, A.; Kawano, Y.; Neyatani, Y.; Nagashima, A.; Nishitani, T.; Konoshima, S.; Tamai, H.; Fujita, T.; Sakamoto, Y.; Kamada, Y.; Ide, S.; Koide, Y.; Takenaga, H.; Kurihara, K.; Sakata, S.; Ozeki, T.; Kawamata, Y.; Miura, Y. M.

    2002-09-15

    The significance of real-time feedback control is emphasized in this paper as an indispensable method to improve and sustain the improved plasma characteristics in JT-60U high fusion performance discharges as well as to operate the fusion reactor under the optimal divertor conditions with respect to the heat load and exhaust pumping. In accordance, substantial improvement in the equivalent fusion amplification gain of over unity has been reproducibly achieved at the JT-60U tokamak in the reversed shear mode of operation with the robust feedback controls, where the value of target density was deliberately optimized for the reliable internal transport barrier formation, and the magneto-hydrodynamic stability control was performed with the stored energy feedback. The feedback control techniques also demonstrated the effectiveness to produce quasi-steady-state high-performance plasmas. In addition, three major parameters associated with the fusion reactor instrumentations, namely the neutron production rate, operating density, and divertor radiation power, were simultaneously feedback controlled in the ELMy H-mode plasmas. Here, the matrix response function was evaluated to identify the limitations involved with the linear combination of independent controls. Other advanced feedback schemes, such as the feedback suppression of the neoclassical tearing mode required to sustain high plasma pressure in a steady-state, are also described. Finally, the controversial issues for the future intelligent plasma control necessary for the advanced steady-stated tokamak reactor are addressed.

  2. Laser-driven fusion reactor

    DOEpatents

    Hedstrom, J.C.

    1973-10-01

    A laser-driven fusion reactor consisting of concentric spherical vessels in which the thermonuclear energy is derived from a deuterium-tritium (D + T) burn within a pellet'', located at the center of the vessels and initiated by a laser pulse. The resulting alpha -particle energy and a small fraction of the neutron energy are deposited within the pellet; this pellet energy is eventually transformed into sensible heat of lithium in a condenser outside the vessels. The remaining neutron energy is dissipated in a lithium blanket, located within the concentric vessels, where the fuel ingredient, tritium, is also produced. The heat content of the blanket and of the condenser lithium is eventually transferred to a conventional thermodynamic plant where the thermal energy is converted to electrical energy in a steam Rankine cycle. (Official Gazette)

  3. Application of railgun principle to high-velocity hydrogen pellet injection for magnetic fusion reactor refueling

    SciTech Connect

    Kim, K.; Honig, J.

    1984-09-01

    Design, construction, testing, and performance evaluation of a small-bore plasma-arc-driven electromagnetic railgun system are described. The railgun system, which is intended for injecting high-velocity hydrogen pellets into the magnetic fusion devices for the purpose of refueling, has two acceleration stages. One consists of a gas gun preaccelerator and the other a railgun booster accelerator. The plasma-arc armature is formed behind the pellet by electrically discharging the propellant gas following the pellet into the railgun from the gas gun.

  4. Application of railgun principle to high-velocity hydrogen pellet injection for magnetic fusion reactor refueling: Technical progress report

    SciTech Connect

    Kim, Kyekyoon

    1987-12-01

    This paper discusses the use of a railgun accelerator to inject hydrogen pellets into a magnetic fusion reactor for refueling purposes. Specific studies in this paper include: 1.5 mm-diameter two-stage fuseless plasma-arc-driven electromagnetic railgun, construction and testing of a 3.2 mm-diameter two-stage railgun and a theoretical analysis of the behavior of a railgun plasma-arc armature inside a railgun. (LSP)

  5. Application of railgun principle to high-velocity hydrogen pellet injection for magnetic fusion reactor refueling

    SciTech Connect

    Kim, K.

    1991-08-01

    This report contains three documents describing the progress made by the University of Illinois electromagnetic railgun program sponsored by the Office of Fusion Energy of the United States Department of Energy during the period from July 16, 1990 to August 16, 1991. The first document contains a brief summary of the tasks initiated, continued, or completed, the status of major tasks, and the research effort distribution, estimated and actual, during the period. The second document contains a description of the work performed on time resolved laser interferometric density measurement of the railgun plasma-arc armature. The third document is an account of research on the spectroscopic measurement of the electron density and temperature of the railgun plasma arc.

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

    SciTech Connect

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

    1983-06-01

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

  7. Ion nitriding of titanium alpha plus beta alloy for fusion reactor applications

    NASA Astrophysics Data System (ADS)

    Rolinski, E.; Sharp, G.; Cowgill, D. F.; Peterman, D. J.

    1998-02-01

    The titanium alpha plus beta alloy, Ti-6Al-4V, has been plasma nitrided at 1033 K in pure nitrogen to produce a 9 μm thick surface nitride zone, followed by a 5 μm thick Al-enriched zone and a 100 μm thick nitrogen diffusion zone. X-ray diffraction and Auger electron spectroscopy indicate that the nitride zone consists of two layers: a thin TiN layer at the surface followed by a Ti 2N layer. In preliminary tests, the deuterium retention of nitrided titanium specimens was determined after exposure to fluences of 4×10 25 m -2, 60 eV D + ions at 440 K and of 2×10 25 m -2, 800 eV D + ions at 623 K. The results are compared to the retention found for similar exposure of the untreated alloy and to model simulations. Implications for the use of these materials in a fusion environment are discussed.

  8. Generic Magnetic Fusion Reactor Revisited

    NASA Astrophysics Data System (ADS)

    Sheffield, John; Milora, Stanley

    2015-11-01

    The original Generic Magnetic Fusion Reactor paper was published in 1986. This update describes what has changed in 30 years. Notably, the construction of ITER is providing important benchmark numbers for technologies and costs. In addition, we use a more conservative neutron wall flux and fluence. But these cost-increasing factors are offset by greater optimism on the thermal-electric conversion efficiency and potential availability. The main examples show the cost of electricity (COE) as a function of aspect ratio and neutron flux to the first wall. The dependence of the COE on availability, thermo-electric efficiency, electrical power output, and the present day's low interest rates is also discussed. Interestingly, at fixed aspect ratio there is a shallow minimum in the COE at neutron flux around 2.5 MW/m2. The possibility of operating with only a small COE penalty at even lower wall loadings (to 1.0 MW/m2 at larger plant size) and the use of niobium-titanium coils are also investigated. J. Sheffield was supported by ORNL subcontract 4000088999 with the University of Tennessee.

  9. Trends in fusion reactor safety research

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Holland, D. F.; Piet, S. J.

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex, with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g., for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions.

  10. Laser-fusion targets for reactors

    DOEpatents

    Nuckolls, John H.; Thiessen, Albert R.

    1987-01-01

    A laser target comprising a thermonuclear fuel capsule composed of a centrally located quantity of fuel surrounded by at least one or more layers or shells of material for forming an atmosphere around the capsule by a low energy laser prepulse. The fuel may be formed as a solid core or hollow shell, and, under certain applications, a pusher-layer or shell is located intermediate the fuel and the atmosphere forming material. The fuel is ignited by symmetrical implosion via energy produced by a laser, or other energy sources such as an electron beam machine or ion beam machine, whereby thermonuclear burn of the fuel capsule creates energy for applications such as generation of electricity via a laser fusion reactor.

  11. IPFR: Integrated Pool Fusion Reactor concept

    SciTech Connect

    Sze, D.K.

    1986-01-01

    The IPFR (Integrated Pool Fusion Reactor) concept is to place a fusion reactor into a pool of molten Flibe. The Flibe will serve the multiple functions of breeding, cooling, shielding, and moderating. Therefore, the only structural material between the superconducting magnets and the plasma is the first wall. The first wall is a stand-alone structure with no coolant connection and is cooled by Flibe at the atmospheric pressure. There is also no need of the primary coolant loop. The design is expected to improve the safety, reliability, and maintainability aspects of the fusion system.

  12. Status and problems of fusion reactor development.

    PubMed

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes. PMID:11402837

  13. Open-ended fusion devices and reactors

    SciTech Connect

    Kawabe, T.; Nariai, H.

    1983-12-01

    Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown.

  14. Generic Stellarator-like Magnetic Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Sheffield, John; Spong, Donald

    2015-11-01

    The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

  15. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    NASA Astrophysics Data System (ADS)

    Abe, K.; Kohyama, A.; Namba, C.; Wiffen, F. W.; Jones, R. H.

    1998-10-01

    A Japan-USA Program of irradiation experiments for fusion research, "JUPITER", has been established as a 6 year program from 1995 to 2000. The goal is to study "the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment". This is phase-three of the collaborative program, which follows RTNS-II Program (Phase-1: 1982-1986) and FFTF/MOTA Program (Phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA Program, JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects.

  16. Evaluation of charcoal sorbents for helium cryopumping in fusion reactors

    SciTech Connect

    Tobin, A.G.; Sedgley, D.W.; Batzer, T.H.; Call, W.R.

    1987-01-01

    Improved methods for cryopumping helium were developed for application to fusion reactors where high helium generation rates are expected. In this study, small coconut charcoal granules were utilized as the sorbent, and braze alloys and low temperature curing cements were used as the bonding agents for attachment to a copper support structure. Problems of scale-up of the bonding agent to a 40 cm diam panel were also investigated. Our results indicate that acceptable helium pumping performance of braze bonded and cement bonded charcoals can be achieved over the range of operating conditions expected in fusion reactors.

  17. Computational mathematics and physics of fusion reactors.

    PubMed

    Garabedian, Paul R

    2003-11-25

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors. PMID:14614129

  18. Computational mathematics and physics of fusion reactors

    PubMed Central

    Garabedian, Paul R.

    2003-01-01

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors. PMID:14614129

  19. Ceramics for fusion applications

    SciTech Connect

    Clinard, F.W. Jr.

    1986-01-01

    Ceramics are required for a variety of uses in both near-term fusion devices and in commercial powerplants. These materials must retain adequate structural and electrical properties under conditions of neutron, particle, and ionizing irradiation; thermal and applied stresses; and physical and chemical sputtering. Ceramics such as Al/sub 2/O/sub 3/, MgAl/sub 2/O/sub 4/, BeO, Si/sub 3/N/sub 4/ and SiC are currently under study for fusion applications, and results to date show widely-varying response to the fusion environment. Materials can be identified today which will meet initial operating requirements, but improvements in physical properties are needed to achieve satisfactory lifetimes for critical applications.

  20. Packed fluidized bed blanket for fusion reactor

    DOEpatents

    Chi, John W. H.

    1984-01-01

    A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

  1. Waste management for JAERI fusion reactors

    NASA Astrophysics Data System (ADS)

    Tobita, K.; Nishio, S.; Konishi, S.; Jitsukawa, S.

    2004-08-01

    In the fusion reactor design study at Japan Atomic Energy Institute (JAERI), several waste management strategies were assessed. The assessed strategies are: (1) reinforced neutron shield to clear the massive ex-shielding components from regulatory control; (2) low aspect ratio tokamak to reduce the total waste; (3) reuse of liquid metal breeding material and neutron shield. Combining these strategies, the weight of disposal waste from a low aspect ratio reactor VECTOR is expected to be comparable with the metal radwaste from a light water reactor (˜4000 t).

  2. Fuel provision for nonbreeding deuterium-tritium fusion reactors

    SciTech Connect

    Jassby, D.L.; Katsurai, M.

    1980-01-01

    Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

  3. Evolution towards Economically Viable Magnetic Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Furth, H. P.

    1996-11-01

    Large pedestrian dinosaurs have long been extinct, while flying dinosaurs have evolved from the archaeopteryx to the common sparrow. Removal of superfluous constraints was the key. In order for soi-disant intelligent life to have emerged on Earth, fusion-power emission from our Sun must have been kept sufficiently feeble and slow-changing (c.f., Bethe's Carbon-Cycle) so as to allow time for non-trivial evolution. By contrast, any economically viable fusion-reactor scheme must use some fast-burning fuel (e.g. D-D,D-T,etc.), so as to elude the economic constraints of excessive single-unit size and cost. The quest for livelier fusion fuel tends to motivate various departures from a strictly thermalized ``Maxwellian'' reactor-plasma distribution. Illustrative material will include specific options for applying the joint resources of the international ``Three-Large-Tokamak Collaboration''.

  4. Neutronic analysis of a fusion hybrid reactor

    SciTech Connect

    Kammash, T.

    2012-07-01

    In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

  5. Beryllium for fusion application - recent results

    NASA Astrophysics Data System (ADS)

    Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

    2002-12-01

    The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

  6. The nano-particle dispersion strengthening of V-4Cr-4Ti alloys for high temperature application in fusion reactors

    NASA Astrophysics Data System (ADS)

    Zheng, Pengfei; Chen, Jiming; Xu, Zengyu; Duan, Xuru

    2013-10-01

    V-4Cr-4Ti was identified as an attractive structural material for Li blanket in fusion reactors. However, both high temperature and irradiation induced degradation are great challenges for this material. It was thought that the nano-particles with high thermal stability can efficiently strengthen the alloy at elevated temperatures, and accommodate the irradiation induced defects at the boundaries. This study is a starting work aiming at improving the creep resistance and reducing the irradiation induced degradation for V-4Cr-4Ti alloy. Currently, we focus on the preparation of some comparative nano-particle dispersion strengthened V-4Cr-4Ti alloys. A mechanical alloying (MA) route is used to fabricate yttrium and carbides added V-4Cr-4Ti alloys. Nano-scale yttria, carbides and other possible particles have a combined dispersion-strengthening effect on the matrices of these MA-fabricated V-4Cr-4Ti alloys. High-temperature annealing is carried out to stabilize the optimized nano-particles. Mechanical properties are tested. Microstructures of the MA-fabricated V-4Cr-4Ti alloys with yttrium and carbide additions are characterized. Based on these results, the thermal stability of different nano-particle agents are classified. ITER related China domestic project 2011GB108007.

  7. First wall for polarized fusion reactors

    DOEpatents

    Greenside, H.S.; Budny, R.V.; Post, D.E. Jr.

    1985-01-29

    A first-wall or first-wall coating for use in a fusion reactor having polarized fuel may be formed of a low-Z non-metallic material having slow spin relaxation, i.e., a depolarization rate greater than 1 sec/sup -1/. Materials having these properties include hydrogenated and deuterated amorphous semiconductors. A method for preventing the rapid depolarization of a polarized plasma in a fusion device may comprise the step of providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec/sup -1/.

  8. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    SciTech Connect

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  9. Superconducting magnets for toroidal fusion reactors

    SciTech Connect

    Haubenreich, P.N.

    1980-01-01

    Fusion reactors will soon be employing superconducting magnets to confine plasma in which deuterium and tritium (D-T) are fused to produce usable energy. At present there is one small confinement experiment with superconducting toroidal field (TF) coils: Tokamak 7 (T-7), in the USSR, which operates at 4 T. By 1983, six different 2.5 x 3.5-m D-shaped coils from six manufacturers in four countries will be assembled in a toroidal array in the Large Coil Test Facility (LCTF) at Oak Ridge National Laboratory (ORNL) for testing at fields up to 8 T. Soon afterwards ELMO Bumpy Torus (EBT-P) will begin operation at Oak Ridge with superconducting TF coils. At the same time there will be tokamaks with superconducting TF coils 2 to 3 m in diameter in the USSR and France. Toroidal field strength in these machines will range from 6 to 9 T. NbTi and Nb/sub 3/Sn, bath cooling and forced flow, cryostable and metastable - various designs are being tried in this period when this new application of superconductivity is growing and maturing.

  10. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

    Cheung, A.; Binderbauer, M.; Liu, F.; Qerushi, A.; Rostoker, N.; Wessel, F. J.

    2004-02-01

    The Colliding Beam Fusion Reactor Space Propulsion System, CBFR-SPS, is an aneutronic, magnetic-field-reversed configuration, fueled by an energetic-ion mixture of hydrogen and boron11 (H-B11). Particle confinement and transport in the CBFR-SPS are classical, hence the system is scaleable. Fusion products are helium ions, α-particles, expelled axially out of the system. α-particles flowing in one direction are decelerated and their energy recovered to ``power'' the system; particles expelled in the opposite direction provide thrust. Since the fusion products are charged particles, the system does not require the use of a massive-radiation shield. This paper describes a 100 MW CBFR-SPS design, including estimates for the propulsion-system parameters and masses. Specific emphasis is placed on the design of a closed-cycle, Brayton-heat engine, consisting of heat-exchangers, turbo-alternator, compressor, and finned radiators.

  11. Choice of coils for a fusion reactor

    PubMed Central

    Alexander, Romeo; Garabedian, Paul R.

    2007-01-01

    In a fusion reactor a hot plasma of deuterium and tritium is confined by a strong magnetic field to produce helium ions and release energetic neutrons. The 3D geometry of a stellarator provides configurations for such a device that reduce net toroidal current that might lead to disruptions. We construct smooth coils generating an external magnetic field designed to prevent the plasma from deteriorating. PMID:17640879

  12. Choice of coils for a fusion reactor.

    PubMed

    Alexander, Romeo; Garabedian, Paul R

    2007-07-24

    In a fusion reactor a hot plasma of deuterium and tritium is confined by a strong magnetic field to produce helium ions and release energetic neutrons. The 3D geometry of a stellarator provides configurations for such a device that reduce net toroidal current that might lead to disruptions. We construct smooth coils generating an external magnetic field designed to prevent the plasma from deteriorating. PMID:17640879

  13. First wall for polarized fusion reactors

    DOEpatents

    Greenside, Henry S.; Budny, Robert V.; Post, Jr., Douglass E.

    1988-01-01

    Depolarization mechanisms arising from the recycling of the polarized fuel at the limiter and the first-wall of a fusion reactor are greater than those mechanisms in the plasma. Rapid depolarization of the plasma is prevented by providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec.sup.-1.

  14. 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.).

  15. 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.

  16. Superconducting magnets for fusion applications

    SciTech Connect

    Henning, C.D.

    1987-07-02

    Fusion magnet technology has made spectacular advances in the past decade; to wit, the Mirror Fusion Test Facility and the Large Coil Project. However, further advances are still required for advanced economical fusion reactors. Higher fields to 14 T and radiation-hardened superconductors and insulators will be necessary. Coupled with high rates of nuclear heating and pulsed losses, the next-generation magnets will need still higher current density, better stability and quench protection. Cable-in-conduit conductors coupled with polyimide insulations and better steels seem to be the appropriate path. Neutron fluences up to 10/sup 19/ neutrons/cm/sup 2/ in niobium tin are achievable. In the future, other amorphous superconductors could raise these limits further to extend reactor life or decrease the neutron shielding and corresponding reactor size.

  17. Diamond neutral particle spectrometer for fusion reactor ITER

    NASA Astrophysics Data System (ADS)

    Krasilnikov, V.; Amosov, V.; Kaschuck, Yu.; Skopintsev, D.

    2014-08-01

    A compact diamond neutral particle spectrometer with digital signal processing has been developed for fast charge-exchange atoms and neutrons measurements at ITER fusion reactor conditions. This spectrometer will play supplementary role for Neutral Particle Analyzer providing 10 ms time and 30 keV energy resolutions for fast particle spectra in non-tritium ITER phase. These data will also be implemented for independent studies of fast ions distribution function evolution in various plasma scenarios with the formation of a single fraction of high-energy ions. In tritium ITER phase the DNPS will measure 14 MeV neutrons spectra. The spectrometer with digital signal processing can operate at peak counting rates reaching a value of 106 cps. Diamond neutral particle spectrometer is applicable to future fusion reactors due to its high radiation hardness, fast response and high energy resolution.

  18. Fusion reactor high vacuum pumping: Charcoal cryosorber tritium exposure results

    SciTech Connect

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M. )

    1991-01-01

    Recent experiments, have shown the practically of using activated charcoal (coconut charcoal) at 4{degrees}K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were shown to be satisfactory. The long term effects of tritium on the charcoal/cement system developed by Grumman and LLNL were not known and a program was undertaken to see what, if any, effect long term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77{degrees}K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately half way through and after the exposure. Modest effects were noted which would not seriously restrict charcoal's use as a cryosorber for fusion reactor high vacuum pumping applications. 4 refs., 8 figs.

  19. Diamond neutral particle spectrometer for fusion reactor ITER

    SciTech Connect

    Krasilnikov, V.; Amosov, V.; Kaschuck, Yu.; Skopintsev, D.

    2014-08-21

    A compact diamond neutral particle spectrometer with digital signal processing has been developed for fast charge-exchange atoms and neutrons measurements at ITER fusion reactor conditions. This spectrometer will play supplementary role for Neutral Particle Analyzer providing 10 ms time and 30 keV energy resolutions for fast particle spectra in non-tritium ITER phase. These data will also be implemented for independent studies of fast ions distribution function evolution in various plasma scenarios with the formation of a single fraction of high-energy ions. In tritium ITER phase the DNPS will measure 14 MeV neutrons spectra. The spectrometer with digital signal processing can operate at peak counting rates reaching a value of 10{sup 6} cps. Diamond neutral particle spectrometer is applicable to future fusion reactors due to its high radiation hardness, fast response and high energy resolution.

  20. Graphene's Viability for Fusion Applications

    NASA Astrophysics Data System (ADS)

    Navarro, Marcos; Hall, Karla; Rojas, Richard; Santarius, John; Kulcinski, Gerald

    2015-11-01

    Graphene is a source of interest for multiple applications due to its unusual electronic and physical properties. As a coating material, it has reduced oxidation of the main substrate, though no effort has been reported of testing it under fusion conditions. A number of experimental studies have established that defect-free graphene is an excellent barrier material for gases. We explore its viability to maintain a significant pressure difference under ion irradiation. Deuterium is used as a projectile on graphene coated silicon over a range of 10-50 keV energies and various fluences. The vacancy yield (amount of damage) and natural resonance for graphene are found at around 1350 cm-1 and 1550 cm-1, respectively. Damage of each sample is quantified via Raman spectroscopy (RS) using the ratio of the intensities at these wavenumbers. Graphene is also tested here as a coating for some fusion components. Though tungsten is a very promising divertor and first wall candidate, after intense irradiation, it is prone to developing fuzz or grass structures, leading to a diminished lifetime. Graphene grown on tungsten is tested under reactor conditions with 30 keV He ions at several fluences, and the sputtering of both materials is studied via RS and Scanning Electron Microscopy. This work was supported by the Graduate Engineering Research Scholars and the TEAM-Science program at the University of Wisconsin-Madison.

  1. A program toward a fusion reactor

    NASA Astrophysics Data System (ADS)

    Rebut, P.-H.; Watkins, M. L.; Gambier, D. J.; Boucher, D.

    1991-08-01

    Near breakeven conditions have been attained in the JET tokamak [Fusion Technol. 11, 13 (1987)], with beryllium as the first-wall material. A fusion triple product (nDτETi) of 8-9×1020 m-3 sec keV has been reached (within a factor of 8 of that required in a fusion reactor). However, this has only been achieved transiently. At high heating powers, an influx of impurities still limits the achievement of better performance and steady-state operation. In parallel, an improved quantitative understanding of fusion plasmas has emerged from the development of a particular plasma model. Good quantitative agreement is obtained between the model and JET data. The main predictions are also consistent with statistical scaling laws. With such a model, a predictive capability begins to emerge to define the parameters and operating conditions of a DEMO, including impurity levels. Present experimental results and model predictions suggest that impurity dilution is a major threat to a reactor. A divertor concept must be developed further to ensure impurity control before a DEMO can be constructed. A New Phase for JET is planned in which an axisymmetric pumped divertor configuration will be used to address the problems of impurity control, plasma fueling, and helium ash exhaust. It should demonstrate a concept of impurity control and the operational domain for such a device. A single Next Step facility (ITER) is a high risk strategy in terms of physics, technology, and management, since it does not provide a sufficiently sound foundation for a DEMO. A Next Step program is proposed, which could comprise several complementary facilities, each optimized with respect to specific clear objectives. In a minimum program, there could be two Next Step tokamaks, and a Materials Test Facility. Such a program would allow division of effort and sharing of risk across the various scientific and technical problems, permit cross comparison, and ensure continuity of results. It could even be

  2. 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.

  3. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

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

    2003-10-28

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

  4. Activation product transport in fusion reactors. [RAPTOR

    SciTech Connect

    Klein, A.C.

    1983-01-01

    Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. Similar problems are experienced around fission reactor systems. The determination of the transport of radioactive corrosion and neutron sputtering products through the system is achieved using the computer code RAPTOR. This code calculates the mass transfer of a number of activation products based on the corrosion and sputtering rates through the system, the deposition and release characteristics of various plant components, the neturon flux spectrum, as well as other plant parameters. RAPTOR assembles a system of first order linear differential equations into a matrix equation based upon the reactor system parameters. Included in the transfer matrix are the deposition and erosion coefficients, and the decay and activation data for the various plant nodes and radioactive isotopes. A source vector supplies the corrosion and neutron sputtering source rates. This matrix equation is then solved using a matrix operator technique to give the specific activity distribution of each radioactive species throughout the plant. Once the amount of mass transfer is determined, the photon transport due to the radioactive corrosion and sputtering product sources can be evaluated, and dose rates around the plant components of interest as a function of time can be determined. This method has been used to estimate the radiation hazards around a number of fusion reactor system designs.

  5. Present status of liquid metal research for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Tabarés, Francisco L.

    2016-01-01

    Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

  6. (Fourth international conference on fusion reactor materials)

    SciTech Connect

    Bloom, E.E.

    1990-01-24

    This report summarizes the International Conference on Fusion Reactor Materials (ICFRM-4) which was held December 4--9, 1989, in Kyoto, Japan, as well as the results of several workshops, planning meetings, and laboratory visits made by the travelers. The ICFRM-4 is the major forum to present and exchange information on materials research and development in support of the world's fusion development efforts. About 360 papers were presented by the 347 conference attendees. Highlights of the conference are presented. A proposal by the United States to host ICFRM-5 was accepted by the International Advisory Committee. ORNL will be the host laboratory. A meeting of the DOE/JAERI Annex I Steering Committee to review the US/Japan Collaborative Testing of First Wall and Blanket Structural Materials with Mixed Spectrum Fission Reactors was held at JAERI Headquarters on December 1. The Japanese emphasized the critical importance of a resumption of HFIR operation. Even though the HFIR outage has lasted three plus years this program has continued to provide new and important data on materials behavior which has particular relevance to ITER.

  7. Design considerations for an inertial confinement fusion reactor power plant

    SciTech Connect

    Massey, J.V.; Simpson, J.E.

    1981-08-10

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept.

  8. Improvement in fusion reactor performance due to ion channeling

    SciTech Connect

    Emmert, G.A.; El-Guebaly, L.A.; Kulcinski, G.L.; Santarius, J.F.; Sviatoslavsky, I.N.; Meade, D.M.

    1994-11-01

    Ion channeling is a recent idea for improving the performance of fusion reactors by increasing the fraction of the fusion power deposited in the ions. In this paper the authors assess the effect of ion channeling on D-T and D-{sup 3}He reactors. The figures of merit used are the fusion power density and the cost of electricity. It is seen that significant ion channeling can lead to about a 50-65% increase in the fusion power density. For the Apollo D-{sup 3}He reactor concept the reduction in the cost of electricity can be as large as 30%.

  9. Multivariable optimization of fusion reactor blankets

    SciTech Connect

    Meier, W.R.

    1984-04-01

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% /sup 6/Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO/sub 2/ breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO/sub 2/ breeding blanket enriched to 34% /sup 6/Li.

  10. Thermomagnetic burn control for magnetic fusion reactor

    DOEpatents

    Rawls, J.M.; Peuron, A.U.

    1980-07-01

    Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma and a toroidal field coil. A mechanism for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

  11. Thermomagnetic burn control for magnetic fusion reactor

    DOEpatents

    Rawls, John M.; Peuron, Unto A.

    1982-01-01

    Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors (30a, 30b, etc.) formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma (12) and a toroidal field coil (18). A mechanism (60) for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

  12. Decommissioning the Tokamak Fusion Test Reactor

    SciTech Connect

    Spampinato, P.T.; Walton, G.R.

    1993-10-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives.

  13. Helium cryopumping for fusion applications

    SciTech Connect

    Sedgley, D.W.; Batzer, T.H.; Call, W.R.

    1988-05-01

    Large quantities of helium and hydrogen isotopes will be exhausted continuously from fusion power reactors. This paper summarizes two development programs undertaken to address vacuum pumping for this application: (i) A continuous duty cryopump for pumping helium and/or hydrogen species using charcoal sorbent and (ii) a cryopump configuration with an alternative shielding arrangement using charcoal sorbent or argon spray. A test program evaluated automatic pumping of helium, helium pumping by charcoal cryosorption and with argon spray, and cryosorption of helium/hydrogen mixtures. The continuous duty cryopump pumped helium continuously and conveniently. Helium pumping speed was 7.7 l/s/cm/sup 2/ of charcoal, compared to 5.8 l/s/cm/sup 2/ for the alternative pump. Helium speed using argon spray was 18% of that obtained by charcoal cryosorption in the same (W-panel) pump. During continuous duty cryopump mixture tests with helium and hydrogen copumped on charcoal, gas was released sporadically. Testing was insufficient to explain this unacceptable event.

  14. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    ERIC Educational Resources Information Center

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  15. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Technical Reports Server (NTRS)

    Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

    1993-01-01

    The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

  16. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91). Final report

    SciTech Connect

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  17. Optical coatings for laser fusion applications

    SciTech Connect

    Lowdermilk, W.H.; Milam, D.; Rainer, F.

    1980-04-24

    Lasers for fusion experiments use thin-film dielectric coatings for reflecting, antireflecting and polarizing surface elements. Coatings are most important to the Nd:glass laser application. The most important requirements of these coatings are accuracy of the average value of reflectance and transmission, uniformity of amplitude and phase front of the reflected or transmitted light, and laser damage threshold. Damage resistance strongly affects the laser's design and performance. The success of advanced lasers for future experiments and for reactor applications requires significant developments in damage resistant coatings for ultraviolet laser radiation.

  18. Engineering the fusion reactor first wall

    SciTech Connect

    Wurden, Glen; Scott, Willms

    2008-01-01

    Recently the National Academy of Engineering published a set of Grand Challenges in Engineering in which the second item listed was entitled 'Provide energy from fusion'. Clearly a key component of this challenge is the science and technology associated with creating and maintaining burning plasmas. This is being vigorously addressed with both magnetic and inertial approaches with various experiments such as ITER and NIF. Considerably less attention is being given to another key component of this challenge, namely engineering the first wall that will contain the burning plasma. This is a daunting problem requiring technologies and materials that can not only survive, but also perform multiple essential functions in this extreme environment. These functions are (1) shield the remainder of the device from radiation. (2) convert of neutron energy to useful heat and (3) breed and extract tritium to maintain the reactor fuel supply. The first wall must not contaminate the plasma with impurities. It must be infused with cooling to maintain acceptable temperatures on plasma facing and structural components. It must not degrade. It must avoid excessive build-up of tritium on surfaces, and, if surface deposits do form, must be receptive to cleaning techniques. All these functions and constraints must be met while being subjected to nuclear and thermal radiation, particle bombardment, high magnetic fields, thermal cycling and occasional impingement of plasma on the surface. And, operating in a nuclear environment, the first wall must be fully maintainable by remotely-operated manipulators. Elements of the first wall challenge have been studied since the 1970' s both in the US and internationally. Considerable foundational work has been performed on plasma facing materials and breeding blanket/shield modules. Work has included neutronics, materials fabrication and joining, fluid flow, tritium breeding, tritium recovery and containment, energy conversion, materials damage and

  19. Fusion reactor breeder material safety compatibility studies

    SciTech Connect

    Jeppson, D.W.; Cohen, S.; Muhlestein, L.D.

    1983-09-01

    Tritium breeder material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Breeder material safety compatibility studies are being conducted to identify and characterize breeder-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate the following. 1. Ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/, and LiTiO/sub 3/) at postulated blanket operating temperatures are chemically compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ reactions with water generate heat, aerosol, and hydrogen. 2. Lithium oxide and 17Li-83Pb alloy react mildly with water requiring special precautions to control hydrogen release. 3. Liquid lithium reacts substantially, while 17Li83Pb alloy reacts mildly with concrete to produce hydrogen. 4. Liquid lithium-air reactions may present some major safety concerns. Additional scoping tests are needed, but the ternary oxides, lithium oxide, and 17Li-83Pb have definite safety advantages over liquid lithium and Li/sub 7/Pb/sub 2/. The ternary oxides present minimal safetyrelated problems when used with water as coolant, air or concrete; but they do require neutron multipliers, which may have safety compatibility concerns with surrounding materials. The combined favorable neutronics and minor safety compatibility concerns of lithium oxide and 17Li-83Pb make them prime candidates as breeder materials. Current safety efforts are directed toward assessing the compatibility of lithium oxide and the lithium-lead alloy with coolants and other materials.

  20. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    NASA Astrophysics Data System (ADS)

    Hançerliogulları, Aybaba; Cini, Mesut

    2013-10-01

    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  1. Alternative fusion concepts and the prospects for improved reactors

    NASA Astrophysics Data System (ADS)

    Krakowski, R. A.

    1985-05-01

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

  2. Helium-3 blankets for tritium breeding in fusion reactors

    NASA Technical Reports Server (NTRS)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  3. Colliding beam fusion reactor space propulsion system

    NASA Astrophysics Data System (ADS)

    Wessel, Frank J.; Binderbauer, Michl W.; Rostoker, Norman; Rahman, Hafiz Ur; O'Toole, Joseph

    2000-01-01

    We describe a space propulsion system based on the Colliding Beam Fusion Reactor (CBFR). The CBFR is a high-beta, field-reversed, magnetic configuration with ion energies in the range of hundreds of keV. Repetitively-pulsed ion beams sustain the plasma distribution and provide current drive. The confinement physics is based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all sources and sinks for energy and particle flow. The mean azimuthal velocities and temperatures of the fuel ion species are equal and the plasma current is unneutralized by the electrons. The resulting distribution functions are thermal in a moving frame of reference. The ion gyro-orbit radius is comparable to the dimensions of the confinement system, hence classical transport of the particles and energy is expected and the device is scaleable. We have analyzed the design over a range of 106-109 Watts of output power (0.15-150 Newtons thrust) with a specific impulse of, Isp~106 sec. A 50 MW propulsion system might involve the following parameters: 4-meters diameter×10-meters length, magnetic field ~7 Tesla, ion beam current ~10 A, and fuels of either D-He3,P-B11,P-Li6,D-Li6, etc. .

  4. Fission-suppressed hybrid reactor: the fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  5. Design study of a G-band FEL amplifier for application to cyclotron resonant heating in magnetic fusion reactors

    NASA Astrophysics Data System (ADS)

    Freund, H. P.; Read, M. E.; Jackson, R. H.; Pershing, D. E.; Taccetti, J. M.

    1995-04-01

    A G-band (140-150 GHz) free-electron laser is described using a coaxial hybrid iron (CHI) wiggler. The CHI wiggler is produced by insertion into a solenoid of a central rod and an outer ring composed of alternating ferrite and nonferrite spacers. The position of the spacers is such that the ferrite (nonferrite) spacers on the central rod are opposite the nonferrite (ferrite) spacers on the outer ring. The field is cylindrically symmetric and exhibits minima in the center of the gap providing for enhanced beam focusing. We describe a tapered wiggler amplifier for plasma heating applications. Preliminary design studies using a nonlinear simulation indicates that output powers of 3.5 MW are possible using a 690 kV/40 A electron beam for a total efficiency of 13%. It is important to note that no beam loss was observed even for realistic values of beam energy spread.

  6. Fusion reactor nucleonics: status and needs

    SciTech Connect

    Lee, J.D.; Engholm, B.A.; Dudziak, D.J.; Haight, R.C.

    1980-01-01

    The national fusion technology effort has made a good start at addressing the basic nucleonics issues, but only a start. No fundamental nucleonics issues are seen as insurmountable barriers to the development of commercial fusion power. To date the fusion nucleonics effort has relied almost exclusively on other programs for nuclear data and codes. But as we progress through and beyond ETF type design studies the fusion program will need to support a broad based nucleonics effort including code development, sensitivity studies, integral experiments, data acquisition etc. It is clear that nucleonics issues are extremely important to fusion development and that we have only scratched the surface.

  7. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    SciTech Connect

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  8. HYLIFE-2 inertial confinement fusion reactor design

    SciTech Connect

    Moir, R.W.

    1990-10-04

    The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x-rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li{sub 2}BeF{sub 4}) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-I, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09$/kW{center dot}h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 12 refs., 9 figs., 5 tabs.

  9. 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.

  10. Mechanical properties and microstructure of copper alloys and copper alloy-stainless steel laminates for fusion reactor high heat flux applications

    NASA Astrophysics Data System (ADS)

    Leedy, Kevin Daniel

    A select group of copper alloys and bonded copper alloy-stainless steel panels are under consideration for heat sink applications in first wall and divertor structures of a planned thermonuclear fusion reactor. Because these materials must retain high strengths and withstand high heat fluxes, their material properties and microstructures must be well understood. Candidate copper alloys include precipitate strengthened CuNiBe and CuCrZr and dispersion strengthened Cu-Alsb2Osb3 (CuAl25). In this study, uniaxial mechanical fatigue tests were conducted on bulk copper alloy materials at temperatures up to 500sp°C in air and vacuum environments. Based on standardized mechanical properties measurement techniques, a series of tests were also implemented to characterize copper alloy-316L stainless steel joints produced by hot isostatic pressing or by explosive bonding. The correlation between mechanical properties and the microstructure of fatigued copper alloys and the interface of copper alloy-stainless steel laminates was examined. Commercial grades of these alloys were used to maintain a degree of standardization in the materials testing. The commercial alloys used were OMG Americas Glidcop CuAl25 and CuAl15; Brush Wellman Hycon 3HP and Trefimetaux CuNiBe; and Kabelmetal Elbrodur and Trefimetaux CuCrZr. CuAl25 and CuNiBe alloys possessed the best combination of fatigue resistance and microstructural stability. The CuAl25 alloy showed only minimal microstructural changes following fatigue while the CuNiBe alloy consistently exhibited the highest fatigue strength. Transmission electron microscopy observations revealed that small matrix grain sizes and high densities of submicron strengthening phases promoted homogeneous slip deformation in the copper alloys. Thus, highly organized fatigue dislocation structure formation, as commonly found in oxygen-free high conductivity Cu, was inhibited. A solid plate of CuAl25 alloy hot isostatically pressed to a 316L stainless steel

  11. Perspectives of SiC-Based Ceramic Composites and Their Applications to Fusion Reactors 2.Fusion Power Reactor Designs Adopting SiC⁄SiC Composite as the Structural Material

    NASA Astrophysics Data System (ADS)

    Nishio, Satoshi

    SiC⁄SiC composite is a promising structural material candidate for fusion power cores and has been considered internationally in several power plant studies. It offers safety advantages arising from it slow induced radioactivity and afterheat, and the possibility of high efficiency of energy conversion through high temperature operation. The latest SiC⁄SiC-based power core design studies are summarized, and the key SiC⁄SiC parameters affecting the performance of power core components are high lighted.

  12. Flibe use in fusion reactors -- An initial safety assessment

    SciTech Connect

    Cadwallader, L.C.; Longhurst, G.R.

    1999-03-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  13. Flibe Use in Fusion Reactors - An Initial Safety Assessment

    SciTech Connect

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  14. Updated comparison of economics of fusion reactors with advanced fission reactors

    SciTech Connect

    Delene, J.G.

    1990-01-01

    The projected cost of electricity (COE) for fusion is compared with that from current and advanced nuclear fission and coal-fired plants. Fusion cost models were adjusted for consistency with advanced fission plants and the calculational methodology and cost factors follow guidelines recommended for cost comparisons of advanced fission reactors. The results show COEs of about 59--74 mills/kWh for the fusion designs considered. In comparison, COEs for future fission reactors are estimated to be in the 43--54 mills/kWh range with coal-fired plant COEs of about 53--69 mills/kWh ($2--3/GJ coal). The principal cost driver for the fusion plants relative to fission plants is the fusion island cost. Although the estimated COEs for fusion are greater than those for fission or coal, the costs are not so high as to preclude fusion's competitiveness as a safe and environmentally sound alternative.

  15. Fuel Encapsulation for Inertial Electrostatic Confinement Nuclear Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Macleod, C.

    Inertial Electrostatic Confinement (IEC) is an approach to nuclear fusion which utilises the properties of electrostatically accelerated ion-beams instead of hot plasmas. The best known device which uses the principle is the Farnsworth-Hirsch fusor. It has been argued that such devices have some potential advantages in spaceflight and in-particular as power-supplies for trans-atmospheric propulsion. This paper builds on previous work in the field and focuses on how the fixing of the fuel for such reactors in a solid, liquid or encapsulated form may provide a high enough energy-density to make such devices practical power sources. Several methods of fixing the fuel are discussed; theoretical calculations are presented and applicable literature is reviewed. Finally, there is a discussion of practical issues and feasibility, together with suggestions for further work.

  16. Evaluation of performance of select fusion experiments and projected reactors

    NASA Technical Reports Server (NTRS)

    Miley, G. H.

    1978-01-01

    The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters.

  17. MINIMARS: an attractive small tandem mirror fusion reactor

    SciTech Connect

    Perkins, L.J.; Logan, B.G.; Doggett, J.N.; Devoto, R.S.; Nelson, W.D.; Lousteau, D.C.; Kulcinski, G.L.; Santarius, J.F.; Gordon, J.D.; Campbell, R.B.

    1985-11-13

    Through the innovative design of a novel end plug scheme employing octopole MHD stabilization, we present the conceptual design of ''MIMIMARS'', a small commercial fusion reactor based on the tandem mirror principle. The current baseline for MINIMARS has a net electric output of 600 MWe and we have configured the design for short construction times, factory-built modules, inherently safe blanket systems, and multiplexing in station sizes of approx. 600 to 2400 MWe. We demonstrate that the compact octopole end cell provides a number of advantages over the more conventional quadrupole (yin-yang) end cell encountered in the MARS tandem mirror reactor study, and enables ignition to be achieved with much shorter central cell lengths. Accordingly, being economic in small sizes, MINIMARS provides an attractive alternative to the more conventional larger conceptual fusion reactors encountered to date, and would contribute significantly to the lowering of utility financial risk in a developing fusion economy.

  18. Application of polarized nuclei to fusion

    SciTech Connect

    Kulsrud, R.M.

    1987-07-01

    It is shown that the d-t fusion reaction can be modified by polarizing nuclear spins. The ways in which this improves reactor performance are mentioned and the feasibility of the process of spin polarization for magnetic fusion is discussed. 18 refs.

  19. 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.

  20. The TITAN Reversed-Field Pinch fusion reactor study

    SciTech Connect

    Not Available

    1988-03-01

    The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m{sup 2} and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m{sup 2}; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings.

  1. 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.

  2. Neutron Dosimetry Tokamak Fusion Test Reactor Lithium Blanket Module

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderl, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment.

  3. Neutron irradiation facilities for fission and fusion reactor materials studies

    SciTech Connect

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs.

  4. Inertial-fusion-reactor studies at Lawrence Livermore National Laboratory

    SciTech Connect

    Monsler, M.J.; Meier, W.R.

    1982-08-01

    We present results of our reactor studies for inertial-fusion energy production. Design studies of liquid-metal wall chambers have led to reactors that are remarkably simple in design, and that promise long life and low cost. Variants of the same basic design, called HYLIFE, can be used for electricity production, as a fissile-fuel factory, a dedicated tritium breeder, or hybrids of each.

  5. Present Status of Vanadium Alloys for Fusion Applications

    SciTech Connect

    Muroga, Takeo; Chen, J. M.; Chernov, V. M.; Kurtz, Richard J.; Le Flem, M.

    2014-12-01

    Vanadium alloys are advanced options for low activation structural materials. After more than two decades of research, V-4Cr-4Ti has been emerged as the leading candidate, and technological progress has been made in reducing the number of critical issues for application of vanadium alloys to fusion reactors. Notable progress has been made in fabricating alloy products and weld joints without degradation of properties. Various efforts are also being made to improve high temperature strength and creep-rupture resistance, low temperature ductility after irradiation, and corrosion resistance in blanket conditions. Future research should focus on clarifying remaining uncertainty in the operating temperature window of V-4Cr-4Ti for application to near to middle term fusion blanket systems, and on further exploration of advanced materials for improved performance for longer-term fusion reactor systems.

  6. Present status of vanadium alloys for fusion applications

    NASA Astrophysics Data System (ADS)

    Muroga, T.; Chen, J. M.; Chernov, V. M.; Kurtz, R. J.; Le Flem, M.

    2014-12-01

    Vanadium alloys are advanced options for low activation structural materials. After more than two decades of research, V-4Cr-4Ti has been emerged as the leading candidate, and technological progress has been made in reducing the number of critical issues for application of vanadium alloys to fusion reactors. Notable progress has been made in fabricating alloy products and weld joints without degradation of properties. Various efforts are also being made to improve high temperature strength and creep-rupture resistance, low temperature ductility after irradiation, and corrosion resistance in blanket conditions. Future research should focus on clarifying remaining uncertainty in the operating temperature window of V-4Cr-4Ti for application to near to middle term fusion blanket systems, and on further exploration of advanced materials for improved performance for longer-term fusion reactor systems.

  7. The neutronics studies of fusion fission hybrid power reactor

    SciTech Connect

    Zheng Youqi; Wu Hongchun; Zu Tiejun; Yang Chao; Cao Liangzhi

    2012-06-19

    In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

  8. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

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

    1999-11-01

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

  9. Current Trends of Blanket Research and Deveopment in Japan 3.Blanket Designs in Fusion Power Reactors

    NASA Astrophysics Data System (ADS)

    Sagara, Akio; Enoeda, Mikio; Nishio, Satoshi; Kozaki, Yasuji

    The main functions of the blanket in fusion power reactors are basically independent of the type of magnetic fusion reactor (tokamak, helical, etc.) and inertia fusion. However, from technical point of view, many candidate designs of blanket have been proposed depending on the particular reactor concepts. Their main features are characterized for the recent typical designs, and key issues are defined.

  10. Z-Pinch Fusion for Energy Applications

    SciTech Connect

    SPIELMAN,RICK B.

    2000-01-01

    Z pinches, the oldest fusion concept, have recently been revisited in light of significant advances in the fields of plasma physics and pulsed power engineering. The possibility exists for z-pinch fusion to play a role in commercial energy applications. We report on work to develop z-pinch fusion concepts, the result of an extensive literature search, and the output for a congressionally-mandated workshop on fusion energy held in Snowmass, Co July 11-23,1999.

  11. Electron microscopy and microanalysis of the fiber-matrix interface in monolithic silicone carbide-based ceramic composite material for use in a fusion reactor application.

    PubMed

    Toplisek, Tea; Drazic, Goran; Novak, Sasa; Kobe, Spomenka

    2008-01-01

    A composite material made from continuous monolithic silicone carbide (SiC) fibers and a SiC-based matrix (SiC(f)/SiC), was prepared using a novel technique, i.e. adapted dip coating and infiltration of SiC fibers with a water suspension containing SiC particles and a sintering additive. This kind of material could be used in the first-wall blanket of a future fusion reactor. Using magnetron sputtering, the SiC fibers were coated with various thin layers (TiC, CrN, CrC, WC, DLC-diamond-like carbon) of the interface material by physical vapor deposition (PVD). Using scanning and transmission electron microscopy and microanalysis, detailed microstructural studies of the fiber-matrix interface were performed. Both samples, with coated and uncoated fibers, were examined under a load. The microcracks introduced by the Vickers indenter continued their path through the fibers, and thus caused the failure of the composite material, in the case of the uncoated fibers or deviated from their primary direction at the fiber-matrix interface in the case of the coated fibers. PMID:18172883

  12. A Compact Nuclear Fusion Reactor for Space Flights

    NASA Astrophysics Data System (ADS)

    Nastoyashchiy, Anatoly F.

    2006-05-01

    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. 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.

  14. Effects of Collisional Dissipation on the "Colliding Beam Fusion Reactor "

    NASA Astrophysics Data System (ADS)

    Lampe, Martin; Manheimer, Wallace M.

    1998-11-01

    Rostoker, Binderbauer and Monkhorst have recently proposed a "colliding beam fusion reactor" (CBFR) for use with the p-B11 reaction. We have examined the various dissipative processes resulting from Coulomb collisions, and have concluded that the CBFR equilibrium cannot be sustained for long enough to permit net fusion gain. There are many collisional processes which occur considerably faster than fusion, and result in particle loss, energy loss, or detuning of the resonant energy for the p-B reaction. Pitch-angle scattering of protons off the boron beam, which occurs 100 times faster than fusion, isotropizes the proton beam and results in proton loss. Energy exchange between protons and boron, which is 20 times faster than fusion, detunes the resonance. Proton-proton scattering, which is faster than fusion for all CBFR scenarios, Maxwellianizes the protons and thus detunes the resonance. Ion-electron collisions lead indirectly to a friction between the two ion beams, which is typically fast compared to the fusion process. Results of Fokker-Planck analyses of each process will be shown.

  15. Characterization of the axial plasma shock in a table top plasma focus after the pinch and its possible application to testing materials for fusion reactors

    SciTech Connect

    Soto, Leopoldo Pavez, Cristian; Moreno, José; Inestrosa-Izurieta, María José; Veloso, Felipe; Gutiérrez, Gonzalo; Vergara, Julio; Clausse, Alejandro; Bruzzone, Horacio; Castillo, Fermín; and others

    2014-12-15

    The characterization of plasma bursts produced after the pinch phase in a plasma focus of hundreds of joules, using pulsed optical refractive techniques, is presented. A pulsed Nd-YAG laser at 532 nm and 8 ns FWHM pulse duration was used to obtain Schlieren images at different times of the plasma dynamics. The energy, interaction time with a target, and power flux of the plasma burst were assessed, providing useful information for the application of plasma focus devices for studying the effects of fusion-relevant pulses on material targets. In particular, it was found that damage factors on targets of the order of 10{sup 4} (W/cm{sup 2})s{sup 1/2} can be obtained with a small plasma focus operating at hundred joules.

  16. Modular stellarator reactor: a fusion power plant

    SciTech Connect

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  17. 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.

  18. 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.

  19. Frontier of Fusion Research: Path to the Steady State Fusion Reactor by Large Helical Device

    NASA Astrophysics Data System (ADS)

    Motojima, Osamu

    2006-12-01

    The ITER, the International Thermonuclear Experimental Reactor, which will be built in Cadarache in France, has finally started this year, 2006. Since the thermal energy produced by fusion reactions divided by the external heating power, i.e., the Q value, will be larger than 10, this is a big step of the fusion research for half a century trying to tame the nuclear fusion for the 6.5 Billion people on the Earth. The source of the Sun's power is lasting steadily and safely for 8 Billion years. As a potentially safe environmentally friendly and economically competitive energy source, fusion should provide a sustainable future energy supply for all mankind for ten thousands of years. At the frontier of fusion research important milestones are recently marked on a long road toward a true prototype fusion reactor. In its own merits, research into harnessing turbulent burning plasmas and thereby controlling fusion reaction, is one of the grand challenges of complex systems science. After a brief overview of a status of world fusion projects, a focus is given on fusion research at the National Institute for Fusion Science (NIFS) in Japan, which is playing a role of the Inter University Institute, the coordinating Center of Excellence for academic fusion research and by the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility. The current status of LHD project is presented focusing on the experimental program and the recent achievements in basic parameters and in steady state operations. Since, its start in a year 1998, a remarkable progress has presently resulted in the temperature of 140 Million degree, the highest density of 500 Thousand Billion/cc with the internal density barrier (IDB) and the highest steady average beta of 4.5% in helical plasma devices and the largest total input energy of 1.6 GJ, in all magnetic confinement fusion devices. Finally, a perspective is given of the ITER Broad Approach program

  20. Overview of Indian activities on fusion reactor materials

    NASA Astrophysics Data System (ADS)

    Banerjee, Srikumar

    2014-12-01

    This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils. Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V-4Cr-4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu-1 wt%Cr-0.1 wt%Zr - an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall - was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER). India has proposed to install and test a lead-lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li2TiO3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb-Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box

  1. Development of fusion blanket technology for the DEMO reactor.

    PubMed

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. PMID:22112596

  2. Summary of HEDL Fusion Reactor Safety Support studies

    SciTech Connect

    Muhlestein, L.D.; Jeppson, D.W.; Barreca, J.R.

    1981-01-01

    The HEDL Fusion Reactor Safety Support studies are focused on characterizing blanket-coolant-material reactions for deuterium-tritium fusion reactor designs. The objective is to determine and examine potential safety and environmental issues associated with proposed blanket/coolant combinations under postulated accident conditions. The first studies considered liquid lithium as both blanket and coolant, and examined liquid lithium-material reactions. Liquid lithium reactions with oxygen, nitrogen, and various concretes have been characterized. Evaluations of lithium reaction extinguishment methods, lithium aerosol generation and collection, and the volatilization and transport of radioactive materials in connection with lithium-air reactions have been completed. Lithium compound blanket material reactions with water, a prime coolant candidate, have been characterized in terms of energy and gas release rates. Blanket materials considered were lithium aluminate, lithium oxide, lithium zirconate, lithium silicate, and lithium lead alloys (Li/sub 7/Pb/sub 2/ and Li/sub 17/Pb/sub 83/).

  3. Thermonuclear Fusion Research Progress and the Way to the Reactor

    NASA Astrophysics Data System (ADS)

    Koch, Raymond

    2006-06-01

    The paper reviews the progress of fusion research and its prospects for electricity generation. It starts with a reminder of the principles of thermonuclear fusion and a brief discussion of its potential role in the future of the world energy production. The reactions allowing energy production by fusion of nuclei in stars and on earth and the conditions required to sustain them are reviewed. At the high temperatures required for fusion (hundred millions kelvins), matter is completely ionized and has reached what is called its 4th state: the plasma state. The possible means to achieve these extreme temperatures is discussed. The remainder of the paper focuses on the most promising of these approaches, magnetic confinement. The operating principles of the presently most efficient machine of this type — the tokamak — is described in some detail. On the road to producing energy with fusion, a number of obstacles have to be overcome. The plasma, a fluid that reacts to electromagnetic forces and carries currents and charges, is a complex medium. Fusion plasma is strongly heated and is therefore a good example of a system far from equilibrium. A wide variety of instabilities can grow in this system and lead to self-organized structures and spontaneous cycles. Turbulence is generated that degrades the confinement and hinders easy achievement of long lasting hot plasmas. Physicists have learned how to quench turbulence, thereby creating sort of insulating bottles inside the plasma itself to circumvent this problem. The recent history of fusion performance is outlined and the prospect of achieving power generation by fusion in a near future is discussed in the light of the development of the "International Tokamak Experimental Reactor" project ITER.

  4. Comparative breeding characteristics of fusion and fast reactors.

    PubMed

    Fortescue, P

    1977-06-17

    Expressions are developed to allow ready comparison of a hybrid fission-fusion plant and a fast breeder with respect to the number of thermal reactors that their fissile production could support, both for their feed requirements and for the new inventory needs of an expanding industry. These relations are expressed in terms of the neutron multiplication factor obtained in the fusion blanket, and the analogous quantities represented by the conversion ratios of the fast and thermal fission associated with the comparison. Results are presented graphically both for the steady state and for industries of arbitrary growth rate, and include the influence of tritium production requirements. Even a modest blanket neutron multiplication factor could enable the hybrid fusion system greatly to outperform the fast breeder on this simple basis of material balances. PMID:17831749

  5. Towards the detection of magnetohydrodynamics instabilities in a fusion reactor

    SciTech Connect

    Sozzi, Carlo Alessi, E. Figini, L. Galperti, G. Lazzaro, E. Marchetto, C. Nowak, S.; Mosconi, M.

    2014-08-21

    Various active control strategies of the Neoclassical tearing modes are being studied in present tokamaks using established detection techniques which exploit the measurements of the fluctuations of the magnetic field and of the electron temperature. The extrapolation of such techniques to the fusion reactor scale is made problematic by the neutron fluence and by the physics conditions related to the high plasma temperature and density which degrade the spatial resolution of such measurements.

  6. Neutral-beam systems for magnetic-fusion reactors

    SciTech Connect

    Fink, J. H.

    1981-08-10

    Neutral beams for magnetic fusion reactors are at an early stage of development, and require considerable effort to make them into the large, reliable, and efficient systems needed for future power plants. To optimize their performance to establish specific goals for component development, systematic analysis of the beamlines is essential. Three ion source characteristics are discussed: arc-cathode life, gas efficiency, and beam divergence, and their significance in a high-energy neutral-beam system is evaluated.

  7. Towards the detection of magnetohydrodynamics instabilities in a fusion reactor

    NASA Astrophysics Data System (ADS)

    Sozzi, Carlo; Alessi, E.; Figini, L.; Galperti, G.; Lazzaro, E.; Marchetto, C.; Mosconi, M.; Nowak, S.

    2014-08-01

    Various active control strategies of the Neoclassical tearing modes are being studied in present tokamaks using established detection techniques which exploit the measurements of the fluctuations of the magnetic field and of the electron temperature. The extrapolation of such techniques to the fusion reactor scale is made problematic by the neutron fluence and by the physics conditions related to the high plasma temperature and density which degrade the spatial resolution of such measurements.

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

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

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

  9. A study on nuclear properties of Zr, Nb, and Ta nuclei used as structural material in fusion reactor

    NASA Astrophysics Data System (ADS)

    Sahan, Halide; Tel, Eyyup; Sahan, Muhittin; Aydin, Abdullah; Hakki Sarpun, Ismail; Kara, Ayhan; Doner, Mesut

    2015-07-01

    Fusion has a practically limitless fuel supply and is attractive as an energy source. The main goal of fusion research is to construct and operate an energy generating system. Fusion researches also contains fusion structural materials used fusion reactors. Material issues are very important for development of fusion reactors. Therefore, a wide range of fusion structural materials have been considered for fusion energy applications. Zirconium (Zr), Niobium (Nb) and Tantalum (Ta) containing alloys are important structural materials for fusion reactors and many other fields. Naturally Zr includes the 90Zr (%51.5), 91Zr (%11.2), 92Zr (%17.1), 94Zr (%17.4), 96Zr (%2.80) isotopes and 93Nb and 181Ta include the 93Nb (%100) and 181Ta (%99.98), respectively. In this study, the charge, mass, proton and neutron densities and the root-mean-square (rms) charge radii, rms nuclear mass radii, rms nuclear proton, and neutron radii have been calculated for 87-102Zr, 93Nb, 181Ta target nuclei isotopes by using the Hartree-Fock method with an effective Skyrme force with SKM*. The calculated results have been compared with those of the compiled experimental taken from Atomic Data and Nuclear Data Tables and theoretical values of other studies.

  10. Pressure drop considerations of a lithium cooled fusion breeder tokamak reactor blanket

    SciTech Connect

    Wong, C.P.C.

    1983-12-06

    Liquid lithium was selected as one of the coolants for the 1983 fusion breeder blanket used on the magnetically confined tokamak fusion reactor, and as a result, the thermal-hydraulic calculations were dominated by magnetohydrodynamic (MHD) considerations. The applicable sets of MHD equations for the engineering thermal-hydraulic design were reviewed and compared. Special attention was given to the MHD calculations for the fertile material zone, a packed bed of composite beryllium and thorium balls, since this region can dominate the thermal-hydraulic behavior of this blanket module. To keep the pressure drops acceptable, fertile fuel balls were omitted in the inboard blanket.

  11. Fire protection system operating experience review for fusion applications

    SciTech Connect

    Cadwallader, L.C.

    1995-12-01

    This report presents a review of fire protection system operating experiences from particle accelerator, fusion experiment, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of fire protection system component failure rates and fire accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with these systems are discussed, including spurious operation. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor.

  12. Cryogenic system operating experience review for fusion applications

    SciTech Connect

    Cadwallader, L.C.

    1992-01-01

    This report presents a review of cryogenic system operating experiences, from particle accelerator, fusion experiment, space research, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of cryogenic component failure rates and accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with cryogenic systems are discussed, including ozone formation, effects of spills, and modeling spill behavior. This information should be useful to fusion system designers and safety analysts, such as the team working on the International Thermonuclear Experimental Reactor design.

  13. Special purpose materials for fusion application

    NASA Astrophysics Data System (ADS)

    Scott, J. L.; Clinard, F. W.; Wiffen, F. W.

    1985-08-01

    Originally in 1978 the Special Purpose Materials Task Group was concerned with tritium breeding materials, coolants, tritium barriers, graphite and silicon carbide, ceramics, heat-sink materials, and magnet components. Since then several other task groups have been created, so that now the category includes only materials for superconducting magnets and ceramics. For the former application copper-stabilized Nb 3Sn (Ti) insulated with polyimides will meet the general requirements, so that testing of prototype components is the priority task. Ceramics are required for several critical components of fusion reactors either as dielectrics or as a structural material. Components near the first wall will receive exposures of 5-20 MW y/m 2. Other ceramic applications are well behind the first wall, with lower damage levels. Most insulators operate near room temperature, but ceramic blanket structures may operate up to 1000°C. Because of a meager data base, one cannot identify optimum ceramics for structural application; but MgAl 2O 4 is an attractive dielectric material.

  14. Special purpose materials for fusion application

    SciTech Connect

    Scott, J.L.; Clinard, F.W. Jr.; Wiffen, F.W.

    1984-01-01

    Originally in 1978 the Special Purpose Materials Task Group was concerned with tritium breeding materials, coolants, tritium barriers, graphite and silicon carbide, ceramics, heat-sink materials, and magnet components. Since then several other task groups have been created, so now the category includes only materials for superconducting magnets and ceramics. For the former application copper-stabilized Nb/sub 3/Sn (Ti) insulated with polyimides will meet the general requirements, so that testing of prototype components is the priority task. Ceramics are required for several critical components of fusion reactors either as dielectrics or as a structural material. Components near the first wall will receive exposures of 5 to 20 MW.year/m/sup 2/. Other ceramic applications are well behind the first wall, with lower damage levels. Most insulators operate near room temperature, but ceramic blanket structures may operate up to 1000/sup 0/C. Because of a meager data base, one cannot identify optimum ceramics for structural application; but MgAl/sub 2/O/sub 4/ is an attractive dielectric material.

  15. Calculations of ( n, α) Cross Sections on Some Structural Fusion Materials for Fusion Reactor Technology

    NASA Astrophysics Data System (ADS)

    Yiğit, M.; Tel, E.; Tanır, G.

    2013-06-01

    The knowledge of cross section for emission of light charged particles ( p, d, t, and α) induced by fast neutrons on structural fusion materials has a critical importance on fusion reactors. The gas production arising from ( n, p) and ( n, α) reactions causes seriously radiation damage in fusion reactor structure. The radiation damage in fusion related materials is a large problem need to be overcome for development of fusion reactor technology. Particularly, the ( n, α) reaction cross section data are required to estimation of the radiation damage effects on structural fusion materials. Therefore, the cross section data for ( n, α) reaction induced by fast neutrons are of increasing importance for the success of future fusion reactors. In this study, reaction model calculations of the cross sections of neutron induced reactions on structural fusion materials such as 29 Si, 30 Si, 48 Ti, 50 Ti, 50 Cr, 54 Cr, 54 Fe and 58 Fe have been investigated. The new calculations on the excitation functions of 29 Si ( n, α) 26 Mg, 30 Si ( n, α) 27 Mg, 48 Ti ( n, α) 45 Ca, 50 Ti ( n, α) 47 Ca, 50 Cr ( n, α) 47 Ti, 54 Cr ( n, α) 51 Ti, 54 Fe ( n, α) 51 Cr and 58 Fe ( n, α) 55 Cr have been carried out for incident neutron energies up to 30 MeV. In these calculations, the pre-equilibrium and equilibrium effects for ( n, α) reactions have been investigated. The pre-equilibrium calculations involve the new evaluated the geometry dependent hybrid model, hybrid model and the cascade exciton model. The equilibrium effects of the excitation functions for the investigated reactions are calculated according to the Weisskopf-Ewing model. Also in the present work, the ( n, α) reaction cross sections have calculated by using evaluated empirical formulas developed by Tel et al. at 14-15 MeV energy. The calculated results have been discussed and compared with the available experimental data and found agreement with each other.

  16. Characteristics of irradiation creep in the first wall of a fusion reactor

    SciTech Connect

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available.

  17. D-He-3 spherical torus fusion reactor system study

    NASA Astrophysics Data System (ADS)

    Macon, William A., Jr.

    1992-04-01

    This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

  18. Material options for a commercial fusion reactor first wall

    SciTech Connect

    Dabiri, A.E.

    1986-05-01

    A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m/sup 2/. A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW/sup 2/, provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys.

  19. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    SciTech Connect

    Tsai, H.; Gomes, I.C.; Smith, D.L.; Palmer, A.J.; Ingram, F.W.; Wiffen, F.W.

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  20. Repair welding of fusion reactor components. Final technical report

    SciTech Connect

    Chin, B.A.; Wang, C.A.

    1997-09-30

    The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials.

  1. Review of progress on fusion materials technology, Harwell, December 1980. Irradiation effects in fusion reactor materials

    NASA Astrophysics Data System (ADS)

    Harries, D. R.

    1981-03-01

    The evolution of the radiation damage structure, void and gas bubble swelling, and surface blistering effects in both model and potential first wall materials for a D-T fusion reactor system of the TOKAMAK type was investigated along with radiation effects in inorganic insulator materials. In addition, investigations were carried out into the effects of irradiation on organic insulators and on the performance of rubber seals. The principal achievements are summarized and a list of 50 references is given.

  2. Repair welding of fusion reactor components

    SciTech Connect

    Chin, B.A.

    1993-05-15

    Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

  3. Electronuclear ion fusion in an ion cyclotron resonance reactor

    SciTech Connect

    Cowgill, Donald F.

    1996-12-01

    A method and apparatus for generating nuclear fusion by ion cyclotron resonance in an ion trap reactor. The reactor includes a cylindrical housing having an axial axis, an internal surface, and first and second ends. First and second end plates that are charged are respectively located at the first and second ends of the cylindrical housing. A gas layer is adsorbed on the internal surface of the cylindrical housing. Ions are desorbed from the gas layer, forming a plasma layer adjacent to the cylindrical housing that includes first ions that have a same charge sign as the first and second end plates. A uniform magnetic field is oriented along the axial axis of the cylindrical housing. Second ions, that are unlike the first ions, but have the same charge sign, are injected into the cylindrical housing along the axial axis of the cylindrical housing. A radio frequency field resonantly accelerates the injected second ions at the cyclotron resonance frequency of the second ions. The second ions circulate in increasing helical orbits and react with the first ions, at the optimum energy for nuclear fusion. The amplitude of the radio frequency field is adjusted to accelerate the second ions at a rate equal to the rate of tangential energy loss of the second ions by nuclear scattering in the first ions, causing the ions to continually interact until fusion occurs.

  4. Tritium retention in fusion reactor plasma facing components

    SciTech Connect

    Langley, R.A.

    1995-03-01

    The IAEA has proposed a coordinated research program to address tritium retention and release in fusion reactor plasma facing components. This program will address materials which are mainly of interest to the design and construction of ITER, namely beryllium, carbon based materials and medium and high-Z metals, e.g. tungsten, vanadium and molybdenum, but will not be limited to these materials. Experimental data are needed for: recycling models, tritium inventory estimates, tritium permeation calculations and hydrogen embrittlement characterization. The ultimate use of the data would be to influence the formation of models for use by fusion reactor designers. Judicious material choices must be made by the designers and accurate predictive codes are required in order to make these choices. The proposed coordinated research program will provide a forum for discussions between experimentalists, theoreticians, modelers and reactor designers, provide financial support for relevant research projects and collect and evaluate experimental and theoretical data. This paper briefly reviews existing data, addresses the data gaps and points out experiments designed to obtain the needed data. 18 refs., 3 figs., 1 tab.

  5. Innovative approaches to inertial confinement fusion reactors: Final report

    SciTech Connect

    Bourque, R.F.; Schultz, K.R.

    1986-11-01

    Three areas of innovative approaches to inertial confinement fusion (ICF) reactor design are given. First, issues pertaining to the Cascade reactor concept are discussed. Then, several innovative concepts are presented which attempt to directly recover the blast energy from a fusion target. Finally, the Turbostar concept for direct recovery of that energy is evaluated. The Cascade issues discussed are combustion of the carbon granules in the event of air ingress, the use of alternate granule materials, and the effect of changes in carbon flow on details of the heat exchanger. Carbon combustion turns out to be a minor problem. Four ICF innovative concepts were considered: a turbine with ablating surfaces, a liquid piston system, a wave generator, and a resonating pump. In the final analysis, none show any real promise. The Turbostar concept of direct recovery is a very interesting idea and appeared technically viable. However, it shows no efficiency gain or any decrease in capital cost compared to reactors with conventional thermal conversion systems. Attempts to improve it by placing a close-in lithium sphere around the target to increase gas generation increased efficiency only slightly. It is concluded that these direct conversion techniques require thermalization of the x-ray and debris energy, and are Carnot limited. They therefore offer no advantage over existing and proposed methods of thermal energy conversion or direct electrical conversion.

  6. Broad-Application Test Reactor

    SciTech Connect

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators.

  7. Biomagnetic effects: a consideration in fusion reactor development.

    PubMed Central

    Mahlum, D D

    1977-01-01

    Fusion reactors will utilize powerful magnetic fields for the confinement and heating of plasma and for the diversion of impurities. Large dipole fields generated by the plasma current and the divertor and transformer coils will radiate outward for several hundred meters, resulting in magnetic fields up to 450 gauss in working areas. Since occupational personnel could be exposed to substantial magnetic fields in a fusion power plant, an attempt has been made to assess the possible biological and health consequences of such exposure, using the existing literature. The available data indicate that magnetic fields can interact with biological material to produce effects, although the reported effects are usually small in magnitude and often unconfirmed. The existing data base is judged to be totally inadequate for assessment of potential health and environmental consequences of magnetic fields and for the establishment of appropriate standards. Requisite studies to provide an adequate data base are outlined. PMID:598345

  8. Radiation Hydrodynamic Simulations of an Inertial Fusion Energy Reactor Chamber

    NASA Astrophysics Data System (ADS)

    Sacks, Ryan Foster

    Inertial fusion energy reactors present great promise for the future as they are capable of providing baseline power with no carbon footprint. Simulation work regarding the chamber response and first wall insult is carried out using the 1-D BUCKY radiation hydrodynamics code for a variety of differing chamber fills, radii, chamber obstructions and first wall materials. Discussion of the first wall temperature rise, x-ray spectrum incident on the wall, shock timing and maximum overpressure are presented. An additional discussion of the impact of different gas opacities and their effect on overall chamber dynamics, including the formation of two shock fronts, is also presented. This work is performed under collaboration with Lawrence Livermore National Laboratory at the University of Wisconsin-Madison's Fusion Technology Institute.

  9. Estimated radiactive and shock loading of fusion reactor armor

    SciTech Connect

    Swift, D C

    2008-11-25

    Inertial confinement fusion (ICF) is of interest as a source of neutrons for proliferation-resistant and high burn-up fission reactor designs. ICF is a transient process, each implosion leading to energy release over a short period, with a continuous series of ICF operations needed to drive the fission reactor. ICF yields energy in the form of MeV-range neutrons and ions, and thermal x-rays. These radiations, particularly the thermal x-rays, can deposit a pulse of energy in the wall of the ICF chamber, inducing loading by isochoric heating (i.e. at constant volume before the material can expand) or by ablation of material from the surface. The explosion of the hot ICF system, and the compression of any fill material in the chamber, may also result in direct mechanical loading by a blast wave (decaying shock) reaching the chamber wall. The chamber wall must be able to survive the repetitive loading events for long enough for the reactor to operate economically. It is thus necessary to understand the loading induced by ICF systems in possible chamber wall designs, and to predict the response and life time of the wall. Estimates are given for the loading induced in the wall armor of the fusion chamber caused by ablative thermal radiation from the fusion plasma and by the hydrodynamic shock. Taking a version of the LIFE design as an example, the ablation pressure was estimated to be {approx}0.6 GPa with an approximately exponential decay with time constant {approx}0.6 ns. Radiation hydrodynamics simulations suggested that ablation of the W armor should be negligible.

  10. 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.

  11. Considerations for tritium protection at a fusion reactor

    SciTech Connect

    Easterly, C.E.

    1981-01-01

    The purpose of this paper is to indicate the general direction of current fusion reactor concepts regarding tritium, and to indicate some options in tritium control strategies. Certain strategies, in addition to providing reduced potential health hazard, afford the potential for engineering alternatives for in-plant tritium control systems. The overall coupling of containment cleanup systems and health protection must continue to develop with increased knowledge of the health effects of different tritium species and the consequent systems options available subsequent to this understanding.

  12. High conductivity Be-Cu alloys for fusion reactors

    SciTech Connect

    Lilley, E.A.; Adachi, Takao; Ishibashi, Yoshiki

    1995-09-01

    The optimum material has not yet been identified. This will result in heat from plasma to the first wall and divertor. That is, because of cracks and melting by thermal power and shock. Today, it is considered to be some kinds of copper, alloys, however, for using, it must have high conductivity. And it is also needed another property, for example, high strength and so on. We have developed some new beryllium copper alloys with high conductivity, high strength, and high endurance. Therefore, we are introducing these new alloys as suitable materials for the heat sink in fusion reactors.

  13. 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.

  14. 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.

  15. The cost of tritium production in a fusion reactor

    SciTech Connect

    Wittenberg, L.J. . Fusion Technology Inst.)

    1992-03-01

    In this paper, a computational model is presented in order to assess the cost of tritium breeding in a fusion power reactor. This model compares the differential cost of the Li-bearing breeder blanket with that of a steel shield and adds the loss of revenue due to the lower energy multiplication of the breeder blanket compared to the steel shield. The cost of tritium production ranges from $215-$300/g for a simple breeder up to $1420/g for a high temperature breeder.

  16. Synfuels from fusion: producing hydrogen with the Tandem Mirror Reactor and thermochemical cycles

    SciTech Connect

    Werner, R.W.; Ribe, F.L.

    1981-01-21

    This volume contains the following sections: (1) the Tandem Mirror fusion driver, (2) the Cauldron blanket module, (3) the flowing microsphere, (4) coupling the reactor to the process, (5) the thermochemical cycles, and (6) chemical reactors and process units. (MOW)

  17. A Z-Pinch Driven Fusion Reactor Concept

    NASA Astrophysics Data System (ADS)

    Derzon, Mark; Rochau, Gregory; Spielman, Rick; Slutz, Stephen; Rochau, G. E.; Peterson, R. R.; Peterson, P. F.

    1999-11-01

    Recent z-pinch target physics progress has encouraged us to consider how a power reactor could be configured based on a fast z-pinch driver. Initial cost estimates show that recyclable transmission lines (RTLs) are economically viable. Providing 'standoff' between the primary power supply and the target, which is what disposable RTLs provide, has historically been the main obstacle to the consideration of pinches as fusion drivers. We will be introducing basic reactor scaling in terms of shot rate, yield, tritium breeding and neutron flux, etc. This concept has advantages in that z-pinches provide a robust mechanical environment, as well as a chamber which does not require low-pressure pumping between shots and the wall lifetime is expected to be limited factors other than neutron damage. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy under contract DE-AC04-94AL85000.

  18. Materials research and development for fusion energy applications

    SciTech Connect

    Zinkle, S.J.; Snead, L.L.

    1998-11-01

    Some of the critical issues associated with materials selection for proposed magnetic fusion reactors are reviewed, with a brief overview of refractory alloys (vanadium, tantalum, molybdenum, tungsten) and primary emphasis on ceramic materials. SiC/SiC composites are under consideration for the first wall and blanket structure, and dielectric insulators will be used for the heating, control and diagnostic measurement of the fusion plasma. Key issues for SiC/SiC composites include radiation-induced degradation in the strength and thermal conductivity. Recent work has focused on the development of radiation-resistant fibers and fiber/matrix interfaces (porous SiC, SiC multilayers) which would also produce improved SiC/SiC performance for applications such as heat engines and aerospace components. The key physical parameters for dielectrics include electrical conductivity, dielectric loss tangent and thermal conductivity. Ionizing radiation can increase the electrical conductivity of insulators by many orders of magnitude, and surface leakage currents can compromise the performance of some fusion energy components. Irradiation can cause a pronounced degradation in the loss tangent and thermal conductivity. Fundamental physical parameter measurements on ceramics which are of interest for both fusion and non-fusion applications are discussed.

  19. Fusion-reactor plasmas with polarized nuclei. II

    SciTech Connect

    Kulsrud, R.M.; Furth, H.P.; Valeo, E.J.; Budny, R.V.; Jassby, D.L.; Micklich, B.J.; Post, D.E.; Goldhaber, M.; Happer, W.

    1982-11-01

    New techniques of bulk polarization could be used to fuel a reactor with polarized hydrogenic atoms, so as to form a plasma of polarized nuclei. Theoretical calculations indicate that, once the nuclei of the plasma are polarized in some preferred state, they can maintain this state with a probability near 100% during their lifetime in the reactor, including possible recycling. There are a number of practical advantages to be gained from the use of polarized plasma in a fusion reactor. The nuclear reaction rates can be increased or decreased, and/or the direction of emission of the reaction products can be controlled. The D-T reaction rate can be enhanced by as much as 50%, with the reaction products emitted perpendicular to the magnetic field. Alternatively, it is possible to direct the reaction products primarily along the field, with no enhancement. In this case of the D-D reaction, the theoretical predictions are somewhat less certain. Enhancement of the reaction rate by a factor of 1.5-2.5 is to be expected. In a different polarization state, suppression of D-D reactions may be feasible - a possibility that would be of interest for a neutron-free D-He/sup 3/ reactor. A quantitative discussion of the relevant nuclear physics as well as of the various mechanisms producing depolarization is given.

  20. Tungsten-based composite materials for fusion reactor shields

    SciTech Connect

    Greenspan, E.; Karni, Y.

    1985-11-01

    Composite tungsten-based materials were recently proposed for the heavy constituent of compact fusion reactor shields. These composite materials will enable the incorporation of tungsten - the most efficient nonfissionable inelastic scattering (as well as good neutron absorbing and very good photon attenuating) material - in the shield in a relatively cheap way and without introducing voids (so as to enable minimizing the shield thickness). It is proposed that these goals be achieved by bonding tungsten powder, which is significantly cheaper than high-density tungsten, with a material having the following properties: good shielding ability and relatively low cost and ease of fabrication. The purpose of this work is to study the effectiveness of the composite materials as a function of their composition, and to estimate the economic benefit that might be gained by the use of these materials. Two materials are being considered for the binder: copper, second to tungsten in its shielding ability, and iron (or stainless steel), the common fusion reactor shield heavy constituent.

  1. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  2. Reactor Application for Coaching Newbies

    SciTech Connect

    2015-06-17

    RACCOON is a Moose based reactor physics application designed to engage undergraduate and first-year graduate students. The code contains capabilities to solve the multi group Neutron Diffusion equation in eigenvalue and fixed source form and will soon have a provision to provide simple thermal feedback. These capabilities are sufficient to solve example problems found in Duderstadt & Hamilton (the typical textbook of senior level reactor physics classes). RACCOON does not contain any advanced capabilities as found in YAK.

  3. Introduction to D-He(3) fusion reactors

    NASA Technical Reports Server (NTRS)

    Vlases, G. C.; Steinhauer, L. C.

    1989-01-01

    A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

  4. Applications of a Nd:YAG Laser for In-Situ Removal of Tritium from Plasma Facing Components in a Fusion Reactor Environment

    NASA Astrophysics Data System (ADS)

    Sedita, Paul; Gentile, Charles; McGahan, Christina

    2009-11-01

    Studies from TFTR and JET have indicated that long term retention of tritium in plasma facing components account for a significant fraction of the tritium inventory; thus requiring periodic removal of T interned within these first wall components. This investigation will determine the efficacy of removing tritium in an in-situ fashion from co-deposits layers in PFC's employing a continuous wave 325 w Nd:YAG laser. The laser is configured to rapidly heat first wall surfaces using a rastering motion to effectively scan the exposed surfaces within the target area. The top 100 μm of the target surface is expected to transiently heat up to ˜ 1000 C within several seconds, subsequently releasing the trapped (un-expended) tritium for re-use within the fusion fuel cycle. The presentation will discuss the efficacy of this method using empirical data (destructive and non-destructive measurements) from pre and post surface laser heating.

  5. Application of railgun principle to high-velocity hydrogen pellet injection for magnetic fusion reactor refueling. Technical progress report, [July 16, 1990--August 16, 1991

    SciTech Connect

    Kim, K.

    1991-08-01

    This report contains three documents describing the progress made by the University of Illinois electromagnetic railgun program sponsored by the Office of Fusion Energy of the United States Department of Energy during the period from July 16, 1990 to August 16, 1991. The first document contains a brief summary of the tasks initiated, continued, or completed, the status of major tasks, and the research effort distribution, estimated and actual, during the period. The second document contains a description of the work performed on time resolved laser interferometric density measurement of the railgun plasma-arc armature. The third document is an account of research on the spectroscopic measurement of the electron density and temperature of the railgun plasma arc.

  6. Programmable AC power supply for simulating power transient expected in fusion reactor

    SciTech Connect

    Halimi, B.; Suh, K. Y.

    2012-07-01

    This paper focus on control engineering of the programmable AC power source which has capability to simulate power transient expected in fusion reactor. To generate the programmable power source, AC-AC power electronics converter is adopted to control the power of a set of heaters to represent the transient phenomena of heat exchangers or heat sources of a fusion reactor. The International Thermonuclear Experimental Reactor (ITER) plasma operation scenario is used as the basic reference for producing this transient power source. (authors)

  7. A Fusion Reactor Design with a Liquid First Wall and Divertor

    SciTech Connect

    Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z; Merrill, B J; Eberle, C; Fogarty, P J; Nelson, B E; Sze, D K; Majeski, R

    2003-11-13

    Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

  8. 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.

  9. Coupled edge-core model of fusion reactor

    NASA Astrophysics Data System (ADS)

    Zagórski, R.; Kulinski, S.; Scholz, M.

    1997-10-01

    A model has been developed which is capable to describe in a self consistent way the plasma dynamics in the centre and edge region of a fusion reactor. The core plasma is treated in the frame of the 0D model in which an empirical scaling law for the energy confinement time is included. The model accounts for energy losses due to Bremsstrahlung and line radiation as well as alpha particle heating. A 1D analytical model for plasma and impurity transport outside the last close magnetic surface (LCMS) is applied. The model accounts for the strong gradients of the plasma parameters along the magnetic field lines in the divertor. The sputtering phenomena at the plate and radiating cooling by injected impurities are treated self consistently in the model. The model has been used to investigate operating regimes of the ignition experiment. Analysis have been performed for different first wall materials (C, Ni, Mo, W) for ITER like tokamak.

  10. Cryogenic system component development for fusion experimental reactor at JAERI

    SciTech Connect

    Kato, T.; Kamiya, S.; Tada, E.; Hiyama, T.; Kawano, K.; Shimamoto, S.

    1986-11-01

    A supercritical helium (SHE) circulation pump, a jet pump, and a cold compressor were designed and manufactured as the first step of cryogenic component development for a large-scale cryogenic system which is required for the Fusion Experimental Reactor (FER). The SHE circulation pump achieved 320-g/s flow rate with an 0.88-MPa pressure head at 4.6 K, making it the biggest cold pump in the world. The jet pump's mass flow ratio was about 1.0 with an 0.07-MPa pressure head at about 10 K. The cold compressor was successfully operated with an inlet vapor pressure of 0.053 MPa (3.7 K), and outlet pressure of 0.12 MPa, and a mass flow rate of 60 g/s. The designs and test results are described in this paper.

  11. Method and apparatus for making uniform pellets for fusion reactors

    DOEpatents

    Budrick, Ronald G.; King, Frank T.; Martin, Alfred J.; Nolen, Jr., Robert L.; Solomon, David E.

    1977-01-01

    A method and apparatus for making uniform pellets for laser driven fusion reactors which comprises selection of a quantity of glass frit which has been accurately classified as to size within a few micrometers and contains an occluded material, such as urea, which gasifies and expands when heated. The sized particles are introduced into an apparatus which includes a heated vertical tube with temperatures ranging from 800.degree. C to 1300.degree. C. The particles are heated during the drop through the tube to molten condition wherein the occluded material gasifies to form hollow microspheres which stabilize in shape and plunge into a collecting liquid at the bottom of the tube. The apparatus includes the vertical heat resistant tube, heaters for the various zones of the tube and means for introducing the frit and collecting the formed microspheres.

  12. The Magnetic Dipole as an Attractive Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Dawson, John M.

    1997-11-01

    Stability for low β plasma confined by closed B field lines is PV^γ = C_0, P = pressure, V = flux tube volume, γ is c_p/cv = 5/3. Kesner(J. Kesner, Innovative Confinement Concepts Workshop, Mar. 3-6, 1997) proposed a levitated current ring with the plasma stabilized by this condition as an alternate fusion reactor. Such a reactor has many attractive features; at radii large compared to the ring radius, V goes like r^4; the stability condition is Pr^20/3 = C_1. If nr^4 = C_2, then interchanges keep the density constant. The temperature can drop according to Tr^8/3 = C_3. If the chamber is ten times the ring radius, the density can drop from 10^14 near the ring to 10^10 at the edge and the temperature can drop from 50 keV near the ring to 100 eV at the edge. This plasma should present no problems for a divertor. Reacting plasma near the ring will heat it, upsetting the stability relation and cause convection to carry burnt plasma out; it will cool as it expands. At the same time the convection will bring in fresh fuel from the outside which will be compressed and heated to ignition. A super conducting ring design that can float in reacting D-He^3 for 16 hours exists(J.M. Dawson, FUSION, edited by Edward Teller, Vol. 1, Magnetic Confinement, Part, Ch. 16, Academic Press, 1981).

  13. Microchannel Reactors for ISRU Applications

    NASA Astrophysics Data System (ADS)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  14. Transmutation analysis of realistic low-activation steels for magnetic fusion reactors and IFMIF

    SciTech Connect

    Cabellos, O; Sanz, J; Garc?a-Herranz, N; D?az, S; Reyes, S; Piedloup, S

    2005-11-22

    A comprehensive transmutation study for steels considered in the selection of structural materials for magnetic and inertial fusion reactors has been performed in the IFMIF neutron irradiation scenario, as well as in the ITER and DEMO ones for comparison purposes. An element-by-element transmutation approach is used in the study, addressing the generation of: (1) H and He and (2) solid transmutants. The IEAF-2001 activation library and the activation code ACAB were applied to the IFMIF transmutation analysis, after proving the applicability of ACAB for transmutation calculations of this kind of intermediate energy systems.

  15. Research and development on vanadium alloys for fusion applications

    SciTech Connect

    Zinkle, S.J.; Rowcliffe, A.F.; Matsui, H.; Abe, K.; Smith, D.L.; Osch, E. van; Kazakov, V.A.

    1998-03-01

    The current status of research and development on unirradiated and irradiated V-Cr-Ti alloys intended for fusion reactor structural applications is reviewed, with particular emphasis on the flow and fracture behavior of neutron-irradiated vanadium alloys. Recent progress on fabrication, joining, oxidation behavior, and the development of insulator coatings is also summarized. Fabrication of large (>500 kg) heats of V-4Cr-4Ti with properties similar to previous small laboratory heats has now been demonstrated. Impressive advances in the joining of thick sections of vanadium alloys using GTA and electron beam welds have been achieved in the past two years, although further improvements are still needed.

  16. A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

    SciTech Connect

    Brad J. Merrill; Richard L. Moore; J. Phillip Sharp

    2008-09-01

    A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium-steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.

  17. Vacuum system operating experience review for fusion applications

    SciTech Connect

    Cadwallader, L.C.

    1994-03-01

    This report presents a review of vacuum system operating experiences from particle accelerator, fusion experiment, space simulation chamber, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of vacuum system component failure rates and accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with vacuum systems are discussed, including personnel safety, foreign material intrusion, and factors relevant to vacuum systems being the primary confinement boundary for tritium and activated dusts. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor.

  18. Reactor prospects of muon-catalyzed fusion of deuterium and tritium concentrated in transition metals

    SciTech Connect

    Stacey, W.M. Jr. . Fusion Research Center )

    1989-09-01

    It is conjectured that the number of fusion events catalyzed by a single muon is orders of magnitude greater for deuterium and tritium concentrated in a transition metal than in gaseous form and that the recent observation of 2.5-MeV neutrons from a D/sub 2/O electrolytic cell with palladium and titanium cathodes can thereby be interpreted in terms of cosmic muon-catalyzed deuterium-deuterium fusion. This suggests a new fusion reactor reactor consisting of deuterium and tritium concentrated in transition metal fuel elements in a fusion core that surrounds an accelerator-produced muon source. The feasibility of net energy production in such a reactor is established in terms of requirements on the number of fusion events catalyzed per muon. The technological implications for a power reactor based on this concept are examined. The potential of such a concept as a neutron source for materials testing and tritium and plutonium production is briefly discussed.

  19. Fusion-reactor blanket-material safety-compatibility studies

    SciTech Connect

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/ and LiTiO/sub 3/) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li/sub 17/Pb/sub 83/ alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li/sub 17/Pb/sub 83/ alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns.

  20. Design of the DEMO Fusion Reactor Following ITER

    PubMed Central

    Garabedian, Paul R.; McFadden, Geoffrey B.

    2009-01-01

    Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

  1. Reactor Application for Coaching Newbies

    Energy Science and Technology Software Center (ESTSC)

    2015-06-17

    RACCOON is a Moose based reactor physics application designed to engage undergraduate and first-year graduate students. The code contains capabilities to solve the multi group Neutron Diffusion equation in eigenvalue and fixed source form and will soon have a provision to provide simple thermal feedback. These capabilities are sufficient to solve example problems found in Duderstadt & Hamilton (the typical textbook of senior level reactor physics classes). RACCOON does not contain any advanced capabilities asmore » found in YAK.« less

  2. rf coupler technology for fusion applications

    SciTech Connect

    Hoffman, D.J.

    1983-01-01

    Radio frequency (rf) oscillations at critical frequencies have successfully provided a means to convey power to fusion plasmas due to the electrical-magnetic properties of the plasma. While large rf systems to couple power to the plasma have been designed, built, and tested, the main link to the plasma, the coupler, is still in an evolutionary stage of development. Design and fabrication of optimal antennas for fusion applications are complicated by incomplete characterizations of the harsh plasma environment and of coupling mechanisms. A brief description of rf coupler technology required for plasma conditions is presented along with an assessment of the status and goals of coupler development.

  3. CECE alternative for upgrading/detritiation in heavy water nuclear reactors and for tritium recovery in fusion reactors

    SciTech Connect

    Spagnolo, D.A.; Miller, A.I.

    1995-10-01

    The Combined Electrolysis Catalytic Exchange (CECE) process, utilizing AECL`s wetproofed catalyst, is ideally suited for extracting tritium from water because of its high isotopic separation factor and near-ambient operating conditions. Several CECE options are compared with the more conventional DW-VPCE arrangements for heavy water upgrading and detritiation of CANDU nuclear reactors and for detritiation of fusion facilities such as ITER. For both applications, CECE offers a more economical alternative over conventional technology. Experimental data on catalyst activity and lifetime are also presented and past commercial applications of the AECL catalyst are reviewed. AECL has recently committed to assembly of a CECE upgrading/detritiation demonstration facility. 15 refs., 5 figs., 1 tab.

  4. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  5. Negative Halogen Ions for Fusion Applications

    SciTech Connect

    Grisham, L.R.; Kwan, J.W.; Hahto, S.K.; Hahto, S.T.; Leung, K.N.; Westenskow, G.

    2006-01-01

    Over the past quarter century, advances in hydrogen negative ion sources have extended the usable range of hydrogen isotope neutral beams to energies suitable for large magnetically confined fusion devices. Recently, drawing upon this experience, negative halogen ions have been proposed as an alternative to positive ions for heavy ion fusion drivers in inertial confinement fusion, because electron accumulation would be prevented in negative ion beams, and if desired, the beams could be photo-detached to neutrals. This paper reports the results of an experiment comparing the current density and beam emittance of Cl+ and Cl- extracted from substantially ion-ion plasmas with that of Ar+ extracted from an ordinary electron-ion plasma, all using the same source, extractor, and emittance scanner. At similar discharge conditions, the Cl- current was typically 85 – 90% of the positive chlorine current, with an e-/ Cl- ratio as low as seven without grid magnets. The Cl- was as much as 76% of the Ar+ current from a discharge with the same RF drive. The minimum normalized beam emittance and inferred ion temperatures of Cl+, Cl-, and Ar+ were all similar, so the current density and optical quality of Cl- appear as suitable for heavy ion fusion driver applications as a positive noble gas ion of similar mass. Since F, I, and Br should all behave similarly in an ion source, they should also be suitable as driver beams.

  6. 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.

  7. High Power Density Blanket Design Study for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Huang, J. H.; Zhu, Y. K.; Deng, P. Zh.

    2003-06-01

    A conceptual design study of a high power density blanket has been carried out. The Fusion Experimental Breeder, FEB, is adopted as the reference reactor. The neutron wall loading is 0.5 MW/m2. The blanket is cooled by 10 MPa helium in tube. The concept of LiPb eutectic/transuranium oxide suspension is adopted. The neutronics design is performed to provide the design basis, and it gives an energy multiplication of 37 and a flattened power density distribution with a peak value of 70 W/m3. Multiple cooling panels are introduced to reduce the peak temperature of the blanket. In spite of up to 15 cooling panels, the blanket module is calculated using the ANSYS code and analytically as well. The results are consistent with each other and can meet the thermal criteria. However, structural calculation results from ANSYS did not satisfy the criterion: The blanket structure design is then improved by using curved cooling panels to model the structure in detail. Temperature distribution is obtained using the Pro/Mechanica code. Detailed structural analyses are also done by this code. Some satisfactory results are obtained.

  8. Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.

    Energy Science and Technology Software Center (ESTSC)

    1999-07-23

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energymore » deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.« less

  9. Technical issues for beryllium use in fusion blanket applications

    SciTech Connect

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented.

  10. The cryogenic diffusion pump; An advanced design for fusion reactor primary pumping and fuel processing

    SciTech Connect

    Hemmerich, J.L. )

    1992-03-01

    This paper reports on a re-evaluation of the characteristics of the intermediate flow regime with simultaneous thermal accommodation has shown the full potential of the Cryogenic Diffusion Pump for Fusion Reactor applications. A device with a characteristic diameter of 1m will have a pumping speed of 150m{sup 3}s{sup {minus}1} for Deuterium at an inlet pressure of 2 {times} 10{sup 2}Pa (Reactor Burn phased) and 400m{sup 3} s{sup {minus}1} at an inlet pressure of 0.1 Pa (Reactor Dwell phase). Simultaneously, it separates impurities, Hydrogen isotopes and Helium and compresses the Helium. The Helium compression ratio (already proven to be {ge}25 for 3% Helium in D{sub 2}) can be further enhanced by additional D{sub 2} or He driven Diffusion Pump and Ejector stages. The latter feature will also simplify pumping requirements for the Helium Glow Discharge scenario: recirculation of Helium at 0.1 Pa (driven by D{sub 2} or He Ejector) and simultaneous removal of DT and impurities by cryocondensation requires no mechanical pump at all or only small turbomolecular-drag pump combinations of He jet drive. The design offers superior tritium compatibility: all metal, fully bakeable, it avoids use of absorbers and argon for helium pumping, thereby reducing overall tritium inventory both in the pump itself and by replacing major fuel clean-up facilities. The advantages of using the Cryogenic Diffusion Pump in a Fusion Reactor Vacuum System are discussed in detail.

  11. Safety and environmental constraints on space applications of fusion energy

    NASA Technical Reports Server (NTRS)

    Roth, J. Reece

    1990-01-01

    Some of the constraints are examined on fusion reactions, plasma confinement systems, and fusion reactors that are intended for such space related missions as manned or unmanned operations in near earth orbit, interplanetary missions, or requirements of the SDI program. Of the many constraints on space power and propulsion systems, those arising from safety and environmental considerations are emphasized since these considerations place severe constraints on some fusion systems and have not been adequately treated in previous studies.

  12. SAR and LIDAR fusion: experiments and applications

    NASA Astrophysics Data System (ADS)

    Edwards, Matthew C.; Zaugg, Evan C.; Bradley, Joshua P.; Bowden, Ryan D.

    2013-05-01

    In recent years ARTEMIS, Inc. has developed a series of compact, versatile Synthetic Aperture Radar (SAR) systems which have been operated on a variety of small manned and unmanned aircraft. The multi-frequency-band SlimSAR has demonstrated a variety of capabilities including maritime and littoral target detection, ground moving target indication, polarimetry, interferometry, change detection, and foliage penetration. ARTEMIS also continues to build upon the radar's capabilities through fusion with other sensors, such as electro-optical and infrared camera gimbals and light detection and ranging (LIDAR) devices. In this paper we focus on experiments and applications employing SAR and LIDAR fusion. LIDAR is similar to radar in that it transmits a signal which, after being reflected or scattered by a target area, is recorded by the sensor. The differences are that a LIDAR uses a laser as a transmitter and optical sensors as a receiver, and the wavelengths used exhibit a very different scattering phenomenology than the microwaves used in radar, making SAR and LIDAR good complementary technologies. LIDAR is used in many applications including agriculture, archeology, geo-science, and surveying. Some typical data products include digital elevation maps of a target area and features and shapes extracted from the data. A set of experiments conducted to demonstrate the fusion of SAR and LIDAR data include a LIDAR DEM used in accurately processing the SAR data of a high relief area (mountainous, urban). Also, feature extraction is used in improving geolocation accuracy of the SAR and LIDAR data.

  13. Implications of polarized DT plasmas for toroidal fusion reactors

    SciTech Connect

    Micklich, B.J.; Jassby, D.L.

    1983-05-01

    Spin polarization of the deuterons and tritons in a reacting plasma can result in an increase in the fusion reactivity and variation of the angular distribution of emission of the fusion neutrons. The increased fusion reactivity relaxes the confinement-temperature conditions for breakeven and ignition. We have determined the effect of varying the angular distribution of the fusion neutrons on the spatial distribution of fusion neturon current and flux at the first wall, on the global tritium breeding ratio, and on the first-wall radiation damage in low-aspect-ratio toroidal geometry.

  14. Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

    SciTech Connect

    McAdams, R.

    2014-02-15

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  15. Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

    PubMed

    McAdams, R

    2014-02-01

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed. PMID:24593596

  16. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    SciTech Connect

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper.

  17. 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.

  18. Electron cyclotron emission imaging and applications in magnetic fusion energy

    NASA Astrophysics Data System (ADS)

    Tobias, Benjamin John

    Energy production through the burning of fossil fuels is an unsustainable practice. Exponentially increasing energy consumption and dwindling natural resources ensure that coal and gas fueled power plants will someday be a thing of the past. However, even before fuel reserves are depleted, our planet may well succumb to disastrous side effects, namely the build up of carbon emissions in the environment triggering world-wide climate change and the countless industrial spills of pollutants that continue to this day. Many alternatives are currently being developed, but none has so much promise as fusion nuclear energy, the energy of the sun. The confinement of hot plasma at temperatures in excess of 100 million Kelvin by a carefully arranged magnetic field for the realization of a self-sustaining fusion power plant requires new technologies and improved understanding of fundamental physical phenomena. Imaging of electron cyclotron radiation lends insight into the spatial and temporal behavior of electron temperature fluctuations and instabilities, providing a powerful diagnostic for investigations into basic plasma physics and nuclear fusion reactor operation. This dissertation presents the design and implementation of a new generation of Electron Cyclotron Emission Imaging (ECEI) diagnostics on toroidal magnetic fusion confinement devices, or tokamaks, around the world. The underlying physics of cyclotron radiation in fusion plasmas is reviewed, and a thorough discussion of millimeter wave imaging techniques and heterodyne radiometry in ECEI follows. The imaging of turbulence and fluid flows has evolved over half a millennium since Leonardo da Vinci's first sketches of cascading water, and applications for ECEI in fusion research are broad ranging. Two areas of physical investigation are discussed in this dissertation: the identification of poloidal shearing in Alfven eigenmode structures predicted by hybrid gyrofluid-magnetohydrodynamic (gyrofluid-MHD) modeling, and

  19. Fusion reactor materials. Semiannual progress report for period ending September 30, 1993

    SciTech Connect

    Rowcliffe, A.F.; Burn, G.L.; Knee`, S.S.; Dowker, C.L.

    1994-02-01

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  20. HYLIFE-II inertial confinement fusion reactor design

    SciTech Connect

    Moir, R.W.

    1990-12-14

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li{sub 2}BeF{sub 4}) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW{center dot}h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 15 refs., 9 figs., 3 tabs.

  1. Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

    NASA Astrophysics Data System (ADS)

    Wang, Shijia; Wang, Shaojie

    2015-04-01

    The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITER inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ˜30 % , with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by 60 % , with the central pressure also significantly raised.

  2. Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

    SciTech Connect

    Wang, Shijia Wang, Shaojie

    2015-04-15

    The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITER inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ∼30%, with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by  60%, with the central pressure also significantly raised.

  3. Fusion

    NASA Astrophysics Data System (ADS)

    Herman, Robin

    1990-10-01

    The book abounds with fascinating anecdotes about fusion's rocky path: the spurious claim by Argentine dictator Juan Peron in 1951 that his country had built a working fusion reactor, the rush by the United States to drop secrecy and publicize its fusion work as a propaganda offensive after the Russian success with Sputnik; the fortune Penthouse magazine publisher Bob Guccione sank into an unconventional fusion device, the skepticism that met an assertion by two University of Utah chemists in 1989 that they had created "cold fusion" in a bottle. Aimed at a general audience, the book describes the scientific basis of controlled fusion--the fusing of atomic nuclei, under conditions hotter than the sun, to release energy. Using personal recollections of scientists involved, it traces the history of this little-known international race that began during the Cold War in secret laboratories in the United States, Great Britain and the Soviet Union, and evolved into an astonishingly open collaboration between East and West.

  4. Fusion reactor materials semiannual progress report for the period ending March 31, 1993

    SciTech Connect

    Not Available

    1993-07-01

    This is the fourteenth in a series of semiannual technical progress reports on fusion reactor materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the US Depart of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. Separate abstracts were prepared for each individual section.

  5. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Technical Reports Server (NTRS)

    Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

    1993-01-01

    We applied three design principles (DPs) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: (1) provide maximum direct access to space for waste radiation, (2) operate components as passive radiators to minimize cooling-system mass, and (3) optimize the plasma fuel, fuel mix, and temperature for best specific Jet power. The three candidate-terrestrial-fusion-reactor configurations are: (1) the thermal-barrier-tandem-mirror (TBTM), (2) field-reversed-mirror (FRM), and (3) levitated-dipole-field (LDF). The resulting three candidate-space-fusion-propulsion systems have their initial-mass-to-LEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System.

  6. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor.

    PubMed

    Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

    2015-10-01

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. PMID:26164282

  7. Applying design principles to fusion reactor configurations for propulsion in space

    NASA Astrophysics Data System (ADS)

    Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

    1993-06-01

    We applied three design principles (DPs) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: (1) provide maximum direct access to space for waste radiation, (2) operate components as passive radiators to minimize cooling-system mass, and (3) optimize the plasma fuel, fuel mix, and temperature for best specific Jet power. The three candidate-terrestrial-fusion-reactor configurations are: (1) the thermal-barrier-tandem-mirror (TBTM), (2) field-reversed-mirror (FRM), and (3) levitated-dipole-field (LDF). The resulting three candidate-space-fusion-propulsion systems have their initial-mass-to-LEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System.

  8. OPERATION OF FUSION REACTORS IN ONE ATMOSPHERE OF AIR INSTEAD OF VACUUM SYSTEMS

    SciTech Connect

    Roth, J. Reece

    2009-07-26

    Engineering design studies of both magnetic and inertial fusion power plants have assumed that the plasma will undergo fusion reactions in a vacuum environment. Operation under vacuum requires an expensive additional major system for the reactor-a vacuum vessel with vacuum pumping, and raises the possibility of sudden unplanned outages if the vacuum containment is breached. It would be desirable in many respects if fusion reactors could be made to operate at one atmosphere with air surrounding the plasma, thus eliminating the requirement of a pressure vessel and vacuum pumping. This would have obvious economic, reliability, and engineering advantages for currently envisaged power plant reactors; it would make possible forms of reactor control not possible under vacuum conditions (i.e. adiabatic compression of the fusion plasma by increasing the pressure of surrounding gas); it would allow reactors used as aircraft engines to operate as turbojets or ramjets in the atmosphere, and it would allow reactors used as fusion rockets to take off from the surface of the earth instead of low earth orbit.

  9. Materials compatibility considerations for a fusion-fission hybrid reactor design

    SciTech Connect

    DeVan, J.H.; Tortorelli, P.F.

    1983-01-01

    The Tandem Mirror Hybrid Reactor is a fusion reactor concept that incorporates a fission-suppressed breeding blanket for the production of /sup 233/U to be used in conventional fission power reactors. The present paper reports on compatibility considerations related to the blanket design. These considerations include solid-solid interactions and liquid metal corrosion. Potential problems are discussed relative to the reference blanket operating temperature (490/sup 0/C) and the recycling time of breeding materials (<1 year).

  10. A Subcritical, Gas-Cooled Fast Transmutation Reactor with a Fusion Neutron Source

    SciTech Connect

    Stacey, W.M.; Beavers, V.L.; Casino, W.A.; Cheatham, J.R.; Friis, Z.W.; Green, R.D.; Hamilton, W.R.; Haufler, K.W.; Hutchinson, J.D.; Lackey, W.J.; Lorio, R.A.; Maddox, J.W.; Mandrekas, J.; Manzoor, A.A.; Noelke, C.A.; Oliveira, C. de; Park, M.; Tedder, D.W.; Terry, M.R.; Hoffman, E.A.

    2005-05-15

    A design is presented for a subcritical, He-cooled fast reactor, driven by a tokamak D-T fusion neutron source, for the transmutation of spent nuclear fuel (SNF). The reactor is fueled with coated transuranic (TRU) particles and is intended for the deep-burn (>90%) transmutation of the TRUs in SNF without reprocessing of the coated fuel particles. The reactor design is based on the materials, fuel, and separations technologies under near-term development in the U.S. Department of Energy (DOE) Nuclear Energy Program and on the plasma physics and fusion technologies under near-term development in the DOE Fusion Energy Sciences Program, with the objective of intermediate-term ({approx}2040) deployment. The physical and performance characteristics and research and development requirements of such a reactor are described.

  11. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    NASA Astrophysics Data System (ADS)

    Indah Rosidah, M.; Suud, Zaki; Yazid, Putranto Ilham

    2015-09-01

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  12. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    SciTech Connect

    Indah Rosidah, M. Suud, Zaki; Yazid, Putranto Ilham

    2015-09-30

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  13. Development of imaging bolometers for magnetic fusion reactors (invited).

    PubMed

    Peterson, Byron J; Parchamy, Homaira; Ashikawa, Naoko; Kawashima, Hisato; Konoshima, Shigeru; Kostryukov, Artem Yu; Miroshnikov, Igor V; Seo, Dongcheol; Omori, T

    2008-10-01

    Imaging bolometers utilize an infrared (IR) video camera to measure the change in temperature of a thin foil exposed to the plasma radiation, thereby avoiding the risks of conventional resistive bolometers related to electric cabling and vacuum feedthroughs in a reactor environment. A prototype of the IR imaging video bolometer (IRVB) has been installed and operated on the JT-60U tokamak demonstrating its applicability to a reactor environment and its ability to provide two-dimensional measurements of the radiation emissivity in a poloidal cross section. In this paper we review this development and present the first results of an upgraded version of this IRVB on JT-60U. This upgrade utilizes a state-of-the-art IR camera (FLIR/Indigo Phoenix-InSb) (3-5 microm, 256 x 360 pixels, 345 Hz, 11 mK) mounted in a neutron/gamma/magnetic shield behind a 3.6 m IR periscope consisting of CaF(2) optics and an aluminum mirror. The IRVB foil is 7 cm x 9 cm x 5 microm tantalum. A noise equivalent power density of 300 microW/cm(2) is achieved with 40 x 24 channels and a time response of 10 ms or 23 microW/cm(2) for 16 x 12 channels and a time response of 33 ms, which is 30 times better than the previous version of the IRVB on JT-60U. PMID:19044463

  14. Development of imaging bolometers for magnetic fusion reactors (invited)

    SciTech Connect

    Peterson, Byron J.; Parchamy, Homaira; Ashikawa, Naoko; Kawashima, Hisato; Konoshima, Shigeru; Kostryukov, Artem Yu.; Miroshnikov, Igor V.; Seo, Dongcheol; Omori, T.

    2008-10-15

    Imaging bolometers utilize an infrared (IR) video camera to measure the change in temperature of a thin foil exposed to the plasma radiation, thereby avoiding the risks of conventional resistive bolometers related to electric cabling and vacuum feedthroughs in a reactor environment. A prototype of the IR imaging video bolometer (IRVB) has been installed and operated on the JT-60U tokamak demonstrating its applicability to a reactor environment and its ability to provide two-dimensional measurements of the radiation emissivity in a poloidal cross section. In this paper we review this development and present the first results of an upgraded version of this IRVB on JT-60U. This upgrade utilizes a state-of-the-art IR camera (FLIR/Indigo Phoenix-InSb) (3-5 {mu}m, 256x360 pixels, 345 Hz, 11 mK) mounted in a neutron/gamma/magnetic shield behind a 3.6 m IR periscope consisting of CaF{sub 2} optics and an aluminum mirror. The IRVB foil is 7 cmx9 cmx5 {mu}m tantalum. A noise equivalent power density of 300 {mu}W/cm{sup 2} is achieved with 40x24 channels and a time response of 10 ms or 23 {mu}W/cm{sup 2} for 16x12 channels and a time response of 33 ms, which is 30 times better than the previous version of the IRVB on JT-60U.

  15. Conceptual study of fusion-driven transmutation reactor with ITER physics and engineering constraints

    NASA Astrophysics Data System (ADS)

    Hong, Bong

    2011-10-01

    A conceptual study of fusion-driven transmutation reactor was performed based on ITER physics and engineering constraints. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

  16. Neoclassical Simulations of Fusion Alpha Particles in Pellet Charge Exchange Experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    Batha, S.H.; Budny, R.V.; Darrow, D.S.; Levinton, F.M.; Redi, M.H.; et al

    1999-02-01

    Neoclassical simulations of alpha particle density profiles in high fusion power plasmas on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 5 (1998) 1577] are found to be in good agreement with measurements of the alpha distribution function made with a sensitive active neutral particle diagnostic. The calculations are carried out in Hamiltonian magnetic coordinates with a fast, particle-following Monte Carlo code which includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss and collisional effects. New global loss and confinement domain calculations allow an estimate of the actual alpha particle densities measured with the pellet charge exchange diagnostic.

  17. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

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

    SciTech Connect

    Not Available

    1990-01-01

    This report discusses research on the titan-1 fusion power core. The major topics covered are: titan-1 fusion-power-core engineering; titan-1 divertor engineering; titan-1 tritium systems; titan-1 safety design and radioactive-waste disposal; and titan-1 maintenance procedures.

  19. Evaluation of Nb-base alloys for the divertor structure in fusion reactors

    SciTech Connect

    Purdy, I.M.

    1996-04-01

    Niobium-base alloys are candidate materials for the divertor structure in fusion reactors. For this application, an alloy should resist aqueous corrosion, hydrogen embrittlement, and radiation damage and should have high thermal conductivity and low thermal expansion. Results of corrosion and embrittlement screening tests of several binary and ternary Nb alloys in high-temperature water indicated the Mb-1Zr, Nb-5MO-1Zr, and Nb-5V-1Z4 (wt %) showed sufficient promise for further investigation. These alloys, together with pure Nb and Zircaloy-4 have been exposed to high purity water containing a low concentration of dissolved oxygen (<12 ppb) at 170, 230, and 300{degrees}C for up to {approx}3200 h. Weight-change data, microstructural observations, and qualitative mechanical-property evaluation reveal that Nb-5V-1Zr is the most promising alloy at higher temperatures. Below {approx}200{degrees}C, the alloys exhibit similiar corrosion behavior.

  20. 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.

  1. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    SciTech Connect

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.

  2. Progress on Gyrotrons for ITER and Future Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Thumm, Manfred K.

    2009-11-01

    The prototype of the Japan 170 GHz ITER gyrotron holds the energy and efficiency world record of 2.88 GJ (0.8 MW, 3600 s, 57%) with 55% efficiency at 1 MW, 800 s, whereas the Russian 170 GHz ITER prototype tube achieved 0.83 MW with a pulse duration of 203 s at 48% efficiency and 1 MW at 116 s and 52%. The record parameters of the European megawatt-class 140 GHz gyrotron for the Stellarator Wendelstein W7-X are: 0.92 MW output power at 1800 s pulse duration, almost 45% efficiency and 97.5% Gaussian mode purity. All these gyrotrons employ a cylindrical cavity, a quasi-optical output coupler, a synthetic diamond window and a single-stage depressed collector (SDC) for energy recovery. In coaxial cavities the existence of the longitudinally corrugated inner conductor reduces the problems of mode competition and limiting current, thus allowing one to use even higher order modes with lower Ohmic attenuation than in cylindrical cavities. Synthetic diamond windows with a transmission capability of 2 MW, continuous wave (CW) are feasible. In order to keep the number of the required gyrotrons and magnets as low as possible, to reduce the costs of the ITER 26 MW, 170 GHz ECRH system and to allow compact upper launchers for plasma stabilization, 2 MW mm-wave power per gyrotron tube is desirable. The FZK pre-prototype tube for an EU 170 GHz, 2 MW ITER gyrotron has achieved 1.8 MW at 28% efficiency (without depressed collector). Design studies for a 4 MW 170 GHz coaxial-cavity gyrotron with two synthetic diamond output windows and two 2 MW mm-wave output beams for future fusion reactors are currently being performed at FZK. The availability of sources with fast frequency tunability (several GHz s-1, tuning in 1.5-2.5% steps for about ten different frequencies) would permit the use of a simple, fixed, non-steerable mirror antenna for local current drive (ECCD) experiments and plasma stabilization. GYCOM in Russia develops in collaboration with IPP Garching and FZK an industrial

  3. Progress on Gyrotrons for ITER and Future Fusion Reactors

    SciTech Connect

    Thumm, Manfred K.

    2009-11-26

    The prototype of the Japan 170 GHz ITER gyrotron holds the energy and efficiency world record of 2.88 GJ (0.8 MW, 3600 s, 57%) with 55% efficiency at 1 MW, 800 s, whereas the Russian 170 GHz ITER prototype tube achieved 0.83 MW with a pulse duration of 203 s at 48% efficiency and 1 MW at 116 s and 52%. The record parameters of the European megawatt-class 140 GHz gyrotron for the Stellarator Wendelstein W7-X are: 0.92 MW output power at 1800 s pulse duration, almost 45% efficiency and 97.5% Gaussian mode purity. All these gyrotrons employ a cylindrical cavity, a quasi-optical output coupler, a synthetic diamond window and a single-stage depressed collector (SDC) for energy recovery. In coaxial cavities the existence of the longitudinally corrugated inner conductor reduces the problems of mode competition and limiting current, thus allowing one to use even higher order modes with lower Ohmic attenuation than in cylindrical cavities. Synthetic diamond windows with a transmission capability of 2 MW, continuous wave (CW) are feasible. In order to keep the number of the required gyrotrons and magnets as low as possible, to reduce the costs of the ITER 26 MW, 170 GHz ECRH system and to allow compact upper launchers for plasma stabilization, 2 MW mm-wave power per gyrotron tube is desirable. The FZK pre-prototype tube for an EU 170 GHz, 2 MW ITER gyrotron has achieved 1.8 MW at 28% efficiency (without depressed collector). Design studies for a 4 MW 170 GHz coaxial-cavity gyrotron with two synthetic diamond output windows and two 2 MW mm-wave output beams for future fusion reactors are currently being performed at FZK. The availability of sources with fast frequency tunability (several GHz s{sup -1}, tuning in 1.5-2.5% steps for about ten different frequencies) would permit the use of a simple, fixed, non-steerable mirror antenna for local current drive (ECCD) experiments and plasma stabilization. GYCOM in Russia develops in collaboration with IPP Garching and FZK an

  4. 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.

  5. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    NASA Astrophysics Data System (ADS)

    Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.

    2001-05-01

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si

  6. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, George P.

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  7. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1987-02-20

    A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

  8. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  9. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  10. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    NASA Astrophysics Data System (ADS)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  11. Thin blanket design for MINIMARS - A compact tandem mirror fusion reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Sawan, M.E.; El-Guebaly, L.A.; Wittenberg, L.J.; Corradini, M.L.; Vogelsang, W.F.; Kulcinski, G.L.

    1986-11-01

    Recent fusion power reactor designs have shown a trend toward lower power, lower cost, higher mass utilization compact configurations with inherent safety, in order to improve the economic aspects of fusion and make them more competitive with other energy sources. Since the blanket thickness directly impacts the size and mass of the remaining reactor components, it is prudent to minimize its thickness while ensuring adequate neutronic and thermal performance. This paper describes the blanket for the MINI-MARS compact tandem mirror fusion power reactor. The blanket which utilizes HT-9 ferritic steel structure, LiPb breeder, Be multiplier/moderator and He gas cooling is only 17 cm thick and is backed up by a steel reflector. Helium gas cools the blanket and reflector in series and the outlet temperature of 575/sup 0/C gives a gross thermal power cycle efficiency of 42.7%.

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

    SciTech Connect

    Not Available

    1990-01-01

    This report discusses the following topics: overview of titan-2 design; titan-2 fusion-power-core engineering; titan-2 divertor engineering; titan-2 tritium systems; titan-2 safety design and radioactive-waste disposal; and titan-2 maintenance procedures.

  13. Dense Hypervelocity Plasma Jets for Fusion Applications

    NASA Astrophysics Data System (ADS)

    Witherspoon, F. Douglas; Thio, Y. C. Francis

    2005-10-01

    High velocity dense plasma jets are being developed for a variety of fusion applications, including refueling, disruption mitigation, High Energy Density Plasmas, magnetized target/magneto-inertial fusion, injection of angular momentum into centrifugally confined mirrors, and others. The technical goal is to accelerate plasma blobs of density >10^17 cm-3 and total mass >100 micrograms to velocities >200 km/s. The approach utilizes symmetrical injection of very high density plasma into a coaxial EM accelerator having a tailored cross-section that prevents formation of the blow-by instability. AFRL MACH2 modeling identified 2 electrode configurations that produce the desired plasma jet parameters. The injected plasma is generated by up to 64 radially oriented capillary discharges arranged uniformly around the circumference of an angled annular injection section. Initial experimental results are presented in which 8 capillaries are fired in parallel with jitter of ˜100 ns. Current focus is on higher voltage operation to reduce jitter to a few 10's of ns, and development of a suite of optical and spectroscopic plasma diagnostics.

  14. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  15. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  16. SAFIRE: A systems analysis code for ICF (inertial confinement fusion) reactor economics

    SciTech Connect

    McCarville, T.J.; Meier, W.R.; Carson, C.F.; Glasgow, B.B.

    1987-01-12

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code incorporates analytical models for scaling the cost and performance of several inertial confinement fusion reactor concepts for electric power. The code allows us to vary design parameters (e.g., driver energy, chamber pulse rate, net electric power) and evaluate the resulting change in capital cost of power plant and the busbar cost of electricity. The SAFIRE code can be used to identify the most attractive operating space and to identify those design parameters with the greatest leverage for improving the economics of inertial confinement fusion electric power plants.

  17. Fusion reactor materials: Semiannual progress report for the period ending March 31, 1987

    SciTech Connect

    none,

    1987-09-01

    This is the second in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities in the following areas: (1) Alloy Development for Irradiation Performance; (2) Damage Analysis and Fundamental Studies; and (3) Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. Separate analytics were prepared for the reports in this volume.

  18. Multidimensional optimization of fusion reactors using heterogenous codes and engineering software

    NASA Astrophysics Data System (ADS)

    Hartwig, Zachary; Olynyk, Geoffrey; Whyte, Dennis

    2012-10-01

    Magnetic confinement fusion reactors are tightly coupled systems. The parameters under a designer's control, such as magnetic field, wall temperature, and blanket thickness, simultaneously affect the behavior, performance, and components of the reactor, leading to complex tradeoffs and design optimizations. In addition, the engineering analyses require non-trivial, self-consistent inputs, such as reactor geometry, to ensure high fidelity between the various physics and engineering design codes. We present a framework for analysis and multidimensional optimization of fusion reactor systems based on the coupling of heterogeneous codes and engineering software. While this approach is widely used in industry, most code-coupling efforts in fusion have been focused on plasma and edge physics. Instead, we use a simplified plasma model to concentrate on how fusion neutrons and heat transfer affect the design of the first wall, breeding blanket, and magnet systems. The framework combines solid modeling, neutronics, and engineering multiphysics codes and software, linked across Windows and Linux clusters. Initial results for optimizing the design of a compact, high-field tokamak reactor based on high-temperature demountable superconducting coils and a liquid blanket are presented.

  19. Application of image fusion techniques in DSA

    NASA Astrophysics Data System (ADS)

    Ye, Feng; Wu, Jian; Cui, Zhiming; Xu, Jing

    2007-12-01

    Digital subtraction angiography (DSA) is an important technology in both medical diagnoses and interposal therapy, which can eliminate the interferential background and give prominence to blood vessels by computer processing. After contrast material is injected into an artery or vein, a physician produces fluoroscopic images. Using these digitized images, a computer subtracts the image made with contrast material from a series of post injection images made without background information. By analyzing the characteristics of DSA medical images, this paper provides a solution of image fusion which is in allusion to the application of DSA subtraction. We fuse the images of angiogram and subtraction, in order to obtain the new image which has more data information. The image that fused by wavelet transform can display the blood vessels and background information clearly, and medical experts gave high score on the effect of it.

  20. Development of the cascade inertial-confinement-fusion reactor

    SciTech Connect

    Pitts, J.H.

    1985-04-15

    Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670/sup 0/K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.

  1. Development of the cascade inertial-confinement-fusion reactor

    SciTech Connect

    Pitts, J.H.

    1985-07-01

    Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.

  2. Adapting computational optimization concepts from aeronautics to nuclear fusion reactor design

    NASA Astrophysics Data System (ADS)

    Dekeyser, W.; Reiter, D.; Baelmans, M.

    2012-10-01

    Even on the most powerful supercomputers available today, computational nuclear fusion reactor divertor design is extremely CPU demanding, not least due to the large number of design variables and the hybrid micro-macro character of the flows. Therefore, automated design methods based on optimization can greatly assist current reactor design studies. Over the past decades, "adjoint methods" for shape optimization have proven their virtue in the field of aerodynamics. Applications include drag reduction for wing and wing-body configurations. Here we demonstrate that also for divertor design, these optimization methods have a large potential. Specifically, we apply the continuous adjoint method to the optimization of the divertor geometry in a 2D poloidal cross section of an axisymmetric tokamak device (as, e.g., JET and ITER), using a simplified model for the plasma edge. The design objective is to spread the target material heat load as much as possible by controlling the shape of the divertor, while maintaining the full helium ash removal capabilities of the vacuum pumping system.

  3. Models and analyses for inertial-confinement fusion-reactor studies

    SciTech Connect

    Bohachevsky, I.O.

    1981-05-01

    This report describes models and analyses devised at Los Alamos National Laboratory to determine the technical characteristics of different inertial confinement fusion (ICF) reactor elements required for component integration into a functional unit. We emphasize the generic properties of the different elements rather than specific designs. The topics discussed are general ICF reactor design considerations; reactor cavity phenomena, including the restoration of interpulse ambient conditions; first-wall temperature increases and material losses; reactor neutronics and hydrodynamic blanket response to neutron energy deposition; and analyses of loads and stresses in the reactor vessel walls, including remarks about the generation and propagation of very short wavelength stress waves. A discussion of analytic approaches useful in integrations and optimizations of ICF reactor systems concludes the report.

  4. Plasma Heating and Current Drive for Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Holtkamp, Norbert

    2010-02-01

    ITER (in Latin ``the way'') is designed to demonstrate the scientific and technological feasibility of fusion energy. Fusion is the process by which two light atomic nuclei combine to form a heavier one and thus release energy. In the fusion process two isotopes of hydrogen - deuterium and tritium - fuse together to form a helium atom and a neutron. Thus fusion could provide large scale energy production without greenhouse effects; essentially limitless fuel would be available all over the world. The principal goals of ITER are to generate 500 megawatts of fusion power for periods of 300 to 500 seconds with a fusion power multiplication factor, Q, of at least 10. Q >= 10 (input power 50 MW / output power 500 MW). In a Tokamak the definition of the functionalities and requirements for the Plasma Heating and Current Drive are relevant in the determination of the overall plant efficiency, the operation cost of the plant and the plant availability. This paper summarise these functionalities and requirements in perspective of the systems under construction in ITER. It discusses the further steps necessary to meet those requirements. Approximately one half of the total heating will be provided by two Neutral Beam injection systems at with energy of 1 MeV and a beam power of 16 MW into the plasma. For ITER specific test facility is being build in order to develop and test the Neutral Beam injectors. Remote handling maintenance scheme for the NB systems, critical during the nuclear phase of the project, will be developed. In addition the paper will give an overview over the general status of ITER. )

  5. Magnetohydrodynamic Vapor Explosions: A Study with Potential Interest to the Safety of Fusion Reactor Project

    NASA Astrophysics Data System (ADS)

    Arias, F. J.

    2010-04-01

    In this paper, the possibility of vapor explosions in superheat liquids in presence of a magnetic field that undergo sudden variation of magnetic field is discussed. This possible phenomenon may play a very important role in the blanket design of future fusion reactors, where transients in magnetic field on liquid metals, could will be a potential hazard for safety.

  6. Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa; Yapıcı, Hüseyin

    2008-09-01

    ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium-tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li2BeF4 as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF4 or UF4) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S 8- P 3 approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF4 or ThF4 and 4% UF4. In addition, higher energy multiplication values were found for the salt with UF4 compared to that with ThF4. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.

  7. Fusion reactor materials semiannual progress report for the period ending March 31, 1991

    SciTech Connect

    none,

    1991-07-01

    This is the tenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: alloy development for irradiation performance; damage analysis and fundamental studies; special purpose materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of program participants, and to provide a means of communicating the efforts of materials scientists to the test of the fusion community, both nationally and worldwide.

  8. Fusion reactor materials semiannual progress report for the period ending September 30, 1988

    SciTech Connect

    none,

    1989-04-01

    This paper discusses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  9. Compatibility of materials for use in liquid-metal blankets of fusion reactors

    SciTech Connect

    Chopra, O.K.; Tortorelli, P.F.

    1983-11-01

    A review of corrosion and environmental effects on the mechanical properties of austenitic and ferritic steels for use with liquid metals in fusion reactors is presented. The mechanisms and kinetics of the corrosion processes in liquid lithium and Pb-17Li systems are examined and their influence on degradation of structural material is discussed. Requirements for additional data are identified.

  10. Fusion Reactor Materials semiannual progress report for the period ending March 31, 1992

    SciTech Connect

    Not Available

    1992-07-01

    This is the twelfth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; and Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  11. Fusion reactor materials: Semiannual progress report for period ending September 30, 1987

    SciTech Connect

    none,

    1988-03-01

    This is the third in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: Alloy Development for Irradiation Performances; Damage Analysis and Fundamental Studies; Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  12. Fusion reactor materials semiannual progress report for the period ending September 30, 1989

    SciTech Connect

    none,

    1989-01-01

    This is the seventh in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: alloy development for irradiation performance, damage analysis and fundamental studies, and special purpose materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  13. 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.

  14. Fusion reactor materials semiannual progress report for period ending September 30, 1990

    SciTech Connect

    Not Available

    1991-04-01

    This is the ninth in series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following technical progress reports: Alloy Development of Irradiation Performance; Damage Analysis and Fundamental Studies; and Special Purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide.

  15. Modular control of fusion power heating applications

    SciTech Connect

    Demers, D. R.

    2012-08-24

    This work is motivated by the growing demand for auxiliary heating on small and large machines worldwide. Numerous present and planned RF experiments (EBW, Lower Hybrid, ICRF, and ECH) are increasingly complex systems. The operational challenges are indicative of a need for components of real-time control that can be implemented with a moderate amount of effort in a time- and cost-effective fashion. Such a system will improve experimental efficiency, enhance experimental quality, and expedite technological advancements. The modular architecture of this control-suite serves multiple purposes. It facilitates construction on various scales from single to multiple controller systems. It enables expandability of control from basic to complex via the addition of modules with varying functionalities. It simplifies the control implementation process by reducing layers of software and electronic development. While conceived with fusion applications in mind, this suite has the potential to serve a broad range of scientific and industrial applications. During the Phase-I research effort we established the overall feasibility of this modular control-suite concept. We developed the fundamental modules needed to implement open-loop active-control and demonstrated their use on a microwave power deposition experiment.

  16. Status of beryllium development for fusion applications

    SciTech Connect

    Billone, M.C.; Donne, M.D.; Macaulay-Newcombe, R.G.

    1994-05-01

    Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma facing component of first wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold isostatic pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation.

  17. Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

    SciTech Connect

    Maziasz, P.J.

    1983-01-01

    Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation.

  18. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

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

    NASA Astrophysics Data System (ADS)

    McGuire, K. M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J. W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, Cris W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M.; Bell, M. G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C. S.; Cheng, C. Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D. S.; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P. C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Fujita, T.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Hawryluk, R. J.; Heidbrink, W.; Herrmann, H. W.; Hill, K. W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Loughlin, M. J.; Machuzak, J.; Majeski, R.; Mansfield, D. K.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M. P.; McChesney, J.; McCormack, B.; McCune, D. C.; McKee, G.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Parks, P.; Paul, S. F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Scillia, R.; Scott, S. D.; Semenov, I.; Senko, T.; Sesnic, S.; Sissingh, R.; Skinner, C. H.; Snipes, J.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Tighe, W.; Timberlake, J. R.; Tobita, K.; Towner, H. H.; Tuszewski, M.; von Halle, A.; Vannoy, C.; Viola, M.; von Goeler, S.; Voorhees, D.; Walters, R. T.; Wester, R.; White, R.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Winston, J.; Wright, K.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zavereev, V.; Zweben, S. J.

    1995-06-01

    After many years of fusion research, the conditions needed for a D-T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D-T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m-3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D-T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed

  20. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 92--94). Final report

    SciTech Connect

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1994-11-01

    This is the Final Report for a three-year (FY 92--94) study of the Environmental, Safety, and Economic (ESE) aspects of fusion energy systems, emphasizing development of computerized approaches suitable for incorporation as modules in fusion system design codes. First, as is reported in Section 2, the authors now have operating a simplified but complete environment and safety evaluation code, BESAFE. The first tests of BESAFE as a module of the SUPERCODE, a design optimization systems code at LLNL, are reported in Section 3. Secondly, as reported in Section 4, the authors have maintained a strong effort in developing fast calculational schemes for activation inventory evaluation. In addition to these major accomplishments, considerable progress has been made on research on specific topics as follows. A tritium modeling code TRIDYN was developed in collaboration with the TSTA group at LANL and the Fusion Nuclear Technology group at UCLA. A simplified algorithm has been derived to calculate the transient temperature profiles in the blanket during accidents. The scheme solves iteratively a system of non-linear ordinary differential equations describing about 10 regions of the blanket by preserving energy balance. The authors have studied the physics and engineering aspects of divertor modeling for safety applications. Several modifications in the automation and characterization of environmental and safety indices have been made. They have applied this work to the environmental and safety comparisons of stainless steel with alternative structural materials for fusion reactors. A methodology in decision analysis utilizing influence and decision diagrams has been developed to model fusion reactor design problems. Most of the work during this funding period has been reported in 26 publications including theses, journal publications, conference papers, and technical reports, as listed in Section 11.

  1. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    Redi, M.H.; Batha, S.H.; Budny, R.V.; Darrow, D.S.; Levinton, F.M.; McCune, D.C.; Medley, S.S.; Petrov, M.P.; von Goeler, S.; White, R.B.; Zarnstorff, M.C.; Zweben, S.J.; TFTR Team

    1999-07-01

    Neoclassical simulations of alpha particle density profiles in high fusion power plasmas on the Tokamak Fusion Test Reactor [Phys. Plasmas {bold 5}, 1577 (1998)] are found to be in good agreement with measurements of the alpha distribution function made with a sensitive active neutral particle diagnostic. The calculations are carried out in Hamiltonian magnetic coordinates with a fast, particle-following Monte Carlo code which includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss and collisional effects. New calculations show that monotonic shear alpha particles are virtually unaffected by toroidal field ripple. The calculations show that in reversed shear the confinement domain is not empty for trapped alphas at birth and allow an estimate of the actual alpha particle densities measured with the pellet charge exchange diagnostic. {copyright} {ital 1999 American Institute of Physics.}

  2. Advances in Multi-Sensor Data Fusion: Algorithms and Applications

    PubMed Central

    Dong, Jiang; Zhuang, Dafang; Huang, Yaohuan; Fu, Jingying

    2009-01-01

    With the development of satellite and remote sensing techniques, more and more image data from airborne/satellite sensors have become available. Multi-sensor image fusion seeks to combine information from different images to obtain more inferences than can be derived from a single sensor. In image-based application fields, image fusion has emerged as a promising research area since the end of the last century. The paper presents an overview of recent advances in multi-sensor satellite image fusion. Firstly, the most popular existing fusion algorithms are introduced, with emphasis on their recent improvements. Advances in main applications fields in remote sensing, including object identification, classification, change detection and maneuvering targets tracking, are described. Both advantages and limitations of those applications are then discussed. Recommendations are addressed, including: (1) Improvements of fusion algorithms; (2) Development of “algorithm fusion” methods; (3) Establishment of an automatic quality assessment scheme. PMID:22408479

  3. Evaluation of hot isostatic pressing for joining of fusion reactor structural components

    NASA Astrophysics Data System (ADS)

    Ivanov, A. D.; Sato, S.; Le Marois, G.

    2000-12-01

    Hot isostatic pressing (HIP) is a promising technology to fabricate the blanket structure of fusion reactors. HIP joining of solid materials has been selected as a reference fabrication method for the shielding blanket/first wall of the international thermonuclear experimental reactor (ITER). On the basis of experimental results obtained in Europe, Japan and Russia, an evaluation of HIP joining for fusion reactor structural components has been carried out. The parameters of HIP fabrication for copper alloys and stainless steels are given. The results of microscopic observations, X-ray microanalysis, tensile, impact toughness, fracture toughness and fatigue tests are presented. Material science criteria for an estimation of quality for joints fabricated by HIP are discussed.

  4. Design of spectrally tailored fusion reactor materials experiments in the HFIR RB∗ capsule irradiation facility

    NASA Astrophysics Data System (ADS)

    Longest, A. W.; Corum, J. E.; Heatherly, D. W.; Thoms, K. R.

    1988-07-01

    Design of four spectrally tailored capsules to irradiate magnetic fusion energy (MFE) materials specimens in the new High Flux Isotope Reactor (HFIR) removable beryllium (RB∗) facility at temperatures of 60, 200, 330 and 400°C, is described. These designs incorporate provisions for removal, examination and reencapsulation of the specimens at intermediate exposure levels en route to a target exposure level of 30 displacements per atom (dpa). With the exception of the 60°C capsule, where the test specimens are in direct contact with the reactor cooling water, the specimen temperatures (monitored by 21 thermocouples) are controlled by varying the thermal conductance of a small gap region between the specimen holder and the containment tube. Hafnium shields are used to tailor the neutron spectrum to closely match the helium production-to-atom displacement ratio ( {14 appm}/{dpa}) expected in an austenitic stainless steel fusion reactor first wall. These MFE capsule irradiations will begin in early 1988.

  5. A revaluation of helium/dpa ratios for fast reactor and thermal reactor data in fission-fusion correlations

    SciTech Connect

    Garner, F.A.; Greenwood, L.R.; Oliver, B.M.

    1996-10-01

    For many years it has been accepted that significant differences exist in the helium/dpa ratios produced in fast reactors and various proposed fusion energy devices. In general, the differences arise from the much larger rate of (n,{alpha}) threshold reactions occurring in fusion devices, reactions which occur for energies {ge} 6 MeV. It now appears, however, that for nickel-containing alloys in fast reactors the difference may not have been as large as was originally anticipated. In stainless steels that have a very long incubation period for swelling, for instance, the average helium concentration over the duration of the transient regime have been demonstrated in an earlier paper to be much larger in the FFTF out-of-core regions than first calculated. The helium/dpa ratios in some experiments conducted near the core edge or just outside of the FFTF core actually increase strongly throughout the irradiation, as {sup 59}Ni slowly forms by transmutation of {sup 58}Ni. This highly exothermic {sup 59}Ni(n,{alpha}) reaction occurs in all fast reactors, but is stronger in the softer spectra of oxide-fueled cores such as FFTF and weaker in the harder spectra of metal-fueled cores such as EBR-II. The formation of {sup 59}Ni also increases strongly in out-of-core unfueled regions where the reactor spectra softens with distance from the core.

  6. Transmutation behaviour of Eurofer under irradiation in the IFMIF test facility and fusion power reactors

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Simakov, S. P.; Wilson, P. P. H.

    2004-08-01

    The transmutation behaviour of the low activation steel Eurofer was analysed for irradiation simulations in the high flux test module (HFTM) of the International Fusion Material Irradiation Facility (IFMIF) neutron source and the first wall of a typical fusion power reactor (FPR) employing helium cooled lithium lead (HCLL) and pebble bed (HCPB) blankets. The transmutation calculations were conducted with the analytical and laplacian adaptive radioactivity analysis (ALARA) code and IEAF-2001 data for the IFMIF and the EASY-2003 system for the fusion power reactor (FPR) irradiations. The analyses showed that the transmutation of the main constituents of Eurofer, including iron and chromium, is not significant. Minor constituents such as Ti, V and Mn increase by 5-15% per irradiation year in the FPR and by 10-35% in the IFMIF HFTM. Other minor constituents such as B, Ta, and W show a different transmutation behaviour resulting in different elemental compositions of the Eurofer steel after high fluence irradiations in IFMIF and fusion power reactors.

  7. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited)

    SciTech Connect

    Smith, Roger J.

    2008-10-15

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B{sub pol} diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T{sub e}, n{sub e}, and B{sub ||} along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n{sub e}B{sub ||} product and higher n{sub e} and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  8. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited).

    PubMed

    Smith, Roger J

    2008-10-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B(pol) diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T(e), n(e), and B(parallel) along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n(e)B(parallel) product and higher n(e) and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means. PMID:19044521

  9. Materials for the plasma-facing components of fusion reactors

    NASA Astrophysics Data System (ADS)

    Bolt, H.; Barabash, V.; Krauss, W.; Linke, J.; Neu, R.; Suzuki, S.; Yoshida, N.; ASD. E. X. Upgrade Team

    2004-08-01

    During reactor operation the plasma-facing materials have to fulfil very complex and sometimes contradicting requirements. At present, tungsten shows the highest promise as plasma-facing material. Experiments in the ASDEX Upgrade tokamak indicate that plasma operation is feasible with walls and divertor surfaces mostly covered with tungsten. Thick tungsten coatings have been deposited by plasma spraying on EUROFER first wall mock-ups and show good adhesion and stability. The performance of tungsten surfaces under intense transient thermal loads is another critical issue, since the formation of a melt layer may favour the generation of highly activated dust particles. Work on `nanocrystalline' tungsten shall improve the mechanical properties under neutron irradiation which is especially important for designs, where tungsten has also to fulfil structural functions. Alternative divertor heat sink materials with very high thermal conductivity like SiC-fibre reinforced copper composites are presently being developed and should allow operation at reactor relevant coolant temperatures.

  10. Radiation-induced electrical breakdown of helium in fusion reactor superconducting magnet systems

    SciTech Connect

    Perkins, L.J.

    1983-12-02

    A comprehensive theoretical study has been performed on the reduction of the electrical breakdown potential of liquid and gaseous helium under neutron and gamma radiation. Extension of the conventional Townsend breakdown theory indicates that radiation fields at the superconducting magnets of a typical fusion reactor are potentially capable of significantly reducing currently established (i.e., unirradiated) helium breakdown voltages. Emphasis is given to the implications of these results including future deployment choices of magnet cryogenic methods (e.g., pool-boiling versus forced-flow), the possible impact on magnet shielding requirements and the analogous situation for radiation-induced electrical breakdown in fusion RF transmission systems.

  11. Application of Reactor Antineutrinos: Neutrinos for Peace

    NASA Astrophysics Data System (ADS)

    Suekane, F.

    2013-02-01

    In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

  12. Multi-Application Small Light Water Reactor

    SciTech Connect

    Pierre Babka

    2002-10-31

    The Multi-Application Small Light Water Reactor (MASLWR ) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objective was to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility.

  13. Polarized Nuclei in a Simple Mirror Fusion Reactor

    NASA Technical Reports Server (NTRS)

    Noever, David A.

    1995-01-01

    The possibility of enhancing the ratio of output to input power Q in a simple mirror machine by polarizing Deuterium-Tritium (D- T) nuclei is evaluated. Taking the Livermore mirror reference design mirror ratio of 6.54, the expected sin(sup 2) upsilon angular distribution of fusion decay products reduces immediate losses of alpha particles to the loss cone by 7.6% and alpha-ion scattering losses by approx. 50%. Based on these findings, alpha- particle confinement times for a polarized plasma should therefore be 1.11 times greater than for isotropic nuclei. Coupling this enhanced alpha-particle heating with the expected greater than 50% D- T reaction cross section, a corresponding power ratio for polarized nuclei, Q(sub polarized), is found to be 1.63 times greater than the classical unpolarized value Q(sub classical). The effects of this increase in Q are assessed for the simple mirror.

  14. Synfuels from fusion: producing hydrogen with the tandem mirror reactor and thermochemical cycles

    SciTech Connect

    Ribe, F.L.; Werner, R.W.

    1981-01-21

    This report examines, for technical merit, the combination of a fusion reactor driver and a thermochemical plant as a means for producing synthetic fuel in the basic form of hydrogen. We studied: (1) one reactor type - the Tandem Mirror Reactor - wishing to use to advantage its simple central cell geometry and its direct electrical output; (2) two reactor blanket module types - a liquid metal cauldron design and a flowing Li/sub 2/O solid microsphere pellet design so as to compare the technology, the thermal-hydraulics, neutronics and tritium control in a high-temperature operating mode (approx. 1200 K); (3) three thermochemical cycles - processes in which water is used as a feedstock along with a high-temperature heat source to produce H/sub 2/ and O/sub 2/.

  15. A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

    NASA Technical Reports Server (NTRS)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1998-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

  16. A spherical torus nuclear fusion reactor space propulsion vehicle concept for fast interplanetary travel

    NASA Astrophysics Data System (ADS)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1999-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a>5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including payload, central truss, nuclear reactor (including diverter and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, and component design.

  17. Development of new generation reduced activation ferritic-martensitic steels for advanced fusion reactors

    NASA Astrophysics Data System (ADS)

    Tan, L.; Snead, L. L.; Katoh, Y.

    2016-09-01

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ∼500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimental results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. The strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9-20Cr oxide dispersion-strengthened ferritic alloys.

  18. Development of new generation reduced activation ferritic-martenstic steels for advanced fusion reactors

    DOE PAGESBeta

    Tan, Lizhen; Snead, Lance Lewis; Katoh, Yutai

    2016-05-26

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ~500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimentalmore » results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. Furthermore, the strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9–20Cr oxide dispersion-strengthened ferritic alloys.« less

  19. Heavy ion beam transport in an inertial confinement fusion reactor

    SciTech Connect

    Barboza, N.

    1995-08-01

    A new code, bimc, is under development to determine if a beam of heavy ions can be focused to the necessary spot-size radius of about 2 mm within an inertial confinement reactor chamber where the background gas densities are on the order of 10{sup 14}--10{sup 15} cm{sup {minus}3} Lithium (or equivalent). Beam transport is expected to be strongly affected by stripping and collective plasma phenomena; however, if propagation is possible in this regime, it could lead to simplified reactor designs. The beam is modeled using a 2 1/2 D particle-in-cell (PIC) simulation code coupled with a Monte Carlo (MC) method for analyzing collisions. The MC code follows collisions between the beam ions and neutral background gas atoms that account for the generation of electrons and background gas ions (ionization), and an increase of the charge state of the beam ions (stripping). The PIC code models the complete dynamics of the interaction of the various charged particle species with the self generated electromagnetic fields. Details of the code model and preliminary results are presented.

  20. 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.

  1. Disassembly of the fusion-1 capsule after irradiation in the BOR-60 reactor

    SciTech Connect

    Tsai, H.; Kazakov, V.A.; Chakin, V.P.

    1997-04-01

    A U.S./Russia (RF) collaborative irradiation experiment, Fusion-1, was completed in June 1996 after reaching a peak exposure of {approx}17 dpa in the BOR-60 fast reactor at the Research Institute of Atomic Reactors (RIAR) in Russia. The specimens were vanadium alloys, mainly of recent heats from both countries. In this reporting period, the capsule was disassembled at the RIAR hot cells and all test specimens were successfully retrieved. For the disassembly, an innovative method of using a heated diffusion oil to melt and separate the lithium bond from the test specimens was adopted. This method proved highly successful.

  2. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    SciTech Connect

    Monsler, M.J.; Meier, W.R.

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate.

  3. The VISTA spacecraft: Advantages of ICF (Inertial Confinement Fusion) for interplanetary fusion propulsion applications

    SciTech Connect

    Orth, C.D.; Klein, G.; Sercel, J.; Hoffman, N.; Murray, K.; Chang-Diaz, F.

    1987-10-02

    Inertial Confinement Fusion (ICF) is an attractive engine power source for interplanetary manned spacecraft, especially for near-term missions requiring minimum flight duration, because ICF has inherent high power-to-mass ratios and high specific impulses. We have developed a new vehicle concept called VISTA that uses ICF and is capable of round-trip manned missions to Mars in 100 days using A.D. 2020 technology. We describe VISTA's engine operation, discuss associated plasma issues, and describe the advantages of DT fuel for near-term applications. Although ICF is potentially superior to non-fusion technologies for near-term interplanetary transport, the performance capabilities of VISTA cannot be meaningfully compared with those of magnetic-fusion systems because of the lack of a comparable study of the magnetic-fusion systems. We urge that such a study be conducted.

  4. The VISTA spacecraft: Advantages of ICF (Inertial Confinement Fusion) for interplanetary fusions propulsion applications

    NASA Technical Reports Server (NTRS)

    Orth, Charles D.; Klein, Gail; Sercel, Joel; Hoffman, Nate; Murray, Kathy; Chang-Diaz, Franklin

    1987-01-01

    Inertial Confinement Fusion (ICF) is an attractive engine power source for interplanetary manned spacecraft, especially for near-term missions requiring minimum flight duration, because ICF has inherent high power-to-mass ratios and high specific impulses. We have developed a new vehicle concept called VISTA that uses ICF and is capable of round-trip manned missions to Mars in 100 days using A.D. 2020 technology. We describe VISTA's engine operation, discuss associated plasma issues, and describe the advantages of DT fuel for near-term applications. Although ICF is potentially superior to non-fusion technologies for near-term interplanetary transport, the performance capabilities of VISTA cannot be meaningfully compared with those of magnetic-fusion systems because of the lack of a comparable study of the magnetic-fusion systems. We urge that such a study be conducted.

  5. Tritium management in a fusion reactor--safety, handling and economical issues--

    SciTech Connect

    Tanabe, Tetsuo

    2009-02-19

    In order to establish a D-T fusion reactor as an energy source, it is not enough to have a DT burning plasma, and economical conversion of fusion energy to electricity and/or heat, a large enough margin of tritium breeding and tritium safety must be simultaneously achieved. In particular, handling of huge amount of tritium needs significant efforts to ensure that the radiation dose of radiological workers and of the public is below the limits specified by the International Commission on Radiological Protection (ICRP). In this paper, after the introduction of tritium as a fuel of DT reactors and as a radioisotope of hydrogen, tritium safety issues in fuel cycle and blanket systems are summarized. In particular, in-vessel tritium inventory, the most important and uncertain tritium safety issue, is discussed in detail.

  6. Tritium production, management and its impact on safety for a D- sup 3 He fusion reactor

    SciTech Connect

    Sze, D.K. ); Herring, S. ); Sawan, M. )

    1991-11-01

    About three percent of the fusion energy produced by a D-{sup 3}He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, {sup 3}He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D-{sup 3}He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs.

  7. Existing and new applications of micropellet injection (MPI) in magnetic fusion

    NASA Astrophysics Data System (ADS)

    Wang, Zhehui; Lunsford, Robert; Mansfield, Dennis K.; Nichols, Jacob H.

    2016-02-01

    > The intense heat and energetic particle fluxes expected in ITER and future magnetic fusion reactors pose prohibitive problems to the design, selection and maintenance of the first wall and divertor. Micropellet injection (MPI) technologies can offer some innovative solutions to the material and extreme heat challenges. Basic physics of micropellet motion, ablation and interactions with high-temperature plasmas and energetic particles are presented first. We then discuss MPI technology options and applications. In addition to plasma diagnostic applications, controlled injection of micropellets of different sizes, velocities and injection frequencies will offer several possibilities: (1) better assessment of the core plasma cooling due to dust produced in situ; (2) better understanding of the plasma-material interaction physics near the wall; (3) new methods for plasma fuelling and impurity control; and (4) techniques for edge cooling with minimal impact on the plasma core. Dedicated small-scale laboratory experiments will complement major fusion experiments in development and applications of MPI.

  8. Existing and new applications of micropellet injection (MPI) in magnetic fusion

    NASA Astrophysics Data System (ADS)

    Wang, Zhehui; Lunsford, Robert; Mansfield, Dennis K.; Nichols, Jacob H.

    2016-04-01

    > The intense heat and energetic particle fluxes expected in ITER and future magnetic fusion reactors pose prohibitive problems to the design, selection and maintenance of the first wall and divertor. Micropellet injection (MPI) technologies can offer some innovative solutions to the material and extreme heat challenges. Basic physics of micropellet motion, ablation and interactions with high-temperature plasmas and energetic particles are presented first. We then discuss MPI technology options and applications. In addition to plasma diagnostic applications, controlled injection of micropellets of different sizes, velocities and injection frequencies will offer several possibilities: (1) better assessment of the core plasma cooling due to dust produced in situ; (2) better understanding of the plasma-material interaction physics near the wall; (3) new methods for plasma fuelling and impurity control; and (4) techniques for edge cooling with minimal impact on the plasma core. Dedicated small-scale laboratory experiments will complement major fusion experiments in development and applications of MPI.

  9. 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}.

  10. Neutron dosimetry qualification experiments for the Tokamak Fusion Test Reactor Lithium Blanket Module program

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderi, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment and the use of such data in neutron spectrum adjustment and in predicting integral performance parameters, e.g., tritium production.

  11. A three-bar model for ratcheting of fusion reactor first wall

    SciTech Connect

    Wolters, J.; Majumdar, S.

    1994-12-01

    First wall structures of fusion reactors are subjected to cyclic bending stresses caused by inhomogeneous temperature distribution during plasma burn cycles and by electromagnetically induced impact loads during plasma disruptions. Such a combination of loading can potentially lead to ratcheting or incremental accumulation of plastic strain with cycles. An elastic-plastic three-bar model is developed to investigate the ratcheting behavior of the first wall.

  12. Manganese-stabilized austenitic stainless steels for fusion applications

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1990-01-01

    An austenitic stainless steel that is comprised of Fe, Cr, Mn, C but no Ni or Nb and minimum N. To enhance strength and fabricability minor alloying additions of Ti, W, V, B and P are made. The resulting alloy is one that can be used in fusion reactor environments because the half-lives of the elements are sufficiently short to allow for handling and disposal.

  13. Manganese-stabilized austenitic stainless steels for fusion applications

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1990-08-07

    An austenitic stainless steel that is comprised of Fe, Cr, Mn, C but no Ni or Nb and minimum N. To enhance strength and fabricability minor alloying additions of Ti, W, V, B and P are made. The resulting alloy is one that can be used in fusion reactor environments because the half-lives of the elements are sufficiently short to allow for handling and disposal.

  14. Neutronic analysis of alternative structural materials for fusion reactor blankets

    NASA Astrophysics Data System (ADS)

    Santos, Raul dos

    1988-07-01

    The neutronic performance of the International Tokamak Reactor (INTOR) blanket was studied when several alternative structural materials were used instead of the INTOR reference structural material, type 316 stainless steel. The alternative structural materials included: ferritic-, vanadium-, titanium-, long range ordered-, manganese austenitic-, and nimonic-alloys. All were treated both with and without a first-wall coating of beryllium or graphite. The tritium breeding ratio, the nuclear heating, and the gas (hydrogen and helium) production rates in the structural materials were calculated for the possible combinations of structural material and first-wall coating. These parameters were compared with those obtained by using SS-316. The nimonic alloy was the only one with worse neutronic performance than the SS-316.

  15. The Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    The complex nonlinear dependencies observed in typical reactor pressure vessel (RPV) material embrittlement data, as well as the inherent large uncertainties and scatter in the radiation embrittlement data, make prediction of radiation embrittlement a difficult task. Conventional statistical and deterministic approaches have only resulted in rather large uncertainties, in part because they do not fully exploit domain-specific mechanisms. The domain models built by researchers in the field, on the other hand, do not fully exploit the statistical and information content of the data. As evidenced in previous studies, it is unlikely that a single method, whether statistical, nonlinear, or domain model, will outperform all others. More generally, considering the complexity of the embrittlement prediction problem, it is highly unlikely that a single best method exists and is tractable, even in theory. In this paper, we propose to combine a number of complementary methods including domain models, neural networks, and nearest neighbor regressions (NNRs). Such a combination of methods has become possible because of recent developments in measurement-based optimal fusers in the area of information fusion. The information fusion technique is used to develop radiation embrittlement prediction models for reactor RPV steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six Cu, Ni, P, neutron fluence, irradiation time, and irradiation-parameters are used in the embrittlement prediction models. The results-temperature indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The

  16. Low power reactor for remote applications

    SciTech Connect

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs.

  17. Low power reactor for remote applications

    NASA Astrophysics Data System (ADS)

    Meier, K. L.; Palmer, R. G.; Kirchner, W. L.

    1985-05-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long term, virtually maintenance free, operation of this reactor for remote applications.

  18. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  19. Activation analysis of the PULSAR-II fusion power reactor

    SciTech Connect

    Khater, H.Y.

    1995-12-31

    The PULSAR-II pulsed tokamak power plant design utilizes a blanket made of the vanadium alloy, V-5Cr-5Ti, and cooled with liquid lithium. The shield is made of a mixture of the low activation austenitic steel (Tenelon) and vanadium. The blanket is assumed to be replaced every 5.6 full power years (FPY) and the shield is assumed to stay in place for 30 FPY. The activity induced in the blanket at the end of its lifetime is higher than the activity induced in the shield after 30 FPY. At shutdown, the blanket and shield activities are 2678 MCi and 1747 MCi, respectively. One year after shutdown the shield activity drops to 18 MCi compared to 84 MCi for the blanket. The total decay heat generated in the blanket at the end of its lifetime is 34.7 MW and drops to 17.6 MW within an hour. At shutdown, 25.3 MW of decay heat are generated in the shield, dropping to only 0.1 MW within the first year. One week after shutdown, the values of the integrated decay heat are 1770 GJ for the blanket and 469 GJ for the shield. The radwaste classification of the reactor structure is evaluated according to both the NRC 10CFR61 and Fetter waste disposal concentration limits. After 5.6 years of irradiation, the blanket will only qualify for Class C low level waste. After 30 years of operation, the shield will also qualify for disposal as Class C waste. Only remote maintenance will be allowed inside the containment building.

  20. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    SciTech Connect

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits.

  1. Surface modifications of fusion reactor relevant materials on exposure to fusion grade plasma in plasma focus device

    NASA Astrophysics Data System (ADS)

    Niranjan, Ram; Rout, R. K.; Srivastava, R.; Chakravarthy, Y.; Mishra, P.; Kaushik, T. C.; Gupta, Satish C.

    2015-11-01

    An 11.5 kJ plasma focus (PF) device was used here to irradiate materials with fusion grade plasma. The surface modifications of different materials (W, Ni, stainless steel, Mo and Cu) were investigated using various available techniques. The prominent features observed through the scanning electron microscope on the sample surfaces were erosions, cracks, blisters and craters after irradiations. The surface roughness of the samples increased multifold after exposure as measured by the surface profilometer. The X-ray diffraction analysis indicated the changes in the microstructures and the structural phase transformation in surface layers of the samples. We observed change in volumes of austenite and ferrite phases in the stainless steel sample. The energy dispersive X-ray spectroscopic analysis suggested alloying of the surface layer of the samples with elements of the PF anode. We report here the comparative analysis of the surface damages of materials with different physical, thermal and mechanical properties. The investigations will be useful to understand the behavior of the perspective materials for future fusion reactors (either in pure form or in alloy) over the long operations.

  2. Magnet operating experience review for fusion applications

    SciTech Connect

    Cadwallader, L.C.

    1991-11-01

    This report presents a review of magnet operating experiences for normal-conducting and superconducting magnets from fusion, particle accelerator, medical technology, and magnetohydrodynamics research areas. Safety relevant magnet operating experiences are presented to provide feedback on field performance of existing designs and to point out the operational safety concerns. Quantitative estimates of magnet component failure rates and accident event frequencies are also presented, based on field experience and on performance of similar components in other industries.

  3. The TITAN Reversed-Field Pinch fusion reactor study: Scoping phase report

    SciTech Connect

    Not Available

    1987-01-01

    The TITAN research program is a multi-institutional effort to determine the potential of the Reversed-Field Pinch (RFP) magnetic fusion concept as a compact, high-power-density, and ''attractive'' fusion energy system from economic (cost of electricity, COE), environmental, and operational viewpoints. In particular, a high neutron wall loading design (18 MW/m/sup 2/) has been chosen as the reference case in order to quantify the issue of engineering practicality, to determine the physics requirements and plasma operating mode, to assess significant benefits of compact systems, and to illuminate the main drawbacks. The program has been divided into two phases, each roughly one year in length: the Scoping Phase and the Design Phase. During the scoping phase, the TITAN design team has defined the parameter space for a high mass power density (MPD) RFP reactor, and explored a variety of approaches to the design of major subsystems. Two major design approaches consistent with high MPD and low COE, the lithium-vanadium blanket design and aqueous loop-in-pool design, have been selected for more detailed engineering evaluation in the design phase. The program has retained a balance in its approach to investigating high MPD systems. On the one hand, parametric investigations of both subsystems and overall system performance are carried out. On the other hand, more detailed analysis and engineering design and integration are performed, appropriate to determining the technical feasibility of the high MPD approach to RFP fusion reactors. This report describes the work of the scoping phase activities of the TITAN program. A synopsis of the principal technical findings and a brief description of the TITAN multiple-design approach is given. The individual chapters on Plasma Physics and Engineering, Parameter Systems Studies, Divertor, Reactor Engineering, and Fusion Power Core Engineering have been cataloged separately.

  4. 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.

  5. Applications and Progress of Dust Injection to Fusion Energy

    SciTech Connect

    Wang Zhehui; Wurden, Glen A.; Mansfield, Dennis K.; Roquemore, Lane A.; Ticos, Catalin M.

    2008-09-07

    Three regimes of dust injection are proposed for different applications to fusion energy. In the 'low-speed' regime (<5 km/s), basic dust transport study, edge plasma diagnostics, edge-localized-mode (ELM) pacing in magnetic fusion devices can be realized by injecting dust of known properties into today's fusion experiments. ELM pacing, as an alternative to mini-pellet injection, is a promising scheme to prevent disruptions and type I ELM's that can cause catastrophic damage to fusion devices. Different schemes are available to inject dust. In the 'intermediate-speed' regime (10-200 km/s), possible applications of dust injection include fueling of the next-step fusion devices, core-diagnostics of the next-step fusion devices, and compression of plasma and solid targets to aid fusion energy production. Promising laboratory results of dust moving at 10-50 km/s do exist. Significant advance in this regime may be expected in the near term to achieve higher dust speeds. In the 'high-speed' regime (>500 km/s), dust injection can potentially be used to directly produce fusion energy through impact. Ideas on how to achieve these extremely high speeds are mostly on paper. No plan exists today to realize them in laboratory. Some experimental results, including electrostatic, electromagnetic, gas-dragged, plasma-dragged, and laser-ablation-based acceleration, are summarized and compared. Some features and limitations of the different acceleration methods will be discussed. A necessary component of all dust injectors is the dust dropper (also known as dust dispenser). A computer-controlled piezoelectric crystals has been developed to dropped dust in a systematic and reproducible manner. Particle fluxes ranges from a few tens of particles per second up to thousands of particles per second by this simple device.

  6. Repair welding of fusion reactor components. Second year technical report

    SciTech Connect

    Chin, B.A.

    1993-05-15

    Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

  7. Radiation damage in ferritic/martensitic steels for fusion reactors: a simulation point of view

    NASA Astrophysics Data System (ADS)

    Schäublin, R.; Baluc, N.

    2007-12-01

    Low activation ferritic/martensitic steels are good candidates for the future fusion reactors, for, relative to austenitic steels, their lower damage accumulation and moderate swelling under irradiation by the 14 MeV neutrons produced by the fusion reaction. Irradiation of these steels, e.g. EUROFER97, is known to produce hardening, loss of ductility, shift in ductile to brittle transition temperature and a reduction of fracture toughness and creep resistance starting at the lowest doses. Helium, produced by transmutation by the 14 MeV neutrons, is known to impact mechanical properties, but its effect at the microstructure level is still unclear. The mechanisms underlying the degradation of mechanical properties are not well understood, despite numerous studies on the evolution of the microstructure under irradiation. This impedes our ability to predict materials' behaviour at higher doses for use in the future fusion reactors. Simulations of these effects are now essential. An overview is presented on molecular dynamics simulations of the primary state of damage in iron and of the mobility of a dislocation, vector of plasticity, in the presence of a defect.

  8. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Fisch, Nathaniel J.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

  9. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Bers, Abraham

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

  10. 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.

  11. Transmutation and activation analysis for divertor materials in a HCLL-type fusion power reactor

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Pereslavtsev, P.; Möslang, A.; Rieth, M.

    2009-04-01

    The activation and transmutation of tungsten and tantalum as plasma facing materials was assessed for a helium cooled divertor irradiated in a typical fusion power reactor based on the use of Helium-cooled Lithium Lead (HCLL) blankets. 3D activation calculations were performed by applying a programme system linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. Special attention was given to the proper treatment of the resonance shielding of tungsten and tantalum by using reaction rates provided directly by MCNP on the basis of continuous energy activation cross-section data. It was shown that the long-term activation behaviour is dominated by activation products of the assumed tramp material while the short-term behaviour is due to the activation of the stable Ta and W isotopes. The recycling limit for remote handling of 100 mSv/h can be achieved after decay times of 10 and 50 years for Ta and W, respectively. The elemental transmutation rates of Ta and W were shown to be on a moderate level for the HCLL-type fusion power reactor.

  12. Low-activation properties of novel Cr-based materials for fusion reactors

    NASA Astrophysics Data System (ADS)

    Zucchetti, Massimo; Merola, Mario

    1996-10-01

    Due to the low neutron-induced radioactivity of their main constituent, two novel Cr-based structural materials are examined, from the low-activation properties viewpoint, as candidate first-wall materials for fusion reactors. The two alloys are Cr sbnd 5Fe sbnd 1Y 2O 3, and Cr sbnd 44Fe sbnd 5Al sbnd 0.3Ti sbnd 0.5Y 2O 3. Their low-activation properties have been compared with those of some reference materials, in particular the austenitic steel AISI 316L, the Mn-based reduced-activation steel OPTSTAB, a V sbnd 5Ti alloy and a ceramic matrix composite (SiC/SiC). Impurities and tramp elements have been included in the calculations. Irradiation conditions envisaged for the SEAFP (Safety and Environmental Assessment of Fusion Power) study have been adopted. Advantages of the proposed materials with respect to steels, either reference alloys or reduced-activation ones, appear evident from the results of the activation analysis. Cr sbnd 5Fe sbnd 1Y 2O 3, in particular, show excellent low-activation characteristics. This material can be a good choice for further development and characterization of first-wall and blanket structural materials for fusion power reactors.

  13. Microwave generation for magnetic fusion energy applications

    SciTech Connect

    Antonsen, T.M. Jr.; Destler, W.W.; Granatstein, V.L.; Levush, B.

    1992-01-01

    This progress report encompasses work on two separate projects, both related to developing sources for electron cyclotron resonance heating of magnetic fusion plasmas. The report is therefore divided into two parts as follows: Free electron laser with small period wigglers; and theory and modeling of high frequency, high power gryotron operation. Task A is experimental and eventually aims at developing continuously tunable, cw sources for ECRH with power per unit as high as 5 megawatts. Task B provides gryotron theory and modeling in support of the gryotron development programs at MIT and Varian.

  14. Analysis of Tokamak Fusion Test Reactor (TFTR) Prototype of International Thermonuclear Experimental Reactor (ITER)‡

    NASA Astrophysics Data System (ADS)

    Hester, Tim; Maglich, Bogdan; Scott, Dan; Calsec Collaboration

    2015-11-01

    TFTR produced world record of 10 million watts of controlled fusion power1 (CFP-1994) was summarized in Review1. We present evidence3 that: (1) TFTR input vs. output was 40 to 10 MW i.e. a power loss. (2) Review claims no experimental evidence for thermonuclear CFP production (only a calculation). (3) Ultra-high vacuum (UHV) required for τE = 0.2 s is 10-9 torr. TFTR had no UHV pumps, resulting in 10-3 torr, restricting τE <10-6 s, << thermalization time; 0.1 s., hence DT plasma did not occur. (4) Carbon ions were presented as D-T plasma. (5) Unknown neutron detector on unexplained neutron diamagnetic effect, measured ``fusion neutron power'' without particle energy identification, energy or coincidence. (6) 8 of 9 parameters claimed were inferred not measured. Quadratic test of TFTR data results2 in zero thermonuclear fusion power contribution to 10 MW: SFP = (0 +/- 1)%. ‡ Submitted to Physics of Plasmas†

  15. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  16. Development of intermetallic coatings for fusion power applications

    SciTech Connect

    Park, J.H.; Domenico, T.; Dragel, G.; Clark, R.

    1994-03-01

    In the design of liquid-metal cooling systems, corrosion resistance of structural materials and magnetohydrodynamic (MHD) force and its subsequent influence on thermal hydraulics and corrosion are major concerns. The objective of this study is to develop stable corrosion-resistant electrical insulator coatings at the liquid-metal/structural-material interface, with emphasis on electrically insulating coatings that prevent adverse MHD-generated currents from passing through the structural walls. Vanadium and V-base alloys are potential materials for structural applications in a fusion reactor. Insulator coatings inside the tubing are required when the system is cooled by liquid metals. Various intermetallic films were produced on V, V-t, and V-20 Ti, V-5Cr-t and V-15Cr-t, and Ti, and Types 304 and 316 stainless steel. The intermetallic layers were developed by exposure of the materials to liquid lithium of 3--5 at.% and containing dissolved metallic solutes at temperatures of 416--880{degrees}C. Subsequently, electrical insulator coatings were produced by reaction of the reactive layers with dissolved nitrogen in liquid lithium or by air oxidation under controlled conditions at 600--1000{degrees}C. These reactions converted the intermetallic layers to electrically insulating oxide/nitride or oxy-nitride layers. This coating method could be applied to a commercial product. The liquid metal can be used over and over because only the solutes are consumed within the liquid metal. The technique can be applied to various shapes because the coating is formed by liquid-phase reaction. This paper will discuss initial results on the nature of the coatings and their in-situ electrical resistivity characteristics in liquid lithium at high temperatures.

  17. Sensor fusion with application to electronic warfare

    NASA Astrophysics Data System (ADS)

    Zanzalari, Robert M.; Van Alstine, Edward

    1999-03-01

    The Night Vision and Electronics Sensors Directorate, Survivability/Camouflage, Concealment and Deception Division mission is to provide affordable aircraft and ground electronic sensor/systems and signature management technologies which enhance survivability and lethality of US and International Forces. Since 1992, efforts have been undertaken in the area of Situational Awareness and Dominant Battlespace Knowledge. These include the Radar Deception and Jamming Advanced Technology Demonstration (ATD), Survivability and Targeting System Integration, Integrated Situation Awareness and Targeting ATD, Combat Identification, Ground Vehicle Situational Awareness, and Combined Electronic Intelligence Target Correlation. This paper will address the Situational Awareness process as it relates to the integration of Electronic Warfare (EW) with targeting and intelligence and information warfare systems. Discussion will be presented on the Sensor Fusion, Situation Assessment and Response Management Strategies. Sensor Fusion includes the association, correlation, and combination of data and information from single and multiple sources to achieve refined position and identity estimates, and complete and timely assessments of situations and threats as well as their significance. Situation Assessment includes the process of interpreting and expressing the environmnet based on situation abstract products and information from technical and doctrinal data bases. Finally, Response Management provides the centralized, adaptable control of all renewable and expendable countermeasure assets resulting in optimization of the response to the threat environment.

  18. On the power and size of tokamak fusion pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.; Hugill, J.; Buxton, P. F.

    2015-03-01

    It is generally accepted that the route to fusion power involves large devices of ITER scale or larger. However, we show, contrary to expectations, that for steady state tokamaks operating at fixed fractions of the density and beta limits, the fusion gain, Qfus, depends mainly on the absolute level of the fusion power and the energy confinement, and only weakly on the device size. Our investigations are carried out using a system code and also by analytical means. Further, we show that for the two qualitatively different global scalings that have been developed to fit the data contained in the ITER ELMy H-mode database, i.e. the normally used beta-dependent IPB98y2 scaling and the alternative beta-independent scalings, the power needed for high fusion performance differs substantially, typically by factors of three to four. Taken together, these two findings imply that lower power, smaller, and hence potentially lower cost, pilot plants and reactors than currently envisaged may be possible. The main parameters of a candidate low power (˜180 MW), high Qfus (˜5), relatively small (˜1.35 m major radius) device are given.

  19. Applications of Fusion Energy Sciences Research - Scientific Discoveries and New Technologies Beyond Fusion

    SciTech Connect

    Wendt, Amy; Callis, Richard; Efthimion, Philip; Foster, John; Keane, Christopher; Onsager, Terry; O'Shea, Patrick

    2015-09-01

    Since the 1950s, scientists and engineers in the U.S. and around the world have worked hard to make an elusive goal to be achieved on Earth: harnessing the reaction that fuels the stars, namely fusion. Practical fusion would be a source of energy that is unlimited, safe, environmentally benign, available to all nations and not dependent on climate or the whims of the weather. Significant resources, most notably from the U.S. Department of Energy (DOE) Office of Fusion Energy Sciences (FES), have been devoted to pursuing that dream, and significant progress is being made in turning it into a reality. However, that is only part of the story. The process of creating a fusion-based energy supply on Earth has led to technological and scientific achievements of far-reaching impact that touch every aspect of our lives. Those largely unanticipated advances, spanning a wide variety of fields in science and technology, are the focus of this report. There are many synergies between research in plasma physics, (the study of charged particles and fluids interacting with self-consistent electric and magnetic fields), high-energy physics, and condensed matter physics dating back many decades. For instance, the formulation of a mathematical theory of solitons, solitary waves which are seen in everything from plasmas to water waves to Bose-Einstein Condensates, has led to an equal span of applications, including the fields of optics, fluid mechanics and biophysics. Another example, the development of a precise criterion for transition to chaos in Hamiltonian systems, has offered insights into a range of phenomena including planetary orbits, two-person games and changes in the weather. Seven distinct areas of fusion energy sciences were identified and reviewed which have had a recent impact on fields of science, technology and engineering not directly associated with fusion energy: Basic plasma science; Low temperature plasmas; Space and astrophysical plasmas; High energy density

  20. Dust Combustion Safety Issues for Fusion Applications

    SciTech Connect

    L. C. Cadwallader

    2003-05-01

    This report summarizes the results of a safety research task to identify the safety issues and phenomenology of metallic dust fires and explosions that are postulated for fusion experiments. There are a variety of metal dusts that are created by plasma erosion and disruptions within the plasma chamber, as well as normal industrial dusts generated in the more conventional equipment in the balance of plant. For fusion, in-vessel dusts are generally mixtures of several elements; that is, the constituent elements in alloys and the variety of elements used for in-vessel materials. For example, in-vessel dust could be composed of beryllium from a first wall coating, tungsten from a divertor plate, copper from a plasma heating antenna or diagnostic, and perhaps some iron and chromium from the steel vessel wall or titanium and vanadium from the vessel wall. Each of these elements has its own unique combustion characteristics, and mixtures of elements must be evaluated for the mixture’s combustion properties. Issues of particle size, dust temperature, and presence of other combustible materials (i.e., deuterium and tritium) also affect combustion in air. Combustion in other gases has also been investigated to determine if there are safety concerns with “inert” atmospheres, such as nitrogen. Several coolants have also been reviewed to determine if coolant breach into the plasma chamber would enhance the combustion threat; for example, in-vessel steam from a water coolant breach will react with metal dust. The results of this review are presented here.

  1. Acceleration of compact toruses and fusion applications

    SciTech Connect

    Hartman, C.W.; Eddleman, J.L.; Hammer, J.H.; Logan, B.G.; McLean, H.S.; Molvik, A.W.

    1990-10-11

    The Compact Torus (Spheromak-type) is a near ideal plasma confinement configuration for acceleration. The fields are mostly generated by internal plasma currents, plasma confinement is toroidal, and the compact torus exhibits resiliency and stability in virtue of the ``rugged`` helicity invariant. Based on these considerations we are developing a coaxial rail-gun type Compact Torus Accelerator (CTA). In the CTA, the CT ring is formed between coaxial electrodes using a magnetized Marshall gun, it is quasistatically ``precompressed`` in a conical electrode section for inductive energy storage, it is accelerated in a straight-coaxial electrode section as in a conventional rail-gun, and it is focused to small size and high energy and power density in a final ``focus`` cone section. The dynamics of slow precompression and acceleration have been demonstrated experimentally in the RACE device with results in good agreement with 2-D MHD code calculations. CT plasma rings with 100 {micro}gms mass have been accelerated to 40 Kj kinetic energy at 20% efficiency with final velocity = 1 X 10{sup 8} cm/s (= 5 KeV/H{sup +}). Preliminary focus tests exhibi dynamics of radius compression, deceleration, and bouncing. Compression ratios of 2-3 have been achieved. A scaled-up 10-100 MJ CTA is predicted to achieve a focus radius of several cm to deliver = 30 MJ ring kinetic energy in 5-10 nsec. This is sufficient energy, power, and power density to enable the CTA to act as a high efficiency, low cost ICF driver. Alternatively, the focused CT can form the basis for an magnetically insulated, inertial confinement fusion (MICF) system. Preliminary calculations of these fusion systems will be discussed.

  2. Tritum recovery system from waste water of fusion reactor using CECE and cryogenic-wall thermal diffusion column

    SciTech Connect

    Arita, T.; Yamanishi, T.; Iwai, Y.; Okuno, K.; Kobayashi, N.; Yamamoto, I.

    1996-12-31

    A system for recovery of tritium in water has been proposed. The system is composed of CECE (Combined Electrolysis Chemical Exchange) and CTD (Cryogenic-wall Thermal Diffusion) columns. A design study was carried out for the two cases: the waste water processing in fusion facilities; and the tritium recovery from heavy water in a fission reactor in Japan. The size and power consumption of the system can greatly be reduced by using the CECE column than the system of WD (Water Distillation) columns. The operation and maintenance of the CTD column are quite easier than the CD (Cryogenic Distillation) column. The proposed system would be applicable for some cases such as the waste water processing in tritium facilities, where the processing flow rate is relatively small. 11 refs., 4 figs., 6 tabs.

  3. Images of plasma disruption effects in the Tokamak Fusion Test Reactor

    SciTech Connect

    Maqueda, R.J.; Wurden, G.A.

    1999-02-01

    Fast-framing imaging of visible radiation from magnetically confined plasmas has lately become a useful tool for both machine operation and physics studies. Using an intensified, commercial Kodak Ektapro imaging system, the effects of a plasma disruption were observed in the Tokamak Fusion Test Reactor (TFTR). The high-energy runaway electrons created soon after the disruption collide with the plasma facing components damaging this surface and producing a shower of debris that traverses the toroidal vessel and falls over the inner bumper limiter.

  4. Modifications Made to the MELCOR Code for Analyzing Lithium Fires in Fusion Reactors

    SciTech Connect

    Merrill, Brad Johnson

    2000-04-01

    This report documents initial modifications made to the MELCOR code that allows MELCOR to predict the consequences of lithium spill accidents for evolving fusion reactor designs. These modifications include thermodynamic and transport properties for lithium, and physical models for predicting the rate of reaction of and energy production from the lithium-air reaction. A benchmarking study was performed with this new MELCOR capability. Two lithium-air reaction tests conducted at the Hanford Engineering Development Laboratory (HEDL) were selected for this benchmark study. Excellent agreement was achieved between MELCOR predictions and measured data. Recommendations for modeling lithium fires with MELCOR and for future work in this area are included in this report.

  5. Modifications made to the MELCOR Code for Analyzing Lithium Fires in Fusion Reactors

    SciTech Connect

    B. J. Merrill

    2000-04-01

    This report documents initial modifications made to the MELCOR code that allows MELCOR to predict the consequences of lithium spill accidents for evolving fusion reactor designs. These modifications include thermodynamic and transport properties for lithium, and physical models for predicting the rate of reaction of and energy production from the lithium-air reaction. A benchmarking study was performed with this new MELCOR capability. Two lithium-air reaction tests conducted at the Hanford Engineering Development Laboratory (HEDL) were selected for this benchmark study. Excellent agreement was achieved between MELCOR predictions and measured data. Recommendations for modeling lithium fires with MELCOR and for future work in this area are included in this report.

  6. Liquid metal MHD heat transfer investigations apllied to fusion Tokamak reactor cooling ducts

    NASA Astrophysics Data System (ADS)

    Sviridov, V. G.; Ivochkin, Yu. P.; Razuvanov, N. G.; Zhilin, V. G.; Genin, L. G.; Ivanova, O. N.; Averianov, K. V.

    2003-12-01

    The liquid metal heat transfer experimental investigations were carried out at the joint MPEI-IIHI magnetohydrodynamic (MHD) complex. The united scientific group examines the liquid metal flow in a horizontal heated tube without and under a longitudinal or a transverse magnetic field. Various configurations of the applied heat flux were taken into consideration. All these cases correspond to the flow in the Tokamak fusion reactor blanket or divertor. Temperature fields, temperature fluctuations field, heat transfer intensities were measured. Strong influence of thermogravitational convection was observed in a horizontal heated tube. Depending on the MHD-configuration, magnetic fields (MF) can enhance or weaken this effect. Tables 2, Figs 5, Refs 8.

  7. 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.

  8. MHD plasma physics in rail accelerators for hydrogen-pellet injection in fusion reactors

    SciTech Connect

    Azzerboni, B.; Becherini, G.; Cardelli, E.; Tellini, A.

    1989-06-01

    In this paper the behavior of the electromagnetic and thermal qualitities in a plasma arc placed between two conducting rails is analyzed. The plasma hydrogen armature drives the hydrogen pellets for the refueling of magnetic fusion reactors. Considering the general equations of electromagnetic and of plasma fluid dynamics and assuming steady-state conditions in a frame which is moving at the same rate as the plasma arc armature, as monodimensional model is deduced. The effects of an applied magnetic field on the behavior of all flow variables are particularly investigated.

  9. Recent results and challenges in development of metallic Hall sensors for fusion reactors

    SciTech Connect

    Ďuran, Ivan; Mušálek, Radek; Kovařík, Karel; Sentkerestiová, Jana; Kohout, Michal

    2014-08-21

    Reliable and precise diagnostic of local magnetic field is crucial for successful operation of future thermonuclear fusion reactors based on magnetic confinement. Magnetic sensors at these devices will experience an extremely demanding operational environment with large radiation and thermal loads in combination with required long term, reliable, and service-free performance. Neither present day commercial nor laboratory measurement systems comply with these requirements. Metallic Hall sensors based on e.g. copper or bismuth could potentially satisfy these needs. We present the technology for manufacturing of such sensors and some initial results on characterization of their properties.

  10. Note: Readout of a micromechanical magnetometer for the ITER fusion reactor

    SciTech Connect

    Rimminen, H.; Kyynaeraeinen, J.

    2013-05-15

    We present readout instrumentation for a MEMS magnetometer, placed 30 m away from the MEMS element. This is particularly useful when sensing is performed in high-radiation environment, where the semiconductors in the readout cannot survive. High bandwidth transimpedance amplifiers are used to cancel the cable capacitances of several nanofarads. A frequency doubling readout scheme is used for crosstalk elimination. Signal-to-noise ratio in the range of 60 dB was achieved and with sub-percent nonlinearity. The presented instrument is intended for the steady-state magnetic field measurements in the ITER fusion reactor.

  11. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

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

  13. Development of tritium breeding blankets for DT-burning fusion reactors

    SciTech Connect

    Clemmer, R.G.

    1980-01-01

    This study examines the status of understanding of blanket tritium recovery and the performance of potentially viable tritium breeding materials under conditions anticipated in a DT-fueled fusion reactor environment. The existing physicochemical, thermophysical, and ceramographic data for candidate liquid and solid breeders are reviewed and appropriate operating conditions defined. It is shown that selection of a breeding material and an appropriate tritium recovery method can impose significant constraints upon blanket design, particularly when considerations of breeder/coolant/structure compatibility and temperature limitations are taken into account.

  14. 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.

  15. Note: Readout of a micromechanical magnetometer for the ITER fusion reactor.

    PubMed

    Rimminen, H; Kyynäräinen, J

    2013-05-01

    We present readout instrumentation for a MEMS magnetometer, placed 30 m away from the MEMS element. This is particularly useful when sensing is performed in high-radiation environment, where the semiconductors in the readout cannot survive. High bandwidth transimpedance amplifiers are used to cancel the cable capacitances of several nanofarads. A frequency doubling readout scheme is used for crosstalk elimination. Signal-to-noise ratio in the range of 60 dB was achieved and with sub-percent nonlinearity. The presented instrument is intended for the steady-state magnetic field measurements in the ITER fusion reactor. PMID:23742608

  16. Fusion reactor materials: Semiannual progress report for period ending September 30, 1986

    SciTech Connect

    none,

    1987-09-01

    These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials program being conducted in support of the Magnetic Fusion Energy Program of the US Department of Energy. The major areas of concern covered in this report are irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters and activation calculations; materials engineering and design requirements; radiation effects; development of structural alloys; solid breeding materials; ceramics and superconducting magnet materials. There are 61 reports cataloged separately. (LSP)

  17. Preliminary evaluation of beta-spodumene as a fusion reactor structural material

    SciTech Connect

    Kelsey, P.V. Jr.; Schmunk, R.E.; Henslee, S.P.

    1981-01-01

    Beta-spodumene was investigated as a candidate material for use in fusion reactor environments. Properties which support the use of beta-spodumene include good thermal shock resistance, a very low coefficient of thermal expansion, a low-Z composition which would result in minimum impact on the plasma, and flexibility in fabrication processes. Specimens were irradiated in the Advanced Test Reactor (ATR) to a fluence of 5.3 x 10/sup 22/ n/m/sup 2/, E > 0.1 MeV, and 4.9 x 10/sup 23/ n/m/sup 2/ thermal fluence in order to obtain a preliminary evaluation of the impact of irradiation on the material. Preliminary data indicate that the mechanical properties of beta-spodumene are little affected by irradiation. Gas production and release have also been investigated.

  18. Organic insulators and the copper stabilizer for fusion-reactor magnets

    SciTech Connect

    Coltman, R.R. Jr.

    1981-11-01

    The materials which compose the large composite superconducting fusion reactor magnets are subjected to mechanical stress, neutron and gamma-ray radiation with broad energy spectra, high magnetic fields, and thermal cycling from 4 to 300 K. Of the materials now considered for use in the magnets, results show that the organic insulators and the Cu stabilizer are the most sensitive to this environment. In response to the need for stabilizer data, magnetoresistivity changes were studied in eight variously prepared specimens of Cu throughout five cycles of an alternate neutron irradiation (4.0 K) and annealing (14 h at 307 K) program. The results were combined with those on the radiation behavior of epoxy and polyimide organic insulators to provide a preliminary assessment of their comparative radiation resistance in a typical magnet location of the Experimental Power Reactor (EPR).

  19. Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

    SciTech Connect

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity.

  20. Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

    SciTech Connect

    Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

    1981-04-01

    Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity.

  1. Applications for reactor-pumped lasers

    SciTech Connect

    Lipinski, R.J.; McArthur, D.A.

    1994-10-01

    Nuclear reactor-pumped lasers (RPLs) have been developed in the US by the Department of Energy for over two decades, with the primary research occurring at Sandia National Laboratories and Idaho National Engineering Laboratory. The US program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1,271, 1,733, 1,792, 2,032, 2,630, 2,650, and 3,370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, and 3-D ceramic lithography. In addition, a ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night.

  2. Application of deuteron-deuteron (D-D) fusion neutrons to 40Ar/39Ar geochronology.

    PubMed

    Renne, Paul R; Knight, Kim B; Nomade, Sébastien; Leung, Ka-Ngo; Lou, Tak-Pui

    2005-01-01

    Neutron irradiation of samples for 40Ar/39Ar dating in a 235U fission reactor requires error-producing corrections for the argon isotopes created from Ca, K, and, to a lesser extent, Cl. The fission spectrum includes neutrons with energies above 2-3 MeV, which are not optimal for the 39K(n,p)39Ar reaction. These higher-energy neutrons are responsible for the largest recoil displacements, which may introduce age artifacts in the case of fine-grained samples. Both interference corrections and recoil displacements would be significantly reduced by irradiation with 2.45 MeV neutrons, which are produced by the deuteron-deuteron (D-D) fusion reaction 2H(d,n)3He. A new generation of D-D reactors should yield sufficiently high neutron fluxes (>10(12) n cm(-2)s(-1)) to be useful for 40Ar/39Ar dating. Modeling indicates that irradiation with D-D neutrons would result in scientific benefits of improved accuracy and broader applicability to fine-grained materials. In addition, radiological safety would be improved, while both maintenance and operational costs would be reduced. Thus, development of high-flux D-D fusion reactors is a worthy goal for 40Ar/39Ar geochronology. PMID:15498681

  3. Feedback stabilization of the resistive shell mode in a tokamak fusion reactor

    SciTech Connect

    Fitzpatrick, R.

    1997-07-01

    Stabilization of the {open_quotes}resistive shell mode{close_quotes} is vital to the success of the {open_quotes}advanced tokamak{close_quotes} concept. The most promising reactor relevant approach is to apply external feedback using, for instance, the previously proposed {open_quotes}fake rotating shell{close_quotes} scheme [R. Fitzpatrick and T. H. Jensen, Phys. Plasmas {bold 3}, 2641 (1996)]. This scheme, like other simple feedback schemes, only works if the feedback controlled conductors are located inside the {open_quotes}critical radius{close_quotes} at which a perfectly conducting shell is just able to stabilize the ideal external kink mode. In general, this is not possible in a reactor, since engineering constraints demand that any feedback controlled conductors be placed {ital outside} the neutron shielding blanket (i.e., relatively far from the edge of the plasma). It is demonstrated that the fake rotating shell feedback scheme can be modified so that it works even when the feedback controlled conductors are located well beyond the critical radius. The gain, bandwidth, current, and total power requirements of such a feedback system for a reactor sized plasma are estimated to be less than 100, a few Hz, a fews tens of kA, and a few MW, respectively. These requirements could easily be met using existing technology. It is concluded that feedback stabilization of the resistive shell mode is possible in a tokamak fusion reactor. {copyright} {ital 1997 American Institute of Physics.}

  4. Dust dynamics and diagnostic applications in quasi-neutral plasmas and magnetic fusion

    NASA Astrophysics Data System (ADS)

    Wang, Zhehui; Ticos, Catalin M.; Si, Jiahe; Delzanno, Gian Luca; Lapenta, Gianni; Wurden, Glen

    2007-11-01

    Little is known about dust dynamics in highly ionized quasi-neutral plasmas with ca. 1.0 e+20 per cubic meter density and ion temperature at a few eV and above, including in magnetic fusion. For example, dust motion in fusion, better known as UFO's, has been observed since 1980's but not explained. Solid understanding of dust dynamics is also important to International Thermonuclear Experimental Reactor (ITER) because of concerns about safety and dust contamination of fusion core. Compared with well studied strongly-coupled dusty plasma regime, new physics may arise in the higher density quasi-neutral plasma regime because of at least four orders of magnitude higher density and two orders of magnitude hotter ion temperature. Our recent laboratory experiments showed that plasma-flow drag force dominates over other forces in a quasi-neutral flowing plasma. In contrast, delicate balance among different forces in dusty plasma has led to many unique phenomena, in particular, the formation of dust crystal. Based on our experiments, we argue that 1) dust crystal will not form in the highly ionized plasmas with flows; 2) the UFO's are moving dust dragged by plasma flows; 3) dust can be used to measure plasma flow. Two diagnostic applications using dust for laboratory quasi-neutral plasmas and magnetic fusion will also be presented.

  5. High-Yield Lithium-Injection Fusion-Energy (HYLIFE) reactor

    SciTech Connect

    Blink, J.A.; Hogam, W.J.; Hovingh, J.; Meier, E.R.; Pitts, J.H.

    1985-12-23

    The High-Yield Lithium-Injection Fusion Energy (HYLIFE) concept to convent inertial confinement fusion energy into electric power has undergone intensive research and refinement at LLNL since 1978. This paper reports on the final HYLIFE design, focusing on five major areas: the HYLIFE reaction chamber (which includes neutronics, liquid-metal jet-array hydrocynamics, and structural design), supporting systems, primary steam system and balance of plant, safety and environmental protection, and costs. An annotated bibliography of reports applicable to HYLIFE is also provided. We conclude that HYLIFE is a particularly viable concept for the safe, clean production of electrical energy. The liquid-metal jet array, HYLIFE's key design feature, protects the surrounding structural components from x-rays, fusion fuel-pellet debris, neutron damage and activation, and high temperatures and stresses, allowing the structure to last for the plant's entire 30-year lifetime without being replaced. 127 refs., 18 figs.

  6. Multimodality Image Fusion-Guided Procedures: Technique, Accuracy, and Applications

    SciTech Connect

    Abi-Jaoudeh, Nadine; Kruecker, Jochen; Kadoury, Samuel; Kobeiter, Hicham; Venkatesan, Aradhana M. Levy, Elliot Wood, Bradford J.

    2012-10-15

    Personalized therapies play an increasingly critical role in cancer care: Image guidance with multimodality image fusion facilitates the targeting of specific tissue for tissue characterization and plays a role in drug discovery and optimization of tailored therapies. Positron-emission tomography (PET), magnetic resonance imaging (MRI), and contrast-enhanced computed tomography (CT) may offer additional information not otherwise available to the operator during minimally invasive image-guided procedures, such as biopsy and ablation. With use of multimodality image fusion for image-guided interventions, navigation with advanced modalities does not require the physical presence of the PET, MRI, or CT imaging system. Several commercially available methods of image-fusion and device navigation are reviewed along with an explanation of common tracking hardware and software. An overview of current clinical applications for multimodality navigation is provided.

  7. Multimodality Image Fusion Guided Procedures: Technique, Accuracy, and Applications

    PubMed Central

    Abi-Jaoudeh, Nadine; Kruecker, Jochen; Kadoury, Samuel; Kobeiter, Hicham; Venkatesan, Aradhana M.; Levy, Elliot; Wood, Bradford J.

    2012-01-01

    Personalized therapies play an increasingly critical role in cancer care; Image guidance with multimodality image fusion facilitates the targeting of specific tissue for tissue characterization, and plays a role in drug discovery and optimization of tailored therapies. PET, MRI and contrast enhanced CT may offer additional information not otherwise available to the operator during minimally invasive image guided procedures such as biopsy and ablation. With use of multimodality image fusion for image-guided interventions, navigation with advanced modalities does not require the physical presence of the PET, MRI, or CT imaging system. Several commercially available methods of image fusion and device navigation are reviewed along with an explanation of common tracking hardware and software. An overview of current clinical applications for multimodality navigation is provided. PMID:22851166

  8. A Study of the Flow Patterns of Expanding Impurity Aerosol Following a Disruption Event in a Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Majumdar, Rudrodip

    The current study focuses on the adiabatic expansion of aerosol impurity in the post-disruption and thermal quench scenario inside the vacuum chamber of a fusion reactor. A pulsed electrothermal plasma (ET) capillary source has been used as a source term simulating the surface ablation of the divertor or other interior critical components of a tokamak fusion reactor under hard disruption-like conditions. The capillary source generates particulates from wall evaporation by depositing transient radiant high heat flux onto the inner liner of the capillary. The particulates form a plasma jet moving towards the capillary exit at high speed and high pressure. The first chapter discusses briefly the relevance of the study pertaining to the impurities in a fusion reactor based on the work available in the form of published literature. The second chapter discusses briefly the operating principle of a pulsed electrothermal plasma source (PEPS), the virtual integration of PEPS with 1-D electrothermal plasma flow solver ETFLOW and the use of capillary plasma sources in various industrial applications. The third chapter discusses about primitive computational work, backed by the data from actual electrothermal source experiments from the in-house facility "PIPE" (Plasma Interactions with Propellants Experiment), that shows the supersonic bulk flow patterns for the temperature, density, pressure, bulk velocity and the flow Mach number of the impurity particulates as they get ejected as a high-pressure, high-temperature and hyper-velocity jet from the simulated source term. It also shows the uniform steady-state subsonic expansion of bulk aerosol inside the expansion chamber. The fourth chapter discusses scaling laws in 1-D for the aforesaid bulk plasma parameters for ranges of axial length traversed by the flow, so that one can retrieve the flow parameters at some preferred locations. The fifth chapter discusses the effect of temperature and the non--linearity of the adiabatic

  9. Final Report (1994 to 1996) Diagnostic of the Spatial and Velocity Distribution of Alpha Particles in Tokamak Fusion Reactor using Beat-wave Generated Lower Hybrid Wave

    SciTech Connect

    Hwang, D.Q.; Horton, R.D.; Evans, R.W.

    1999-06-03

    The alpha particles in a fusion reactor play a key role in the sustaining the fusion reaction. It is the heating provided by the alpha particles that help a fusion reactor operating in the ignition regime. It is, therefore, essential to understand the behavior of the alpha population both in real space and velocity space in order to design the optimal confinement device for fusion application. Moreover, the alphas represent a strong source of free energy that may generate plasma instabilities. Theoretical studies has identified the Toroidal Alfven Eigenmode (TAE) as an instability that can be excited by the alpha population in a toroidal device. Since the alpha has an energy of 3.5 MeV, a good confinement device will retain it in the interior of the plasma. Therefore, alpha measurement system need to probe the interior of a high density plasma. Due to the conducting nature of a plasma, wave with frequencies below the plasma frequency can not penetrate into the interior of the plasma where the alphas reside. This project uses a wave that can interact with the perpendicular motion of the alphas to probe its characteristics. However, this wave (the lower hybrid wave) is below the plasma frequency and can not be directly launched from the plasma edge. This project was designed to non-linearly excite the lower hybrid in the interior of a magnetized plasma and measure its interaction with a fast ion population.

  10. Analysis of Induced Gamma Activation by D-T Neutrons in Selected Fusion Reactor Relevant Materials with EAF-2010

    NASA Astrophysics Data System (ADS)

    Klix, Axel; Fischer, Ulrich; Gehre, Daniel

    2016-02-01

    Samples of lanthanum, erbium and titanium which are constituents of structural materials, insulating coatings and tritium breeder for blankets of fusion reactor designs have been irradiated in a fusion peak neutron field. The induced gamma activities were measured and the results were used to check calculations with the European activation system EASY-2010. Good agreement for the prediction of major contributors to the contact dose rate of the materials was found, but for minor contributors the calculation deviated up to 50%.

  11. 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.

  12. Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

    NASA Astrophysics Data System (ADS)

    Ragheb, Magdi M. H.; Maynard, Charles W.

    1981-10-01

    A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.

  13. Investigation of oxidation resistance of carbon based first-wall liner materials of fusion reactors

    NASA Astrophysics Data System (ADS)

    Moormann, R.; Hinssen, H. K.; Krüssenberg, A.-K.; Stauch, B.; Wu, C. H.

    1994-09-01

    One important aspect in selection of carbon based first-wall liner materials in fusion reactors is a sufficient oxidation resistance against steam and oxygen; this is because during accidents like loss of coolant into vacuum or loss of vacuum these oxidizing media can enter the vacuum vessel and may cause some corrosion of carbon followed by release of adsorbed tritium; in addition other consequences of oxidation like formation of burnable gases and their explosions have to be examined. Based on extensive experience on nuclear graphite oxidation in HTRs KFA has started in cooperation with NET some experimental investigations on oxidation of fusion reactor carbons. Results of first experiments on CFCs, Ti- and Si-doped carbons and graphites in steam (1273-1423 K) and oxygen (973 K) are reported. It was found that most materials have a similar reactivity as HTR nuclear graphites (which is much smaller than those of usual technical carbons); Si-doped CFCs however have a remarkably better oxidation resistance than those, which is probably due to the formation of a protecting layer of SiO 2. The measured kinetic data will be used in safety analyses for above mentioned accidents.

  14. Advanced Fuel Cycles for Fusion Reactors: Passive Safety and Zero-Waste Options

    NASA Astrophysics Data System (ADS)

    Zucchetti, Massimo; Sugiyama, Linda E.

    2006-05-01

    Nuclear fusion is seen as a much ''cleaner'' energy source than fission. Most of the studies and experiments on nuclear fusion are currently devoted to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition. The recent stress on safety by the world's community has stimulated the research on other fuel cycles than the DT one, based on 'advanced' reactions, such as the Deuterium-Helium-3 (DHe) one. These reactions pose problems, such as the availability of 3He and the attainment of the higher plasma parameters that are required for burning. However, they have many advantages, like for instance the very low neutron activation, while it is unnecessary to breed and fuel tritium. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) has been studied. Results show that Candor does reach the passive safety and zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate response to the environmental requirements for future nuclear power plants.

  15. Comparative study of survived displacement damage defects in iron irradiated in IFMIF and fusion power reactors

    NASA Astrophysics Data System (ADS)

    Simakov, S. P.; Konobeyev, A. Yu.; Fischer, U.; Heinzel, V.

    2009-04-01

    The assessment of the primary survived defects rates in iron such as vacancies-interstitials pairs and simplest clusters have been performed for the IFMIF, fusion power plant and research reactor. This was achieved by a modified version of the NJOY code, when processing evaluated nuclear cross section file. The modifications accounted for the reduction of the available damage energy predicted by the standard NRT model by the surviving defects fractions. These fractions were picked-up from the molecular dynamics and binary collisions simulation results available in the literature. The number of primary survived and clustered defects in the α-iron irradiated in the high flux test module of IFMIF was estimated as 10 and 6 dpa/fpy or several times less than standard NRT estimates at the level of 30 dpa/fpy. The comparison with damages in iron calculated for irradiation in the first wall of fusion power plant gave however the same reduction factors, that supports the qualification of IFMIF as a fusion material irradiation facility.

  16. Nuclear fuels for low-beta fusion reactors: Lithium resources revisited

    NASA Astrophysics Data System (ADS)

    Eckhartt, Dieter

    1995-12-01

    In searching to attain optimum conditions for the controlled release of nuclear energy by fusion processes, the stationary confinement of low-pressure ring-shaped plasmas by strong magnetic fields is now regarded as the most promising approach. We consider a number of fuel combinations that could be operated in such low-beta reactor systems and look upon the relevant fuel reserves. The “classical” D-T-Li cycle will be used as a standard and is extensively discussed therefore. It could supply most of mankind's future long-term power needs—but only on condition that the required lithium fuel can be extracted from seawater at reasonable expenses. The estimated landbound lithium reserves are too small to that end, they will last for about 500 years at most, depending on forecasts of future energy consumption and on assumptions about exploitable resources. Recovery of lithium from seawater would extend the possible range by a factor of 300 or so, provided that extraction technologies which are at present available in the laboratory, could be extended to a very large and industrial scale. Deuterium is abundant on earth but D-D fusion is difficult, if not impossible, to be achieved in the low-beta systems presently investigated for D-T fusion. The same arguments apply to so-called “advanced” concepts, such as the D-3He and the D-6Li cycles.

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

    NASA Astrophysics Data System (ADS)

    Hawryluk, R. J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J. W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C. W.; Barnes, G.; Batchelor, D. B.; Bateman, G.; Batha, S.; Baylor, L. A.; Beer, M.; Bell, M. G.; Biglow, T. S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C. Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D. S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P. C.; Ernst, D.; Faunce, J.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R. C.; Golian, T.; Gorelenkov, N.; Gouge, M. J.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Heidbrink, W.; Hermann, H. W.; Hill, K. W.; Hirshman, S.; Hoffman, D. J.; Hosea, J.; Hulse, R. A.; Hsuan, H.; Jaeger, E. F.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Laughlin, M. J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D. E.; Marchlik, M.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M. P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D. C.; McGuire, K. M.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Milora, S. L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Paul, S. F.; Pavlov, Yu. I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Pitcher, S.; Popovichev, S.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Sabbagh, S. A.; Sasao, M.; Scharer, J.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Scillia, R.; Scott, S. D.; Senko, T.; Sissingh, R.; Skinner, C.; Snipes, J.; Snook, P.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Timberlake, J. R.; Towner, H. H.; Ulrickson, M.; von Halle, A.; Vannoy, C.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Wright, K.; Wong, D.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yeun, A.; Yoshikawa, S.; Young, K. M.; Zakharov, L.; Zarnstorff, M. C.; Zweben, S. J.

    1994-05-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM 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 ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium-tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed.

  18. 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.

  19. Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited)a)

    NASA Astrophysics Data System (ADS)

    Smith, Roger J.

    2008-10-01

    A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local Bpol diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local Te, ne, and B∥ along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher neB∥ product and higher ne and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

  20. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    SciTech Connect

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250/sup 0/C).

  1. Applications of computer modeling to fusion research

    SciTech Connect

    Dawson, J.M.

    1989-01-01

    Progress achieved during this report period is presented on the following topics: Development and application of gyrokinetic particle codes to tokamak transport, development of techniques to take advantage of parallel computers; model dynamo and bootstrap current drive; and in general maintain our broad-based program in basic plasma physics and computer modeling.

  2. PREFACE: The fifth International Conference on Inertial Fusion Sciences and Applications (IFSA2007)

    NASA Astrophysics Data System (ADS)

    Azechi, Hiroshi; Hammel, Bruce; Gauthier, Jean-Claude

    2008-06-01

    by fast-ignition and related ultra-intense laser interaction. Progress in direct drive over the past few years resulted in the achievement of high-density cryogenic implosions at OMEGA. Continuous progresses in hohlraum physics gave confidence in the achievement of ignition at NIF and LMJ. Advances in Z-pinch included double-hohlraum irradiation symmetry and the PW laser beam for the Z-facility. Progress of laser material development for IFE driver was a very interesting topic of inertial fusion energy drivers, including KrF and DPSSL lasers and particle beams. Of special interest, a future session was focused on strategy of inertial fusion energy development. Laboratory tours were held in the middle of the Conference. The Laser for Fusion EXperiments (LFEX), a new high-energy petawatt laser at ILE, was one of the key attractions of IFSA 2007. 83 participants toured LFEX and GEKKO XII lasers, and 35 joined a tour of KPSA-JAEA. In parallel to the tour, the `Symposium on Academics-Industries Cooperation for Applications of High-Power Lasers' was held with more than 90 participants mostly from the industrial community. These Proceedings start with special chapters on the keynote and focus speeches and the Teller lectures. The keynotes and focus give an overview of progress in inertial fusion in Asia, North America, and Europe. The Teller lectures show the contributions of this year's two winners: Brian Thomas of AWE, UK and Kunioki Mima of ILE. The remainder of the Proceedings is divided into three parts. Part A covers the physics of inertial fusion; Part B covers laser, particle beams, and fusion technology including IFE reactors and target fabrication; and Part C covers science and technology applications such as laboratory astrophysics, laser particle acceleration, x-ray and EUV sources, and new applications of intense lasers. These parts are further divided into chapters covering specific areas of science or technology. Within each chapter the talks relevant to that

  3. Applications of high-speed dust injection to magnetic fusion

    SciTech Connect

    Wang, Zhehui; Li, Yangfang

    2012-08-08

    It is now an established fact that a significant amount of dust is produced in magnetic fusion devices due to plasma-wall interactions. Dust inventory must be controlled, in particular for the next-generation steady-state fusion machines like ITER, as it can pose significant safety hazards and degrade performance. Safety concerns are due to tritium retention, dust radioactivity, toxicity, and flammability. Performance concerns include high-Z impurities carried by dust to the fusion core that can reduce plasma temperature and may even induce sudden termination of the plasma. We have recognized that dust transport, dust-plasma interactions in magnetic fusion devices can be effectively studied experimentally by injection of dust with known properties into fusion plasmas. Other applications of injected dust include diagnosis of fusion plasmas and edge localized mode (ELM)'s pacing. In diagnostic applications, dust can be regarded as a source of transient neutrals before complete ionization. ELM's pacing is a promising scheme to prevent disruptions and type I ELM's that can cause catastrophic damage to fusion machines. Different implementation schemes are available depending on applications of dust injection. One of the simplest dust injection schemes is through gravitational acceleration of dust in vacuum. Experiments at Los Alamos and Princeton will be described, both of which use piezoelectric shakers to deliver dust to plasma. In Princeton experiments, spherical particles (40 micron) have been dropped in a systematic and reproducible manner using a computer-controlled piezoelectric bending actuator operating at an acoustic (0,2) resonance. The circular actuator was constructed with a 2.5 mm diameter central hole. At resonance ({approx} 2 kHz) an applied sinusoidal voltage has been used to control the flux of particles exiting the hole. A simple screw throttle located {approx}1mm above the hole has been used to set the magnitude of the flux achieved for a given

  4. Opacity project - Astrophysical and fusion applications

    NASA Technical Reports Server (NTRS)

    Pradhan, A. K.

    1987-01-01

    An overview is presented of a project to calculate large quantities of accurate atomic data for radiative processes of importance in the precise determination of opacities in stellar atmospheres, and for astrophysical and laboratory applications in general. Work is in progress on the oscillator strengths, photoionization cross sections, damping constants, etc., for all atoms and ions in hydrogen through neon isoelectronic sequences going up to iron.

  5. Peaceful Uses of Fusion

    DOE R&D Accomplishments Database

    Teller, E.

    1958-07-03

    Applications of thermonuclear energy for peaceful and constructive purposes are surveyed. Developments and problems in the release and control of fusion energy are reviewed. It is pointed out that the future of thermonuclear power reactors will depend upon the construction of a machine that produces more electric energy than it consumes. The fuel for thermonuclear reactors is cheap and practically inexhaustible. Thermonuclear reactors produce less dangerous radioactive materials than fission reactors and, when once brought under control, are not as likely to be subject to dangerous excursions. The interaction of the hot plasma with magnetic fields opens the way for the direct production of electricity. It is possible that explosive fusion energy released underground may be harnessed for the production of electricity before the same feat is accomplished in controlled fusion processes. Applications of underground detonations of fission devices in mining and for the enhancement of oil flow in large low-specific-yield formations are also suggested.

  6. Neutron measurements for biomedical and fusion technology applications

    NASA Astrophysics Data System (ADS)

    Barschall, H. H.

    1985-01-01

    Measurements of reaction cross sections of neutrons of energy above 5 MeV yield important information about reaction mechanisms. The main impetus for such measurements has, however, recently come from applications. Measurements on light elements are needed for neutron dosimetry, primarily for radiotherapy. Measurements on heavier nuclides provide information for fusion technology, both for the assessment of radiation damage and for the management of radioactive wastes.

  7. Sensor fusion by pseudo information measure: a mobile robot application.

    PubMed

    Asharif, Mohammad Reza; Moshiri, Behzad; HoseinNezhad, Reza

    2002-07-01

    In any autonomous mobile robot, one of the most important issues to be designed and implemented is environment perception. In this paper, a new approach is formulated in order to perform sensory data integration for generation of an occupancy grid map of the environment. This method is an extended version of the Bayesian fusion method for independent sources of information. The performance of the proposed method of fusion and its sensitivity are discussed. Map building simulation for a cylindrical robot with eight ultrasonic sensors and mapping implementation for a Khepera robot have been separately tried in simulation and experimental works. A new neural structure is introduced for conversion of proximity data that are given by Khepera IR sensors to occupancy probabilities. Path planning experiments have also been applied to the resulting maps. For each map, two factors are considered and calculated: the fitness and the augmented occupancy of the map with respect to the ideal map. The length and the least distance to obstacles were the other two factors that were calculated for the routes that are resulted by path planning experiments. Experimental and simulation results show that by using the new fusion formulas, more informative maps of the environment are obtained. By these maps more appropriate routes could be achieved. Actually, there is a tradeoff between the length of the resulting routes and their safety and by choosing the proper fusion function, this tradeoff is suitably tuned for different map building applications. PMID:12160343

  8. Modelling Thermodynamics of Alloys for Fusion Application

    SciTech Connect

    Caro, A; Sadigh, B; Turchi, P A; Caro, M; Lopasso, E; Crowson, D

    2006-01-26

    This research has two main objectives: (1) On one side is the development of computational tools to evaluate alloy properties, using the information contained in thermodynamic functions to improve the ability of classic potentials to account for complex alloy behavior. (2) On the other hand, to apply the tools so developed to predict properties of alloys under irradiation. Atomistic simulations of alloys at the empirical level face the challenge of correctly modeling basic thermodynamic properties. In this work we develop a methodology to generalize many-body classic potentials to incorporate complex formation energy curves. Application to Fe-Cr allows us to predict the implications of the ab initio results of formation energy on the phase diagram of this alloy.

  9. Modelling Thermodynamics of Alloys for Fusion Application

    SciTech Connect

    Caro, A; Erhart, P; de Caro, M S; Sadigh, B; Srinivasan, S G; Stukowski, A

    2009-07-29

    This research has two main objectives: (1) The development of computational tools to evaluate alloy properties, using the information contained in thermodynamic functions. We aim at improving the ability of classical potentials to account for complex alloy behavior, and (2) The application of these tools to predict properties of alloys under irradiation, in particular the FeCr system. This semester has been very productive in the developments of both tools and algorithms. Our work aims at developing theoretical and numerical methodologies that are directly applicable to multi-scale modeling addressing the specific issues related to multi-component, multi-phase systems in non-equilibrium states, such as solid-solution hardening, point defect-solute interactions, stoichiometry effects, static and dynamic strain aging, dislocation-solute interactions, and in general the aspects of microstructure evolution that are affected by irradiation. At its present stage of development, we have been able to predict numerous thermodynamic properties of FeCr mainly related to ordering and precipitation; we have found new intermetallic phases and suggested the existence of a dependence of the solubility limit on the degree of order of the alloy. At present, we are studying dislocation mobility in the solid solution and the heterogeneous phase, and we are developing a new algorithm to perform Monte Carlo simulations inside the miscibility gap, a technique that will allow us to study interfacial energies and nucleation sizes. We develop a strategy to model radiation damage in FeCr alloys, system in which magnetism introduces an anomaly in the heat of formation of the solid solution that is at the basis of its unique behavior. Magnetism has implications for the precipitation of excess Cr in the a phase in the presence of heterogeneities. These complexities pose many challenges for atomistic (empirical) methods. To address such issues we develop a modified, many-body potential by

  10. A carbon-carbon composite materials development program for fusion energy applications

    SciTech Connect

    Burchell, T.D.; Eatherly, W.P. ); Engle, G.B. ); Hollenberg, G.W. )

    1992-10-01

    Carbon-carbon composites increasingly are being used for plasma-facing component (PFC) applications in magnetic-confinement plasma-fusion devices. They offer substantial advantages such as enhanced physical and mechanical properties and superior thermal shock resistance compared to the previously favored bulk graphite. Next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) and the Burning Plasma Experiment (BPX), will require advanced carbon-carbon composites possessing extremely high thermal conductivity to manage the anticipated extreme thermal heat loads. This report outlines a program that will facilitate the development of advanced carbon-carbon composites specifically tailored to meet the requirements of ITER and BPX. A strategy for developing the necessary associated design data base is described. Materials property needs, i.e., high thermal conductivity, radiation stability, tritium retention, etc., are assessed and prioritized through a systems analysis of the functional, operational, and component requirements for plasma-facing applications. The current Department of Energy (DOE) Office of Fusion Energy Program on carbon-carbon composites is summarized. Realistic property goals are set based upon our current understanding. The architectures of candidate PFC carbon-carbon composite materials are outlined, and architectural features considered desirable for maximum irradiation stability are described. The European and Japanese carbon-carbon composite development and irradiation programs are described. The Working Group conclusions and recommendations are listed. It is recommended that developmental carbon-carbon composite materials from the commercial sector be procured via request for proposal/request for quotation (RFP/RFQ) as soon as possible.

  11. 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

  12. 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

  13. Evidence for Critical Energy for Ion Confinement in Magnetic Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Maglich, Bogdan; Hester, Tim; Scott, Dan; Calsec Collaboration

    2015-03-01

    It is shown here that fusion test reactors could not ignite for half-a-century because trials were conducted at thermonuclear ion energies 10-30 KeV, an order of magnitude lower than critical energy, Ec ~ 200 KeV. At subcritical energies, plasma is destroyed by neutralization of ions via overlooked atomic (non-nuclear) charge transfer collisions with giant cross-section, 109 barns, 100 times greater than that for ionization collisions that counters neutralization. Neutral injection sets limit on ion magnetic confinement time <10-6 s vs. >1 s required for ignition. In contrast, at energies above Ec, ionization prevails; near ~ 1 MeV, stable confinement of 20 s was routinely observed with charged injection. - To render ITER viable, ion energy must be increased to >/ = 1 MeV; neutral radioactive DT fuel replaced with charged, nonradioactive deuterium, giving rise to compact aneutronicreactor with direct conversion into RF power.

  14. Divertor conditions relevant for fusion reactors achieved with linear plasma generator

    SciTech Connect

    Eck, H. J. N. van; Lof, A.; Meiden, H. J. van der; Rooij, G. J. van; Scholten, J.; Zeijlmans van Emmichoven, P. A.; Kleyn, A. W.

    2012-11-26

    Intense magnetized hydrogen and deuterium plasmas have been produced with electron densities up to 3.6 Multiplication-Sign 10{sup 20} m{sup -3} and electron temperatures up to 3.7 eV with a linear plasma generator. Exposure of a W target has led to average heat and particle flux densities well in excess of 4 MW m{sup -2} and 10{sup 24} m{sup -2} s{sup -1}, respectively. We have shown that the plasma surface interactions are dominated by the incoming ions. The achieved conditions correspond very well to the projected conditions at the divertor strike zones of fusion reactors such as ITER. In addition, the machine has an unprecedented high gas efficiency.

  15. Calculation procedures for the analysis of integral experiments for fusion-reactor design

    NASA Astrophysics Data System (ADS)

    Santoro, R. T.; Barnes, J. M.; Alsmiller, R. G., Jr.; Oblow, E. M.

    1981-07-01

    The calculational models, nuclear data, and radiation transport codes that are used in the analysis of integral measurements of the transport of approxmately 14 MeV neutrons through laminated slabs of materials typical of those found in fusion reactor shields are described. The two dimensional discrete ordinates calculations to optimize the experimental configuration of reducing the neutron and gamma ray background levels and for obtaining and equivalent, reduced geometry of the calculational model to reduce computer core storage and running times are also presented. The equations and data to determine the energy-angle relations to neutrons produced in the reactions of 250 keV deuterons in a titanium tritide target are given. The procedures used to collapse the 17ln 36gamma VATAMIN C cross section library to a 53n 21 gamma broad group library are described.

  16. Modeling of neutral hydrogen velocities in the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Stotler, D. P.; Skinner, C. H.; Budny, R. V.; Ramsey, A. T.; Ruzic, D. N.; Turkot, R. B., Jr.

    1996-11-01

    Monte Carlo neutral transport simulations of hydrogen velocities in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] are compared with experiment using the Doppler-broadened Balmer-α spectral line profile. Good agreement is obtained under a range of conditions, validating the treatment of charge exchange, molecular dissociation, surface reflection, and sputtering in the neutral gas code DEGAS [D. Heifetz et al., J. Comput. Phys. 46, 309 (1982)]. A residual deficiency of 10-100 eV neutrals in most of the simulations indicates that further study of the energetics of H+2 dissociation for electron energies in excess of 100 eV is needed.

  17. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    SciTech Connect

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S.

    2015-03-15

    The objective of this work is to establish the tritium-permeation properties of structural alloys considered for Fusion systems and very high temperature reactors (VHTR). A description of the work performed to set up an apparatus to measure permeation rates of hydrogen and tritium in 304L stainless steel is presented. Following successful commissioning with hydrogen, the test apparatus was commissioned with tritium. Commissioning tests with tritium suggest the need for a reduction step that is capable of removing the oxide layer from the test sample surfaces before accurate tritium-permeation data can be obtained. Work is also on-going to clearly establish the temperature profile of the sample to correctly estimate the tritium-permeability data.

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

    NASA Astrophysics Data System (ADS)

    Zhu, Limin; Liu, Haifeng; Wang, Xianqu

    2016-09-01

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

  19. Activation characteristics of a solid breeder blanket for a fusion power demonstration reactor

    NASA Astrophysics Data System (ADS)

    Fischer, Ulrich; Tsige-Tamirat, Haileyesus

    2002-12-01

    Activation characteristics have been assessed for a helium cooled solid breeder blanket on the basis of three-dimensional activation calculations for a 2200 MW fusion power demonstration reactor. FISPACT inventory calculations were performed for the beryllium neutron multiplier, the Li 4SiO 4 breeder ceramics and the Eurofer low activation steel. Neutron flux spectra distributions were provided by a previous MCNP calculation. Detailed spatial distributions have been obtained for the nuclide inventories and related quantities such as activity, decay heat and contact dose rate. These data are available form the authors upon request. On the basis of the calculated contact gamma dose rates, the waste quality was assessed with regard to a possible re-use of the activated materials following the remote or the hands-on handling recycling options.

  20. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    SciTech Connect

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method.

  1. The role of risk management in the design of diagnostics for fusion reactors

    NASA Astrophysics Data System (ADS)

    Ingesson, L. C.; F4E Diagnostic Project Team

    2014-08-01

    A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial.

  2. Conceptual design study of Fusion Experimental Reactor (FY86 FER): Safety

    NASA Astrophysics Data System (ADS)

    Seki, Yasushi; Iida, Hiromasa; Honda, Tsutomu

    1987-08-01

    This report describes the study on safety for FER (Fusion Experimental Reactor) which has been designed as a next step machine to the JT-60. Though the final purpose of this study is to have an image of design base accident, maximum credible accident and to assess their risk or probability, etc., as FER plant system, the emphasis of this years study is placed on fuel-gas circulation system where the tritium inventory is maximum. The report consists of two chapters. The first chapter summarizes the FER system and describes FMEA (Failure Mode and Effect Analysis) and related accident progression sequence for FER plant system as a whole. The second chapter of this report is focused on fuel-gas circulation system including purification, isotope separation and storage. Probability of risk is assessed by the probabilistic risk analysis (PRA) procedure based on FMEA, ETA and FTA.

  3. Dielectronic satellite spectra of hydrogenlike iron from TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Decaux, V. ); Bitter, M.; Hsuan, H.; von Goeler, S.; Hill, K.W.; Hulse, R.A.; Taylor, G.; Park, H. . Plasma Physics Lab.); Bhalla, C.P. . Dept. of Physics)

    1990-08-01

    Spectra of hydrogenlike iron, Fe26, have been observed from Tokamak Fusion Test Reactor (TFTR) plasmas with a high-resolution crystal spectrometer. The experimental arrangement permits simultaneous observation of the Fe26 Ly-{alpha}{sub 1}and Ly-{alpha}{sub 2} lines and the associated dielectronic satellites, which are due to transitions 1snl-2pnl{prime} with n {ge} 2, as well as the heliumlike 1s{sup 2}({sup 1}S{sub 0}){minus}1s4p({sup 1}P{sub 1})and both hydrogenlike Ly-{beta}{sub 1} and Ly-{beta}{sub 2} lines from chromium. Relative wavelengths and line intensities can be determined very accurately. The spectral data are in very good agreement with theoretical calculations. The observed spectra have also been used to estimate the total dielectronic recombination rate coefficient of Fe26. 30 refs., 4 figs., 3 tabs.

  4. Removing tritium and other impurities during industrial recycling of beryllium from a fusion reactor

    SciTech Connect

    Dylst, K.; Seghers, J.; Druyts, F.; Braet, J.

    2008-07-15

    Recycling beryllium used in a fusion reactor might be a good way to overcome problems related to the disposal of neutron irradiated beryllium. The critical issues for the recycling of used first wall beryllium are the presence of tritium and (transuranic) impurities. High temperature annealing seems to be the most promising technique for detritiation. Purification of the de-tritiated beryllium can be achieved by chlorination of the irradiated beryllium and the subsequent reduction of beryllium chloride to highly pure metallic beryllium. After that, the beryllium can be re-fabricated into first wall tiles via powder metallurgy which is already a mature industrial practice. This paper outlines the path to define the experimental needs for beryllium recycling and tackles problems related to the detritiation and the purification via the chlorine route. (authors)

  5. An integrated approach to assessing the fracture safe margins of fusion reactor structures

    SciTech Connect

    Odette, G.R.

    1996-10-01

    Design and operation of fusion reactor structures will require an appropriate data base closely coupled to a reliable failure analysis method to safely manage irradiation embrittlement. However, ongoing irradiation programs will not provide the information on embrittlement necessary to accomplish these objectives. A new engineering approach is proposed based on the concept of a master toughness-temperature curve indexed on an absolute temperature scale using shifts to account for variables such as size scales, crack geometry and loading rates as well as embrittlement. While providing a simple practical engineering expedient, the proposed method can also be greatly enhanced by fundamental mechanism based models of fracture and embrittlement. Indeed, such understanding is required for the effective use of small specimen test methods, which is a integral element in developing the necessary data base.

  6. Effects of solutes on thermodynamic activity of tritium in liquid lithium blanket of fusion reactor

    SciTech Connect

    Lyublinski, I.E.; Evtikhin, V.A.; Krassine, V.P.

    1995-10-01

    The study of tritium dissolved in liquid lithium systems containing metallic impurities is of technological interest for tritium removal processes and suppression of the tritium leakages from the blanket in a fusion reactor. The interaction parameter formalism and coordination cluster theory have been used to calculate the tritium activity coefficients in Li-T-Al, Li-T-Mg, Li-T-Si, Li-T-Y and Li-T-La systems. Calculations performed demonstrated that silicon, aluminum and magnesium have no influence on the tritium activity coefficients in dilute lithium solutions within the temperature range 400-800{degree}C, but yttrium and lanthanum may be used to effectively decrease this coefficient in lithium melts. 13 refs., 3 figs., 2 tabs.

  7. Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Kennedy, Daniel; Jaworski, Michael

    2014-10-01

    Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

  8. The role of risk management in the design of diagnostics for fusion reactors

    SciTech Connect

    Ingesson, L. C.; Collaboration: F4E Diagnostic Project Team

    2014-08-21

    A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial.

  9. Negative ion source development for fusion application (invited).

    PubMed

    Takeiri, Yasuhiko

    2010-02-01

    Giant negative ion sources, producing high-current of several tens amps with high energy of several hundreds keV to 1 MeV, are required for a neutral beam injector (NBI) in a fusion device. The giant negative ion sources are cesium-seeded plasma sources, in which the negative ions are produced on the cesium-covered surface. Their characteristic features are discussed with the views of large-volume plasma production, large-area beam acceleration, and high-voltage dc holding. The international thermonuclear experimental reactor NBI employs a 1 MeV-40 A of deuterium negative ion source, and intensive development programs for the rf-driven source plasma production and the multistage electrostatic acceleration are in progress, including the long pulse operation for 3600 s. Present status of the development, as well as the achievements of the giant negative ion sources in the working injectors, is also summarized. PMID:20192420

  10. Negative ion source development for fusion application (invited)

    SciTech Connect

    Takeiri, Yasuhiko

    2010-02-15

    Giant negative ion sources, producing high-current of several tens amps with high energy of several hundreds keV to 1 MeV, are required for a neutral beam injector (NBI) in a fusion device. The giant negative ion sources are cesium-seeded plasma sources, in which the negative ions are produced on the cesium-covered surface. Their characteristic features are discussed with the views of large-volume plasma production, large-area beam acceleration, and high-voltage dc holding. The international thermonuclear experimental reactor NBI employs a 1 MeV-40 A of deuterium negative ion source, and intensive development programs for the rf-driven source plasma production and the multistage electrostatic acceleration are in progress, including the long pulse operation for 3600 s. Present status of the development, as well as the achievements of the giant negative ion sources in the working injectors, is also summarized.

  11. Inertial Fusion Sciences and Applications 2003: State of the Art 2003, Published by the American Nuclear Society

    SciTech Connect

    Editors: B. A. Hammel; D. D. Meyerhofer; J. Meyer-ter-Vehn; H. Azechi. Organizing Chair: W. J. Hogan

    2004-06-01

    Collection of all papers presented and submitted at the IFSA2003 conference. Topics included target design and performance, fast ignition, plasma instabilities, laser technology, fusion reactor technology

  12. LBB application in the US operating and advanced reactors

    SciTech Connect

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  13. 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.

  14. Study of Tungsten as a Plasma-facing Material for a Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Khripunov, B. I.; Koidan, V. S.; Ryazanov, A. I.; Gureev, V. M.; Kornienko, S. N.; Latushkin, S. T.; Rupyshev, A. S.; Semenov, E. V.; Kulikauskas, V. S.; Zatekin, V. V.

    The paper presents the development of experimental investigations and recent results of the impact on tungsten at high level of radiation damage under steady-state deuterium plasma. Tungsten is considered as a plasma facing material for a fusion reactor. The effect of fusion neutron impact is simulated by surrogate irradiations with high-energy ions. The primary defects at 1-100 dpa were produced in tungsten samples by He and C ions accelerated in the Kurchatov cyclotron to 3-10 MeV at the total fluence of 1017-1019 cm-2. The irradiated material was studied in deuterium plasma on the LENTA linear divertor simulator at the plasma fluence 1021-1022 D/cm2. Erosion dynamics, development of the surface microstructure and deuterium retention were analyzed. Increased deuterium retention detected previously in tungsten pre-irradiated by He ions was also registered (ERDA) on C-irradiated samples at 2-3 dpa. In contrast, a significant decrease in the D uptake has been observed on those samples operated in the experiments at 500 °C.

  15. NUCLEAR CHEMISTRY OF WATER-COOLED FUSION REACTORS: ISSUES AND SOLUTIONS

    SciTech Connect

    Petrov, Andrei Y; Flanagan, George F

    2010-01-01

    ITER is an experimental Tokamak fusion energy reactor that is being built in Cadarache, France, in collaboration with seven agencies representing China, the European Union, India, Japan, Republic of Korea, the Russian Federation, and the United States. The main objective of ITER is to demonstrate the scientific and technical feasibility of a controlled fusion reaction An important U.S. contribution is the design, fabrication, and delivery of the Tokamak Cooling Water System (TCWS). This paper describes the main sources of radioactivity in TCWS water, which are the nitrogen isotopes 16N and 17N, tritium, activated corrosion products, and the carbon isotope 14C; the relative contribution of each of these sources to the total radioactive contamination of water; issues related to excess accumulation of these species; and methods to control TCWS radioactivity within acceptable limits. Among these methods are: (1) water purification to minimize corrosion of materials in contact with TCWS water; (2) monitoring of vital chemistry parameters and control of water chemistry; (3) design of proper building structure and/or TCWS loop/geometry configuration; and (4) design of an ITER liquid radwaste facility tailored to TCWS operational requirements. Design of TCWS nuclear chemistry control is crucial to ensuring that the inventory of radioactive species is consistent with the principle of 'As Low as Reasonably Achievable.'

  16. Cryogenic system component development for the fusion experimental reactor at JAERI

    SciTech Connect

    Kato, T.; Kamiya, S.; Tada, E.; Hiyama, T.; Kawano, K.; Shimamoto, S.

    1986-01-01

    The major objective of fusion R and D at the Japan Atomic Energy Research Institute (JAERI) is to construct the Fusion Experimental Reactor (FER) to follow JT-60. The construction of FER inevitably requires development of a large, reliable, and efficient helium liquefier/refrigerator and the more advanced cryogenic technology for cooling superconducting toroidal and poloidal coils. Typical characteristics required for the cryogenic system of FER are 10 to 20 kW at 4 K as one unit, reliability for > 8000 h, a stable pulsed heat load, and high-energy efficiency of > 1/500. In this cryogenic system, the major components such as the helium compressor, turbo-expander, cold circulation pump for supercritical helium, and cold compressor to reduce operating temperature below 4 K should be scaled up to a mass flow rate of > 1000 g/s. For this purpose, JAERI has developed cryogenics since 1980 in accordance with the development program in which the scaling up of the major components mentioned above are involved as well as cooling technology development.

  17. Materials recycle and waste management in fusion power reactors. Progress report for 1982

    SciTech Connect

    Vogler, S.; Jung, J.; Steindler, M.J.; Maya, I.; Levine, H.E.; Peterman, D.D.; Strausburg, S.; Schultz, K.R.

    1983-01-01

    Several components of a STARFIRE fusion reactor have been studied. The breeding ratios were calculated as a function of lithium enrichment and neutron multiplier for systems containing either Li/sub 2/O or LiAlO/sub 2/. The lithium requirements for a fusion economy were also estimated for those cases and the current US resources were found to be adequate. However, competition with other lithium demands in the future emphasizes the need for recovering and reusing lithium. The radioactivities induced in the breeder and the impurities responsible for their formation were determined. The residual radioactivities of several low-activation structural materials were compared with the radioactivity from the prime candidate alloy (PCA) a titanium modified Type 316 stainless steel used in STARFIRE. The impurities responsible for the radioactivity levels were identified. From these radioactive impurity levels it was determined that V15Cr5Ti could meet the requirements for shallow land burial as specified by the Nuclear Regulatory Commission (10CFR61), whereas PCA would require a more restrictive disposal mode, i.e. in a geologic medium. The costs for each of these disposal modes were then estimated.

  18. 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.

  19. Radioactive waste produced by DEMO and commerical fusion reactors extrapolated from ITER and advanced data bases

    SciTech Connect

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

    1994-12-31

    The potential for providing energy with minimal environmental impact is a powerful motivation for the development of fusion and is the long-term objective of most fusion programs. However, the societal acceptability of magnetic fusion may well be decided in the near-term when decisions are taken on the construction of DEMO to follow ITER (if not when the construction decision is taken on ITER). Component wastes were calculated for DEMOs based on each data base by first calculating reactor sizes needed to satisfy the physics, stress and radiation attenuation requirements, and then calculating component replacement rates based on radiation damage and erosion limits. Then, radioactive inventories were calculated and compared to a number of international criteria for {open_quote}near-surface{close_quote} burial. None of the components in either type of design would meet the Japanese LLW criterion (<1 Ci/m{sup 3}) within 10 years of shutdown, although the advanced (V/Li) blanket would do so soon afterwards. The vanadium first wall, divertor and blanket would satisfy the IAEA LLW criterion (<2 mSv/h contact dose) within about 10 years after shutdown, but none of the stainless steel or copper components would. All the components in the advanced data base designs except the stainless steel vacuum vessel and shield readily satisfy the US extended 10CFR61 intruder dose criterion, but none of the components in the {open_quotes}ITER data base{close_quotes} designs do so. It seems unlikely that a stainless steel first wall or a copper divertor plate could satisfy the US (class C) criterion for near surface burial, much less the more stringent international, criteria. On the other hand, the first wall, divertor and blanket of the V/Li system would still satisfy the intruder dose concentration limits even if the dose criterion was reduced by two orders of magnitude.

  20. Ventilation Systems Operating Experience Review for Fusion Applications

    SciTech Connect

    Cadwallader, Lee Charles

    1999-12-01

    This report is a collection and review of system operation and failure experiences for air ventilation systems in nuclear facilities. These experiences are applicable for magnetic and inertial fusion facilities since air ventilation systems are support systems that can be considered generic to nuclear facilities. The report contains descriptions of ventilation system components, operating experiences with these systems, component failure rates, and component repair times. Since ventilation systems have a role in mitigating accident releases in nuclear facilities, these data are useful in safety analysis and risk assessment of public safety. An effort has also been given to identifying any safety issues with personnel operating or maintaining ventilation systems. Finally, the recommended failure data were compared to an independent data set to determine the accuracy of individual values. This comparison is useful for the International Energy Agency task on fusion component failure rate data collection.

  1. Fusion Power Program biannual progress report, April-September 1979

    SciTech Connect

    Not Available

    1980-02-01

    This biannual report summarizes the Argonne National Laboratory work performed for the Office of Fusion Energy during the April-September 1979 quarter in the following research and development areas: materials; energy storage and transfer; tritium containment, recovery and control; advanced reactor design; atomic data; reactor safety; fusion-fission hybrid systems; alternate applications of fusion energy; and other work related to fusion power. Separate abstracts were prepared for three sections. (MOW)

  2. 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

  3. Conceptual design of a laser-fusion power plant. Part II. Two technical options: 1. JADE reactor; 2. Heat transfer by heat pipes

    SciTech Connect

    Not Available

    1981-07-01

    A laser fusion reactor concept is described that employs liquid metal walls. The concept envisions a porous medium, called the JADE, of specific geometry lining the reactor cavity. Some advantages and disadvantages of the concept are pointed out. The possibility of using heat pipes for passive cooling in ICF reactors is discussed. Some of the problems are outlined. (MOW)

  4. 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.

  5. A target station for plasma exposure of neutron irradiated fusion material samples to reactor relevant conditions

    NASA Astrophysics Data System (ADS)

    Rapp, Juergen; Giuliano, Dominic; Ellis, Ronald; Howard, Richard; Lore, Jeremy; Lumsdaine, Arnold; Lessard, Timothy; McGinnis, William; Meitner, Steven; Owen, Larry; Varma, Venugopal

    2015-11-01

    The Material Plasma Exposure eXperiment (MPEX) is a device planned to address scientific and technological gaps for the development of viable plasma facing components for fusion reactor conditions (FNSF, DEMO). It will have to address the relevant plasma conditions in a reactor divertor (electron density, electron temperature, ion fluxes) and it needs to be able to expose a-priori neutron irradiated samples. A pre design of a target station able to handle activated materials will be presented. This includes detailed MCNP as well as SCALE and MAVRIC calculations for all potential plasma-facing materials to estimate dose rates. Details on the remote handling schemes for the material samples will be presented. 2 point modeling of the linear plasma transport has been used to scope out the parameter range of the anticipated power fluxes to the target. This has been used to design the cooling capability of the target. The operational conditions of surface temperatures, plasma conditions, and oblique angle of incidence of magnetic field to target surface will be discussed. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under contract DE-AC-05-00OR22725.

  6. Technical evaluation of major candidate blanket systems for fusion power reactor

    NASA Astrophysics Data System (ADS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are ; (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li/sub 2/O/H/sub 2/O/Be, Mo-alloy/Li/sub 2/O/He/Be, Mo-alloy/LiAlO/sub 2//He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies.

  7. Development and testing ov danadium alloys for fusion applications

    SciTech Connect

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1996-10-01

    V base alloys have advantages for fusion reactor first-wall and blanket structure. To screen candidate alloys and optimize a V-base alloy, physical and mechanical properties of V-Ti, V-Cr-Ti, and V-Ti- Si alloys were studied before and after irradiation in Li environment in fast fission reactors. V-4Cr-4Ti containing 500-1000 wppM Si and <1000 wppM O+N+C was investigated as the most promising alloy, and more testing is being done. Major results of the work are presented in this paper. The reference V-4Cr-4Ti had the most attractive combination of the mechanical and physical properties that are prerequisite for first-wall and blanket structures: good thermal creep, good tensile strength/ductility, high impact energy, excellent resistance to swelling, and very low ductile-brittle transition temperature before and after irradiation. The alloy was highly resistant to irradiation-induced embrittlement in Li at 420-600 C, and the effects of dynamically charged He on swelling and mechanical properties were insignificant. However, several important issues remain unresolved: welding, low-temperature irradiation, He effect at high dose and high He concentration, irradiation creep, and irradiation performance in air or He. Initial results of investigation of some of these issues are also given.

  8. MYRRHA a multi-purpose hybrid research reactor for high-tech applications

    SciTech Connect

    Abderrahim, H. A.; Baeten, P.

    2012-07-01

    MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

  9. Fusion protein technologies for biopharmaceuticals: Applications and challenges

    PubMed Central

    Berger, Sven; Lowe, Peter; Tesar, Michael

    2015-01-01

    Stefan R. Schmidt consolidates the hugely diverse field of fusion proteins and their application in the creation of biopharmaceuticals. The text is replete with case studies and clinical data that inform and intrigue the reader as to the myriad possibilities available when considering the creation of a fusion protein. This valuable text will serve the novice as a broad introduction or the seasoned professional as a thorough review of the state of the art. The first marketed therapeutic recombinant protein was human insulin (Humulin® R). Its approval in 1982 was followed by other such products, including erythropoietin (EPO), interferon (IFN), and tissue plasminogen activator (tPa). Since the 1980s, the number and general availability of recombinant products that replace natural proteins harvested from animal or human sources has increased considerably. Following the initial success, researchers started de novo designs of therapeutic proteins that do not occur in nature. The first of these new drugs to be approved was etanercept (Enbrel®), a fusion portion containing a section of the tumor necrosis factor (TNF) receptor fused to the Fc portion of human IgG1.

  10. Integrated continuous dissolution, refolding and tag removal of fusion proteins from inclusion bodies in a tubular reactor.

    PubMed

    Pan, Siqi; Zelger, Monika; Jungbauer, Alois; Hahn, Rainer

    2014-09-20

    An integrated continuous tubular reactor system was developed for processing an autoprotease expressed as inclusion bodies. The inclusion bodies were suspended and fed into the tubular reactor system for continuous dissolving, refolding and precipitation. During refolding, the dissolved autoprotease cleaves itself, separating the fusion tag from the target peptide. Subsequently, the cleaved fusion tag and any uncleaved autoprotease were precipitated out in the precipitation step. The processed exiting solution results in the purified soluble target peptide. Refolding and precipitation yields performed in the tubular reactor were similar to batch reactor and process was stable for at least 20 h. The authenticity of purified peptide was also verified by mass spectroscopy. Productivity (in mg/l/h and mg/h) calculated in the tubular process was twice and 1.5 times of the batch process, respectively. Although it is more complex to setup a tubular than a batch reactor, it offers faster mixing, higher productivity and better integration to other bioprocessing steps. With increasing interest of integrated continuous biomanufacturing, the use of tubular reactors in industrial settings offers clear advantages. PMID:24950296

  11. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    SciTech Connect

    Donohue, M.L.; Price, M.E.

    1984-07-01

    Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen. (MOW)

  12. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  13. 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.

  14. Application of the JDL data fusion process model for cyber security

    NASA Astrophysics Data System (ADS)

    Giacobe, Nicklaus A.

    2010-04-01

    A number of cyber security technologies have proposed the use of data fusion to enhance the defensive capabilities of the network and aid in the development of situational awareness for the security analyst. While there have been advances in fusion technologies and the application of fusion in intrusion detection systems (IDSs), in particular, additional progress can be made by gaining a better understanding of a variety of data fusion processes and applying them to the cyber security application domain. This research explores the underlying processes identified in the Joint Directors of Laboratories (JDL) data fusion process model and further describes them in a cyber security context.

  15. Application of the JDL data fusion process model to hard/soft information fusion in the condition monitoring of aircraft

    NASA Astrophysics Data System (ADS)

    Bernardo, Joseph T.

    2014-05-01

    Hard/soft information fusion has been proposed as a way to enhance diagnostic capability for the condition monitoring of machinery. However, there is a limited understanding of where hard/soft information fusion could and should be applied in the condition monitoring of aircraft. Condition-based maintenance refers to the philosophy of performing maintenance when the need arises, based upon indicators of deterioration in the condition of the machinery. The addition of the multisensory capability of human cognition to electronic sensors may create a fuller picture of machinery condition. Since 1988, the Joint Directors of Laboratories (JDL) data fusion process model has served as a framework for information fusion research. Advances are described in the application of hard/soft information fusion in condition monitoring using terms that condition-based maintenance professionals in aviation will recognize. Emerging literature on hard/soft information fusion in condition monitoring is organized into the levels of the JDL data fusion process model. Gaps in the literature are identified, and the author's ongoing research is discussed. Future efforts will focus on building domain-specific frameworks and experimental design, which may provide a foundation for improving flight safety, increasing mission readiness, and reducing the cost of maintenance operations.

  16. Small reactor power system for space application

    NASA Technical Reports Server (NTRS)

    Shirbacheh, M.

    1987-01-01

    A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

  17. Evaluation of permeable and non-permeable tritium in normal condition in a fusion reactor

    NASA Astrophysics Data System (ADS)

    Marta, V.; A, S. Luis; Manuel, P. J.

    2008-05-01

    The tritium cycle, technologies of process and control of the tritium in the plant will constitute a fraction of the environmental impact of the first generation of DT fusion reactors. The efforts of conceptual development of the tritium cycle are centered in the Internal Regenerator Cycle. The tritium could be recovered from a flow of He gas, or directly from solid breeder. The limits of transfers to the atmosphere are assumed ~ 1 gr-T/a (~20 Ci/a) (without species distinction). In the case of ITER, for example, we have global demands of control of 5 orders of magnitude have been demonstrated at experimental level. The transfer limits determine the key parameters in tritium Cycle (HT, HTO, as dominant, and T2, T2O as marginal). Presently, the transfer from the cycle to the environment is assumed through the exchange system of the power plant (primary to secondary). That transport is due to the permeation through HT, T2, or leakage to the coolant in the primary system. It is key the chemical optimitation in the primary system, that needs to be reanalyzed in terms of radiological impact both for permeable, HT, T2, and non-permeable HTO, T2O. It is necessary considered the pathway of tritium from the reactor to the atmosphere, these processes are modelled adequately. Results of the assessments were early and chronic doses which have been evaluated for the Most Exposed Individual at particular distance bands from the release point. The impact evaluations will be performed with the computational tools (NORMTRI), besides national regulatory models, internationally accepted computer these code for dosimetric evaluations of tritiated effluents in operational conditions.

  18. SIRIUS-P: An inertially confined direct drive laser fusion power reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Kulcinski, G.L.; Moses, G.A.; Bruggink, D.; Engelstad, R.L.; Khater, H.Y.; Larsen, E.M.; Lovell, E.G.; MacFarlane, J.J.; Mogahed, E.A.; Peterson, R.R.; Sawan, M.E.; Wang, P.; Wittenberg, L.J.

    1993-03-01

    The SIRIUS-P conceptual design study is of a 1000 MWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. The reference driver is a KrF laser; however, any other laser capable of delivering short wavelength energy can be substituted. Sixty beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz and a target gain of 118. The spherical chamber has an internal diameter of 6.5 m and consists of two independent components, a first wall assembly fabricated from a c/c composite and a blanket assembly made of SiC. First wall protection is provided by a xenon buffer gas at a pressure of 0.5 torr. The chamber is cooled by a flowing granular bed of solid ceramic material, TiO[sub 2] for the first wall assembly and Li[sub 2]O for the blanket assembly. The chamber is housed within a 42 m radius cylindrical reactor building which is 86 m high and which shares the same vacuum space as the chamber. All the laser beams are brought in at the bottom of the building, first onto a dielectrically coated final focusing mirror and finally onto a metallic grazing incidence mirror which reflects them into the chamber through beam ports open to the building. Neutron traps behind the grazing incidence mirrors are used to prolong the lifetimes of the final focusing optics. The nominal cost of electricity from this system is 65 mills/kwh assuming an 8% interest rate on capital.

  19. SIRIUS-P: An inertially confined direct drive laser fusion power reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Kulcinski, G.L.; Moses, G.A.; Bruggink, D.; Engelstad, R.L.; Khater, H.Y.; Larsen, E.M.; Lovell, E.G.; MacFarlane, J.J.; Mogahed, E.A.; Peterson, R.R.; Sawan, M.E.; Wang, P.; Wittenberg, L.J.

    1993-03-01

    The SIRIUS-P conceptual design study is of a 1000 MWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. The reference driver is a KrF laser; however, any other laser capable of delivering short wavelength energy can be substituted. Sixty beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz and a target gain of 118. The spherical chamber has an internal diameter of 6.5 m and consists of two independent components, a first wall assembly fabricated from a c/c composite and a blanket assembly made of SiC. First wall protection is provided by a xenon buffer gas at a pressure of 0.5 torr. The chamber is cooled by a flowing granular bed of solid ceramic material, TiO{sub 2} for the first wall assembly and Li{sub 2}O for the blanket assembly. The chamber is housed within a 42 m radius cylindrical reactor building which is 86 m high and which shares the same vacuum space as the chamber. All the laser beams are brought in at the bottom of the building, first onto a dielectrically coated final focusing mirror and finally onto a metallic grazing incidence mirror which reflects them into the chamber through beam ports open to the building. Neutron traps behind the grazing incidence mirrors are used to prolong the lifetimes of the final focusing optics. The nominal cost of electricity from this system is 65 mills/kwh assuming an 8% interest rate on capital.

  20. Tritium production, management and its impact on safety for a D-{sup 3}He fusion reactor

    SciTech Connect

    Sze, D.K.; Herring, S.; Sawan, M.

    1991-11-01

    About three percent of the fusion energy produced by a D-{sup 3}He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, {sup 3}He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D-{sup 3}He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs.

  1. Silicon Carbide and Silicon Carbide Composites for Fusion Reactor Applications

    SciTech Connect

    Hinoki, Tatsuya; Hasegawa, Akira; Katoh, Yutai; Snead, Lance Lewis; Jung, H.C.; Katsui, Hirokazu; Kondo, Sosuke; Zhong, Z. H.; Park, Y. H.; Shih, Chunghao; Ozawa, Kazumi; Parish, Chad M; Meisner, Roberta Ann

    2013-01-01

    This paper reviews recent achievements as to "nuclear-grade" SiC composites in particular for materials-system integration. SiC composite component development are reviewed including VHTR control rod scale model and compact intermediate heat exchanger scale mode by current joining and assembly techniques. Joining methods for SiC to metal and results of characterization of joint shear strength by the torsion tests using small specimens were also reviewed. The recent results of neutron irradiation experiments were also reviewed including detailed analysis of mechanical properties, irradiation creep and preliminary results on tritium behavior in SiC.

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

    SciTech Connect

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

    1987-07-01

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

  3. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    SciTech Connect

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs.

  4. Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

    SciTech Connect

    Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi )

    1993-06-01

    If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basic phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. The authors plan a scaled-model test and a full-scale model test for the LOVE.

  5. Thermal analysis of coatings and substrate materials during a disruption in fusion reactors

    SciTech Connect

    Hassanein, A.

    1993-06-01

    In a tokamak fusion reactor, the frequency of occurrence and the severity of a plasma disruption event will determine the lifetime of the plasma facing components. Disruptions are plasma instabilities which result in rapid loss of confinement and termination of plasma current Intense energy fluxes to components like the rust wall and the divertor plate are expected during the disruptions. This high energy deposition in short times may cause severe surface erosion of these components resulting from melting and vaporization. Coatings and tile materials are proposed to protect and maintain the integrity of the underneath, structural materials from both erosion losses as well as from high thermal stresses encountered during a disruption. The coating thickness should be large enough to withstand both erosion losses and to reduce the temperature rise in the substrate structural material. The coating thickness should be minimized to enhance the structural integrity, to reduce potential problems from radioactivity, and to minimize materials cost. Tile materials such as graphite and coating materials such as beryllium and tungsten on structural materials like copper, steel, and vanadium are analyzed and compared as potential diverter and first wall design options. The effect of the sprayed coating properties during the disruption is investigated. Porous sprayed material may be found to protect the structure better than condensed phase properties. The minimum coating thickness required to protect the structural material during disruption is discussed. The impact of self shielding effect by the eroded material oil the response of both the type/coating and the substrate is discussed.

  6. Ion source development for a photoneutralization based NBI system for fusion reactors

    NASA Astrophysics Data System (ADS)

    Simonin, A.; de Esch, H. P. L.; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A.

    2015-04-01

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D- beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R&D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

  7. Development of optical components for in-vessel viewing systems used for fusion experimental reactor

    NASA Astrophysics Data System (ADS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Tada, Eisuke; Morita, Yosuke; Seki, Masahiro

    1994-12-01

    Optical components including imagefiber, periscope, glass, reflecting mirror and adhesive for lens are essential elements of in-vessel viewing system use for fusion experimental reactor and extensive of gamma irradiation tests have been conducted. These components were irradiated in the range of 1 MGy - 100 MGy under the average exposure dose rate of 1 X 106 R/h. As a result, the observation limit of the imagefiber specially fabricated for radiation hard is obtained to be 12 MGy at a illuminance of 8500 lx. Deterioration of transmissivity of three kinds of glass (alkaline barium glass, lead glass and synthetic quartz glass) is small compared with standard glass for commercial periscope. A periscope which was made of these glasses is visible even after 20 MGy at 8500 lx and in case of the standard periscope, the observation limit is 1 kGy at 8500 lx. Decrease in the reflectance on chromium nitride coated reflecting mirror is extremely small than aluminum coated and platinum coated mirrors at accumulated dose of 100 MGy. Two types of adhesive made of polyester resin and epoxy resin became discolored and exfoliated after 50 MGy.

  8. Ion source development for a photoneutralization based NBI system for fusion reactors

    SciTech Connect

    Simonin, A.; Esch, H. P. L. de; Garibaldi, P.; Grand, C.; Bechu, S.; Bès, A.; Lacoste, A.

    2015-04-08

    The next step after ITER is to demonstrate the viability and generation of electricity by a future fusion reactor (DEMO). The specifications required to operate an NBI system on DEMO are very demanding. The system has to provide a very high level of power and energy, ~100MW of D° beam at 1MeV, including high wall-plug efficiency (η > 60%). For this purpose, a new injector concept, called Siphore, is under investigation between CEA and French universities. Siphore is based on the stripping of the accelerated negative ions by photo-detachment provided by several Fabry-Perot cavities (3.5MW of light power per cavity) implemented along the D{sup −} beam. The beamline is designed to be tall and narrow in order that the photon flux overlaps the entire negative ion beam. The paper will describe the present R and D at CEA which addresses the development of an ion source and pre-accelerator prototypes for Siphore, the main goal being to produce an intense negative ion beam sheet. The negative ion source Cybele is based on a magnetized plasma column where hot electrons are emitted from the source center. Parametric studies of the source are performed using Langmuir probes in order to characterize the plasma and to compare with numerical models being developed in French universities.

  9. Comparison of implantation-driven permeation characteristics of fusion reactor structural materials

    SciTech Connect

    Longhurst, G.R.; Anderl, R.A.; Struttmann, D.A.

    1986-04-04

    Implantation-driven permeation experiments have been conducted on samples of the ferritic steel HT-9, the austenitic Primary Candidate Alloy (PCA) and the vanadium alloy V-15Cr-5Ti using D/sub 3//sup +/ ions under conditions that simulate charge-exchange neutral loading on a fusion reactor first wall. The steels all exhibited an initially intense permeation ''spike'' followed by an exponential decrease to low steady-state values. That spike was not evident in the V-15Cr-5Ti experiments. Steady-state permeation was highest in the vanadium alloy and lowest in the austenitic steel. Though permeation rates in the HT-9 were lower than those in V-15Cr-5Ti, permeation transients were much faster in HT-9 than in other materials tested. Ion-beam sputtering of the surface in the steel experiments resulted in enhanced remission at the front surface, whereas in the vanadium tests, recombination and diffusivity both appeared to diminish as the deuterium concentration rose. This may be due to a phase change in the material. We conclude that for conditions comparable to those of these experiments, tritium retention and loss in first wall structures made of steels will be less than in structures made of V-15Cr-5Ti.

  10. Modeling of liquid-metal corrosion/deposition in a fusion reactor blanket

    SciTech Connect

    Malang, S.; Smith, D.L.

    1984-04-01

    A model has been developed for the investigation of the liquid-metal corrosion and the corrosion product transport in a liquid-metal-cooled fusion reactor blanket. The model describes the two-dimensional transport of wall material in the liquid-metal flow and is based on the following assumptions: (1) parallel flow in a straight circular tube; (2) transport of wall material perpendicular to the flow direction by diffusion and turbulent exchange; in flow direction by the flow motion only; (3) magnetic field causes uniform velocity profile with thin boundary layer and suppresses turbulent mass exchange; and (4) liquid metal at the interface is saturated with wall material. A computer code based on this model has been used to analyze the corrosion of ferritic steel by lithium lead and the deposition of wall material in the cooler part of a loop. Three cases have been investigated: (1) ANL forced convection corrosion experiment (without magnetic field); (2) corrosion in the MARS liquid-metal-cooled blanket (with magnetic field); and (3) deposition of wall material in the corrosion product cleanup system of the MARS blanket loop.

  11. Low-temperature irradiation creep of fusion reactor structural materials*1

    NASA Astrophysics Data System (ADS)

    Grossbeck, M. L.; Mansur, L. K.

    1991-03-01

    Irradiation creep has been investigated in the Oak Ridge Research Reactor in an assembly spectrally tailored to achieve a He : dpa ratio of 12-14:1 appm/dpa in austenitic stainless steels. Temperatures of 60-400°C were investigated to address the requirements of near term fusion devices. It was found that austenitic alloys, especially PCA, have higher creep rates at 60°C than at 330 and 400°C. Since this phenomenon could not be explained by existing theoretical models, a new mechanism was proposed and a corresponding theoretical model was developed. Since vacancy migration times can be a few orders of magnitude longer than the irradiation times in this temperature regime, the immobile vacancies do not cancel climb produced by mobile interstitials absorbed at dislocations. The result is a high climb rate independent of stress-induced preferred absorption (SIPA) mechanisms. Preliminary calculations indicate that this mechanism coupled with preferred-absorption-driven glide at higher temperatures predicts a high creep rate at low temperatures and a weak temperature dependence of irradiation creep over the entire temperature range investigated.

  12. Thin blanket design for MINIMARS - a compact tandem mirror fusion reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Sawan, M.E.; El-Guebaly, L.A.; Wittenberg, L.J.; Corradini, M.L.; Vogelsang, W.F.; Kulcinski, G.L.

    1986-01-01

    A primary goal in the MINIMARS fusion power reactor design is to achieve the lowest possible cost of electricity and highest mass utilization while maintaining credibility and passive safety. Because the blanket impacts many components, reducing its thickness-while achieving adequate breeding and a high energy multiplication-was of prime importance. The MINIMARS blanket is a helium-gas-cooled design using 17Li-83Pb (LiPb) breeder, HT-9 structure, and beryllium moderator/multiplier. The helium gas is contained in small tubes that are immersed in a close-packed matrix of beryllium balls and LiPb. The result is a compact blanket only 18 cm thick in which only the tubes are operated in a stressed condition, but the blanket structure is designed to withstand a helium gas leak in one of the tubes. By circulating the helium gas from the blanket into the reflector, the reflector energy is recovered at a high temperature giving a gross power cycle efficiency of 42.7% while maintaining a low interface temperature between the breeding material and structure.

  13. Radioactivation of structural material of the superconducting magnet for a fusion reactor

    SciTech Connect

    Seki, Y.; Kawasaki, H.; Yamada, K.; Yamauchi, I.

    1983-08-01

    Radioactivation of five types of candidate steel alloys for the structural materials of superconducting toroidal field coils (TFC) of a D-T fusion reactor has been comparatively studied. As a result, the use of a high Mn steel in place of 316 SS is shown to reduce the dose rate at the He vessel of the TFC to about1/3 the value with 316 SS at 1 day after shutdown, and to about1/1000 at 10 years after shutdown. These reductions are mostly caused by the 0.28 wt% Co assumed to be included in 316 SS but none in the high Mn steel. Newly defined dose rate sensitivities of constituent elements are shown to be useful in identifying the cause of dose rate change brought on by the steel composition change. They can also be utilized in estimating the dose rate change brought on by the replacement of 316 SS with any new steel alloy with similar composition.

  14. Radioactivation of structural material of the superconducting magnet for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Seki, Yasushi; Yamauchi, Isamu; Yamada, Koubun; Kawasaki, Hiromitsu

    1983-08-01

    Radioactivation of five types of candidate steel alloys for the structural materials of superconducting toroidal field coils (TFC) of a D-T fusion reactor has been comparatively studied. As a result, the use of a high Mn steel in place of 316 SS is shown to reduce the dose rate at the He vessel of the TFC to ˜ 1/3 the value with 316 SS at 1 day after shutdown, and to ˜ 1/1000 at 10 years after shutdown. These reductions are mostly caused by the 0.28 wt% Co assumed to be included in 316 SS but none in the high Mn steel. Newly defined dose rate sensitivities of constituent elements are shown to be useful in identifying the cause of dose rate change brought on by the steel composition change. They can also be utilized in estimating the dose rate change brought on by the replacement of 316 SS with any new steel alloy with similar composition.

  15. Radiation Damage and Tritium Breeding Study in a Fusion Reactor Using a Liquid Wall of Various Thorium Molten Salts

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa

    2007-12-01

    A new magnetic fusion reactor design, called APEX uses a liquid wall between fusion plasma and solid first wall to reach high neutron wall loads and eliminate the replacement of the first wall structure during the reactor's operation due to the radiation damage. In this paper, radiation damage behavior of the inboard and outboard first walls made of a ferritic steel, 9Cr-2WVTa, in the APEX blanket using various thorium molten salts, 75% LiF-25% ThF4, 75% LiF-24% ThF4-1% 233UF4 and 75% LiF-23% ThF4-2% 233UF4 was investigated. Furthermore, tritium breeding potential of these salts in such a blanket was also examined. Computations were carried out using the code Scale 4.3 by solving Boltzmann neutron transport equation. Numerical results brought out that only the liquid wall containing the molten salt, 75% LiF-23% ThF4-2% 233UF4 and having a thickness of ≥38 cm would be suitable to be used in the APEX reactor with respect to radiation damage criteria for the first wall structures and tritium self-sufficiency for the (DT) fusion driver.

  16. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  17. Response to FESAC survey, non-fusion connections to Fusion Energy Sciences. Applications of the FES-supported beam and plasma simulation code, Warp

    SciTech Connect

    Friedman, A.; Grote, D. P.; Vay, J. L.

    2015-05-29

    The Fusion Energy Sciences Advisory Committee’s subcommittee on non-fusion applications (FESAC NFA) is conducting a survey to obtain information from the fusion community about non-fusion work that has resulted from their DOE-funded fusion research. The subcommittee has requested that members of the community describe recent developments connected to the activities of the DOE Office of Fusion Energy Sciences. Two questions in particular were posed by the subcommittee. This document contains the authors’ responses to those questions.

  18. Reactor for boron fusion with picosecond ultrahigh power laser pulses and ultrahigh magnetic field trapping

    NASA Astrophysics Data System (ADS)

    Miley, G. H.; Hora, H.; Kirchhoff, G.

    2016-05-01

    Compared with the deuterium tritium (DT) fusion, the environmentally clean fusion of protons with 11B is extremely difficult. When instead of nanosecond laser pulses for thermal-ablating driven ignition, picosecond pulses are used, a drastic change by nonlinearity results in ultrahigh acceleration of plasma blocks. This radically changes to economic boron fusion by a measured new avalanche ignition.

  19. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    NASA Astrophysics Data System (ADS)

    Pilan, N.; Antoni, V.; De Lorenzi, A.; Chitarin, G.; Veltri, P.; Sartori, E.

    2016-02-01

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.

  20. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices.

    PubMed

    Pilan, N; Antoni, V; De Lorenzi, A; Chitarin, G; Veltri, P; Sartori, E

    2016-02-01

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming. PMID:26932053

  1. Dynamic reactor modeling with applications to SPR and ZEDNA.

    SciTech Connect

    Suo-Anttila, Ahti Jorma

    2011-12-01

    A dynamic reactor model has been developed for pulse-type reactor applications. The model predicts reactor power, axial and radial fuel expansion, prompt and delayed neutron population, and prompt and delayed gamma population. All model predictions are made as a function of time. The model includes the reactivity effect of fuel expansion on a dynamic timescale as a feedback mechanism for reactor power. All inputs to the model are calculated from first principles, either directly by solving systems of equations, or indirectly from Monte Carlo N-Particle Transport Code (MCNP) derived results. The model does not include any empirical parameters that can be adjusted to match experimental data. Comparisons of model predictions to actual Sandia Pulse Reactor SPR-III pulses show very good agreement for a full range of pulse magnitudes. The model is also applied to Z-pinch externally driven neutron assembly (ZEDNA) type reactor designs to model both normal and off-normal ZEDNA operations.

  2. Glassy materials investigated for nuclear reactor applications

    NASA Technical Reports Server (NTRS)

    Lynch, E. D.

    1968-01-01

    Studies determine the feasibility of preparing fuel-bearing glasses and glasses bearing neutron-absorbing materials for use as crystalline fuel and control rods for reactors. Properties investigated were devitrification resistance, urania solubility, and density.

  3. Heat pipe reactors for space power applications

    NASA Technical Reports Server (NTRS)

    Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  4. Artificial intelligence program in a computer application supporting reactor operations

    SciTech Connect

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II.

  5. Surface analysis of 1984/1985 Tokamak Fusion Test Reactor limiter tiles

    SciTech Connect

    Pontau, A.E.; Wampler, W.R.; Mills, B.E.; Doyle, B.L.; Wright, A.F.; Ulrickson, M.A.; LaMarche, P.H.; Dylla, H.F.; Fukuda, S.

    1986-05-01

    Bare POCO AXF-5Q graphite tiles were installed as the plasma-facing surface of the Tokamak Fusion Test Reactor (TFTR) movable limiter for a series of approx. 2700 high power discharges (600 with up to 6 MW neutral beams). During this operating phase, erosion and deposition processes modified the surface of the limiter. In the regions of the most intense plasma contact, which reached temperatures over 2400/sup 0/C, only small amounts of metallic impurities (<5 x 10/sup 16/ atoms/cm/sup 2/) are observed during subsequent beta backscattering and proton-induced x-ray emission (PIXE) analysis. Also observed in these regions are several small areas of surface crazing, and scattered droplets of metals (approx. 1 mm diam) presumed to originate from melted internal hardware. In regions more removed from direct plasma contact, thicker metallic deposits are observed (approx. 2 x 10/sup 18/ atoms/cm/sup 2/). There are more of these deposits on the side facing the electron drift direction than on the ion drift side. The metallic deposits are composed of approx. 55% Ni, 25% Cr, and 20% Fe and several trace elements. Near-surface retained deuterium levels vary from 10/sup 16/ to 10/sup 18//cm/sup 2/ along one ion-side/electron-side pair of tiles. Deuterium from TFTR operations in the bulk of a tile may have been as much as 5 ppm. Hydrogen in the tile was approx.1 at. %. Surface roughness measured outside the intense plasma contact region is much more than the original POCO AXF-5Q graphite, whereas within the plasma contact region, tile surfaces are somewhat smoother than they were initially. Two of the tiles in high heat flux regions broke during operation. The resulting geometry change led to areas of enhanced erosion and nearby redeposition of carbon.

  6. Range shortening, radiation transport, and Rayleigh-Taylor instability phenomena in ion-beam-driven inertial-fusion-reactor-size targets: Implosion, ignition, and burn phases

    SciTech Connect

    Long, K.A.; Tahir, N.A.

    1987-03-15

    In this paper we present an analysis of the theory of the energy deposition of ions in cold materials and hot dense plasmas together with numerical calculations for heavy and light ions of interest to ion-beam fusion. We have used the g-smcapso-smcapsr-smcapsg-smcapso-smcapsn-smcaps computer code of Long, Moritz, and Tahir (which is an extension of the code originally written for protons by Nardi, Peleg, and Zinamon) to carry out these calculations. The energy-deposition data calculated in this manner has been used in the design of heavy-ion-beam-driven fusion targets suitable for a reactor, by its inclusion in the m-smcapse-smcapsd-smcapsu-smcapss-smcapsa-smcaps code of Christiansen, Ashby, and Roberts as extended by Tahir and Long. A number of other improvements have been made in this code and these are also discussed. Various aspects of the theoretical analysis of such targets are discussed including the calculation of the hydrodynamic stability, the hydrodynamic efficiency, and the gain. Various different target designs have been used, some of them new. In general these targets are driven by Bi/sup +/ ions of energy 8--12 GeV, with an input energy of 4--6.5 MJ, with output energies in the range 600--900 MJ, and with gains in the range 120--180. The peak powers are in the range of 500--750 TW. We present detailed calculations of the ablation, compression, ignition, and burn phases. By the application of a new stability analysis which includes ablation and density-gradient effects we show that these targets appear to implode in a stable manner. Thus the targets designed offer working examples suited for use in a future inertial-confinement fusion reactor.

  7. Homeland security application of the Army Soft Target Exploitation and Fusion (STEF) system

    NASA Astrophysics Data System (ADS)

    Antony, Richard T.; Karakowski, Joseph A.

    2010-04-01

    A fusion system that accommodates both text-based extracted information along with more conventional sensor-derived input has been developed and demonstrated in a terrorist attack scenario as part of the Empire Challenge (EC) 09 Exercise. Although the fusion system was developed to support Army military analysts, the system, based on a set of foundational fusion principles, has direct applicability to department of homeland security (DHS) & defense, law enforcement, and other applications. Several novel fusion technologies and applications were demonstrated in EC09. One such technology is location normalization that accommodates both fuzzy semantic expressions such as behind Library A, across the street from the market place, as well as traditional spatial representations. Additionally, the fusion system provides a range of fusion products not supported by traditional fusion algorithms. Many of these additional capabilities have direct applicability to DHS. A formal test of the fusion system was performed during the EC09 exercise. The system demonstrated that it was able to (1) automatically form tracks, (2) help analysts visualize behavior of individuals over time, (3) link key individuals based on both explicit message-based information as well as discovered (fusion-derived) implicit relationships, and (4) suggest possible individuals of interest based on their association with High Value Individuals (HVI) and user-defined key locations.

  8. Preliminary Evaluation of the Adequacy of Lithium Resources of the World and China for D-T Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Wang, Yongliang; Ni, Muyi; Jiang, Jieqiong; Wu, Yican; FDS-Team

    2012-07-01

    This paper studied the adequacy of the World and China lithium resources, considering the most promising uses in the future, involving nuclear fusion and electric-vehicles. The lithium recycle model for D-T fusion power plant and electric-vehicles, and the logistic growth prediction model of the primary energy for the World and China were constructed. Based on these models, preliminary evaluation of lithium resources adequacy of the World and China for D-T fusion reactors was presented under certain assumptions. Results show that: a. The world terrestrial reserves of lithium seems too limited to support a significant D-T power program, but the lithium reserves of China are relatively abundant, compared with the world case. b. The lithium resources contained in the oceans can be called the “permanent" energy. c. The change in 6Li enrichment has no obvious effect on the availability period of the lithium resources using FDS-II (Liquid Pb-17Li breeder blanket) type of reactors, but it has a stronger effect when PPCS-B (Solid Li4 SiO4 ceramics breeder blanket) is used.

  9. A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Piloted and Robotic Missions

    NASA Astrophysics Data System (ADS)

    Williams, C. H.; Borowski, S. K.; Dudzinski, L. A.; Juhasz, A. J.

    1999-11-01

    A conceptual space vehicle concept to support NASA's 21^st century requirements was designed to enable human, multi-month travel throughout the outer solar system. The design was predicated on an ignited, spherical torus fusion reactor (R=2.5 m; a=1.25 m) burning spin polarized D^3He fuel and operating at high beta (30%). Peaked plasma temperature (50 keV) and number density (5×10^20 m-3) profiles were used. Engineering design was performed on all major vehicle systems including fusion reactor, fast wave plasma heating, power conversion, magnetic nozzle (for direct plasma propulsion), tankage and others, with emphasis on 1D fusion power balance, operation physics, first wall, toroidal field coils, and heat transfer. Two related proof-of-concept experiments at OSU, LANL, and PPPL are discussed. Results showed a 108 mt crew habitat payload could be delivered to Saturn rendezvous in 214 days using 6,145 MW of plasma jet power.

  10. Heavy ion fusion science research for high energy density physics and fusion applications

    SciTech Connect

    LOGAN, B.G.; Logan, B.G.; Bieniosek, F.M.; Barnard, J.J.; Cohen, R.H.; Coleman, J.E.; Davidson, R.C.; Efthimion, P.C.; Friedman, A.; Gilson, E.P.; Greenway, W.G.; Grisham, L.; Grote, D.P.; Henestroza, E.; Hoffmann, D.H.H.; Kaganovich, I.D.; Kireeff Covo, M.; Kwan, J.W.; LaFortune, K.N.; Lee, E.P.; Leitner, M.; Lund, S.M.; Molvik, A.W.; Ni, P.; Penn, G.E.; Perkins, L.J.; Qin, H.; Roy, P.K.; Sefkow, A.B.; Seidl, P.A.; Sharp, W.; Startsev, E.A.; Varentsov, D.; Vay, J.-L.; Waldron, W.L.; Wurtele, J.S.; Welch, D.; Westenskow, G.A.; Yu, S.S.

    2007-06-25

    During the past two years, the U.S. heavy ion fusion science program has made significant experimental and theoretical progress in simultaneous transverse and longitudinal beam compression, ion-beam-driven warm dense matter targets, high brightness beam transport, advanced theory and numerical simulations, and heavy ion target designs for fusion. First experiments combining radial and longitudinal compression of intense ion beams propagating through background plasma resulted in on-axis beam densities increased by 700X at the focal plane. With further improvements planned in 2007, these results will enable initial ion beam target experiments in warm dense matter to begin next year at LBNL. We are assessing how these new techniques apply to low-cost modular fusion drivers and higher-gain direct-drive targets for inertial fusion energy.

  11. Characterization of scintillator materials for fast-ion loss detectors in nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Jiménez-Ramos, M. C.; García López, J.; García-Muñoz, M.; Rodríguez-Ramos, M.; Carmona Gázquez, M.; Zurro, B.

    2014-08-01

    In fusion plasma reactors, fast ion generated by heating systems and fusion born particles must be well confined. The presence of magnetohydrodynamic (MHD) instabilities can lead to a significant loss of these ions, which may reduce drastically the heating efficiency and may cause damage to plasma facing components in the vacuum vessel. In order to understand the physics underlying the fast ion loss mechanism, scintillator based detectors have been installed in several fusion devices. In this work we present the absolute photon yield and its degradation with ion fluence in terms of the number of photons emitted per incident ion of several scintillators thin coatings: SrGa2S4:Eu2+ (TG-Green), Y3Al5O12:Ce3+ (P46) and Y2O3:Eu3+ (P56) when irradiated with light ions of different masses (deuterium ions, protons and α-particles) at energies between approximately 575 keV and 3 MeV. The photon yield will be discussed in terms of the energy deposited by the particles into the scintillator. For that, the actual composition and thickness of the thin layers were determined by Rutherford Backscattering Spectrometry (RBS). A collimator with 1 mm of diameter, which defines the beam size for the experiments, placed at the entrance of the chamber. An electrically isolated sample holder biased to +300 V to collect the secondary electrons, connected to a digital current integrator (model 439 by Ortec) to measure the incident beam current. A home made device has been used to store the real-time evolution of the beam current in a computer file allowing the correction of the IL yields due to the current fluctuations. The target holder is a rectangle of 150 × 112 mm2 and can be tilted. The X and Y movements are controlled through stepping motors, which permits a fine control of the beam spot positioning as well as the study of several samples without venting the chamber. A silica optical fiber of 1 mm diameter fixed to the vacuum chamber, which collects the light from the scintillators

  12. Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    SciTech Connect

    Bourham, Mohamed A.; Gilligan, John G.

    1999-08-14

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

  13. Analysis of a Helium Brayton Power Cycle for a Direct-Drive Inertial Fusion Energy Power Reactor

    NASA Astrophysics Data System (ADS)

    Wagner, Scott; Gentile, Charles; Parsells, Robert; Priniski, Craig

    2008-11-01

    Presented is a thermodynamic model analysis and optimization of a helium Brayton power cycle for direct-drive inertial fusion energy (IFE) reactor. Preliminary reactor design goals include production of 2GW of thermal power and an estimated 700MW of electricity using a tertiary indirect helium Brayton cycle. A thermodynamic analysis of the proposed helium Brayton cycle is performed using baseline technology specifications and generalized thermodynamic assumptions. Analytic equations are developed using first and second law analysis. The model constraints are the turbine inlet temperature and pressure set by the reactor temperature of ˜700^oC and current turbine specifications of 7MPa, respectively. Optimization of this model is then performed using iterative numerical programming for key variables. Previous analysis shows a 51% cycle efficiency using current technology; best estimates of near-term technology increase the cycle efficiency to 64%. Results will be presented. R. Schleicher, A. R. Raffray, C. P. Wong, ``An Assessment of the Brayton Cycle for High Performance Power Plant,'' Fusion Technology, 39 (2), 823-827, March 2001.

  14. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  15. RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1996-04-30

    Version 00 CCC-388/RACC was specifically developed to compute the radioactivity and radioactivity-related parameters (e.g., afterheat, biological hazard potential, etc.) due to neutron activation within Inertial Fusion Energy and Magnetic Fusion Energy reactor systems. It can also be utilized to compute the radioactivity in fission, accelerator or any other neutron generating and neutron source system. This new version designated RACC-PULSE is based on CCC-388 and has the capability to model irradiation histories of varying flux levelsmore » having varying pulse widths (on times) and dwell periods (off times) and varying maintenance periods. This provides the user with the flexibility of modeling most any complexity of irradiation history beginning with simple steady state operating systems to complex multi-flux level pulse/intermittent operating systems.« less

  16. Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

    SciTech Connect

    Zhao, M. L.; Chen, Y. P.; Li, G. Q.; Luo, Z. P.; Guo, H. Y.; Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031; General Atomics, P.O. Box 85608, San Diego, California 92186 ; Ye, M. Y.; Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 ; Tendler, M.

    2014-05-15

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

  17. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 1, Final report

    SciTech Connect

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for inertial confinement reactors. The first of three volumes briefly discusses the following: Introduction; Key objectives, requirements, and assumptions; Systems modeling and trade studies; Prometheus-L reactor plant design overview; Prometheus-H reactor plant design overview; Key technical issues and R&D requirements; Comparison of IFE designs; and study conclusions.

  18. Characterization of scintillator materials for fast-ion loss detectors in nuclear fusion reactors

    NASA Astrophysics Data System (ADS)

    Jiménez-Ramos, M. C.; García López, J.; García-Muñoz, M.; Rodríguez-Ramos, M.; Carmona Gázquez, M.; Zurro, B.

    2014-08-01

    In fusion plasma reactors, fast ion generated by heating systems and fusion born particles must be well confined. The presence of magnetohydrodynamic (MHD) instabilities can lead to a significant loss of these ions, which may reduce drastically the heating efficiency and may cause damage to plasma facing components in the vacuum vessel. In order to understand the physics underlying the fast ion loss mechanism, scintillator based detectors have been installed in several fusion devices. In this work we present the absolute photon yield and its degradation with ion fluence in terms of the number of photons emitted per incident ion of several scintillators thin coatings: SrGa2S4:Eu2+ (TG-Green), Y3Al5O12:Ce3+ (P46) and Y2O3:Eu3+ (P56) when irradiated with light ions of different masses (deuterium ions, protons and α-particles) at energies between approximately 575 keV and 3 MeV. The photon yield will be discussed in terms of the energy deposited by the particles into the scintillator. For that, the actual composition and thickness of the thin layers were determined by Rutherford Backscattering Spectrometry (RBS). A collimator with 1 mm of diameter, which defines the beam size for the experiments, placed at the entrance of the chamber. An electrically isolated sample holder biased to +300 V to collect the secondary electrons, connected to a digital current integrator (model 439 by Ortec) to measure the incident beam current. A home made device has been used to store the real-time evolution of the beam current in a computer file allowing the correction of the IL yields due to the current fluctuations. The target holder is a rectangle of 150 × 112 mm2 and can be tilted. The X and Y movements are controlled through stepping motors, which permits a fine control of the beam spot positioning as well as the study of several samples without venting the chamber. A silica optical fiber of 1 mm diameter fixed to the vacuum chamber, which collects the light from the scintillators

  19. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  20. Polonium aspects associated with the use of lead-lithium blankets in fusion applications

    SciTech Connect

    Hoffman, N.J.; Blink, J.A.; Meier, W.R.; Murray, K.A.; Vogelsang, W.F.

    1985-07-01

    Polonium, an alpha-emitting sulfur-like element, is formed by neutron irradiation of lead or bismuth impurity in lead. Design studies of both the Pulse*Star inertial confinement fusion (ICF) reactor and the MARS mirror fusion reactor postulated use of 83Pb-17Li melt as the tritium breeding blanket and coolant. Comparison of the amounts of polonium in the melt at plant shutdown indicated that Pulse*Star would have a far higher level of polonium in the melt. Neutronic considerations and the polonium distribution between the vacuum cleanup system and 83Pb-17Li melt for the two reactors are explored in this paper. Sample neutronics runs showed that the codes used by each design team were not the source of the difference in polonium content.

  1. 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. PMID:14525368

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

    SciTech Connect

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

    2003-03-26

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

  3. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    SciTech Connect

    Rowcliffe, A.F.; Tsai, H.C.; Smith, D.L.

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

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

    NASA Astrophysics Data System (ADS)

    Hartman, Charles W.; Reisman, David B.; McLean, Harry S.; Thomas, John

    2008-06-01

    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. 60 /proj/ads/abstracts/absload/springerxml.pl* /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9121: total 28 28 10894_2007_Article_9121.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9122: total 16 16 10894_2007_Article_9122.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9123: total 12 12 10894_2007_Article_9123.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9124: total 36 36 10894_2007_Article_9124.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9125: total 32 32 10894_2007_Article_9125.xml.meta /proj

  5. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  6. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  7. Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa

    2009-09-01

    Selection of lithium containing materials is very important in the design of a deuterium-tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid-solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.

  8. Monte Carlo Particle Transport Capability for Inertial Confinement Fusion Applications

    SciTech Connect

    Brantley, P S; Stuart, L M

    2006-11-06

    A time-dependent massively-parallel Monte Carlo particle transport calculational module (ParticleMC) for inertial confinement fusion (ICF) applications is described. The ParticleMC package is designed with the long-term goal of transporting neutrons, charged particles, and gamma rays created during the simulation of ICF targets and surrounding materials, although currently the package treats neutrons and gamma rays. Neutrons created during thermonuclear burn provide a source of neutrons to the ParticleMC package. Other user-defined sources of particles are also available. The module is used within the context of a hydrodynamics client code, and the particle tracking is performed on the same computational mesh as used in the broader simulation. The module uses domain-decomposition and the MPI message passing interface to achieve parallel scaling for large numbers of computational cells. The Doppler effects of bulk hydrodynamic motion and the thermal effects due to the high temperatures encountered in ICF plasmas are directly included in the simulation. Numerical results for a three-dimensional benchmark test problem are presented in 3D XYZ geometry as a verification of the basic transport capability. In the full paper, additional numerical results including a prototype ICF simulation will be presented.

  9. Molecular potential energy surfaces for interstellar chemistry and fusion applications

    NASA Astrophysics Data System (ADS)

    Braams, Bastiaan J.; Huang, Xinchuan; Jin, Zhong; Xie, Zhen; Zhang, Xiubin; Bowman, Joel M.; Sharma, Amit Raj; Scheider, Ralf

    2006-04-01

    In the Born-Oppenheimer approximation the electronic Schr"odinger equation is solved given the nuclear positions as parameters, and this defines the potential energy surface. We have used computational invariant theory and the MAGMA computer algebra system as an aid to develop representations for the potential energy and dipole moment surfaces that are fully invariant under permutations of like nuclei, extending an approach that for 3-body and 4-body systems has a long history, e.g. [J. N. Murrell et al. Molecular Potential Energy Functions, Wiley, 1984]. A many-body (cluster) expansion is used to describe reaction complexes. The methods have been applied in an almost routine way for systems of up to 7 nuclei, including several molecules that are of interest for interstellar chemistry and for the issue of hydrocarbon breakdown in fusion edge plasma: H5^+, CH5, CH5^+, C2H3^+, and their fragments, with C2H5^+ on the way. The mathematical and computional methods and the hydrocarbon applications will be presented.

  10. Development of ODS FeCrAl for compatibility in fusion and fission energy applications

    DOE PAGESBeta

    Pint, Bruce A.; Dryepondt, Sebastien N.; Unocic, Kinga A.; Hoelzer, David T.

    2014-11-15

    In this paper, oxide dispersion strengthened (ODS) FeCrAl alloys with 12–15% Cr are being evaluated for improved compatibility with Pb-Li for a fusion energy application and with high temperature steam for a more accident-tolerant light water reactor fuel cladding application. A 12% Cr content alloy showed low mass losses in static Pb-Li at 700°C, where a LiAlO2 surface oxide formed and inhibited dissolution into the liquid metal. All the evaluated compositions formed a protective scale in steam at 1200°C, which is not possible with ODS FeCr alloys. However, most of the compositions were not protective at 1400°C, which is amore » general and somewhat surprising problem with ODS FeCrAl alloys that is still being studied. More work is needed to optimize the alloy composition, microstructure and oxide dispersion, but initial promising tensile and creep results have been obtained with mixed oxide additions, i.e. Y2O3 with ZrO2, HfO2 or TiO2.« less

  11. Development of ODS FeCrAl for compatibility in fusion and fission energy applications

    SciTech Connect

    Pint, Bruce A.; Dryepondt, Sebastien N.; Unocic, Kinga A.; Hoelzer, David T.

    2014-11-15

    In this paper, oxide dispersion strengthened (ODS) FeCrAl alloys with 12–15% Cr are being evaluated for improved compatibility with Pb-Li for a fusion energy application and with high temperature steam for a more accident-tolerant light water reactor fuel cladding application. A 12% Cr content alloy showed low mass losses in static Pb-Li at 700°C, where a LiAlO2 surface oxide formed and inhibited dissolution into the liquid metal. All the evaluated compositions formed a protective scale in steam at 1200°C, which is not possible with ODS FeCr alloys. However, most of the compositions were not protective at 1400°C, which is a general and somewhat surprising problem with ODS FeCrAl alloys that is still being studied. More work is needed to optimize the alloy composition, microstructure and oxide dispersion, but initial promising tensile and creep results have been obtained with mixed oxide additions, i.e. Y2O3 with ZrO2, HfO2 or TiO2.

  12. Development of ODS FeCrAl for Compatibility in Fusion and Fission Energy Applications

    NASA Astrophysics Data System (ADS)

    Pint, B. A.; Dryepondt, S.; Unocic, K. A.; Hoelzer, D. T.

    2014-12-01

    Oxide dispersion strengthened (ODS) FeCrAl alloys with 12-15% Cr are being evaluated for improved compatibility with Pb-Li for a fusion energy application and with high temperature steam for a more accident-tolerant light water reactor fuel cladding application. A 12% Cr content alloy showed low mass losses in static Pb-Li at 700°C, where a LiAlO2 surface oxide formed and inhibited dissolution into the liquid metal. All the evaluated compositions formed a protective scale in steam at 1200°C, which is not possible with ODS FeCr alloys. However, most of the compositions were not protective at 1400°C, which is a general and somewhat surprising problem with ODS FeCrAl alloys that is still being studied. More work is needed to optimize the alloy composition, microstructure and oxide dispersion, but initial promising tensile and creep results have been obtained with mixed oxide additions, i.e. Y2O3 with ZrO2, HfO2 or TiO2.

  13. Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

    NASA Astrophysics Data System (ADS)

    Kawamura, H.; Ishitsuka, E.; Tsuchiya, K.; Nakamichi, M.; Uchida, M.; Yamada, H.; Nakamura, K.; Ito, H.; Nakazawa, T.; Takahashi, H.; Tanaka, S.; Yoshida, N.; Kato, S.; Ito, Y.

    2003-08-01

    The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

  14. Development status of CLAM steel for fusion application

    NASA Astrophysics Data System (ADS)

    Huang, Qunying

    2014-12-01

    The China low activation martensitic (CLAM) steel is being developed at the Institute of Nuclear Energy Safety Technology (INEST) under wide collaboration within China. Significant R&D work on CLAM steel was carried out to help make it suitable for industrial applications. The effect of refining processes and thermal aging on composition, microstructures and mechanical properties were investigated. Material properties before irradiation including impact, fracture toughness, thermal aging, creep and fatigue properties etc. were assessed. A series of irradiation tests in the fission reactor HFETR in Chengdu up to 2 dpa and in the spallation neutron source SINQ in Paul Scherrer Institute up to 20 dpa were performed. PbLi corrosion tests for more than 10,000 h were done in the DRAGON-I and PICOLO loops. Fabrication techniques for a test blanket module (TBM) are being developed and a 1/3 scale TBM prototype is being fabricated with CLAM steel. Recent progresses on the development status of this steel are presented here. The code qualification of CLAM steel is under plan for its final application in ITER-TBM and DEMO in the future.

  15. Spatial Statistical Data Fusion for Remote Sensing Applications

    NASA Technical Reports Server (NTRS)

    Nguyen, Hai

    2010-01-01

    Data fusion is the process of combining information from heterogeneous sources into a single composite picture of the relevant process, such that the composite picture is generally more accurate and complete than that derived from any single source alone. Data collection is often incomplete, sparse, and yields incompatible information. Fusion techniques can make optimal use of such data. When investment in data collection is high, fusion gives the best return. Our study uses data from two satellites: (1) Multiangle Imaging SpectroRadiometer (MISR), (2) Moderate Resolution Imaging Spectroradiometer (MODIS).

  16. Fiber-Optical Sensors: Basics and Applications in Multiphase Reactors

    PubMed Central

    Li, Xiangyang; Yang, Chao; Yang, Shifang; Li, Guozheng

    2012-01-01

    This work presents a brief introduction on the basics of fiber-optical sensors and an overview focused on the applications to measurements in multiphase reactors. The most commonly principle utilized is laser back scattering, which is also the foundation for almost all current probes used in multiphase reactors. The fiber-optical probe techniques in two-phase reactors are more developed than those in three-phase reactors. There are many studies on the measurement of gas holdup using fiber-optical probes in three-phase fluidized beds, but negative interference of particles on probe function was less studied. The interactions between solids and probe tips were less studied because glass beads etc. were always used as the solid phase. The vision probes may be the most promising for simultaneous measurements of gas dispersion and solids suspension in three-phase reactors. Thus, the following techniques of the fiber-optical probes in multiphase reactors should be developed further: (1) online measuring techniques under nearly industrial operating conditions; (2) corresponding signal data processing techniques; (3) joint application with other measuring techniques.

  17. Final report SI 08-SI-004: Fusion application targets

    SciTech Connect

    Biener, J; Kucheyev, S O; Wang, M Y; Dawedeit, C; Worsley, M A; Kim, S H; Walton, C; Gilmer, G; Zepeda-Ruiz, L; Chernov, A A; Lee, J I; Willey, T M; Biener, M M; van Buuren, T; Wu, K J; Satcher, J H; Hamza, A V

    2010-12-03

    Complex target structures are necessary to take full advantage of the unique laboratory environment created by inertial confinement fusion experiments. For example, uses-of-ignition targets that contain a thin layer of a low density nanoporous material inside a spherical ablator shell allow placing dopants in direct contact with the DT fuel. The ideal foam for this application is a low-density hydrocarbon foam that is strong enough to survive wetting with cryogenic hydrogen, and low enough in density (density less than {approx}30 mg/cc) to not reduce the yield of the target. Here, we discuss the fabrication foam-lined uses-of-ignition targets, and the development of low-density foams that can be used for this application. Much effort has been directed over the last 20 years toward the development of spherical foam targets for direct-drive and fast-ignition experiments. In these targets, the spherical foam shell is used to define the shape of the cryogenic DT fuel layer, or acts as a surrogate to simulate the cryogenic fuel layer. These targets are fabricated from relatively high-density aerogels (>100 mg/cc) and coated with a few micron thick permeation barrier. With exception of the above mentioned fast ignition targets, the wall of these targets is typically larger than 100 microns. In contrast, the fusion application targets for indirect-drive experiments on NIF will require a much thinner foam shell surrounded by a much thicker ablator shell. The design requirements for both types of targets are compared in Table 1. The foam shell targets for direct-drive experiments can be made in large quantities and with reasonably high yields using an encapsulation technique pioneered by Takagi et al. in the early 90's. In this approach, targets are made by first generating unsupported foam shells using a triple-orifice droplet generator, followed by coating the dried foam shells with a thin permeation barrier. However, this approach is difficult, if not impossible, to

  18. Magnetic-Nozzle Studies for Fusion Propulsion Applications: Gigawatt Plasma Source Operation and Magnetic Nozzle Analysis

    NASA Technical Reports Server (NTRS)

    Gilland, James H.; Mikekkides, Ioannis; Mikellides, Pavlos; Gregorek, Gerald; Marriott, Darin

    2004-01-01

    This project has been a multiyear effort to assess the feasibility of a key process inherent to virtually all fusion propulsion concepts: the expansion of a fusion-grade plasma through a diverging magnetic field. Current fusion energy research touches on this process only indirectly through studies of plasma divertors designed to remove the fusion products from a reactor. This project was aimed at directly addressing propulsion system issues, without the expense of constructing a fusion reactor. Instead, the program designed, constructed, and operated a facility suitable for simulating fusion reactor grade edge plasmas, and to examine their expansion in an expanding magnetic nozzle. The approach was to create and accelerate a dense (up to l0(exp 20)/m) plasma, stagnate it in a converging magnetic field to convert kinetic energy to thermal energy, and examine the subsequent expansion of the hot (100's eV) plasma in a subsequent magnetic nozzle. Throughout the project, there has been a parallel effort between theoretical and numerical design and modelling of the experiment and the experiment itself. In particular, the MACH2 code was used to design and predict the performance of the magnetoplasmadynamic (MPD) plasma accelerator, and to design and predict the design and expected behavior for the magnetic field coils that could be added later. Progress to date includes the theoretical accelerator design and construction, development of the power and vacuum systems to accommodate the powers and mass flow rates of interest to out research, operation of the accelerator and comparison to theoretical predictions, and computational analysis of future magnetic field coils and the expected performance of an integrated source-nozzle experiment.

  19. 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''.

  20. Research on the HYLIFE liquid-first-wall concept for future laser-fusion reactors. Final report No. 5

    SciTech Connect

    Hoffman, M.A.

    1980-09-01

    It has been proposed to protect the structural walls of a future laser fusion reactor with a curtain or fluid-wall of liquid lithium jets. As part of the investigation of this concept, experiments have been performed on planar sheet water jets issuing vertically downward from slit nozzles. The nozzles were subjected to transverse forced harmonic excitation to simulate the vibrational environment of the laser fusion reactor, and experiments were run at both 1 atm and at lower ambient pressures. Linear temporal stability theory is shown to predict the onset of the unstable regime and the initial spatial growth rates quite well for the cases where the amplitudes of the nozzle vibration are not too large and the waveform is nearly sinusoidal. In addition, both the linear theory and a simplified trajectory theory are shown to predict the initial wave envelope amplitudes very well. For larger amplitude nozzle excitation, the waveform becomes highly nonlinear and non-sinusoidal and can resemble a sawtooth waveform in some cases; these latter experimental results can only be partially explained by existing theories at the present time.