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Sample records for confinement fusion reactor

  1. Designing the Cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1987-01-01

    The primary goal in designing inertial confinement fusion (ICF) reactors is to produce electrical power as inexpensively as possible, with minimum activation and without compromising safety. This paper discusses a method for designing the Cascade rotating ceramic-granule-blanket reactor (Pitts, 1985) and its associated power plant (Pitts and Maya, 1985). Although focus is on the cascade reactor, the design method and issues presented are applicable to most other ICF reactors

  2. Trends and developments in magnetic confinement fusion reactor concepts

    International Nuclear Information System (INIS)

    Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

    1981-01-01

    An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts. Emphasis is placed on reactors that operate on the deuterium/tritium/lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts. The paper emphasizes recent developments of these concepts within the last two to three years

  3. Confinement inertial fusion. Power reactors of nuclear fusion by lasers

    International Nuclear Information System (INIS)

    Velarde, G.; Ahnert, C.; Aragones, J.M.; Leira, G; Martinez-Val, J.M.

    1980-01-01

    The energy crisis and the need of the nuclear fusion energy are analized. The nuclear processes in the laser interation with the ablator material are studied, as well as the thermohydrodinamic processes in the implossion, and the neutronics of the fusion. The fusion reactor components are described and the economic and social impact of its introduction in the future energetic strategies.(author)

  4. Fuel and helium confinement in fusion reactors

    International Nuclear Information System (INIS)

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

    1993-01-01

    An expanded macroscopic model for particle confinement is used to investigate both fuel and helium confinement in reactor plasmas. The authors illustrate the relative effects of external sources of fuel, divertor pumping, and wall and divertory recycle on core, edge and scrape-off layer densities by using separate particle confinement times for open-quote core close-quote fueling (deep pellet or beam penetration, τ c ), open-quote shallow close-quote fueling (shallow pellet penetration or neutral atoms that penetrate the scrape-off layer, τ s ) and fueling in the scrape-off layer (τ sol ). Because τ s is determined by the parallel flow velocity and characteristic distance to the divertor plate, it can be orders of magnitude lower than either τ c or τ sol . A dense scrape-off region, desirable for reduced divertor erosion, leads to a high fraction of the recycled neutrals being ionized in the scrape-off region and poor core fueling efficiency. The overall fueling efficiency can then be dramatically improved with either shallow or deep auxillary fueling. Helium recycle is nearly always coupled to the scrape-off region and does not lead to strong core accumulation unless the helium pumping efficiency is much less than the fuel pumping efficiency, or the plasma preferentially retains helium over hydrogenic ions. Differences between the results of this model, single-τ p macroscopic models, and 1-D and 2-D models are discussed in terms of assumptions and boundary conditions

  5. Pulse Star inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Blink, J.A.; Hogan, W.J.

    1985-01-01

    Pulse Star is a pool-type ICF reactor that emphasizes low cost and high safety levels. The reactor consists of a vacuum chamber (belljar) submerged in a compact liquid metal (Li 17 Pb 83 or lithium) pool which also contains the heat exchangers and liquid metal pumps. The shielding efficiency of the liquid metal pool is high enough to allow hands-on maintenance of (removed) pumps and heat exchangers. Liquid metal is allowed to spray through the 5.5 m radius belljar at a controlled rate, but is prohibited from the target region by a 4 m radius mesh first wall. The wetted first wall absorbs the fusion x-rays and debris while the spray region absorbs the fusion neutrons. The mesh allows vaporized liquid metal to blow through to the spray region where it can quickly cool and condense. Preliminary calculations show that a 2 m thick first wall could handle the mechanical (support, buckling, and x-ray-induced hoop) loads. Wetting and gas flow issues are in an initial investigation stage

  6. Reactor design considerations for inertial confinement fusion

    International Nuclear Information System (INIS)

    Booth, L.A.

    1979-01-01

    The most challenging reactor design consideration is protection of the cavity wall from the various energy forms as released by the pellet and as affected by the reaction-chamber phenomena. These phenomena depend on both the design and the yield of the pellet, as well as on ambient conditions in the chamber at the time of the pellet microexplosion. The effects on pellet energy-release mechanisms of various reaction chamber atmosphere options are summarized

  7. Heat transfer in inertial confinement fusion reactor systems

    International Nuclear Information System (INIS)

    Hovingh, J.

    1979-01-01

    The transfer of energy produced by the interaction of the intense pulses of short-ranged fusion microexplosion products with materials is one of the most difficult problems in inertially-confined fusion (ICF) reactor design. The short time and deposition distance for the energy results in local peak power densities on the order of 10 18 watts/m 3 . High local power densities may cause change of state or spall in the reactor materials. This will limit the structure lifetimes for ICF reactors of economic physical sizes, increasing operating costs including structure replacement and radioactive waste management. Four basic first wall protection methods have evolved: a dry-wall, a wet-wall, a magnetically shielded wall, and a fluid wall. These approaches are distinguished by the way the reactor wall interfaces with fusion debris as well as the way the ambient cavity conditions modify the fusion energy forms and spectra at the first wall. Each of these approaches requires different heat transfer considerations

  8. Analysis of Confinement Strategies for a Tokamak Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

    2001-01-15

    The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established.

  9. Analysis of Confinement Strategies for a Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

    2001-01-01

    The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established

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

    International Nuclear Information System (INIS)

    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

  11. Electromagnetic pumping of liquid lithium in inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Baker, R.S.; Blink, J.A.; Tessier, M.J.

    1983-01-01

    The basic operating principles and geometries of ten electromagnetic pumps are described. Two candidate pumps, the annular-linear-induction pump and the helical-rotor electromagnetic pump, are compared for possible use in a full-scale liquid-lithium inertial confinement fusion reactor. A parametric design study completed for the helical-rotor pump is shown to be valid when applied to an experimental sodium pump. Based upon the preliminary HYLIFE requirements for a lithium flow rate per pump of 8.08 m 3 /s at a head of 82.5 kPa, a complete set of 70 variables are specified for a helical-rotor pump with either a normally conducting or a superconducting winding. The two alternative designs are expected to perform with efficiencies of 50 and 60%, respectively

  12. Present status of inertial confinement fusion reactor design

    International Nuclear Information System (INIS)

    Mima, Kunioki; Ido, Shunji; Nakai, Sadao.

    1986-01-01

    Since inertial nuclear fusion reactors do not require high vacuum and high magnetic field, the structure of the reactor cavity becomes markedly simple as compared with tokamak type fusion reactors. In particular, since high vacuum is not necessary, liquid metals such as lithium and lead can be used for the first wall, and the damage of reactor structures by neutrons can be prevented. As for the core, the energy efficiency of lasers is not very high, accordingly it must be designed so that the pellet gain due to nuclear fusion becomes sufficiently high, and typically, the gain coefficient from 100 to 200 is necessary. In this paper, the perspective of pellet gain, the plan from the present status to the practical reactors, and the conceptual design of the practical reactors are discussed. The plan of fuel ignition, energy break-even and high gain by the implosion mode, of which the uncertain factor due to uneven irradiation and instability was limited to the minimum, was clarified. The scenario of the development of laser nuclear fusion reactors is presented, and the concept of the reactor system is shown. The various types of nuclear fusion-fission hybrid reactors are explained. As for the design of inertial fusion power reactors, the engineering characteristics of the core, the conceptual design, water fall type reactors and DD fuel reactors are discussed. (Kako, I.)

  13. Performance of the Cascade inertial-confinement-fusion conceptual reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1984-01-01

    A 4.5-m-radius rotating fusion reactor made of silicon carbide and containing a moving 1-m-thick lithium-ceramic granular blanket can produce 3000 MW/sub t/. The blanket operates at high temperature (>1200 K) leading to gross plant efficiencies of up to 60% using a combined helium-gas turbine (Brayton cycle) with a vapor bottoming cycle

  14. Liquid jet experiments: relevance to inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1981-01-01

    In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

  15. SAFIRE: A systems analysis code for ICF [inertial confinement fusion] reactor economics

    International Nuclear Information System (INIS)

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

    1987-01-01

    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

  16. Inertial confinement fusion (ICF)

    International Nuclear Information System (INIS)

    Nuckolls, J.

    1977-01-01

    The principal goal of the inertial confinement fusion program is the development of a practical fusion power plant in this century. Rapid progress has been made in the four major areas of ICF--targets, drivers, fusion experiments, and reactors. High gain targets have been designed. Laser, electron beam, and heavy ion accelerator drivers appear to be feasible. Record-breaking thermonuclear conditions have been experimentally achieved. Detailed diagnostics of laser implosions have confirmed predictions of the LASNEX computer program. Experimental facilities are being planned and constructed capable of igniting high gain fusion microexplosions in the mid 1980's. A low cost long lifetime reactor design has been developed

  17. HYLIFE-II inertial confinement fusion reactor design

    International Nuclear Information System (INIS)

    Moir, R.W.

    1990-01-01

    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 2 BeF 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·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

  18. Initiative taken by India in magnetically confined fusion reactor

    International Nuclear Information System (INIS)

    Bora, Dhiraj

    2017-01-01

    There is a growing gap between demand and supply of energy in the world. Any attempt to develop new and cleaner sources of energy to meet the future global requirement is welcome. Therefore, it is attractive to think of having fusion as an alternate clean source of energy to contribute in the energy mix towards the second half of the century, with a virtually inexhaustible fuel supply. The environmental impact of fusion would be acceptable and relatively safe. These advantages have driven the world fusion research programme since its inception. Indian progress in fusion science and technology and participation in ITER will be discussed during the talk

  19. Fusion, magnetic confinement

    International Nuclear Information System (INIS)

    Berk, H.L.

    1992-01-01

    An overview is presented of the principles of magnetic confinement of plasmas for the purpose of achieving controlled fusion conditions. Sec. 1 discusses the different nuclear fusion reactions which can be exploited in prospective fusion reactors and explains why special technologies need to be developed for the supply of tritium or 3 He, the probable fuels. In Sec. 2 the Lawson condition, a criterion that is a measure of the quality of confinement relative to achieving fusion conditions, is explained. In Sec. 3 fluid equations are used to describe plasma confinement. Specific confinement configurations are considered. In Sec. 4 the orbits of particle sin magneti and electric fields are discussed. In Sec. 5 stability considerations are discussed. It is noted that confinement systems usually need to satisfy stability constraints imposed by ideal magnetohydrodynamic (MHD) theory. The paper culminates with a summary of experimental progress in magnetic confinement. Present experiments in tokamaks have reached the point that the conditions necessary to achieve fusion are being satisfied

  20. Maintenance and waste treatment of tritium existing in and out of the fusion reactor systems. 6. Study of tritium confinement in the facility of a fusion reactor

    International Nuclear Information System (INIS)

    Kobayashi, Kazuhiro

    2000-01-01

    In a future fusion reactor, tritium confinement is one of the key issues for safety. Large amount of tritium (a few grams to a hundred grams level) has been handled safely at the several facilities in the world for fusion research under the multiple confinement concept. In this chapter, the study of tritium behavior in large space such as the building is described using the Caisson Assembly for Tritium Safety (CATS) study such as the final confinement and the present R and D status concerning the tritium confinement is reviewed. (author)

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

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1985-01-01

    Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li 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 2 granules. Average blanket exit temperature is 1670 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

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

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1985-01-01

    Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li 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 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

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    Fisch, N.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

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

    International Nuclear Information System (INIS)

    Bers, A.

    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

  6. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  7. The fusion reactor

    International Nuclear Information System (INIS)

    Brennan, M.H.

    1974-01-01

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  8. Fourth annual progress report on special-purpose materials for magnetically confined fusion reactors

    International Nuclear Information System (INIS)

    1982-08-01

    The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. The Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits

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

    International Nuclear Information System (INIS)

    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 (ICF) 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 which 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

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

    International Nuclear Information System (INIS)

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

    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 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

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

    International Nuclear Information System (INIS)

    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

  12. Special-purpose materials for magnetically confined fusion reactors. Third annual progress report

    International Nuclear Information System (INIS)

    1981-11-01

    The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. It is recognized that there will be numerous materials problems that will arise during the design and construction of large magnetic-fusion energy devices such as the Engineering Test Facility (ETF) and Demonstration Reactor (DEMO). Most of these problems will be specific to a particular design or project and are the responsibility of the project, not the Materials and Radiation Effects Branch. Consequently, the Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits

  13. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    Energy Technology Data Exchange (ETDEWEB)

    Aleksandrova, I. V.; Koresheva, E. R., E-mail: elena.koresheva@gmail.com; Krokhin, O. N. [Russian Academy of Sciences, Lebedev Physical Institute (Russian Federation); Osipov, I. E. [Power Efficiency Centre, Inter RAO UES (Russian Federation)

    2016-12-15

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  14. Terrestrial and extraterrestrial superresonators as drivers for an inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Vath, W.

    1992-01-01

    This paper reports on the recirculating power fraction of a laser-driven inertial confinement fusion (ICF) reactor which can be reduced by using laser diodes to pump a neodymium solid-state laser. To overcome the high costs of two-dimensional arrays of laser diodes, two types of superresonators are proposed: a terrestrially based one and an extraterrestrially based one on a geostationary orbit. Both are designed in such a way that a sequence of short laser pulses (10 to 20 ns wide), each with an energy of 5 to 10 MJ and a frequency of 10 Hz, are produced to trigger a deuterium-tritium ICF reactor. The terrestrial superresonator needs a much smaller number of two-dimensional laser diode arrays than a conventionally pumped once-through solid-state laser system, and the extraterrestrial resonator is pumped by means of concentrated solar radiation. In practice, at least an order of magnitude fewer laser diodes and crystalline calcium fluoride gain media are needed to meet the requirements of a laser driver for an ICF reactor. If, finally, a liquid neodymium laser system could be used for an ICF reactor, the cooling of the gain slabs would be facilitated substantially

  15. The role of edge physics and confinement issues in the fusion reactor

    International Nuclear Information System (INIS)

    Lackner, K.; Schneider, R.

    1993-01-01

    The compatibility of the power exhaust to the walls and target plates with an adequate survival time of these structures and a sufficiently low impurity concentration in the bulk plasma, on the one hand, and the scaling of the energy confinement time with the tokamak parameters, on the other, are presently seen as two of the most critical issues determining the viability and minimum size of a fusion reactor. Statistical analysis of the world-wide data base of divertor tokamaks and experimental progress in superposing different forms of confinement improvements instil confidence that tokamak devices with the dimensions of ITER can meet the energy confinement requirements for ignition when operating in the H-regime. The critical problem consists, however, in simultaneously satisfying both the impurity and energy confinement related requirements. Regimes of improved confinement also tend to show a reduced rate of outward flow of impurities, thus increasing (for a given production rate at the walls or - in the case of helium - in the interior) their expected concentration. The confinement improvement in the H-regime also extends beyond the separatrix, leading to a thinner scrape-off layer with enhanced power concentration at the target plate and an enhanced penetration of impurities into the bulk plasma. These issues have therefore become the focal point of divertor and H-regime studies. Experiments on ASDEX, DIII-D and JET have successfully demonstrated that the impurity concentration in the H-regime can be effectively limited by the spontaneous cyclic destruction of the transport barrier at the edge (through the so-called ELMs) with only partial sacrifice of the energy confinement improvement. Expectations of handling the power load issues hinge on the capability of the divertor solution to decouple plasma properties in the edge of the bulk plasma from those near the target plates and the possibility of converting the energy flux arriving in the divertor into

  16. Reactor potential of the Magnetically Insulated Inertial Confinement Fusion (MICF) system

    International Nuclear Information System (INIS)

    Kammash, T.; Galbraith, D.L.

    1987-01-01

    In this paper a quasi one dimensional, time dependent set of particle and energy balance equations for the thermal species, namely, electrons, ions and thermal alphas which also allows for an appropriate set of fast alpha groups is utilized to assess the reactor prospects of a DT-burning Magnetically Insulated Inertial Confinement Fusion (MICF) system. A reference reactor consisting of an initial plasma with density of 10 21 cm -3 , temperature of keV, a radius of 0.25 cm is shown to ignite and yield an energy multiplication factor ''Q'' of about 60 when the plasma is allowed to burn for 2 microseconds. When the burntime is extended to 9 microseconds for the same initial conditions our calculations show that Q almost doubles just before the final radius becomes equal to the inner radius of the shell. These preliminary results seem to indicate that MICF does indeed have the potential for a reactor although some relevant physics issues need to be addressed first. 42 refs., 6 figs

  17. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  18. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  19. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  20. Pulse*Star Inertial Confinement Fusion Reactor: heat transfer loop and balance of plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Murray, K.A.

    1984-01-01

    A conceptual heat transfer loop and balance of plant design for the Pulse*Star Inertial Confinement Fusion Reactor has been investigated and results are presented. The Pulse*Star reaction vessel, a perforated steel bell jar approximately 11 m in diameter, is immersed in Li 17 Pb 83 coolant which flows through the perforations and forms a 1.5 m thick plenum of droplets around an 8 m diameter inner chamber. The reactor and associated pumps, piping, and steam generators are contained within a 17 m diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops with flow rates of 5.5 m 3 /s each are necessary to transfer 3300 MWt of power. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. Power balance calculations based on an improved electrical conversion efficiency revealed a net electrical output of 1260 MWe to the bus bar and a resulting net efficiency of 39%. Suggested balance-of-plant layouts are also presented

  1. Primary heat transfer loop design for the Cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Murray, K.A.; McDowell, M.W.

    1984-05-01

    This study investigates a heat exchanger and balance of plant design to accompany the Cascade inertial confinement fusion reaction chamber concept. The concept uses solid Li 2 O or other lithium-ceramic granules, held to the wall of a rotating reaction chamber by centrifugal action, as a tritium breeding blanket and first wall protection. The Li 2 O granules enter the chamber at 800 K and exit at 1200 K after absorbing the thermal energy produced by the fusion process

  2. SIRIUS-P: An inertially confined direct drive laser fusion power reactor

    International Nuclear Information System (INIS)

    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 2 for the first wall assembly and Li 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

  3. Pulse Star Inertial Confinement Fusion Reactor: Heat transfer loop and balance-of-plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Blink, J.A.; Curlander, K.A.

    1983-01-01

    A conceptual heat transfer loop and balance-of-plant design for the Pulse Star Inertial Confinement Fusion Reactor has been investigated and the results are presented. The Pulse Star reaction vessel, a perforated steel bell jar about11 m in diameter, is immersed in Li 17 Pb 83 coolant, which flows through the perforations and forms a 1.5-m-thick plenum of droplets around a 8-m-diameter inner chamber. The bell jar and associated pumps, piping, and steam generators are contained within a 17-m-diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops, each with a flow rate of 5.5 m 3 /s, are necessary to transfer 3300 MWt of power. Liquid metal is pumped to the top of the pool, where it flows downward through eight vertical steam generators. Double-walled tubes are used in the steam generators to assure tritium containment without intermediate heat transfer loops. Each pump is a mixed flow type and has a required NPSH of 3.4 m, a speed of 278 rpm, and an impeller diameter of 1.2 m. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. The design minimizes the total cost (heat exchanger area plus pumping) for the plant lifetime. The power required for the pumps is 36 MWe. Each resulting steam generator is 12 m high and 1.6 m in diameter, with 2360 tubes. The steam generators and pumps fit easily in the pool between the reactor chamber and the pool wall

  4. Condensation of ablated first-wall materials in the cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Ladd, A.J.C.

    1985-01-01

    This report concerns problems involved in recondensing first-wall materials vaporized by x rays and pellet debris in the Cascade inertial confinement fusion reactor. It examines three proposed first-wall materials, beryllium oxide (BeO), silicon carbide (SiO), and pyrolytic graphite (C), paying particular attention to the chemical equilibrium and kinetics of the vaporized gases. The major results of this study are as follows. Ceramic materials composed of diatomic molecules, such as BeO and SiC, exist as highly dissociated species after vaporization. The low gas density precludes significant recombination during times of interest (i.e., less than 0.1 s). The dissociated species (Be, O, Si, and C) are, except for carbon, quite volatile and are thermodynamically stable as a vapor under the high temperature and low density found in Cascade. These materials are thus unsuitable as first-wall materials. This difficulty is avoided with pyrolytic graphite. Since the condensation coefficient of monatomic carbon vapor (approx. 0.5) is greater than that of the polyatomic vapor (<0.1), recondensation is assisted by the expected high degree of dissociation. The proposed 10-layer granular carbon bed is sufficient to condense all the carbon vapor before it penetrates to the BeO layer below. The effective condensation coefficient of the porous bed is about 50% greater than that of a smooth wall. An estimate of the mass flux leaving the chamber results in a condensation time for a carbon first wall of about 30 to 50 ms. An experiment to investigate condensation in a Cascade-like chamber is proposed

  5. Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Werner, R.W.; Carlson, G.A.; Cornish, D.N.

    1976-01-01

    Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements. Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated

  6. Technology requirements for fusion--fission reactors based on magnetic-mirror confinement

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    Technology requirements for mirror hybrid reactors are discussed. The required 120-keV neutral beams can use positive ions. The magnetic fields are 8 T or under and can use NbTi superconductors. The value of Q (where Q is the ratio of fusion power to injection power) should be in the range of 1 to 2 for economic reasons relating to the cost of recirculating power. The wall loading of 14-MeV neutrons should be in the range of 1 to 2 MW/m 2 for economic reasons. Five-times higher wall loading will likely be needed if fusion reactors are to be economical. The magnetic mirror experiments 2XIIB, TMX, and MFTF are described

  7. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    1975-01-01

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  8. Inertial confinement fusion

    International Nuclear Information System (INIS)

    Nuckolls, J.H.; Wood, L.L.

    1988-01-01

    Edward Teller has been a strong proponent of harnessing nuclear explosions for peaceful purposes. There are two approaches: Plowshare, which utilizes macro- explosions, and inertial confinement fusion, which utilizes microexplosions. The development of practical fusion power plants is a principal goal of the inertial program. It is remarkable that Teller's original thermonuclear problem, how to make super high yield nuclear explosions, and the opposite problem, how to make ultra low yield nuclear explosions, may both be solved by Teller's radiation implosion scheme. This paper reports on the essential physics of these two thermonuclear domains, which are separated by nine orders of magnitude in yield, provided by Teller's similarity theorem and its exceptions. Higher density makes possible thermonuclear burn of smaller masses of fuel. The leverage is high: the scale of the explosion diminishes with the square of the increase in density. The extraordinary compressibility of matter, first noticed by Teller during the Los Alamos atomic bomb program, provides an almost incredible opportunity to harness fusion. The energy density of thermonuclear fuels isentropically compressed to super high-- -densities---even to ten thousand times solid density---is small compared to the energy density at thermonuclear ignition temperatures. In small masses of fuel imploded to these super high matter densities, the energy required to achieve ignition may be greatly reduced by exploiting thermonuclear propagation from a relatively small hot spot

  9. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  10. LIBRA - a light ion beam inertial confinement fusion reactor conceptual design

    International Nuclear Information System (INIS)

    Moses, G.A.; Kulcinski, G.L.; Bruggink, D.

    1989-01-01

    The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4 MJ shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbide tubes with Li 17 Pb 83 flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is a right circular cylinder, 8.7 meters in diameter. The target chamber is 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with liquid that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequency without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of LIBRA is estimated to be $2200/kWe. (author)

  11. Inertial confinement fusion target

    International Nuclear Information System (INIS)

    Bourdier, A.

    2001-12-01

    A simple, zero-dimensional model describing the temporal behaviour of an imploding-shell, magnetized fuel inertial confinement fusion target is formulated. The addition of a magnetic field to the fuel reduces thermal conduction losses. As a consequence, it might lead to high gains and reduce the driver requirements. This beneficial effect of the magnetic field on thermonuclear gains is confirmed qualitatively by the zero-dimensional model results. Still, the extent of the initial-condition space for which significant gains can occur is not, by far, as large as previously reported. One-dimensional CEA code simulations which confirm this results are also presented. Finally, we suggest to study the approach proposed by Hasegawa. In this scheme, the laser target is not imploded, and the life-time of the plasma can be very much increased. (author)

  12. Modeling of a confinement bypass accident with CONSEN, a fast-running code for safety analyses in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it [Sapienza University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Roma (Italy); Giannetti, Fabio [Sapienza University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Roma (Italy); Porfiri, Maria Teresa [ENEA FUS C.R. Frascati, Via Enrico Fermi, 45, 00044 Frascati, Roma (Italy)

    2013-12-15

    Highlights: • The CONSEN code for thermal-hydraulic transients in fusion plants is introduced. • A magnet induced confinement bypass accident in ITER has been simulated. • A comparison with previous MELCOR results for the accident is presented. -- Abstract: The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.

  13. Review of mirror fusion reactor designs

    International Nuclear Information System (INIS)

    Bender, D.J.

    1977-01-01

    Three magnetic confinement concepts, based on the mirror principle, are described. These mirror concepts are summarized as follows: (1) fusion-fission hybrid reactor, (2) tandem mirror reactor, and (3) reversed field mirror reactor

  14. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

  15. Prospects for spheromak fusion reactors

    International Nuclear Information System (INIS)

    Fowler, T.K.; Hua, D.D.

    1995-01-01

    The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on physics principles confirmed in CTX experiments in many respects. Most uncertain was the energy confinement time and the role of magnetic turbulence inherent in the concept. In this paper, a one-dimensional model of heat confinement, calibrated by CTX, predicts negligible heat loss by magnetic turbulence at reactor scale

  16. Commercial applications of inertial confinement fusion

    International Nuclear Information System (INIS)

    Booth, L.A.; Frank, T.G.

    1977-05-01

    This report describes the fundamentals of inertial-confinement fusion, some laser-fusion reactor (LFR) concepts, and attendant means of utilizing the thermonuclear energy for commercial electric power generation. In addition, other commercial energy-related applications, such as the production of fissionable fuels, of synthetic hydrocarbon-based fuels, and of process heat for a variety of uses, as well as the environmental and safety aspects of fusion energy, are discussed. Finally, the requirements for commercialization of laser fusion technologies are described

  17. Twenty years of ''Nuclear Fusion''. Inertial confinement

    International Nuclear Information System (INIS)

    Yamanaka, C.

    1980-01-01

    Inertial confinement (ICF) fusion research is directed towards demonstrating the feasibility of very rapidly heating and compressing small pellets of suitable fuel until conditions exist where thermonuclear fusion can occur and useful amounts of power can be produced. Major problems which have to be solved are the following: 1) pellet design based on driver-plasma coupling; 2) the technology of energy drivers; 3) feasibility of ICF reactor systems

  18. Vapor condensation on the surface of a liquid blanket jet in an inertial-confinement fusion reactor

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Inoue, Akira; Fujinuma, Hajime; Tsukui, Jun.

    1991-01-01

    As the fundamental study on lithium jet cooling of an inertial-confinement fusion reactor, the experiment was performed to investigate for the steady condensation of saturated steam on a vertical downward water jet. The experimental parameters were the nozzle diameter of 3 and 5 mm, the jet length of 60∼316 mm, the outlet velocity of 2∼12 m/s, the outlet temperature of 30∼70degC, and the pressure of 0.03∼0.44 MPa, which corresponds to the Reynolds number of 1.35 x 10 4 ∼2.71 x 10 5 and the Prandtl number of 1.0∼5.2. As the Reynolds number or the jet length is increased, the Stanton number decreases and then increases again. As the steam pressure is increased, it increases monotonously. These characteristics of condensation heat transfer have been classical into four regions based on the criteria for jet break-up and surface disturbance, or entrainment. The empirical correlations for the Stanton number have been obtained for these regions, and the validity was confirmed by comparing them with the previous correlations. (author)

  19. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    Steinhauer, L.C.; Quimby, D.C.

    1978-01-01

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  20. Material for fusion reactor

    International Nuclear Information System (INIS)

    Abhishek, Anuj; Ranjan, Prem

    2011-01-01

    To make nuclear fusion power a reality, the scientists are working restlessly to find the materials which can confine the power generated by the fusion of two atomic nuclei. A little success in this field has been achieved, though there are still miles to go. Fusion reaction is a special kind of reaction which must occur at very high density and temperature to develop extremely large amount of energy, which is very hard to control and confine within using the present techniques. As a whole it requires the physical condition that rarely exists on the earth to carry out in an efficient manner. As per the growing demand and present scenario of the world energy, scientists are working round the clock to make effective fusion reactions to real. In this paper the work presently going on is considered in this regard. The progress of the Joint European Torus 2010, ITER 2005, HiPER and minor works have been studied to make the paper more object oriented. A detailed study of the technological and material requirement has been discussed in the paper and a possible suggestion is provided to make a contribution in the field of building first ever nuclear fusion reactor

  1. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  2. Hydrogen isotopes confinement in the over-dusted layers of fusion reactor candidate materials

    International Nuclear Information System (INIS)

    Klepikov, A.Kh.; Tazhibaeva, I.L.; Shestakov, V.P.; Lisitsyn, V.N.; Tuleushev, Yu.Zh.

    2001-01-01

    In the work the experiments on gas-emission determination from samples of sputtered beryllium, graphite, tungsten, jointly sputtered graphite and tungsten obtained by the magnetron sputtering method at the 'Argamak' facility (National Nuclear Center of the Republic of Kazakhstan), as well as the samples processed on the 'OSPA' plasma accelerator (TRINITI, Russia). The gas-release curves were obtained for indicated samples under different heating velocities within temperature range from 300 up to 1200 K. Gas-release parameters and hydrogen isotopes confinement in these layers were determined. Simulation of hydrogen isotopes gas-emission from samples sputtered layers on the base of obtained experiments with application of simulating programs and TMAP code was carried out

  3. Fusion reactor in which the hydrogen plasma is heated, confined, and stabilized by oscillating magnetic fields in the center of a spherical or toroidal reaction vessel

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The two most difficult problems in the tests for controlled nuclear fusion are heating of the plasma to fusion temperature as well as the safe confinement in a magnetic field. According to the invention, low-resistance, low-inductive, iron-free AC compression coils are provided for dealing with these two problems the coils being arranged on the reactor vessel in such a way that both effective heating and a sufficient enclosure time are possible. It is of essential importance in this connection that the coils are fed with a relatively rectangular alternating current, which is variable with respect to frequency and power. (orig./GG) [de

  4. Alternate fusion -- continuous inertial confinement

    International Nuclear Information System (INIS)

    Barnes, D.C.; Turner, L.; Nebel, R.A.

    1993-01-01

    The authors argue that alternate approaches to large tokamak confinement are appropriate for fusion applications if: (1) They do not require magnetic confinement of a much higher quality than demonstrated in tokamaks; (2) Their physics basis may be succinctly stated and experimentally tested; (3) They offer near-term applications to important technical problems; and (4) Their cost to proof-of-principle is low enough to be consistent with current budget realities. An approach satisfying all of these criteria is presented. Fusion systems based on continuous inertial confinement are described. In these approaches, the inertia of a nonequilibrium plasma is used to produce local concentrations of plasma density in space and/or time. One implementation (inertial electrostatic confinement) which has been investigated both experimentally and theoretically uses a system of electrostatic grids to accelerate plasma ions toward a spherical focus. This system produced a steady 2 x 10 10 D-T neutrons/second with an overall fusion gain of 10 -5 in a sphere of about 9 cm radius. Recent theoretical developments show how to raise the fusion gain to order unity or greater by replacing the internal grids by a combination of applied magnetic and electrostatic fields. In these approaches, useful thermonuclear conditions may be produced in a system as small as a few mm radius. Confinement is that of a nonneutralized plasma. A pure electron plasma with a radial beam velocity distribution is absolutely confined by an applied Penning trap field. Spherical convergence of the confined electrons forms a deep virtual cathode near r = 0, in which thermonuclear ions are absolutely confined at useful densities. The authors have examined the equilibrium, stability, and classical relaxation of such systems, and obtained many positive physics results. Equilibria exist for both pure electron and partially charge-neutralized systems with arbitrarily high core-plasma densities

  5. Fusion reactor design studies

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. Advanced fusion reactor

    International Nuclear Information System (INIS)

    Tomita, Yukihiro

    2003-01-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p- 6 Li and p- 11 B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D- 3 He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D- 3 He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of 3 He per a year. On the other hand, 1 million tons of 3 He is estimated to be in the moon. The 3 He of about 10 23 kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  7. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  8. Fusion and technology: An introduction to the physics and technology of magnetic confinment fusion

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1984-01-01

    This book is an introduction covering all aspects of magnetic fusion and magnetic fusion technology. Physical property data relevant to fusion technology and a summary of fusion reactor design parameters are provided. Topics covered include: basic properties; equilibrium and transport confinement concepts; plasma heating; plasma wall interaction; magnetics; energy storage and transfer; interaction of radiation with matter; primary energy conversion and tritium breeding blanket; tritium and vacuum; and Fusion Reactor Design

  9. Inertial-confinement-fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1981-01-01

    Inertial confinement fusion (ICF) targets are made as simple flat discs, as hollow shells or as complicated multilayer structures. Many techniques have been devised for producing the targets. Glass and metal shells are made by using drop and bubble techniques. Solid hydrogen shells are also produced by adapting old methods to the solution of modern problems. Some of these techniques, problems and solutions are discussed. In addition, the applications of many of the techniques to fabrication of ICF targets is presented

  10. Fusion reactors - types - problems

    International Nuclear Information System (INIS)

    Schmitter, K.H.

    1979-07-01

    A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

  11. Conceptual design of inertial confinement fusion power plant

    International Nuclear Information System (INIS)

    Mima, Kunioki; Yamanaka, Tatsuhiko; Nakai, Sadao

    1994-01-01

    Presented is the status of the conceptual design studies of inertial confinement fusion reactors. The recent achievements of the laser fusion research enable us to refine the conceptual design of the power plant. In the paper, main features of several new conceptual designs of ICF reactor; KOYO, SIRIUS-P, HYLIFE-II and so on are summarized. In particular, the target design and the reactor chamber design are described. Finally, the overview of the laser fusion reactor and the irradiation system is also described. (author)

  12. Inertial confinement fusion at NRL

    International Nuclear Information System (INIS)

    Bodner, S.E.; Boris, J.P.; Cooperstein, G.

    1979-01-01

    The NRL Inertial Confinement Fusion Program's emphasis has moved toward pellet concepts which use longer (approximately 10ns) lower intensity driver pulses than previously assumed. For laser drivers, this change was motivated by recent experiments at NRL with enhanced stimulated Brillouin backscatter. For ion drivers, the motivation is the possibility that substantial energy at 10-ns pulse lengths may soon be available. To accept these 10-ns pulses, it may be necessary to consider pellets of larger radius and thinner shell. The computational studies of Rayleigh-Taylor instability at NRL indicate the possibility of a dynamic stabilization of these thinner shells. (author)

  13. Summary on inertial confinement fusion

    International Nuclear Information System (INIS)

    Meyer-Ter-Vehn, J.

    1995-01-01

    Highlights on inertial confinement during the fifteenth international conference on plasma physics and controlled nuclear fusion are briefly summarized. Specifically the following topics are discussed: the US National Ignition Facility presently planned by the US Department of Energy; demonstration of diagnostics for hot spot formation; declassification of Hohlraum target design; fusion targets, in particular, the Hohlraum target design for the National Ignition Facility (NIF), Hohlraum experiments, direct drive implosions, ablative Rayleigh-Taylor instabilities, laser imprinting (of perturbations by the laser on the laser target surface), hot spot formation and mixing, hot spot implosion experiments at Lawrence Livermore National Laboratory, Livermore, USA, time resolving hot spot dynamics at the Institute of Laser Engineering (ILE), Osaka, Japan, laser-plasma interaction

  14. Pulsed power systems for inertial confinement fusion

    International Nuclear Information System (INIS)

    VanDevender, J.P.

    1979-01-01

    Sandis's Particle Beam Fusion Program is investigating pulsed electron and light ion beam accelerators with the goal of demonstrating the practical application of such drivers as igniters in inertial confinement fusion (ICF) reactors. The power and energy requirements for net energy gain are 10 14 to 10 15 W and 1 to 10 MJ. Recent advances in pulsed power and power flow technologies permit suitable accelerators to be built. The first accelerator of this new generation is PBFA I. It operates at 2 MV, 15 MA, 30 TW for 35 ns and is scheduled for completion in June 1980. The principles of this new accelerator technology and their application to ICF will be presented

  15. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  16. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  17. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  18. Inertial confinement fusion and related topics

    International Nuclear Information System (INIS)

    Starodub, A. N.

    2007-01-01

    The current state of different approaches (laser fusion, light and heavy ions, electron beam) to the realization of inertial confinement fusion is considered. From comparative analysis a conclusion is made that from the viewpoint of physics, technology, safety, and economics the most realistic way to future energetics is an electric power plant based on a hybrid fission-fusion reactor which consists of an external source of neutrons (based on laser fusion) and a subcritical two-cascade nuclear blanket, which yields the energy under the action of 14 MeV neutrons. The main topics on inertial confinement fusion such as the energy driver, the interaction between plasmas and driver beam, the target design are discussed. New concept of creation of a laser driver for IFE based on generation and amplification of radiation with controllable coherence is reported. The performed studies demonstrate that the laser based on generation and amplification of radiation with controllable coherence (CCR laser) has a number of advantages as compared to conventional schemes of lasers. The carried out experiments have shown a possibility of suppression of small-scale self-focusing, formation of laser radiation pulses with required characteristics, simplification of an optical scheme of the laser, good matching of laser-target system and achievement of homogeneous irradiation and high output laser energy density without using traditional correcting systems (phase plates, adaptive optics, space filters etc.). The results of the latest experiments to reach ultimate energy characteristics of the developed laser system are also reported. Recent results from the experiments aimed at studying of the physical processes in targets under illumination by the laser with controllable coherence of radiation are presented and discussed, especially such important laser-matter interaction phenomena as absorption and scattering of the laser radiation, the laser radiation harmonic generation, X

  19. Magnetic confinement fusion energy research

    International Nuclear Information System (INIS)

    Grad, H.

    1977-03-01

    Controlled Thermonuclear Fusion offers probably the only relatively clean energy solution with completely inexhaustible fuel and unlimited power capacity. The scientific and technological problem consists in magnetically confining a hot, dense plasma (pressure several to hundreds of atmospheres, temperature 10 8 degrees or more) for an appreciable fraction of a second. The scientific and mathematical problem is to describe the behavior, such as confinement, stability, flow, compression, heating, energy transfer and diffusion of this medium in the presence of electromagnetic fields just as we now can for air or steam. Some of the extant theory consists of applications, routine or ingenious, of known mathematical structures in the theory of differential equations and in traditional analysis. Other applications of known mathematical structures offer surprises and new insights: the coordination between sub-supersonic and elliptic-hyperbolic is fractured; supersonic propagation goes upstream; etc. Other completely nonstandard mathematical structures with significant theory are being rapidly uncovered (and somewhat less rapidly understood) such as non-elliptic variational equations and new types of weak solutions. It is these new mathematical structures which one should expect to supply the foundation for the next generation's pure mathematics, if history is a guide. Despite the substantial effort over a period of some twenty years, there are still basic and important scintific and mathematical discoveries to be made, lying just beneath the surface

  20. Microencapsulation and fabrication of fuel pellets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Nolen, R.L. Jr.; Kool, L.B.

    1981-01-01

    Various microencapsulation techniques were evaluated for fabrication of thermonuclear fuel pellets for use in existing experimental facilities studying inertial confinement fusion and in future fusion-power reactors. Coacervation, spray drying, in situ polymerization, and physical microencapsulation methods were employed. Highly spherical, hollow polymeric shells were fabricated ranging in size from 20 to 7000 micron. In situ polymerization microencapsulation with poly(methyl methacrylate) provided large shells, but problems with local wall defects still must be solved. Extension to other polymeric systems met with limited success. Requirements for inertial confinement fusion targets are described, as are the methods that were used

  1. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2001-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  2. Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

    Science.gov (United States)

    Woolley, Robert D.

    2002-01-01

    A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

  3. Materials for fusion reactors

    International Nuclear Information System (INIS)

    Ehrlich, K.; Kaletta, D.

    1978-03-01

    The following report describes five papers which were given during the IMF seminar series summer 1977. The purpose of this series was to discuss especially the irradiation behaviour of materials intended for the first wall of future fusion reactors. The first paper deals with the basic understanding of plasma physics relating to the fusion reactor and presents the current state of art of fusion technology. The next two talks discuss the metals intended for the first wall and structural components of a fusion reactor. Since 14 MeV neutrons play an important part in the process of irradiation damage their role is discussed in detail. The question which machines are presently available to simulate irradiation damage under conditions similar to the ones found in a fusion reactor are investigated in the fourth talk which also presents the limitations of the different methods of simulation. In this context also discussed is the importance future intensive neutron sources and materials test reactors will have for this problem area. The closing paper has as a theme the review of the present status of research of metallic and non-metallic materials in view of the quite different requirements for different fusion systems; a closing topic is the world supply on rare materials required for fusion reactors. (orig) [de

  4. Possible fusion reactor

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1976-05-01

    A scheme to improve performance characteristics of a tokamak-type fusion reactor is proposed. Basically, the tokamak-type plasma could be moved around so that the plasma could be heated by compression, brought to the region where the blanket surrounds the plasma, and moved so as to keep wall loading below the acceptable limit. This idea should be able to help to economize a fusion reactor

  5. Fusion reactor materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

  6. Fusion reactor materials

    International Nuclear Information System (INIS)

    Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

    1980-01-01

    Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

  7. Fusion reactor safety

    International Nuclear Information System (INIS)

    1987-12-01

    Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

  8. Compact magnetic confinement fusion: Spherical torus and compact torus

    Directory of Open Access Journals (Sweden)

    Zhe Gao

    2016-05-01

    Full Text Available The spherical torus (ST and compact torus (CT are two kinds of alternative magnetic confinement fusion concepts with compact geometry. The ST is actually a sub-category of tokamak with a low aspect ratio; while the CT is a toroidal magnetic configuration with a simply-connected geometry including spheromak and field reversed pinch. The ST and CT have potential advantages for ultimate fusion reactor; while at present they can also provide unique fusion science and technology contributions for mainstream fusion research. However, some critical scientific and technology issues should be extensively investigated.

  9. Fusion reactor materials

    International Nuclear Information System (INIS)

    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

  10. Inertial confinement fusion: present status and future potential

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1984-01-01

    Power from inertial confinement fusion holds much promise for society. This paper points out many of the benefits relative to combustion of hydrocarbon fuels and fission power. Potential problems are also identified and put in perspective. The progress toward achieving inertial fusion power is described and results of recent work at the Lawrence Livermore National Laboratory are presented. Key phenomenological uncertainties are described and experimental goals for the Nova laser system are given. Several ICF reactor designs are discussed

  11. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  12. West European magnetic confinement fusion research

    International Nuclear Information System (INIS)

    McKenney, B.L.; McGrain, M.; Hogan, J.T.; Porkolab, M.; Thomassen, K.I.

    1990-01-01

    This report presents a technical assessment and review of the West European program in magnetic confinement fusion by a panel of US scientists and engineers active in fusion research. Findings are based on the scientific and technical literature, on laboratory reports and preprints, and on the personal experiences and collaborations of the panel members. Concerned primarily with developments during the past 10 years, from 1979 to 1989, the report assesses West European fusion research in seven technical areas: tokamak experiments; magnetic confinement technology and engineering; fusion nuclear technology; alternate concepts; theory; fusion computations; and program organization. The main conclusion emerging from the analysis is that West European fusion research has attained a position of leadership in the international fusion program. This distinction reflects in large measure the remarkable achievements of the Joint European Torus (JET). However, West European fusion prominence extends beyond tokamak experimental physics: the program has demonstrated a breadth of skill in fusion science and technology that is not excelled in the international effort. It is expected that the West European primacy in central areas of confinement physics will be maintained or even increased during the early 1990s. The program's maturity and commitment kindle expectations of dramatic West European advances toward the fusion energy goal. For example, achievement of fusion breakeven is expected first in JET, before 1995

  13. Magnetic fusion reactor economics

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    An almost primordial trend in the conversion and use of energy is an increased complexity and cost of conversion systems designed to utilize cheaper and more-abundant fuels; this trend is exemplified by the progression fossil fission → fusion. The present projections of the latter indicate that capital costs of the fusion ''burner'' far exceed any commensurate savings associated with the cheapest and most-abundant of fuels. These projections suggest competitive fusion power only if internal costs associate with the use of fossil or fission fuels emerge to make them either uneconomic, unacceptable, or both with respect to expensive fusion systems. This ''implementation-by-default'' plan for fusion is re-examined by identifying in general terms fusion power-plant embodiments that might compete favorably under conditions where internal costs (both economic and environmental) of fossil and/or fission are not as great as is needed to justify the contemporary vision for fusion power. Competitive fusion power in this context will require a significant broadening of an overly focused program to explore the physics and simbiotic technologies leading to more compact, simplified, and efficient plasma-confinement configurations that reside at the heart of an attractive fusion power plant

  14. Fusion reactor radioactive waste management

    International Nuclear Information System (INIS)

    Kaser, J.D.; Postma, A.K.; Bradley, D.J.

    1976-01-01

    Quantities and compositions of non-tritium radioactive waste are estimated for some current conceptual fusion reactor designs, and disposal of large amounts of radioactive waste appears necessary. Although the initial radioactivity of fusion reactor and fission reactor wastes are comparable, the radionuclides in fusion reactor wastes are less hazardous and have shorter half-lives. Areas requiring further research are discussed

  15. Controlled thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Walstrom, P.L.

    1976-01-01

    Controlled production of energy by fusion of light nuclei has been the goal of a large portion of the physics community since the 1950's. In order for a fusion reaction to take place, the fuel must be heated to a temperature of 100 million degrees Celsius. At this temperature, matter can exist only in the form of an almost fully ionized plasma. In order for the reaction to produce net power, the product of the density and energy confinement time must exceed a minimum value of 10 20 sec m -3 , the so-called Lawson criterion. Basically, two approaches are being taken to meet this criterion: inertial confinement and magnetic confinement. Inertial confinement is the basis of the laser fusion approach; a fuel pellet is imploded by intense laser beams from all sides and ignites. Magnetic confinement devices, which exist in a variety of geometries, rely upon electromagnetic forces on the charged particles of the plasma to keep the hot plasma from expanding. Of these devices, the most encouraging results have been achieved with a class of devices known as tokamaks. Recent successes with these devices have given plasma physicists confidence that scientific feasibility will be demonstrated in the next generation of tokamaks; however, an even larger effort will be required to make fusion power commercially feasible. As a result, emphasis in the controlled thermonuclear research program is beginning to shift from plasma physics to a new branch of nuclear engineering which can be called fusion engineering, in which instrumentation and control engineers will play a major role. Among the new problem areas they will deal with are plasma diagnostics and superconducting coil instrumentation

  16. Review of Inertial Confinement Fusion

    Science.gov (United States)

    Haines, M. G.

    The physics of inertial confinement fusion is reviewed. The trend to short-wavelength lasers is argued, and the distinction between direct and indirect (soft X-ray) drive is made. Key present issues include the non-linear growth of Rayleigh-Taylor (R-T) instabilities, the seeding of this instability by the initial laser imprint, the relevance of self-generated magnetic fields, and the importance of parametric instabilities (stimulated Brillouin and Raman scattering) in gas-filled hohlraums. Experiments are reviewed which explore the R-T instability in both planar and converging geometry. The employment of various optical smoothing techniques is contrasted with the overcoating of the capsule by gold coated plastic foams to reduce considerably the imprint problem. The role of spontaneously generated magnetic fields in non-symmetric plasmas is discussed. Recent hohlraum compression results are presented together with gas bag targets which replicate the long-scale-length low density plasmas expected in NIF gas filled hohlraums. The onset of first Brillouin and then Raman scattering is observed. The fast ignitor scheme is a proposal to use an intense short pulse laser to drill a hole through the coronal plasma and then, with laser excited fast electrons, create a propagating thermonuclear spark in a dense, relatively cold laser-compressed target. Some preliminary results of laser hole drilling and 2-D and 3-D PIC simulations of this and the > 10^8 Gauss self-generated magnetic fields are presented. The proposed National Ignition Facility (NIF) is described.

  17. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

    Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  18. Inertial fusion reactors and magnetic fields

    International Nuclear Information System (INIS)

    Cornwell, J.B.; Pendergrass, J.H.

    1985-01-01

    The application of magnetic fields of simple configurations and modest strengths to direct target debris ions out of cavities can alleviate recognized shortcomings of several classes of inertial confinement fusion (ICF) reactors. Complex fringes of the strong magnetic fields of heavy-ion fusion (HIF) focusing magnets may intrude into reactor cavities and significantly affect the trajectories of target debris ions. The results of an assessment of potential benefits from the use of magnetic fields in ICF reactors and of potential problems with focusing-magnet fields in HIF reactors conducted to set priorities for continuing studies are reported. Computational tools are described and some preliminary results are presented

  19. Review of fission-fusion pellet designs and inertial confinement system studies at EIR

    Energy Technology Data Exchange (ETDEWEB)

    Seifriz, W [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1978-01-01

    The article summarizes the work done so far at the Swiss Federal Institute for Reactor Research (EIR) in the field of the inertial confinement fusion technique. The following subjects are reviewed: a) fission fusion pellet designs using fissionable triggers, b) uranium tampered pellets, c) tampered pellets recycling unwanted actinide wastes from fission reactors in beam-driven micro-explosion reactors, and d) symbiotic fusion/fission reactor studies.

  20. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2000-01-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  1. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Schultz, K.R.; Smith, A.C. Jr.

    1978-01-01

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  2. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  3. Fusion reactor critical issues

    International Nuclear Information System (INIS)

    1987-11-01

    The document summarizes the results of a series of INTOR-related meetings organized by the IAEA in 1985-1986 with the following topics: Impurity control modelling, non-inductive current-drive, confinement in tokamaks with intense heating and DEMO requirements. These results are useful to the specialists involved in research on large fusion machines or in the design activity on the next generation tokamaks. Refs, figs and tabs

  4. Radiolytic production of chemical fuels in fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fish, J D

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.

  5. Radiolytic production of chemical fuels in fusion reactor systems

    International Nuclear Information System (INIS)

    Fish, J.D.

    1977-06-01

    Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered

  6. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Fusion reactor materials

    International Nuclear Information System (INIS)

    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

  8. Fusion reactor wastes

    International Nuclear Information System (INIS)

    Young, J.R.

    1976-01-01

    The fusion reactor currently is being developed as a clean source of electricity with an essentially infinite source of fuel. These reactors are visualized as using a fusion reaction to generate large quantities of high temperature energy which can be used as process heat or for the generation of electricity. The energy would be created primarily as the kinetic energy of neutrons or other reaction products. Neutron energy could be converted to high-temperature heat by moderation and capture of the neutrons. The energy of other reaction products could be converted to high-temperature heat by capture, or directly to electricity by direct conversion electrostatic equipment. An analysis to determine the wastes released as a result of operation of fusion power plants is presented

  9. Measurement of inertial confinement fusion reaction rate

    International Nuclear Information System (INIS)

    Peng Xiaoshi; Wang Feng; Tang Daorun; Liu Shenye; Huang Tianxuan; Liu Yonggang; Xu Tao; Chen Ming; Mei Yu

    2011-01-01

    Fusion reaction rate is an important parameter for measuring compression during the implosion in inertial confinement fusion experiment. We have developed a system for fusion reaction history measurement with high temporal resolution. The system is composed of plastic scintillator and nose cone, optical system and streak camera. We have applied this system on the SG-III prototype for fusion reaction rate measuring. For the first time, fusion reaction rate history have been measured for deuterium-tritium filled targets with neutrons yields about 10 10 . We have analyzed possible influence factor during fusion reaction rate measuring. It indicates that the instrument measures fusion reaction bang time at temporal resolutions as low as 30 ps.(authors)

  10. Fusion reactor problems

    International Nuclear Information System (INIS)

    Carruthers, R.

    It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

  11. Fusion reactor pumped laser

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1988-01-01

    A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

  12. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

    Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion ractors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  13. Continuous Electrode Inertial Electrostatic Confinement Fusion

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA recognizes within its roadmaps (specifically TA 3.1.6) that development of aneutronic fusion (such as p-11B) reactors with direct energy conversion (>80%)...

  14. Modular Stellarator Fusion Reactor concept

    International Nuclear Information System (INIS)

    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

  15. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    International Nuclear Information System (INIS)

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods

  16. Inertial fusion reactor designs

    International Nuclear Information System (INIS)

    Meier, W.

    1987-01-01

    In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

  17. Advanced spheromak fusion reactor

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1996-01-01

    The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

  18. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Moons, F.

    1998-01-01

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  19. Open-ended fusion devices and reactors

    International Nuclear Information System (INIS)

    Kawabe, T.; Nariai, H.

    1983-01-01

    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

  20. Fusion an introduction to the physics and technology of magnetic confinement fusion

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    This second edition of a popular textbook is thoroughly revised with around 25% new and updated content.It provides an introduction to both plasma physics and fusion technology at a level that can be understood by advanced undergraduates and graduate students in the physical sciences and related engineering disciplines.As such, the contents cover various plasma confinement concepts, the support technologies needed to confine the plasma, and the designs of ITER as well as future fusion reactors.With end of chapter problems for use in courses.

  1. Inertial Confinement Fusion at Los Alamos

    International Nuclear Information System (INIS)

    Cartwright, D.C.

    1989-09-01

    This report discusses the following topics on Inertial Confinement Fusion: ICF contributions to science and technology; target fabrication; laser-target interaction; KrF laser development; advanced KrF lasers; KrF laser technology; and plasma physics for light-ion program

  2. Inertial Confinement Fusion at Los Alamos

    International Nuclear Information System (INIS)

    Cartwright, D.C.

    1989-09-01

    This report discusses the following topics on inertial confinement fusion: distribution of electron-beam energy in KrF laser media; electron collision processes in KrF laser media; Krf laser kinetics; and properties of the KrF laser medium

  3. Nuclear diagnostics for inertial confinement fusion implosions

    International Nuclear Information System (INIS)

    Murphy, T.J.

    1997-01-01

    This abstract contains viewgraphs on nuclear diagnostic techniques for inertial confinement fusion implosions. The viewgraphs contain information on: reactions of interest in ICF; advantages and disadvantages of these methods; the properties nuclear techniques can measure; and some specifics on the detectors used

  4. Hohlraum manufacture for inertial confinement fusion

    International Nuclear Information System (INIS)

    Foreman, L.R.; Gobby, P.; Bartos, J.

    1994-01-01

    Hohlraums are an integral part of indirect drive targets for Inertial Confinement Fusion (ICF) research. Hohlraums are made by an electroforming process that combines elements of micromachining and coating technology. The authors describe how these target element are made and extension of the method that allow fabrication of other, more complex target components

  5. Progress in high gain inertial confinement fusion

    International Nuclear Information System (INIS)

    Sun Jingwen

    2001-01-01

    The author reviews the progress in laboratory high gain inertial confinement fusion (ICF), including ICF capsule physics, high-energy-density science, inertial fusion energy, the National Ignition Facility (NIF) and its design of ignition targets and the peta watt laser breakthrough. High power laser, particle beam, and pulsed power facilities around the world have established the new laboratory field of high-energy- density plasma physics and have furthered development of inertial fusion. New capabilities such as those provided by high-brightness peta watt lasers have enabled the study of matter feasible in conditions previously unachievable on earth. Science and technology developed in inertial fusion research have found near-term commercial use and have enabled steady progress toward the goal of fusion ignition and high gain in the laboratory, and have opened up new fields of study for the 21 st century

  6. The fusion reactor - a chance to solve the energy problem

    International Nuclear Information System (INIS)

    Wienecke, R.

    1975-01-01

    The work deals with the physical fundamentals of nuclear fusion and the properties of the necessary plasma and gives a survey on the arrangements used today for magnetic confinement such as tokamak, stellarator, high-beta experiments and laser fusion. Finally, the technology of the fusion reactor and its potential advantages are explained. (RW/LH) [de

  7. Inertial-confinement-fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1982-01-01

    Much of the research in laser fusion has been done using simple ball on-stalk targets filled with a deuterium-tritium mixture. The targets operated in the exploding pusher mode in which the laser energy was delivered in a very short time (approx. 100 ps or less) and was absorbed by the glass wall of the target. The high energy density in the glass literally exploded the shell with the inward moving glass compressing the DT fuel to high temperatures and moderate densities. Temperatures achieved were high enough to produce DT reactions and accompanying thermonuclear neutrons and alpha particles. The primary criteria imposed on the target builders were: (1) wall thickness, (2) sphere diameter, and (3) fuel in the sphere

  8. Repetitive pulsed power technology for inertial-confinement fusion

    International Nuclear Information System (INIS)

    Prestwich, K.R.; Buttram, M.T.

    1983-01-01

    The pulsed power requirements for inertial-confinement fusion reactors are defined for ion-beam and laser drivers. Several megajoule beams with 100's of terrawatt peak powers must be delivered to the reactor chamber 1 to 10 times per second. Ion-beam drivers are relatively efficient requiring less energy storage in the pulsed-power system but more time compression in the power flow chain than gas lasers. These high peak powers imply very large numbers of components for conventional pulse-power systems. A new design that significantly reduces the number of components is presented

  9. Directions for improved fusion reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

    Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts

  10. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  11. Coatings for fusion reactor environments

    International Nuclear Information System (INIS)

    Mattox, D.M.

    1979-01-01

    The internal surfaces of a tokamak fusion reactor control the impurity injection and gas recycling into the fusion plasma. Coating of internal surfaces may provide a desirable and possibly necessary design flexibility for achieving the temperatures, ion densities and containment times necessary for net energy production from fusion reactions to take place. In this paper the reactor environments seen by various componentare reviewed along with possible materials responses. Characteristics of coating-substrate systems, important to fusion applications, are delineated and the present status of coating development for fusion applications is reviewed. Coating development for fusion applications is just beginning and poses a unique and important challenge for materials development

  12. Possible in-lattice confinement fusion (LCF). Dynamic application of atomic and nuclear data

    International Nuclear Information System (INIS)

    Kawarasaki, Yuuki

    1995-01-01

    New scheme of a nuclear fusion reactor system is proposed, the basic concept of which comes from ingenious combination of hitherto developed techniques and verified facts; 1) so-called cold fusion (CF), 2) plasma of both magnetic confinement fusion (MCF) and inertial confinement fusion (ICF), and 3) accelerator-based D-T(D) neutron source. Details of the LCF reactor physics require dynamics of atomic data as well as nuclear data; interaction of ions with matters in solid and the problems of radiation damage. (author)

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

    Science.gov (United States)

    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…

  14. Status of fusion reactor concept development in Japan

    International Nuclear Information System (INIS)

    Tsuji-Iio, Shunji

    1996-01-01

    Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

  15. Target support for inertial confinement fusion

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1995-08-01

    General Atomics (GA) plays an important industrial support role for the US Inertial Confinement Fusion (ICF) program in the area of target technology. This includes three major activities: target fabrication support, target handling systems development, and target chamber design. The work includes target fabrication for existing ICF experiments, target and target system development for future experiments, and target research and target chamber design for experiments on future machines, such as the National Ignition Facility (NIF)

  16. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    1987-01-01

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  17. Past, present and future of the fusion reactors

    International Nuclear Information System (INIS)

    Rosenbaum P, M.

    1992-01-01

    Among the alternate technologies that have acquired a special interest in the present decade, we find the nuclear fusion. Within this, the fusion reactors by magnetic confinement of the Tokamak type have shown an increasing technological progress during this period. For this reason, a new strategy, coordinated at international level, has been implemented for the specific development of the nuclear fusion reactors, aimed to face those scientific and technological aspects which still remain, and which will determine their future economic feasibility. (Author)

  18. Radioisotope production in fusion reactors

    International Nuclear Information System (INIS)

    Engholm, B.A.; Cheng, E.T.; Schultz, K.R.

    1986-01-01

    Radioisotope production in fusion reactors is being investigated as part of the Fusion Applications and Market Evaluation (FAME) study. /sup 60/Co is the most promising such product identified to date, since the /sup 60/Co demand for medical and food sterilization is strong and the potential output from a fusion reactor is high. Some of the other radioisotopes considered are /sup 99/Tc, /sup 131/l, several Eu isotopes, and /sup 210/Po. Among the stable isotopes of interest are /sup 197/Au, /sup 103/Rh and Os. In all cases, heat or electricity can be co-produced from the fusion reactor, with overall attractive economics

  19. Laser drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Holzrichter, J.F.

    1983-01-01

    Inertial Confinement Fusion (ICF) is the technology that we are developing to access the vast stored energy potential of deuterium fuel located in the world's water supply. This form of fusion is accomplished by compressing and heating small volumes of D-T fuel to very high temperatures (greater than 100M 0 C) and to very high densities (greater than 1000 times the normal liquid density). Under these fuel conditions, a thermonuclear reaction can occur, leading to a net energy release compared to the energy used to heat the fuel initially. To accomplish the condition where fusion reactions begin, effective drivers are required. These are lasers or particle beam accelerators which can provide greater than 10 14 W/cm 2 over millimeter scale targets with an appropriately programmed intensity vs time. At present, we are using research lasers to obtain an understanding of the physics and engineering of fuel compression

  20. Hydrogen production in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated. (author).

  1. Hydrogen production in fusion reactors

    Science.gov (United States)

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    As one of the methods of innovative energy production in fusion reactors (that do not include a conventional turbine-type generator), the efficient use of fusion-reactor radiation and semiconductors to supply clean fuel in the form of hydrogen gas is studied. Taking the reactor candidates such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a plant system concept are investigated.

  2. Hydrogen production in fusion reactors

    International Nuclear Information System (INIS)

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated. (author)

  3. Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Nohara, Kiyohiko

    2009-01-01

    The structural material is one of key issues for the development of reliable superconducting magnets and peripheral equipments of fusion reactors. Standard stainless steels like SUS 304 and 316 steels available at present do not meet requirements. We are developing a new austenitic steel that has proposed target properties named 'JAERI BOX'. Additions of N and V at different amounts were tested to improve strength and fracture toughness of a base alloy SUS316LN at 4.2 K. Mechanical properties of the developed steel were examined. It is found that the charpy absorbed energy and the fracture toughness of the developed steel at 4.2 K are within JAERI BOX. (T.I.)

  4. Reinforced confinement in a nuclear reactor

    International Nuclear Information System (INIS)

    Norman, H.

    1988-01-01

    The present invention concerns a nuclear reactor containing a reactor core, a swimming pool space that is filled and pressurized with a neutron-absorbing solution, a reactor tank, at least one heat exchanger, at least one inlet line, at least one return line and at least one circulation pump, where the said reactor tank is confined in the said swimming pool space and designed to be cooled with the aid of relatively pure water, which is fed by means of the said at least one circulating pump to the said reactor tank from the said heat exchanger via the said at least one inlet line and is returned to the heat exchanger via the said at least one return line. The problem that is to be solved by the invention is to design a reactor of the above type in such a way that a complete confinement of the primary circuit of the reactor is achieved at relatively low extra cost. This problem is solved by providing the reactor with a special confinement space that confines the heat exchanger, but not the reactor tank, with the confinement space and the swimming pool space being fashioned in the same concrete body

  5. Cermet coatings for magnetic fusion reactors

    International Nuclear Information System (INIS)

    Smith, M.F.; Whitley, J.B.; McDonald, J.M.

    1984-01-01

    Cermet coatings consisting of SiC particles in an aluminum matrix were produced by a low pressure chamber plasma spray process. Properties of these coatings are being investigated to evaluate their suitability for use in the next generation of magnetic confinement fusion reactors. Although this preliminary study has focused primarily upon SiC-Al cermets, the deposition process can be adapted to other ceramic-metal combinations. Potential applications for cermet coatings in magnetic fusion devices are presented along with experimental results from thermal tests of candidate coatings. (Auth.)

  6. Advanced fuels for nuclear fusion reactors

    International Nuclear Information System (INIS)

    McNally, J.R. Jr.

    1974-01-01

    Should magnetic confinement of hot plasma prove satisfactory at high β (16 πnkT//sub B 2 / greater than 0.1), thermonuclear fusion fuels other than D.T may be contemplated for future fusion reactors. The prospect of the advanced fusion fuels D.D and 6 Li.D for fusion reactors is quite promising provided the system is large, well reflected and possesses a high β. The first generation reactions produce the very active, energy-rich fuels t and 3 He which exhibit a high burnup probability in very hot plasmas. Steady state burning of D.D can ensue in a 60 kG field, 5 m reactor for β approximately 0.2 and reflectivity R/sub mu/ = 0.9 provided the confinement time is about 38 sec. The feasibility of steady state burning of 6 Li.D has not yet been demonstrated but many important features of such systems still need to be incorporated in the reactivity code. In particular, there is a need for new and improved nuclear cross section data for over 80 reaction possibilities

  7. High-density-plasma diagnostics in magnetic-confinement fusion

    International Nuclear Information System (INIS)

    Jahoda, F.C.

    1982-01-01

    The lectures will begin by defining high density in the context of magnetic confinement fusion research and listing some alternative reactor concepts, ranging from n/sub e/ approx. 2 x 10 14 cm -3 to several orders of magnitude greater, that offer potential advantages over the main-line, n/sub e/ approx. 1 x 10 14 cm -3 , Tokamak reactor designs. The high density scalings of several major diagnostic techniques, some favorable and some disadvantageous, will be discussed. Special emphasis will be given to interferometric methods, both electronic and photographic, for which integral n/sub e/dl measurements and associated techniques are accessible with low wavelength lasers. Reactor relevant experience from higher density, smaller dimension devices exists. High density implies high β, which implies economies of scale. The specialized features of high β diagnostics will be discussed

  8. Magnetic confinement fusion plasma theory, Task 1

    International Nuclear Information System (INIS)

    Callen, J.D.

    1991-07-01

    The research performed under this grant during the current year has concentrated on a few key tokamak plasma confinement and heating theory issues: extensive development of a new Chapman-Enskog-like fluid/kinetic hybrid approach to deriving rigorously valid fluid moment equations; applications (neoclassical viscous force, instabilities in the banana-plateau collisionality regime, nonlinear gyroviscous force, unified plasma microinstability equations and their implications, semi-collisional presheath modeling, etc.) of this new formalism; interactions of fluctuating bootstrap-current-driven magnetic islands; determination of net transport processes and equations for a tokamak; and some other topics (extracting more information from heat-pulse-propagation data, modeling of BES fluctuation data, exploring sawtooth effects on energy confinement in DIII-D, divertor X-point modeling). Recent progress and publications in these areas, and in the management of the local NERSC node and fusion theory DECstation 5000 at UW-Madison are summarized briefly in this report

  9. Electrostatic-Dipole (ED) Fusion Confinement Studies

    Science.gov (United States)

    Miley, George H.; Shrestha, Prajakti J.; Yang, Yang; Thomas, Robert

    2004-11-01

    The Electrostatic-Dipole (ED) concept significantly differs from a "pure" dipole confinement device [1] in that the charged particles are preferentially confined to the high-pressure region interior of the dipole coil by the assistance of a surrounding spherical electrostatic grid. In present ED experiments, a current carrying coil is embedded inside the grid of an IEC such as to produce a magnetic dipole field. Charged particles are injected axisymmetrically from an ion gun (or duo-plasmatron) into the center of the ED confinement grid/dipole ring where they oscillate along the magnetic field lines and pass the peak field region at the center of the dipole region. As particles begin accelerating away from the center region towards the outer electrostatic grid region, they encounter a strong electrostatic potential (order of 10's of kilovolts) retarding force. The particles then decelerate, reverse direction and re-enter the dipole field region where again magnetic confinement dominates. This process continues, emulating a complex harmonic oscillator motion. The resulting pressure profile averaged over the field curvature offers good plasma stability in the ED configuration. The basic concept and results from preliminary experiments will be described. [1] M.E. Mauel, et al. "Dipole Equilibrium and Stability," 18th IAEA Conference of Plasma Phys. and Control. Nuclear Fusion, Varenna, Italy 2000, IAEA-F1-CN-70/TH

  10. Optical design considerations for laser fusion reactors

    International Nuclear Information System (INIS)

    Monsler, M.J.; Maniscalco, J.A.

    1977-09-01

    The plan for the development of commercial inertial confinement fusion (ICF) power plants is discussed, emphasizing the utilization of the unique features of laser fusion to arrive at conceptual designs for reactors and optical systems which minimize the need for advanced materials and techniques requiring expensive test facilities. A conceptual design for a liquid lithium fall reactor is described which successfully deals with the hostile x-ray and neutron environment and promises to last the 30 year plant lifetime. Schemes for protecting the final focusing optics are described which are both compatible with this reactor system and show promise of surviving a full year in order to minimize costly downtime. Damage mechanisms and protection techniques are discussed, and a recommendation is made for a high f-number metal mirror final focusing system

  11. Hydrodynamic instabilities in inertial confinement fusion

    International Nuclear Information System (INIS)

    Hoffman, N.M.

    1995-01-01

    The focus of these (two) lectures is on buoyancy-driven instabilities of the Rayleigh-Taylor type, which are commonly regarded as the most important kind of hydrodynamic instability in inertial-confinement-fusion implosions. The paper is intended to be pedagogical rather than research-oriented, and so is by no means a comprehensive review of work in this field. Rather, it is hoped that the student will find here a foundation on which to build an understanding of current research, and the experienced researcher will find a compilation of useful results. (author)

  12. Micromachining of inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Gobby, P.L.; Salzer, L.J.; Day, R.D.

    1996-01-01

    Many experiments conducted on today's largest inertial confinement fusion drive lasers require target components with sub-millimeter dimensions, precisions of a micron or less and surface finishes measured in nanometers. For metal and plastic, techniques using direct machining with diamond tools have been developed that yield the desired parts. New techniques that will be discussed include the quick-flip locator, a magnetically held kinematic mount that has allowed the direct machining of millimeter-sized beryllium hemishells whose inside and outside surface are concentric to within 0.25 micron, and an electronic version of a tracer lathe which has produced precise azimuthal variations of less than a micron

  13. Generalized Lawson Criteria for Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Tipton, Robert E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-08-27

    The Lawson Criterion was proposed by John D. Lawson in 1955 as a general measure of the conditions necessary for a magnetic fusion device to reach thermonuclear ignition. Over the years, similar ignition criteria have been proposed which would be suitable for Inertial Confinement Fusion (ICF) designs. This paper will compare and contrast several ICF ignition criteria based on Lawson’s original ideas. Both analytical and numerical results will be presented which will demonstrate that although the various criteria differ in some details, they are closely related and perform similarly as ignition criteria. A simple approximation will also be presented which allows the inference of each ignition parameter directly from the measured data taken on most shots fired at the National Ignition Facility (NIF) with a minimum reliance on computer simulations. Evidence will be presented which indicates that the experimentally inferred ignition parameters on the best NIF shots are very close to the ignition threshold.

  14. Hydrodynamic instabilities in inertial confinement fusion

    International Nuclear Information System (INIS)

    Freeman, J.R.

    1977-01-01

    Inertial confinement fusion targets generally consist of hollow high-density spheres filled with low density thermonuclear fuel. Targets driven ablatively by electrons, ions, or lasers are potentially unstable during the initial acceleration phase. Later in time, the relatively low density fuel decelerates the dense inner portion of the sphere (termed the pusher), permitting unstable growth at the fuel-pusher interface. The instabilities are of the Rayleigh-Taylor variety, modified by thermal and viscous diffusion and convection. These problems have been analyzed by many in recent years using both linearized perturbation methods and direct numerical simulation. Examples of two-dimensional simulations of the fuel-pusher instability in electron beam fusion targets will be presented, along with a review of possible stabilization mechanisms

  15. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

  16. Definition of Ignition in Inertial Confinement Fusion

    Science.gov (United States)

    Christopherson, A. R.; Betti, R.

    2017-10-01

    Defining ignition in inertial confinement fusion (ICF) is an unresolved problem. In ICF, a distinction must be made between the ignition of the hot spot and the propagation of the burn wave in the surrounding dense fuel. Burn propagation requires that the hot spot is robustly ignited and the dense shell exhibits enough areal density. Since most of the energy gain comes from burning the dense shell, in a scale of increasing yields, hot-spot ignition comes before high gains. Identifying this transition from hot-spot ignition to burn-wave propagation is key to defining ignition in general terms applicable to all fusion approaches that use solid DT fuel. Ad hoc definitions such as gain = 1 or doubling the temperature are not generally valid. In this work, we show that it is possible to identify the onset of ignition through a unique value of the yield amplification defined as the ratio of the fusion yield including alpha-particle deposition to the fusion yield without alphas. Since the yield amplification is a function of the fractional alpha energy fα =EαEα 2Ehs 2Ehs (a measurable quantity), it appears possible not only to define ignition but also to measure the onset of ignition by the experimental inference of the fractional alpha energy and yield amplification. This material is based upon work supported by the Department of Energy Office of Fusion Energy Services under Award Number DE-FC02-04ER54789 and National Nuclear Security Administration under Award Number DE-NA0001944.

  17. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  18. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

  19. ELMO Bumpy Torus fusion-reactor design study

    International Nuclear Information System (INIS)

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

    1981-01-01

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is described that emphasizes those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are generic to magnetic fusion being adopted from past, more extensive tokamak reactor designs

  20. Intense ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    Mehlhorn, T.A.

    1997-01-01

    Intense beams of light of heavy ions are being studied as inertial confinement fusion (ICF) drivers for high yield and energy. Heavy and light ions have common interests in beam transport, targets, and alternative accelerators. Self-pinched transport is being jointly studied. This article reviews the development of intense ion beams for ICF. Light-ion drivers are highlighted because they are compact, modular, efficient and low cost. Issues facing light ions are: (1) decreasing beam divergence; (2) increasing beam brightness; and (3) demonstrating self-pinched transport. Applied-B ion diodes are favored because of efficiency, beam brightness, perceived scalability, achievable focal intensity, and multistage capability. A light-ion concept addressing these issues uses: (1) an injector divergence of ≤ 24 mrad at 9 MeV; (2) two-stage acceleration to reduce divergence to ≤ 12 mrad at 35 MeV; and (3) self-pinched transport accepting divergences up to 12 mrad. Substantial progress in ion-driven target physics and repetitive ion diode technology is also presented. Z-pinch drivers are being pursued as the shortest pulsed power path to target physics experiments and high-yield fusion. However, light ions remain the pulsed power ICF driver of choice for high-yield fusion energy applications that require driver standoff and repetitive operation. 100 refs

  1. Fusion Plasma Theory: Task 1, Magnetic confinement Fusion Plasma Theory

    International Nuclear Information System (INIS)

    Callen, J.D.

    1993-01-01

    The research performed under this grant during the current year has concentrated on few tokamak plasma confinement issues: applications of our new Chapman-Enskog-like approach for developing hybrid fluid/kinetic descriptions of tokamak plasmas; multi-faceted studies as part of our development of a new interacting island paradigm for the tokamak equilibrium'' and transport; investigations of the resolution power of BES and ECE diagnostics for measuring core plasma fluctuations; and studies of net transport in the presence of fluctuating surfaces. Recent progress and publications in these areas, and in the management of the NERSC node and the fusion theory workstations are summarized briefly in this report

  2. Fusion reactors and the environment

    International Nuclear Information System (INIS)

    Hancox, R.

    1990-04-01

    Fusion power, based on the nuclear fusion of light elements to yield a net gain of energy, has the potential to extend the world's resources in a way which is environmentally attractive. Nevertheless, the easiest route to fusion - the reaction between deuterium and tritium - involves hazards from the use of tritium and the neutron activation of the structural materials. These hazards have been considered on the basis of simple conceptual reactor designs, both in relation to normal operation and decommissioning and to potential accident situations. Results from several studies are reviewed and suggest that fusion reactors appear to have an inherently lower environmental impact than fission reactors. However, the realization of this potential has yet to be demonstrated. (author)

  3. Technical issues in fusion reactors

    International Nuclear Information System (INIS)

    Rohatgi, V.K.; Vijayan, T.

    1989-01-01

    In this paper the issues in fusion reactor technology are examined. Rapid progress in fusion technology research in recent years can be attributed to the advances in various technologies. The commercial generation of fusion power greatly depends on the evolution and improvements in these technologies. With better understanding of plasma physics, fusion reactor designs are becoming more and more realistic and comprehensive. It is now possible to compare various concepts within the framework of established technologies. The technological issues needing better understanding and solutions to problem areas are identified. Various instabilities and energy losses are major problem areas. Extensive developments in reactor-relevant advanced materials, compact and powerful superconducting magnets, high-power systems, and plasma heating drivers need to be undertaken and emphasized

  4. Polymer materials for fusion reactors

    International Nuclear Information System (INIS)

    Yamaoka, H.

    1993-01-01

    The radiation-resistant polymer materials have recently drawn much attention from the viewpoint of components for fusion reactors. These are mainly applied to electrical insulators, thermal insulators and structural supports of superconducting magnets in fusion reactors. The polymer materials used for these purposes are required to withstand the synergetic effects of high mechanical loads, cryogenic temperatures and intense nuclear radiation. The objective of this review is to summarize the anticipated performance of candidate materials including polymer composites for fusion magnets. The cryogenic properties and the radiation effects of polymer materials are separately reviewed, because there is only limited investigation on the above-mentioned synergetic effects. Additional information on advanced polymer materials for fusion reactors is also introduced with emphasis on recent developments. (orig.)

  5. Environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    Coffman, F.E.; Williams, J.M.

    1975-01-01

    With the continued depletion of fossil and uranium resources in the coming decades, the U. S. will be forced to look more toward renewable energy resources (e.g., wind, tidal, geothermal, and solar power) and toward such longer-term and nondepletable energy resources as fissile fast breeder reactors and fusion power. Several reference reactor designs have been completed for full-scale fusion power reactors that indicate that the environmental impacts from construction, operation, and eventual decommissioning of fusion reactors will be quite small. The principal environmental impact from fusion reactor operation will be from thermal discharges. Some of the safety and environmental characteristics that make fusion reactors appear attractive include an effectively infinite fuel supply at low cost, inherent incapability for a ''nuclear explosion'' or a ''nuclear runaway,'' the absence of fission products, the flexibility of selecting low neutron-cross-section structural materials so that emergency core cooling for a loss-of-coolant or other accident will not be necesary, and the absence of special nuclear materials such as 235 U or 239 Pu, so that diversion of nuclear weapons materials will not be possible and nuclear blackmail will not be a serious concern

  6. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Present trends in magnetic fusion research and development indicate the promise of commercialization of one of a limited number of inexhaustible energy options early in the next century. Operation of the large-scale fusion experiments, such as the Joint European Torus (JET) and Takamak Fusion Test Reactor (TFTR) now under construction, are expected to achieve the scientific break even point. Early design concepts of power producing reactors have provided problem definition, whereas the latest concepts, such as STARFIRE, provide a desirable set of answers for commercialization. Safety and environmental concerns have been considered early in the development of magnetic fusion reactor concepts and recognition of proplem areas, coupled with a program to solve these problems, is expected to provide the basis for safe and environmentally acceptable commercial reactors. First generation reactors addressed in this paper are expected to burn deuterium and tritium fuel because of the relatively high reaction rates at lower temperatures compared to advanced fuels such as deuterium-deuterium. This paper presents an overwiew of the safety and environmental problems presently perceived, together with some of the programs and techniques planned and/or underway to solve these problems. A preliminary risk assessment of fusion technology relative to other energy technologies is made. Improvements based on material selection are discussed. Tritium and neutron activation products representing potential radiological hazards in fusion reactor are discussed, and energy sources that can lead to the release of radioactivity from fusion reactors under accident conditions are examined. The handling and disposal of radioactive waste are discussed; the status of biological effects of magnetic fields are referenced; and release mechanisms for tritium and activation products, including analytical methods, are presented. (orig./GG)

  7. Density dependence of reactor performance with thermal confinement scalings

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1992-03-01

    Energy confinement scalings for the thermal component of the plasma published thus far have a different dependence on plasma density and input power than do scalings for the total plasma energy. With such thermal scalings, reactor performance (measured by Q, the ratio of the fusion power to the sum of the ohmic and auxiliary input powers) worsens with increasing density. This dependence is the opposite of that found using scalings based on the total plasma energy, indicating that reactor operation concepts may need to be altered if this density dependence is confirmed in future research

  8. Fast ignition schemes for inertial confinement fusion

    International Nuclear Information System (INIS)

    Deutsch, C.

    2003-01-01

    The controlled production of a local hot spot in super-compressed deuterium + tritium fuel is examined in details. Relativistic electron beams (REB) in the MeV and proton beams in the few tens MeV energy range produced by PW-lasers are respectively considered. A strong emphasis is given to the propagation issues due to large density gradients in the outer core of compressed fuel. A specific attention is also paid to the final and complete particle stopping resulting in hot spot generation as well as to the interplay of collective vs. particle stopping at the entrance channel on the low density side in plasma target. Moreover, REB production and fast acceleration mechanisms are also given their due attention. Proton fast ignition looks promising as well as the wedged (cone angle) approach circumventing most of transport uncertainties between critical layer and hot spot. Global engineering perspectives for fast ignition scenario (FIS) driven inertial confinement fusion are also detailed. (author)

  9. Survey of fusion reactor technology

    International Nuclear Information System (INIS)

    Chung, M.K.; Kang, H.D.; Oh, Y.K.; Lee, K.W.; In, S.Y.; Kim, Y.C.

    1983-01-01

    The present object of the fusion research is to accomplish the scientific break even by the year of 1986. In view of current progress in the field of Fusion reactor development, we decided to carry out the conceptual design of Tokamak-type fusion reactor during the year of 82-86 in order to acquire the principles of the fusion devices, find the engineering problems and establish the basic capabilities to develop the key techniques with originality. In this year the methods for calculating the locations of the poloidal coils and distribution of the magnetic field, which is one of the most essential and complicated task in the fusion reactor design works, were established. Study on the optimization of the design method of toroidal field coil was also done. Through this work, we established the logic for the design of the toroidal field coil in tokamak and utilize this technique to the design of small compact tokamak. Apart from the development work as to the design technology of tokamak, accelerating column and high voltage power supply (200 KVDC, 100 mA) for intense D-T neutron generator were constructed and now beam transport systems are under construction. This device will be used to develop the materials and the components for the tokamak fusion reactor. (Author)

  10. Inertial confinement fusion with light ion beams

    International Nuclear Information System (INIS)

    VanDevender, J.P.; Cook, D.L.

    1986-01-01

    The Particle Beam Fusion Accelerator II (PBFA II) is presently under construction and is the only existing facility with the potential of igniting thermonuclear fuel in the laboratory. The accelerator will generate up to 5 megamperes of lithium ions at 30 million electron volts and will focus them onto an inertial confinement fusion (ICF) target after beam production and focusing have been optimized. Since its inception, the light ion approach to ICF has been considered the one that combines low cost, high risk, and high payoff. The beams are of such high density that their self-generated electric and magnetic fields were thought to prohibit high focal intensities. Recent advances in beam production and focusing demonstrate that these self-forces can be controlled to the degree required for ignition, break-even, and high gain experiments. ICF has been pursued primarily for its potential military applications. However, the high efficiency and cost-effectiveness of the light ion approach enhance its potential for commercial energy application as well

  11. Thermonuclear plasma physic: inertial confinement fusion

    International Nuclear Information System (INIS)

    Bayer, Ch.; Juraszek, D.

    2001-01-01

    Inertial Confinement Fusion (ICF) is an approach to thermonuclear fusion in which the fuel contained in a spherical capsule is strongly compressed and heated to achieve ignition and burn. The released thermonuclear energy can be much higher than the driver energy, making energetic applications attractive. Many complex physical phenomena are involved by the compression process, but it is possible to use simple analytical models to analyze the main critical points. We first determine the conditions to obtain fuel ignition. High thermonuclear gains are achieved if only a small fraction of the fuel called hot spot is used to trigger burn in the main fuel compressed on a low isentrope. A simple hot spot model will be described. The high pressure needed to drive the capsule compression are obtained by the ablation process. A simple Rocket model describe the main features of the implosion phase. Several parameters have to be controlled during the compression: irradiation symmetry, hydrodynamical stability and when the driver is a laser, the problems arising from interaction of the EM wave with the plasma. Two different schemes are examined: Indirect Drive which uses X-ray generated in a cavity to drive the implosion and the Fast Ignitor concept using a ultra intense laser beam to create the hot spot. At the end we present the Laser Megajoule (LMJ) project. LMJ is scaled to a thermonuclear gain of the order of ten. (authors)

  12. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    Schulte, S.C.; Willke, T.L.; Young, J.R.

    1978-05-01

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  13. Reactor concepts for laser fusion

    International Nuclear Information System (INIS)

    Meier, W.R.; Maniscalco, J.A.

    1977-07-01

    Scoping studies were initiated to identify attractive reactor concepts for producing electric power with laser fusion. Several exploratory reactor concepts were developed and are being subjected to our criteria for comparing long-range sources of electrical energy: abundance, social costs, technical feasibility, and economic competitiveness. The exploratory concepts include: a liquid-lithium-cooled stainless steel manifold, a gas-cooled graphite manifold, and fluidized wall concepts, such as a liquid lithium ''waterfall'', and a ceramic-lithium pellet ''waterfall''. Two of the major reactor vessel problems affecting the technical feasibility of a laser fusion power plant are: the effects of high-energy neutrons and cyclical stresses on the blanket structure and the effects of x-rays and debris from the fusion microexplosion on the first-wall. The liquid lithium ''waterfall'' concept is presented here in more detail as an approach which effectively deals with these damaging effects

  14. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

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

    1981-01-01

    A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-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. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

  15. Reactor potential for magnetized target fusion

    International Nuclear Information System (INIS)

    Dahlin, J.E.

    2001-06-01

    Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well

  16. Reactor potential for magnetized target fusion

    Energy Technology Data Exchange (ETDEWEB)

    Dahlin, J.E

    2001-06-01

    Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well.

  17. The tritium and the controlled fusion reactors

    International Nuclear Information System (INIS)

    Leger, D.; Rouyer, J.L.

    1986-04-01

    It is shown how tritium is used how it is circulating in a fusion reactor. The great functions of tritium circuits are detailed: reprocessing of burnt gases, reprocessing of gases coming from neutral injectors, reprocessing from gaseous wastes, detritiation of cooling fluids. Current technologic developments are quoted. Then tritium confinement and containment, in normal or accidental situations, are displayed. Limitation devices of effluents and release for normal operating (noticeably the reprocessing systems of atmosphere) and safety and protection systems in case of accident are described [fr

  18. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    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

  19. Review of fusion DEMO reactor study

    International Nuclear Information System (INIS)

    Seki, Yasushi

    1996-01-01

    Fusion DEMO Reactor is defined and the Steady State Tokamak Reactor (SSTR) concept is introduced as a typical example of a DEMO reactor. Recent DEMO reactor studies in Japan and abroad are introduced. The DREAM Reactor concept is introduced as an ultimate target of fusion research. (author)

  20. Advances in fusion reactor design

    International Nuclear Information System (INIS)

    Baker, C.C.

    1987-01-01

    The author addresses the tokamak as a power reactor. Contrary to popular opinion, there are still a few people that think a tokamak might make a good fusion power reactor. In thinking about advances in fusion reactor design, in the U.S., at least, that generally means advances relevant to the Starfire design. He reviews some of the features of Starfire. Starfire is the last major study done of the tokamak as a reactor in this country. It is now over eight years old in the sense that eight years ago was really the time in which major decisions were made as to its features. Starfire was a tokamak with a major radius of seven meters, about twice the linear dimensions of a machine like TIBER

  1. Assessment of the critical engineering data needs for the commercialization of magnetic confinement fusion

    International Nuclear Information System (INIS)

    Waganer, L.M.; Zuckerman, D.S.

    1983-01-01

    A survey of twenty-two recent conceptual fusion reactor designs was conducted to ascertain both generic and specific engineering data needs critical for the commercialization of magnetic confinement fusion (MCF). Design experts or advocates for each concept were queried as to the more critical engineering issues and data needs affecting the achievement of commercialization. For each concept, the technical issues were identified and the data needs quantified. Issues and data needs were then ranked based upon the experts' perceptions of the relative importance of each to the concept. The issues encompassed all aspects of the fusion reactor plant design including materials, performance, maintainability, operability, cost, safety and resources

  2. Fusion. An introduction of the physics and technology of magnetic confinement fusion. 2. compl. rev. and enl. ed.

    International Nuclear Information System (INIS)

    Stacey, Weston M.

    2010-01-01

    This second edition of a popular textbook is thoroughly revised with around 25% new and updated content. It provides an introduction to both plasma physics and fusion technology at a level that can be understood by advanced undergraduates and graduate students in the physical sciences and related engineering disciplines. As such, the contents cover various plasma confinement concepts, the support technologies needed to confine the plasma, and the designs of ITER as well as future fusion reactors. With end of chapter problems for use in courses. (orig.)

  3. The spheromak as a compact fusion reactor

    International Nuclear Information System (INIS)

    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

  4. The spheromak as a compact fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Materials needs for compact fusion reactors

    International Nuclear Information System (INIS)

    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 3 versus 0.3 to 0.5 MW/m 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.)

  6. Inertia-confining thermonuclear molten salt reactors

    International Nuclear Information System (INIS)

    Furukawa, Kazuo; Yamanaka, Chiyoe; Nakai, Sadao; Imon, Shunji; Nakajima, Hidenori; Nakamura, Norio; Kato, Yoshio.

    1984-01-01

    Purpose: To increase the heat generating efficiency while improving the reactor safety and thereby maintaining the energy balance throughout the reactor. Constitution: In an inertia-confining type D-T thermonuclear reactor, the blanket is made of lithium-containing fluoride molten salts (LiF.BeF 2 , LiF.NaF.KF, LiF.KF, etc) which are cascaded downwardly in a large thickness (50 - 100 cm) along the inner wall of the thermonuclear reaction vessel, and neutrons generated by explosive compression are absorbed to lithium in the molten salts to produce tritium, Heat transportation is carried out by the molten salts. (Ikeda, J.)

  7. Fusion reactors and the environment

    International Nuclear Information System (INIS)

    Wrixon, A.D.

    1976-01-01

    A summary is given of the report of a study group set up in 1971 by the Director of the UKAEA Culham Laboratory to investigate environmental and safety aspects of future commercial fusion reactors (1975, Carruthers, R., Dunster, H.J., Smith, R.D., Watson, C.J.H., and Mitchell, J.T.D., Culham Study Group Report on Fusion Reactors and the Environment, CLM-R148, HMSO, London). This report was originally issued in 1973 under limited distribution, but has only recently been made available for open circulation. Deuterium/tritium fusion is thought to be the most likely reaction to be used in the first generation of reactors. Estimates were made of the local and world-wide population hazards from the release of tritium, both under normal operating conditions and in the event of an accident. One serious type of accident would be a lithium metal fire in the blanket region of the reactor. The use of a fusible lithium salt (FLIBE), eliminating the lithium fire risk, is considered but the report concentrates on lithium metal in the blanket region. The main hazards to operating staff arise both from tritium and from neutron activation of the construction materials. Remote servicing of the reactor structure will be essential, but radioactive waste management seems less onerous than for fission reactors. Meaningful comparison of the overall hazards associated with fusion and fission power programmes is not yet possible. The study group emphasized the need for more data to aid the safety assessments, and the need for such assessments to keep pace with fusion power station design. (U.K.)

  8. Technology requirements for commercial applications of inertial confinement fusion

    International Nuclear Information System (INIS)

    Frank, T.G.; Rossi, C.E.

    1981-01-01

    Current inertial confinement fusion (ICF) research is directed primarily at physics experiments intended to provide confidence in the scientific feasibility of the basic concept. In conjunction with these experiments, a variety of laser and particle beam drivers having potential for eventual use in fusion power plants is being developed. Expectations are that the scientific feasibility of ICF will be demonstrated in the latter part of the 1980s. At that time, the emphasis of the program will shift to engineering, economic, environmental, and licensing issues with the necessary technology development effort continuing into the early part of the next century. This paper discusses the technology requirements for the successive phases of engineering development leading to commercial application of ICF. The engineering areas requiring significant effort for ICF application include high average power driver development; pulsed high-energy power supply development; reactor cavity and heat transport system design; tritium extraction and control; commercial pellet development; pellet injection, tracking, and targeting systems design; materials radiation, fatigue, and corrosion behavior; and reactor plant systems integration and demonstration

  9. Fusion energy

    International Nuclear Information System (INIS)

    Gross, R.A.

    1984-01-01

    This textbook covers the physics and technology upon which future fusion power reactors will be based. It reviews the history of fusion, reaction physics, plasma physics, heating, and confinement. Descriptions of commercial plants and design concepts are included. Topics covered include: fusion reactions and fuel resources; reaction rates; ignition, and confinement; basic plasma directory; Tokamak confinement physics; fusion technology; STARFIRE: A commercial Tokamak fusion power plant. MARS: A tandem-mirror fusion power plant; and other fusion reactor concepts

  10. Helium cooling of fusion reactors

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

    1994-01-01

    On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

  11. On fusion and fission breeder reactors

    International Nuclear Information System (INIS)

    Brandt, B.; Schuurman, W.; Klippel, H.Th.

    1981-02-01

    Fast breeder reactors and fusion reactors are suitable candidates for centralized, long-term energy production, their fuel reserves being practically unlimited. The technology of a durable and economical fusion reactor is still to be developed. Such a development parallel with the fast breeder is valuable by reasons of safety, proliferation, new fuel reserves, and by the very broad potential of the development of the fusion reactor. In order to facilitate a discussion of these aspects, the fusion reactor and the fast breeder reactor were compared in the IIASA-report. Aspects of both reactor systems are compared

  12. Radiation shielding for fusion reactors

    International Nuclear Information System (INIS)

    Santoro, R.T.

    2000-01-01

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

  13. Prospects for toroidal fusion reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.D.

    1994-01-01

    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, ∼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

  14. Inertial confinement fusion and fast ignitor studies

    International Nuclear Information System (INIS)

    Willi, O.; Barringer, L.; Bell, A.

    1999-01-01

    The paper discusses inertial confinement fusion research carried out at several different laser facilities including the VULCAN laser at the Rutherford Appleton Laboratory, the TRIDENT laser at the Los Alamos National Laboratory and the PHEBUS laser at Limeil. Low density foam targets were irradiated either with nanosecond laser or soft x-ray pulses. Laser imprinting was studied and in particular saturation of areal density perturbations induced by near-single mode laser imprinting has been observed. Several issues important for the foam buffered direct drive scheme were investigated. These studies included measurements of the absolute levels of Stimulated Brillouin and Raman Scattering observed from laser irradiated low density foam targets either bare or overcoated with a thin layer of gold. A novel scheme is proposed to increase the pressure in indirectly driven targets. Low density foams that are mounted onto a foil target are heated with an intense pulse of soft x-ray radiation. If the foam is heated supersonically the pressure generated is not only the ablation pressure but the combined pressure due to ablation at the foam/foil interface and the heated foam material. The scheme was confirmed on planar targets. Brominated foil targets overcoated with a low density foam were irradiated by a soft x-ray pulse emitted from a hohlraum. The pressure was obtained by comparing the rear side trajectory of the driven target observed by soft x-ray radiography to one dimensional radiation hydrodynamic simulations. Further, measurements were carried out to observe the transition from super- to subsonic propagation of an ionisation front in low density chlorinated foam targets irradiated by an intense soft x-ray pulse both in open and confined geometry. The diagnostic for these measurements was K-shell point projection absorption spectroscopy. In the fast ignitor area the channeling and guiding of picosecond laser pulses through underdense plasmas, preformed density

  15. Inertial confinement fusion and fast ignitor studies

    International Nuclear Information System (INIS)

    Willi, O.; Barringer, L.; Bell, A.

    2001-01-01

    The paper discusses inertial confinement fusion research carried out at several different laser facilities including the VULCAN laser at the Rutherford Appleton Laboratory, the TRIDENT laser at the Los Alamos National Laboratory and the PHEBUS laser at Limeil. Low density foam targets were irradiated either with nanosecond laser or soft x-ray pulses. Laser imprinting was studied and in particular saturation of areal density perturbations induced by near-single mode laser imprinting has been observed. Several issues important for the foam buffered direct drive scheme were investigated. These studies included measurements of the absolute levels of Stimulated Brillouin and Raman Scattering observed from laser irradiated low density foam targets either bare or overcoated with a thin layer of gold. A novel scheme is proposed to increase the pressure in indirectly driven targets. Low density foams that are mounted onto a foil target are heated with an intense pulse of soft x-ray radiation. If the foam is heated supersonically the pressure generated is not only the ablation pressure but the combined pressure due to ablation at the foam/foil interface and the heated foam material. The scheme was confirmed on planar targets. Brominated foil targets overcoated with a low density foam were irradiated by a soft x-ray pulse emitted from a hohlraum. The pressure was obtained by comparing the rear side trajectory of the driven target observed by soft x-ray radiography to one dimensional radiation hydrodynamic simulations. Further, measurements were carried out to observe the transition from super- to subsonic propagation of an ionisation front in low density chlorinated foam targets irradiated by an intense soft x-ray pulse both in open and confined geometry. The diagnostic for these measurements was K-shell point projection absorption spectroscopy. In the fast ignitor area the channeling and guiding of picosecond laser pulses through underdense plasmas, preformed density

  16. Physical and engineering aspects of a fusion engineering test facility based on mirror confinement

    International Nuclear Information System (INIS)

    Kawabe, T.; Hirayama, S.; Hojo, H.; Kozaki, Y.; Yoshikawa, K.

    1986-01-01

    Controlled fusion research has accomplished great progress in the field of confinement of high-density and high-temperature plasmas and breakeven experiments are expected before the end of the 1980s. Many experiments have been proposed as the next step for fusion research. Among them is the study of ignited plasmas and another is the study of fusion engineering. Some of the important studies in fusion engineering are the integrated test in a fusion reactor environment as well as tests of first-wall materials and of the reactor structures, and test for tritium breeding and blanket modules or submodules. An ideal neutron source for the study of fusion engineering is the deuterium-tritium (D-T) fusion plasma itself. A neutron facility based on a D-T-burning plasma consists of all of the components that a real fusion power reactor would have, so eventually the integrated test for fusion reactor engineering can be done as well as the tests for each engineering component

  17. Plan for the development and commercialization of inertial confinement fusion

    International Nuclear Information System (INIS)

    Willke, T.; Dingee, D.; Ault, L.; Bampton, M.; Bickford, W.; Hartman, J.; Rockwood, A.; Simonen, E.; Teofilo, V.; Frank, T.

    1978-01-01

    An engineering development program strategy to take inertial confinement fusion (ICF) from the milestone of scientific feasibility to a point where its commercial viability can be determined is described. The ICF program objectives and basic program strategy are discussed

  18. Tokamak Fusion Test Reactor. Final conceptual design report

    International Nuclear Information System (INIS)

    1976-02-01

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included

  19. Polarization beam smoothing for inertial confinement fusion

    International Nuclear Information System (INIS)

    Rothenberg, Joshua E.

    2000-01-01

    For both direct and indirect drive approaches to inertial confinement fusion (ICF) it is imperative to obtain the best possible drive beam uniformity. The approach chosen for the National Ignition Facility uses a random-phase plate to generate a speckle pattern with a precisely controlled envelope on target. A number of temporal smoothing techniques can then be employed to utilize bandwidth to rapidly change the speckle pattern, and thus average out the small-scale speckle structure. One technique which generally can supplement other smoothing methods is polarization smoothing (PS): the illumination of the target with two distinct and orthogonally polarized speckle patterns. Since these two polarizations do not interfere, the intensity patterns add incoherently, and the rms nonuniformity can be reduced by a factor of (√2). A number of PS schemes are described and compared on the basis of the aggregate rms and the spatial spectrum of the focused illumination distribution. The (√2) rms nonuniformity reduction of PS is present on an instantaneous basis and is, therefore, of particular interest for the suppression of laser plasma instabilities, which have a very rapid response time. When combining PS and temporal methods, such as smoothing by spectral dispersion (SSD), PS can reduce the rms of the temporally smoothed illumination by an additional factor of (√2). However, it has generally been thought that in order to achieve this reduction of (√2), the increased divergence of the beam from PS must exceed the divergence of SSD. It is also shown here that, over the time scales of interest to direct or indirect drive ICF, under some conditions PS can reduce the smoothed illumination rms by nearly (√2) even when the PS divergence is much smaller than that of SSD. (c) 2000 American Institute of Physics

  20. Antiproton fast ignition for inertial confinement fusion

    International Nuclear Information System (INIS)

    Perkins, L.J.

    1999-01-01

    With 180 MJ/microg, antiprotons offer the highest stored energy per unit mass of any known entity. The use of antiprotons to promote fast ignition in an inertial confinement fusion (ICF) capsule and produce high target gains with only modest compression of the main fuel is investigated. Unlike standard fast ignition where the ignition energy is supplied by energetic, short pulse laser, the energy here is supplied through the ionization energy deposited when antiprotons annihilate at the center of a compressed fuel capsule. This can be considered in-situ fast ignition as it obviates the need for the external injection of the ignition energy. In the first of two candidate schemes, the antiproton package is delivered by a low-energy ion beam. In the second, autocatalytic scheme, the antiprotons are preemplaced at the center of the capsule prior to compression. In both schemes, the author estimates that ∼10 12 antiprotons are required to initiate fast ignition in a typical ICF capsule and show that incorporation of a thin, heavy metal shell is desirable to enhance energy deposition within the ignitor zone. In addition to eliminating the need for a second, energetic fast laser and vulnerable final optics, this scheme would achieve central ignition without reliance on laser channeling through halo plasma or Hohlraum debris. However, in addition to the practical difficulties of storage and manipulation of antiprotons at low energy, the other large uncertainty for the practicality of such a speculative scheme is the ultimate efficiency of antiproton production in an external, optimized facility. Estimates suggest that the electrical wall plug energy per pulse required for the separate production of the antiprotons is of the same order as that required for the conventional slow compression driver

  1. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    Victoria, M.; Baluc, N.; Spaetig, P.

    2001-01-01

    In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

  2. Synfuels production from fusion reactors

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.; Steinberg, M.

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies by high temperature electrolysis of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets

  3. Consideration on nuclear fusion in plasma by the magnetic confinement as a heat engine

    International Nuclear Information System (INIS)

    Tsuji, Yoshio

    1990-01-01

    In comparing nuclear fusion in plasma by the magnetic confinement with nuclear fission and chemical reactions, the power density and the function of a heat engine are discussed using a new parameter G introduced as an eigenvalue of a reaction and the value of q introduced to estimate the thermal efficiency of a heat engine. It is shown that the fusion reactor by the magnetic confinement is very difficult to be a modern heat engine because of the lack of some indispensable functions as a modern heat engine. The value of G and q have the important role in the consideration. (author)

  4. Fusion reactor horizontal versus vertical maintenance approach

    International Nuclear Information System (INIS)

    Charruyer, Ph.; Djerassi, H.; Leger, D.; Maupou, M.; Rouillard, J.; Salpietro, E.; Holloway, C.; Suppan, A.

    1987-01-01

    This paper concerns the comparison of horizontal versus vertical maintenance options of internal components (blanket and segment) of fusion reactors NET (Next European Torus) and INTOR Design. The described mechanical options are taken to ensure the handling of internals with the required precision, taking into account the problems raised by the safety and confinement requirements. Handling is obviously performed remotely. The option comparisons are performed according to the criteria of feasibility, building size, duration of maintenance operations, safety, flexibility, availability and cost. The first conclusions point on that the vertical handling option offers advantages, as regards the ease of handling and confinement possibilities. From the building size point of view, the two solutions are almost equivalent, while other criteria do not provide a basis for choice. It is emphasized that the confinement option C.T.U. (Containment Transfer Unit) or T.I.C. (Tight Intermediate Confinement) should be the major factor in determining the best options. In additions, a cost comparative analysis emphasizes the best cost/benefit ratio for the different options studied

  5. Vacuum engineering for fusion research and fusion reactors

    International Nuclear Information System (INIS)

    Pittenger, L.C.

    1976-01-01

    The following topics are described: (1) surface pumping by cryogenic condensation, (2) operation of large condensing cryopumps, (3) pumping for large fusion experiments, and (4) vacuum technology for fusion reactors

  6. Fusion reactor systems studies

    International Nuclear Information System (INIS)

    1993-01-01

    Fusion Technology Institute personnel actively participated in the ARIES/PULSAR project during the present contract period. Numerous presentations were made at PULSAR project meetings, major contributions were written for the ARIES-II/IV Final Report presentations and papers were given at technical conferences contributions were written for the ARIES Lessons Learned report and a very large number of electronic-mail and regular-mail communications were sent. The remaining sections of this progress report win summarize the work accomplished and in progress for the PULSAR project during the contract period. The main areas of effort are: PULSAR Research; ARIES-II/IV Report Contributions; ARIES Lessons Learned Report Contributions; and Stellarator Study

  7. Comparative assessment of world research efforts on magnetic confinement fusion

    International Nuclear Information System (INIS)

    McKenney, B.L.; McGrain, M.; Rutherford, P.H.

    1990-02-01

    This report presents a comparative assessment of the world's four major research efforts on magnetic confinement fusion, including a comparison of the capabilities in the Soviet Union, the European Community (Western Europe), Japan, and the United States. A comparative evaluation is provided in six areas: tokamak confinement; alternate confinement approaches; plasma technology and engineering; and fusion computations. The panel members are involved actively in fusion-related research, and have extensive experience in previous assessments and reviews of the world's four major fusion programs. Although the world's four major fusion efforts are roughly comparable in overall capabilities, two conclusions of this report are inescapable. First, the Soviet fusion effort is presently the weakest of the four programs in most areas of the assessment. Second, if present trends continue, the United States, once unambiguously the world leader in fusion research, will soon lose its position of leadership to the West European and Japanese fusion programs. Indeed, before the middle 1990s, the upgraded large-tokamak facilities, JT-60U (Japan) and JET (Western Europe), are likely to explore plasma conditions and operating regimes well beyond the capabilities of the TFTR tokamak (United States). In addition, if present trends continue in the areas of fusion nuclear technology and materials, and plasma technology and materials, and plasma technology development, the capabilities of Japan and Western Europe in these areas (both with regard to test facilities and fusion-specific industrial capabilities) will surpass those of the United States by a substantial margin before the middle 1990s

  8. Civilian applications of particle-beam-initiated inertial confinement fusion technology

    International Nuclear Information System (INIS)

    Varnado, S.G.; Mitchiner, J.L.

    1977-05-01

    Electrical power generation by controlled fusion may provide a partial solution to the world's long-term energy supply problem. Achievement of a fusion reaction requires the confinement of an extremely hot plasma for a time long enough to allow fuel burnup. Inertial confinement of the plasma may be possible through the use of tightly focused, relativistic electron or ion beams to compress a fuel pellet. The Sandia Particle Beam Fusion program is developing the particle-beam accelerators necessary to achieve fuel ignition. In this report we review the status of the particle-beam fusion technology development program and identify several potential civilian applications for this technology. We describe program objectives, discuss the specific accelerators presently under development, and briefly review the results of beam-focusing and target-irradiation experiments. Then we identify and discuss applications for the beam technology and for the fusion neutrons. The applications are grouped into near-term, intermediate-term, and long-term categories. Near-term applications for the beam technology include electron-beam (e-beam) pumping of gas lasers and several commercial applications. Intermediate-term applications (pellet gain less than 50) include hybrid reactors for electrical power production and fissile fuel breeding, pure fusion reactors for electrical power production, and medical therapy using ion accelerators. In the long term, complex, high-gain pellets may be used in pure fusion reactors

  9. The sensitivity theory for inertial confinement pellet fusion system

    International Nuclear Information System (INIS)

    Cai, Shaohui; Zhang, Yuquan.

    1986-01-01

    A sensitivity theory for inertial confinement pellet fusion system is developed based on a physical model similar to that embodied in the laser fusion code MEDUSA. The theory presented here can be an efficient tool for estimating the effects of many alternations in the data field. Our result is different from Greenspan's work in 1980. (author)

  10. Fusion reactor high vacuum pumping

    International Nuclear Information System (INIS)

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

    1992-01-01

    This paper reports on recent experiments which have shown the practicality of using activated carbon (coconut charcoal) at 4K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were satisfactory. The long-term effects of tritium on the charcoal/cement system developed by Grumman and LLNL was now known; therefore 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 K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately one-third way through, and after the exposure. Modest effects were noted which would not seriously restrict the use of charcoal as a cryosorber for fusion reactor high-vacuum pumping applications

  11. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.; Talbot, J.B.

    1980-01-01

    When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

  12. Assessment of fusion reactor development. Proceedings

    International Nuclear Information System (INIS)

    Inoue, N.; Tazima, T.

    1994-04-01

    Symposium on assessment of fusion reactor development was held to make clear critical issues, which should be resolved for the commercial fusion reactor as a major energy source in the next century. Discussing items were as follows. (1) The motive force of fusion power development from viewpoints of future energy demand, energy resources and earth environment for 'Sustainable Development'. (2) Comparison of characteristics with other alternative energy sources, i.e. fission power and solar cell power. (3) Future planning of fusion research and advanced fuel fusion (D 3 He). (4) Critical issues of fusion reactor development such as Li extraction from the sea water, structural material and safety. (author)

  13. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  14. Nuclear data requirements for fusion reactor nucleonics

    International Nuclear Information System (INIS)

    Bhat, M.R.; Abdou, M.A.

    1980-01-01

    Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future

  15. Generic magnetic fusion reactor cost assessment

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    The Fusion Energy Division of the Oak Ridge National Laboratory discusses ''generic'' magnetic fusion reactors. The author comments on DT burning magnetic fusion reactor models being possibly operational in the 21st century. Representative parameters from D-T reactor studies are given, as well as a shematic diagram of a generic fusion reactor. Values are given for winding pack current density for existing and future superconducting coils. Topics included are the variation of the cost of electricity (COE), the dependence of the COE on the net electric power of the reactor, and COE formula definitions

  16. Scaling laws for inertial confinement fusion

    International Nuclear Information System (INIS)

    Brueckner, K.A.

    1978-01-01

    The fusion yield due to a spherically expanding burning front in a compressed fuel pellet is obtained. The pellet gain and beam energy for a laser system and an ion or electron beam driven system are compared. The results suggest an interesting possibility for heavy-ion fusion with driver parameters far below those usually considered

  17. ITER, the 'Broader Approach', a DEMO fusion reactor

    International Nuclear Information System (INIS)

    Janeschitz, G.; Bahm, W.

    2007-01-01

    Fusion is a very promising future energy option, which is characterized by almost unlimited fuel reserves, favourable safety features and environmental sustainability. The aim of the worldwide fusion research is a fusion power station which imitates the process taking place in the sun and thus gains energy from the fusion of light atomic nuclei. The experimental reactor ITER which will be built in Cadarache, France, marks a breakthrough in the worldwide fusion research: For the first time an energy multiplication factor of at least 10 will be achieved, the factor by which the fusion power exceeds the external plasma heating. Partners in this project are the European Union, Japan, the Russian Federation, USA, China, South Korea and India as well as Brazil as associated partner. The facility is supposed to demonstrate a long burning, reactor-typical plasma and to test techniques such as plasma heating, plasma confinement by superconducting magnets, fuel cycle as well as energy transition, tritium breeding and remote handling technologies. The next step beyond ITER will be the demonstration power station DEMO which requires further developments in order to create the basis for its design and construction. The roadmap to fusion energy is described. It consists of several elements which are needed to develop the knowledge required for a commercial fusion reactor. The DEMO time schedule depends on the efforts in terms of personnel and budget resources the society is willing to invest in fusion taking into account the long term energy supply and its environmental impact. (orig.)

  18. Inertial-confinement fusion with lasers

    International Nuclear Information System (INIS)

    Betti, R.; Hurricane, O. A.

    2016-01-01

    The quest for controlled fusion energy has been ongoing for over a half century. The demonstration of ignition and energy gain from thermonuclear fuels in the laboratory has been a major goal of fusion research for decades. Thermonuclear ignition is widely considered a milestone in the development of fusion energy, as well as a major scientific achievement with important applications to national security and basic sciences. The U.S. is arguably the world leader in the inertial con fment approach to fusion and has invested in large facilities to pursue it with the objective of establishing the science related to the safety and reliability of the stockpile of nuclear weapons. Even though significant progress has been made in recent years, major challenges still remain in the quest for thermonuclear ignition via laser fusion

  19. Thermonuclear plasma physic: inertial confinement fusion; Physique des plasmas thermonucleaires: la fusion par confinement inertiel

    Energy Technology Data Exchange (ETDEWEB)

    Bayer, Ch.; Juraszek, D

    2001-07-01

    Inertial Confinement Fusion (ICF) is an approach to thermonuclear fusion in which the fuel contained in a spherical capsule is strongly compressed and heated to achieve ignition and burn. The released thermonuclear energy can be much higher than the driver energy, making energetic applications attractive. Many complex physical phenomena are involved by the compression process, but it is possible to use simple analytical models to analyze the main critical points. We first determine the conditions to obtain fuel ignition. High thermonuclear gains are achieved if only a small fraction of the fuel called hot spot is used to trigger burn in the main fuel compressed on a low isentrope. A simple hot spot model will be described. The high pressure needed to drive the capsule compression are obtained by the ablation process. A simple Rocket model describe the main features of the implosion phase. Several parameters have to be controlled during the compression: irradiation symmetry, hydrodynamical stability and when the driver is a laser, the problems arising from interaction of the EM wave with the plasma. Two different schemes are examined: Indirect Drive which uses X-ray generated in a cavity to drive the implosion and the Fast Ignitor concept using a ultra intense laser beam to create the hot spot. At the end we present the Laser Megajoule (LMJ) project. LMJ is scaled to a thermonuclear gain of the order of ten. (authors)

  20. Prospect of realizing nuclear fusion reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

  1. Radiation control in fusion plasmas by magnetic confinement

    International Nuclear Information System (INIS)

    Dachicourt, R.

    2012-10-01

    The present work addresses two important issues for the industrial use of fusion: plasma radiation control, as a part of the more general power handling issue, and high density tokamak operation. These two issues will be most critical in the demonstration reactor, called DEMO, intermediate step between ITER and a future commercial reactor. For DEMO, the need to radiate a large fraction of the power so as to limit the peak power load on the divertor will be a key constraint. High confinement will have to be combined with high radiated power fraction, and the required level of plasma purity. The main achievement of this thesis is to have shown experimental evidence of the existence of a stable plasma regime meeting the most critical requirements of a DEMO scenario: an electron density up to 40% above the Greenwald value, together with a fraction of radiated power close to 80%, with a good energy confinement and limited dilution. The plasma is additionally heated with ion cyclotron waves in a central electron heating scenario, featuring alpha particle heating. The original observations reported in this work bring highly valuable new pieces of information both to the physics of the tokamak edge layer and to the construction of an 'integrated operational scenario' required to successfully operate fusion devices. In the way for getting high density plasmas, the new observations involve the following topics. First, the formation of a poloidal asymmetry in the edge electron density profile, with a maximum density located close to toroidal pumped limiter. This asymmetry occurs inside the separatrix, with a constant plasma pressure on magnetic surfaces. Secondly, a correlative decrease of the electron temperature in the same edge region. Thirdly, the excellent coupling capabilities of the ICRH waves, up to a central line averaged electron density of 1.4 times the Greenwald density. Fourthly, a poloidally asymmetric edge radiation region, providing the dissipation of 80% of

  2. Inertial Confinement Fusion R and D and Nuclear Proliferation

    International Nuclear Information System (INIS)

    Goldston, Robert J.

    2011-01-01

    In a few months, or a few years, the National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory may achieve fusion gain using 192 powerful lasers to generate x-rays that will compress and heat a small target containing isotopes of hydrogen. This event would mark a major milestone after decades of research on inertial confinement fusion (ICF). It might also mark the beginning of an accelerated global effort to harness fusion energy based on this science and technology. Unlike magnetic confinement fusion (ITER, 2011), in which hot fusion fuel is confined continuously by strong magnetic fields, inertial confinement fusion involves repetitive fusion explosions, taking advantage of some aspects of the science learned from the design and testing of hydrogen bombs. The NIF was built primarily because of the information it would provide on weapons physics, helping the United States to steward its stockpile of nuclear weapons without further underground testing. The U.S. National Academies' National Research Council is now hosting a study to assess the prospects for energy from inertial confinement fusion. While this study has a classified sub-panel on target physics, it has not been charged with examining the potential nuclear proliferation risks associated with ICF R and D. We argue here that this question urgently requires direct and transparent examination, so that means to mitigate risks can be assessed, and the potential residual risks can be balanced against the potential benefits, now being assessed by the NRC. This concern is not new (Holdren, 1978), but its urgency is now higher than ever before.

  3. Tokamak Fusion Test Reactor D-T results

    International Nuclear Information System (INIS)

    Meade, D.M.

    1995-01-01

    Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, α confinement, α heating and possible α-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20MW of tritium and 14MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2MW. The fusion power density in the core of the plasma was about 1.8MWm -3 , approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of τ E ∝A 0.6 . The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6keV can be attributed to electron heating by the α particles. The approximately 5% loss of α particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy α particles and the resultant α ash density. At fusion power levels of 7.5MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional α loss due to the fluctuations was observed. (orig.)

  4. Progress in toroidal confinement and fusion research

    International Nuclear Information System (INIS)

    Furth, H.P.

    1987-10-01

    During the past 30 years, the characteristic T/sub i/n tau/sub E/-value of toroidal-confinement experiments has advanced by more than seven orders of magnitude. Part of this advance has been due to an increase of gross machine parameters. Most of this advance has been due to an increase of gross machine parameters. Most of the advance is associated with improvements in the ''quality of plasma confinement.'' The combined evidence of spherator and tokamak research clarifies the role of magnetic-field geometry in determining confinement and points to the importance of shielding out plasma edge effects. A true physical understanding of anomalous transport remains to be achieved. 39 refs., 11 figs., 1 tab

  5. What have fusion reactor studies done for you today?

    International Nuclear Information System (INIS)

    Kulchinski, G.L.

    1985-01-01

    The University of Wisconsin examines the fusion program and puts into perspective what return is being made on investments in fusion reactor studies. Illustations show financial support for fusion research from the four major programs, FY'82 expenditures on fusion research, and the total expenditures on fusion research since 1951. Topics discussed include the estimated number of scientists conducting fusion research, the conceptual design study of a fusion reactor, scoping study of a reactor, the chronology of fusion reactor design studies, published fusion reactor studies 1967-1983, conceptual fusion reactor design studies, STARFIRE reference design, MARS central cell, HYLIFE reaction chamber, and selected contributions of reactor design studies to base programs

  6. Vanadium recycling for fusion reactors

    International Nuclear Information System (INIS)

    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

  7. Muon catalyzed fusion - fission reactor driven by a recirculating beam

    International Nuclear Information System (INIS)

    Eliezer, S.; Tajima, T.; Rosenbluth, M.N.

    1986-01-01

    The recent experimentally inferred value of multiplicity of fusion of deuterium and tritium catalyzed by muons has rekindled interest in its application to reactors. Since the main energy expended is in pion (and consequent muon) productions, we try to minimize the pion loss by magnetically confining pions where they are created. Although it appears at this moment not possible to achieve energy gain by pure fusion, it is possible to gain energy by combining catalyzed fusion with fission blankets. We present two new ideas that improve the muon fusion reactor concept. The first idea is to combine the target, the converter of pions into muons, and the synthesizer into one (the synergetic concept). This is accomplished by injecting a tritium or deuterium beam of 1 GeV/nucleon into DT fuel contained in a magnetic mirror. The confined pions slow down and decay into muons, which are confined in the fuel causing little muon loss. The necessary quantity of tritium to keep the reactor viable has been derived. The second idea is that the beam passing through the target is collected for reuse and recirculated, while the strongly interacted portion of the beam is directed to electronuclear blankets. The present concepts are based on known technologies and on known physical processes and data. 29 refs., 6 figs., 4 tabs

  8. Critical plasma-materials issues for fusion reactor designs

    International Nuclear Information System (INIS)

    Wilson, K.L.; Bauer, W.

    1983-01-01

    Plasma-materials interactions are a dominant driving force in the design of fusion power reactors. This paper presents a summary of plasma-materials interactions research. Emphasis is placed on critical aspects related to reactor design. Particular issues to be addressed are plasma edge characterization, hydrogen recycle, impurity introduction, and coating development. Typical wall fluxes in operating magnetically confined devices are summarized. Recent calculations of tritium inventory and first wall permeation, based on laboratory measurements of hydrogen recycling, are given for various reactor operating scenarios. Impurity introduction/wall erosion mechanisms considered include sputtering, chemical erosion, and evaporation (melting). Finally, the advanced material development for in-vessel components is discussed. (author)

  9. Tritium production in fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.

    1981-08-01

    The present analyses on the possibilities of extracting tritium from the liquid and solid fusion reactor blankets show up many problems. A consistent ensemble of materials and devices for extracting the heat and the tritium has not yet been integrated in a fusion reactor blanket project. The dimensioning of the many pipes required for shifting the tritium can only be done very approximately and the volume taken up by the blanket is difficult to evaluate, etc. The utilization of present data leads to over-dimensioning the installations by prudence and perhaps rejecting the best solutions. In order to measure the parameters of the most promising materials, work must be carried out on well defined samples and not only determine the base physical-chemical coefficients, such as thermal conductivity, scattering coefficients, Sievert parameters, but also the kinetic parameters conventional in chemical engineering, such as the hourly space rates of degassing. It is also necessary to perform long duration experiments under radiation and at operating temperatures, or above, in order to study the ageing of the bodies employed [fr

  10. Towards diagnostics for a fusion reactor

    International Nuclear Information System (INIS)

    Costley, A. E.

    2009-01-01

    The requirements for measurements on modern tokamak fusion plasmas are outlined, and the techniques and systems used to make the measurements, usually referred to as 'diagnostics', are introduced. The basics of three particular diagnostics - magnetics, neutron systems and a laser based optical system - are outlined as examples of modern diagnostic systems, and the implementation of these diagnostics on a current tokamak (JET) are described. The next major step in magnetic confinement fusion is the construction and operation of the International Thermonuclear Experimental Reactor (ITER), which is a joint project of China, Europe, Japan, India, Korea, the Russian Federation, and the United States. Construction has begun in Cadarache, France. It is expected that ITER will operate at the 500 MW level. Because of the harsh environment in the vacuum vessel where many diagnostic components are located, the development of diagnostics for ITER is a major challenge - arguably the most difficult challenge ever undertaken in the field of diagnostics. The main elements in the diagnostic step are outlined using the three chosen techniques as examples. Finally, the step beyond ITER to a demonstration reactor, DEMO, that is expected to produce several GWs of fusion power is considered and the impact on diagnostics outlined. It is shown that the applicability and development steps needed for the individual diagnostics techniques will differ. The challenges for DEMO diagnostics are substantial and a dedicated effort should be made to find and develop new techniques, and especially techniques appropriate to the DEMO environment. It is argued that the limitations and difficulties in diagnostics should be a consideration in the optimization and designs of DEMO. (author)

  11. The TITAN Reversed-Field Pinch fusion reactor study

    International Nuclear Information System (INIS)

    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 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 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

  12. Review of compact, alternate concepts for magnetic confinement fusion

    International Nuclear Information System (INIS)

    Nickerson, S.B.; Shmayda, W.T.; Dinner, P.J.; Gierszewski, P.

    1984-06-01

    This report documents a study of compact alternate magnetic confinement fusion experiments and conceptual reactor designs. The purpose of this study is to identify those devices with a potential to burn tritium in the near future. The bulk of the report is made up of a review of the following compact alternates: compact toroids, high power density tokamaks, linear magnetic systems, compact mirrors, reversed field pinches and some miscellaneous concepts. Bumpy toruses and stellarators were initially reviewed but were not pursued since no compact variations were found. Several of the concepts show promise of either burning tritium or evolving into tritium burning devices by the early 1990's: RIGGATRON, Ignitor, OHTE, Frascati Tokamak upgrade, several driven (low or negative net power) mirror experiments and several Reversed Field Pinch experiments that may begin operation around 1990. Of the above only the Frascati Tokamak Upgrade has had funds allocated. Also identified in this report are groups who may have tritium burning experiments in the mid to late 1990's. There is a discussion of the differences between the reviewed devices and the mainline tokamak experiments. This discussion forms the basis of recommendations for R and D aimed at the compact alternates and the applicability of the present CFFTP program to the needs of the compact alternates. These recommendations will be presented in a subsequent report

  13. Safety and environmental aspects of the HYLIFE-II and ARIES fusion reactor designs

    International Nuclear Information System (INIS)

    Dolan, T.J.; Longhurst, G.R.; Herring, J.S.

    1993-01-01

    The HYLIFE-II inertial confinement fusion reactor design uses jets of Flibe molten salt to protect the blast chamber walls and to breed tritium. It has a low tritium inventory and effective tritium removal. The issue with this design is not one of safety but of economics. The ARIES reactor designs have safety concerns associated with fires. These reactors designs are described

  14. Laser thermonuclear fusion with force confinement of hot plasma

    International Nuclear Information System (INIS)

    Korobkin, V.V.; Romanovsky, M.Y.

    1994-01-01

    The possibility of the utilization of laser radiation for plasma heating up to thermonuclear temperatures with its simultaneous confinement by ponderomotive force is investigated. The plasma is located inside a powerful laser beam with a tubelike section or inside a cavity of duct section, formed by several intersecting beams focused by cylindrical lenses. The impact of various physical processes upon plasma confinement is studied and the criteria of plasma confinement and maintaining of plasma temperature are derived. Plasma and laser beam stability is considered. Estimates of laser radiation energy necessary for thermonuclear fusion are presented

  15. Magnetic and inertial confinement fusion - an overview

    International Nuclear Information System (INIS)

    Murtaza, G.

    1993-01-01

    This paper describes the status of ICF which output is given in terms of number of emitted neutrons and fusion energy as a function of the estimated input energy from the implosion driver. Results from the highest energy laser experiments are summarised. The theoretical targets and the proposed upgrades of NOVA, GEKKO and OMEGA are also shown. The promised ICF paradise will therefore be approached through a process of interpolation between two known extremes. The objectives of this experiment was to produce more than one MW of fusion power in a controlled way, to demonstrate the technology related to tritium usage tritium injection, its tracking, monitoring/recovery and to establish safe procedures for handling tritium in compliance with the regulatory requirements. (A.B)

  16. Heavy ion drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Keefe, D.

    1983-01-01

    The advantages of heavy ion beams as a way of delivering the needed energy and power to an inertial fusion target are surveyed. The existing broad technology base of particle accelerators provides an important foundation for designing, costing, and evaluating proposed systems. The sequence of steps needed for the verification of the heavy ion approach is described; recent research results are even more encouraging than had been assumed hitherto

  17. Heavy ion drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Keefe, D.

    1983-12-01

    The advantages of heavy ion beams as a way of delivering the needed energy and power to an inertial fusion target are surveyed. The existing broad technology base of particle accelerators provides an important foundation for designing, costing, and evaluating proposed systems. The sequence of steps needed for the verification of the heavy ion approach is described; recent research results are even more encouraging than had been assumed hitherto

  18. Inertial confinement fusion at the Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Lindman, E.; Baker, D.; Barnes, C.; Bauer, B.; Beck, J.B.

    1997-01-01

    The Los Alamos National Laboratory is contributing to the resolution of key issues in the US Inertial-Confinement-Fusion Program and plans to play a strong role in the experimental program at the National Ignition Facility when it is completed

  19. Colliding beam fusion reactor space propulsion system

    International Nuclear Information System (INIS)

    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 10 6 -10 9 Watts of output power (0.15-150 Newtons thrust) with a specific impulse of, I sp ∼10 6 sec. A 50 MW propulsion system might involve the following parameters: 4-meters diameterx10-meters length, magnetic field ∼7 Tesla, ion beam current ∼10 A, and fuels of either D-He 3 ,P-B 11 ,P-Li 6 ,D-Li 6 , etc

  20. Reversed-field pinch fusion reactor

    International Nuclear Information System (INIS)

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

    1980-01-01

    A conceptual engineering design of a fusion reactor based on plasma confinement in a toroidal Reversed-Field Pinch (RFP) configuration is described. The plasma is ohmically ignited by toroidal plasma currents which also inherently provide the confining magnetic fields in a toroidal chamber having major and minor radii of 12.7 and 1.5 m, respectively. The DT plasma ignites in 2 to 3 s and undergoes a transient, unrefueled burn at 10 to 20 keV for approx. 20 s to give a DT burnup of approx. 50%. The 5-s dwell period between burn pulses for plasma quench and refueling allows steady-state operation of all thermal systems outside the first wall; no auxiliary thermal capacity is required. Tritium breeding occurs in a granular Li 2 O blanket which is packed around an array of radially oriented water/steam coolant tubes. The slightly superheated steam emerging from this blanket directly drives a turbine that produces electrical power at an efficiency of 30%. A borated-water shield is located immediately outside the thermal blanket to protect the superconducting magnet coils. Both the superconducting poloidal and toroidal field coils are energized by homopolar motor/generators. Accounting for all major energy sinks yields a cost-optimized system with a recirculating power fraction of 0.17; the power output is 750 MWe

  1. Superconducting magnets for toroidal fusion reactors

    International Nuclear Information System (INIS)

    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 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

  2. Heavy ion beam propagation through a gas-filled chamber for inertial confinement fusion

    International Nuclear Information System (INIS)

    Barboza, N.O.

    1996-10-01

    The work presented here evaluates the dynamics of a beam of heavy ions propagating through a chamber filled with gas. The motivation for this research stems from the possibility of using heavy ion beams as a driver in inertial confinement fusion reactors for the purpose of generating electricity. Such a study is important in determining the constraints on the beam which limit its focus to the small radius necessary for the ignition of thermonuclear microexplosions which are the source of fusion energy. Nuclear fusion is the process of combining light nuclei to form heavier ones. One possible fusion reaction combines two isotopes of hydrogen, deuterium and tritium, to form an alpha particle and a neutron, with an accompanying release of ∼17.6 MeV of energy. Generating electricity from fusion requires that we create such reactions in an efficient and controlled fashion, and harness the resulting energy. In the inertial confinement fusion (ICF) approach to energy production, a small spherical target, a few millimeters in radius, of deuterium and tritium fuel is compressed so that the density and temperature of the fuel are high enough, ∼200 g/cm 3 and ∼20 keV, that a substantial number of fusion reactions occur; the pellet microexplosion typically releases ∼350 MJ of energy in optimized power plant scenarios

  3. Inertial Confinement Fusion Annual Report 1997

    International Nuclear Information System (INIS)

    Correll, D

    1998-01-01

    The ICF Annual Report provides documentation of the achievements of the LLNL ICF Program during the fiscal year by the use of two formats: (1) an Overview that is a narrative summary of important results for the fiscal year and (2) a compilation of the articles that previously appeared in the ICF Quarterly Report that year. Both the Overview and Quarterly Report are also on the Web at http://lasers.llnl.gov/lasers/pubs/icfq.html. Beginning in Fiscal Year 1997, the fourth quarter issue of the ICF Quarterly was no longer printed as a separate document but rather included in the ICF Annual. This change provided a more efficient process of documenting our accomplishments with-out unnecessary duplication of printing. In addition we introduced a new document, the ICF Program Monthly Highlights. Starting with the September 1997 issue and each month following, the Monthly Highlights will provide a brief description of noteworthy activities of interest to our DOE sponsors and our stakeholders. The underlying theme for LLNL's ICF Program research continues to be defined within DOE's Defense Programs missions and goals. In support of these missions and goals, the ICF Program advances research and technology development in major interrelated areas that include fusion target theory and design, target fabrication, target experiments, and laser and optical science and technology. While in pursuit of its goal of demonstrating thermonuclear fusion ignition and energy gain in the laboratory, the ICF Program provides research and development opportunities in fundamental high-energy-density physics and supports the necessary research base for the possible long-term application of inertial fusion energy for civilian power production. ICF technologies continue to have spin-off applications for additional government and industrial use. In addition to these topics, the ICF Annual Report covers non-ICF funded, but related, laser research and development and associated applications. We also

  4. Dynamic behaviour of the high confinement mode of fusion plasmas

    International Nuclear Information System (INIS)

    Zohm, H.

    1995-05-01

    This paper describes the dynamic behaviour of the High Confinement mode (H-mode) of fusion plasmas, which is one of the most promising regimes of enhanced energy confinement in magnetic fusion research. The physics of the H-mode is not yet fully understood, and the detailed behaviour is complex. However, we establish a simple physics picture of the phenomenon. Although a first principles theory of the anomalous transport processes in a fusion plasma has not yet been given, we show that within the picture developed here, it is possible to describe the dynamic behaviour of the H-mode, namely the dynamics of the L-H transition and the occurrence of edge localized modes (ELMs). (orig.)

  5. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs, 90 Sr, 129 I, 99 Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R = 1.0 to 3.0) requirements. These studies also indicate that masses on the order of 1.0 g at densities of rho greater than or equal to 500.0 g/cm 3 are required for a practical fusion-based fission product transmutation system

  6. Fusion-product energy loss in inertial confinement fusion plasmas with applications to target burns

    International Nuclear Information System (INIS)

    Harris, D.B.; Miley, G.H.

    1984-01-01

    Inertial confinement fusion (ICF) has been proposed as a competitor to magnetic fusion in the drive towards energy production, but ICF target performance still contains many uncertainties. One such area is the energy-loss rate of fusion products. This situation is due in part to the unique plasma parameters encountered in ICF plasmas which are compressed to more than one-thousand times solid density. The work presented here investigates three aspects of this uncertainty

  7. General principles of magnetic fusion confinement

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1980-01-01

    A few of the areas are described in which there is close interaction between atomic/molecular (A and M) and magnetic fusion physics. The comparisons between predictions of neoclassical transport theory and experiment depend on knowledge of ionization and recombination rate coefficients. Modeling of divertor/scrapeoff plasmas requires better low energy charge exchange cross sections for H + A/sup n+/ collisions. The range of validity of neutral beam trapping cross sections must be broadened, both to encompass the energies typical of present injection experiments and to deal with the problem of prompt trapping of highly excited beam atoms at high energy. Plasma fueling models present certain anomalies that could be resolved by calculation and measurement of low energy (<1 keV) charge exchange cross sections

  8. Advanced nuclear reactor and nuclear fusion power generation

    International Nuclear Information System (INIS)

    2000-04-01

    This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

  9. Lower activation materials and magnetic fusion reactors

    International Nuclear Information System (INIS)

    Conn, R.W.; Bloom, E.E.; Davis, J.W.; Gold, R.E.; Little, R.; Schultz, K.R.; Smith, D.L.; Wiffen, F.W.

    1984-01-01

    Radioactivity in fusion reactors can be effectively controlled by materials selection. The detailed relationship between the use of a material for construction of a magnetic fusion reactor and the material's characteristics important to waste disposal, safety, and system maintainability has been studied. The quantitative levels of radioactivation are presented for many materials and alloys, including the role of impurities, and for various design alternatives. A major outcome has been the development of quantitative definitions to characterize materials based on their radioactivation properties. Another key result is a four-level classification scheme to categorize fusion reactors based on quantitative criteria for waste management, system maintenance, and safety. A recommended minimum goal for fusion reactor development is a reference reactor that (a) meets the requirements for Class C shallow land burial of waste materials, (b) permits limited hands-on maintenance outside the magnet's shield within 2 days of a shutdown, and (c) meets all requirements for engineered safety. The achievement of a fusion reactor with at least the characteristics of the reference reactor is a realistic goal. Therefore, in making design choices or in developing particular materials or alloys for fusion reactor applications, consideration must be given to both the activation characteristics of a material and its engineering practicality for a given application

  10. Activation product transport in fusion reactors

    International Nuclear Information System (INIS)

    Klein, A.C.; Vogelsang, W.F.

    1984-01-01

    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. A computer code, RAPTOR, has been developed to determine the transport of these products in fusion reactor coolant/tritium breeding materials. Without special treatment, it is likely that fusion reactor power plant operators could experience dose rates as high as 8 rem per hour around a number of plant components after only a few years of operation. (orig.)

  11. Fusion radioactivity confinement and application to postulated ITER accidents

    International Nuclear Information System (INIS)

    Piet, S.J.; Brereton, S.J.

    1991-01-01

    An assessment of the ITER radioactivity confinement shows reduction of potential accidental releases to the environment by two orders of magnitude. Important credits are the 1% volume/day confinement leakage rate, radioactivity decay for short-lived isotopes, resumption of detritiation/negative pressure within seven days of the accident, and wind meander during the slow confinement leakage. Achieving this two order of magnitude credit in practice requires appropriate design details, especially the leakage rate and detritiation/negative pressure equipment, and research to validate some key assumptions. The confinement maximizes dependence on passive safety features, thereby working toward using fusion's potential safety advantages. The confinement includes several confinement zones with varying human access allowances. Some confinement areas are normally isolated from the environment, the closed ventilation zone. Some areas have an inert cover gas to inhibit combustion. If future assessments of accidental overpressure show the need, we propose a filter/vent system. This report documents our work for the ITER Conceptual Design Activity (CDA). The report is consistent with the final CDA design reports and descriptions, except that our analysis includes a filter/vent. For gaseous or vapor tritium and for most activated aerosols, the reference release fraction is about 2%. For short-lived tungsten-rhenium aerosols, the reference release fraction is somewhat lower, as low as 0.5% for some accident scenarios. Even without resumption of detritiation/decontamination or negative pressure within seven days of the accident, the release fraction for stays below 4%

  12. Inertially confined fusion using heavy ion drivers

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.; Bangerter, R.O.; Bock, R.; Hogan, W.J.; Lindl, J.D.

    1991-10-01

    The various technical issues of HIF will be briefly reviewed in this paper. It will be seen that there are numerous areas in common in all the approaches to HIF. In the recent International Symposium on Heavy Ion Inertial Fusion, the attendees met in specialized workshop sessions to consider the needs for research in each area. Each of the workshop groups considered the key questions of this report: (1) Is this an appropriate time for international collaboration in HIF? (2) Which problems are most appropriate for such collaboration? (3) Can the sharing of target design information be set aside until other driver and systems issues are better resolved, by which time it might be supposed that there could be a relaxation of classification of target issues? (4) What form(s) of collaboration are most appropriate, e.g., bilateral or multilateral? (5) Can international collaboration be sensibly attempted without significant increases in funding for HIF? The authors of this report share the conviction that collaboration on a broad scale is mandatory for HIF to have the resources, both financial and personnel, to progress to a demonstration experiment. Ultimately it may be possible for a single driver with the energy, power, focusibility, and pulse shape to satisfy the needs of the international community for target physics research. Such a facility could service multiple experimental chambers with a variety of beam geometries and target concepts

  13. Future directions in inertial confinement fusion

    International Nuclear Information System (INIS)

    Bodner, S.E.

    1992-01-01

    The author discusses future directions for the ICF program. At this time there is still uncertainty on a number of key issues necessary to decide on what type of a National Ignition Facility should be constructed. Mechanisms are in place to answer these questions. The author offers his opinions of where the program is likely to proceed. Technology wise indications are that direct drive heating has the best chance of reaching ignition and high gain. This has the advantage of making all three major user programs happy, namely weapons physics, weapons effects, and electrical energy. The demand for and price of energy in the country will have a major impact on the way the program is developed. From the laser fusion side the most promising drivers at present seem to be KrF lasers, and a major concern for these systems is whether the peak to valley nonuniformities can be reduced to the 1 to 2% level when delivered to the target in order to avoid driving instabilities

  14. Deceleration phase of inertial confinement fusion implosions

    International Nuclear Information System (INIS)

    Betti, R.; Anderson, K.; Goncharov, V.N.; McCrory, R.L.; Meyerhofer, D.D.; Skupsky, S.; Town, R.P.J.

    2002-01-01

    A model for the deceleration phase and marginal ignition of imploding capsules is derived by solving a set of ordinary differential equations describing the hot-spot energy balance and the shell dynamics including the return shock propagation. It is found that heat flux leaving the hot spot goes back in the form of internal energy and PdV work of the material ablated off the inner-shell surface. Though the hot-spot temperature is reduced by the heat conduction losses, the hot-spot density increases due to the ablated material in such a way that the hot-spot pressure is approximately independent of heat conduction. For hot-spot temperatures exceeding approximately 7 keV, the ignition conditions are not affected by heat conduction losses that are recycled into the hot spot by ablation. Instead, the only significant internal energy loss is due to the hot-spot expansion tamped by the surrounding shell. The change of adiabat induced by the shock is also calculated for marginally igniting shells, and the relation between the in-flight and stagnation adiabats is in general agreement with the numerical fit of LASNEX simulations by Herrmann et al. [Nucl. Fusion 41, 99 (2001)] and the self-similar solution of Kemp et al. [Phys. Rev. Lett. 15, 3336 (2001)]. The minimum kinetic energy required for ignition is also calculated from the same model and shown to be in good agreement with the numerical fit of LASNEX simulations. It is also found that mass ablation leads to a significant reduction of the deceleration phase Rayleigh-Taylor instability growth rates and to the suppression of short wavelength modes

  15. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs, 90 Sr, 129 I, 99 Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R=1.0 to 3.0) requirements

  16. Progress of nuclear fusion research and review on development of fusion reactors

    International Nuclear Information System (INIS)

    1976-01-01

    Set up in October 1971, the ad hoc Committee on Survey of Nuclear Fusion Reactors has worked on overall fusion reactor aspects and definition of the future problems under four working groups of core, nuclear heat, materials and system. The presect volume is intended to provide reference materials in the field of fusion reactor engineering, prepared by members of the committee. Contents are broadly the following: concept of the nuclear fusion reactor, fusion core engineering, fusion reactor blanket engineering, fusion reactor materials engineering, and system problems in development of fusion reactors. (Mori, K.)

  17. Diagnostic measurements related to laser driven inertial confinement fusion

    International Nuclear Information System (INIS)

    Campbell, D.E.

    1979-01-01

    Scientists at the Lawrence Livermore Laboratory have been conducting laser driven inertial confinement fusion experiments for over five years. The first proof of the thermonuclear burn came at the Janus target irradiation facility in the spring of 1975. Since that time three succeedingly higher energy facilities have been constructed at Livermore, Cyclops, Argus and Shiva, where increased fusion efficiency has been demonstrated. A new facility, called Nova, is now in the construction phase and we are hopeful that scientific break even (energy released compared to incident laser energy on target) will be demonstrated here in early 1980's. Projected progress of the Livermore program is shown

  18. Introduction to magnetic fusion reactor design

    International Nuclear Information System (INIS)

    Watanabe, Kenji

    1988-01-01

    Trend of the tokamak reactor design works so far carried out is reviewed, and method of conceptual design for commercial fusion reactor is critically considered concerning the black-box conpepts. System-framework of the engineering of magnetic fusion (commercial) reactor design is proposed as four steps. Based on it the next design studies are recommended in parallel approaches for making real-overcome of reactor material problem, from the view point of technological realization and not from the economical one. Real trials are involved. (author)

  19. Safety and environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    Kilic, H.; Jensen, B.

    1982-01-01

    This paper deals with those problems concerning safety and environmental aspects of the future fusion reactors (e.g. fuel cycle, magnetic failure, after heat disturbances, radioactive waste and magnetic field)

  20. Fuel cycle problems in fusion reactors

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1976-01-01

    Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known

  1. Transport of heavy ions in inertial confinement fusion

    International Nuclear Information System (INIS)

    Parvazian, A.; Shahbandari Gouchani, A.

    2007-01-01

    In this article we have investigated the interaction of heavy ions (U) with a target (Au). In inertial confinement fusion method Interaction between heavy ion beam and target was simulated, Numerical analysis of the Boltzmann Fokker Planck equation used in order to optimize the material of the target and Energy deposition of ion beam to electrons and ions of target and The thickness of the target were calculated.

  2. Health physics appraisal guidelines for fusion/confinement devices

    International Nuclear Information System (INIS)

    Neeson, P.M.

    1987-01-01

    Several types of fusion/confinement devices have been developed for a variety of research applications. The health physics considerations for these devices can vary, depending on a number of parameters. This paper presents guidelines for health physics appraisal of such devices, which can be tailored to apply to specific systems. The guidelines can also be useful for establishing ongoing health physics programs for safe operation of the devices

  3. The Swedish fusion research programme on magnetic confinement 1978

    International Nuclear Information System (INIS)

    Lehnert, B.

    1978-02-01

    A review is given on the activities and plans for research on plasma physics and controlled fusion at the Royal Institute of Technology in Stockholm, with descriptions and motivations of the research lines being conducted. These activities include investigations on plasma-neutral gas interaction, development of special principles for plasma stabilization, magnetic confinement schemes being based mainly on poloidal fields, as well as the generation, heating, and diagnostics of plasmas being ''impermeable'' to neutral gas. (author)

  4. Inertial confinement fusion systems using heavy ion accelerators as drivers

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.; Godlove, T.F.; Keefe, D.

    1980-03-01

    Heavy ion accelerators are the most recent entrants in the effort to identify a practical driver for inertial confinement fusion. They are of interest because of the expected efficient coupling of ion kinetic energy to the thermal energy needed to implode the pellet and because of the good electrical efficiency of high intensity particle accelerators. The beam intensities required, while formidable, lie within the range that can be studied by extensions of the theories and the technology of modern high energy accelerators

  5. Fusion reactor safety studies, FY 1977

    International Nuclear Information System (INIS)

    Darby, J.B. Jr.

    1978-04-01

    This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

  6. A comprehensive alpha-heating model for inertial confinement fusion

    Science.gov (United States)

    Christopherson, A. R.; Betti, R.; Bose, A.; Howard, J.; Woo, K. M.; Campbell, E. M.; Sanz, J.; Spears, B. K.

    2018-01-01

    A comprehensive model is developed to study alpha-heating in inertially confined plasmas. It describes the time evolution of a central low-density hot spot confined by a compressible shell, heated by fusion alphas, and cooled by radiation and thermal losses. The model includes the deceleration, stagnation, and burn phases of inertial confinement fusion implosions, and is valid for sub-ignited targets with ≤10 × amplification of the fusion yield from alpha-heating. The results of radiation-hydrodynamic simulations are used to derive realistic initial conditions and dimensionless parameters for the model. It is found that most of the alpha energy (˜90%) produced before bang time is deposited within the hot spot mass, while a small fraction (˜10%) drives mass ablation off the inner shell surface and its energy is recycled back into the hot spot. Of the bremsstrahlung radiation emission, ˜40% is deposited in the hot spot, ˜40% is recycled back in the hot spot by ablation off the shell, and ˜20% leaves the hot spot. We show here that the hot spot, shocked shell, and outer shell trajectories from this analytical model are in good agreement with simulations. A detailed discussion of the effect of alpha-heating on the hydrodynamics is also presented.

  7. High temperature fusion reactor design

    International Nuclear Information System (INIS)

    Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

    1979-01-01

    Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

  8. Trends in fusion reactor safety research

    International Nuclear Information System (INIS)

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

    1991-01-01

    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. 18 refs

  9. FELIX experiments and computational needs for eddy current analysis of fusion reactors

    International Nuclear Information System (INIS)

    Turner, L.R.

    1984-01-01

    In a fusion reactor, changing magnetic fields are closely coupled to the electrically-conducting metal structure. This coupling is particularly pronounced in a tokamak reactor in which magnetic fields are used to confine, stabilize, drive, and heat the plasma. Electromagnetic effects in future fusion reactors will have far-reaching implications in the configuration, operation, and maintenance of the reactors. This paper describes the impact of eddy-current effects on future reactors, the requirements of computer codes for analyzing those effects, and the FELIX experiments which will provide needed data for code validation

  10. Fusion reactor design studies: standard unit costs and cost scaling rules

    International Nuclear Information System (INIS)

    Schulte, S.C.; Bickford, W.E.; Willingham, C.E.; Ghose, S.K.; Walker, M.G.

    1979-09-01

    This report establishes standard unit costs and scaling rules for estimating costs of material, equipment, land, and labor components used in magnetic confinement fusion reactor plant construction and operation. Use of the standard unit costs and scaling rules will add uniformity to cost estimates, and thus allow valid comparison of the economic characteristics of various reactor concepts

  11. An experiment to test centrifugal confinement for fusion

    International Nuclear Information System (INIS)

    Ellis, R.F.; Hassam, A.B.; Messer, S.; Osborn, B.R.

    2001-01-01

    The basic idea of centrifugal confinement is to use centrifugal forces from supersonic rotation to augment conventional magnetic confinement. Optimizing this 'knob' results in a fusion device that features four advantages: steady state, no disruptions, superior cross-field confinement, and a simpler coil configuration. The idea rests on two prongs: first, centrifugal forces can confine plasmas to desired regions of shaped magnetic fields; second, the accompanying large velocity shear can stabilize even magnetohydrodynamic (MHD) instabilities. A third feature is that the velocity shear also viscously heats the plasma; no auxiliary heating is necessary to reach fusion temperatures. Regarding transport, the velocity shear can also quell microturbulence, leading to fully classical confinement, as there are no neoclassical effects. Classical parallel electron transport then sets the confinement time. These losses are minimized by a large Pastukhov factor resulting from the deep centrifugal potential well: at Mach 4-5, the Lawson criterion is accessible. One key issue is whether velocity shear will be sufficient by itself to stabilize MHD interchanges. Numerical simulations indicate that laminar equilibria can be obtained at Mach numbers of 4-5 but that the progression toward laminarity with increasing Mach number is accompanied by residual convection from the interchanges. The central goal of the Maryland Centrifugal Torus (MCT) [R. F. Ellis et al., Bull. Am. Phys. Soc. 44, 48 (1998)] is to obtain MHD stability from velocity shear. As an assist to accessing laminarity, MCT will incorporate two unique features: plasma elongation and toroidal magnetic field. The former raises velocity shear efficiency, and modest magnetic shear should suppress residual convection

  12. Inertial electrostatic confinement (IEC) fusion fundamentals and applications

    CERN Document Server

    Miley, George H

    2014-01-01

    This book provides readers with an introductory understanding of Inertial Electrostatic Confinement (IEC), a type of fusion meant to retain plasma using an electrostatic field. IEC provides a unique approach for plasma confinement, as it offers a number of spin-off applications, such as a small neutron source for Neutron Activity Analysis (NAA), that all work towards creating fusion power. The IEC has been identified in recent times as an ideal fusion power unit because of its ability to burn aneutronic fuels like p-B11 as a result of its non-Maxwellian plasma dominated by beam-like ions. This type of fusion also takes place in a simple mechanical structure small in size, which also contributes to its viability as a source of power. This book posits that the ability to study the physics of IEC in very small volume plasmas makes it possible to rapidly investigate a design to create a power-producing device on a much larger scale. Along with this hypothesis the book also includes a conceptual experiment propose...

  13. Tritium problems in fusion reactor systems

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1975-01-01

    A brief introduction is given to the role tritium will play in the development of fusion power. The biological and worldwide environmental behavior of tritium is reviewed. The tritium problems expected in fusion power reactors are outlined. A few thoughts on tritium permeation and recent results for tritium cleanup and CT 4 accumulation are presented. Problems involving the recovery of tritium from the breeding blanket in fusion power reactors are also considered, including the possible effect of impurities in lithium blankets and the use of lithium as a regenerable getter pump. (auth)

  14. Composites as structural materials in fusion reactors

    International Nuclear Information System (INIS)

    Megusar, J.

    1989-01-01

    In fusion reactors, materials are used under extreme conditions of temperature, stress, irradiation, and chemical environment. The absence of adequate materials will seriously impede the development of fusion reactors and might ultimately be one of the major difficulties. Some of the current materials problems can be solved by proper design features. For others, the solution will have to rely on materials development. A parallel and balanced effort between the research in plasma physics and fusion-related technology and in materials research is, therefore, the best strategy to ultimately achieve economic, safe, and environmentally acceptable fusion. The essential steps in developing composites for structural components of fusion reactors include optimization of mechanical properties followed by testing under fusion-reactor-relevant conditions. In optimizing the mechanical behavior of composite materials, a wealth of experience can be drawn from the research on ceramic matrix and metal matrix composite materials sponsored by the Department of Defense. The particular aspects of this research relevant to fusion materials development are methodology of the composite materials design and studies of new processing routes to develop composite materials with specific properties. Most notable examples are the synthesis of fibers, coatings, and ceramic materials in their final shapes form polymeric precursors and the infiltration of fibrous preforms by molten metals

  15. HYPERFUSE: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs or 90 Sr. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n, 2n), (n, α), etc.) that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  16. Hyper fuse: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1979-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with a target in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs or 90 Sr. The 14 MeV fusion neutrons released during the pellet burn cause transmutation reactions [e.g., (n, 2n), (n, α), etc.] that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  17. Studies of conceptual spheromak fusion reactors

    International Nuclear Information System (INIS)

    Katsurai, M.; Yamada, M.

    1982-01-01

    Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

  18. Direct-drive inertial confinement fusion: A review

    International Nuclear Information System (INIS)

    Craxton, R. S.; Anderson, K. S.; Boehly, T. R.; Goncharov, V. N.; Harding, D. R.; Knauer, J. P.; McKenty, P. W.; Myatt, J. F.; Short, R. W.; Skupsky, S.; Theobald, W.; Collins, T. J. B.; Delettrez, J. A.; Hu, S. X.; Marozas, J. A.; Maximov, A. V.; Michel, D. T.; Radha, P. B.; Regan, S. P.; Sangster, T. C.

    2015-01-01

    The direct-drive, laser-based approach to inertial confinement fusion (ICF) is reviewed from its inception following the demonstration of the first laser to its implementation on the present generation of high-power lasers. The review focuses on the evolution of scientific understanding gained from target-physics experiments in many areas, identifying problems that were demonstrated and the solutions implemented. The review starts with the basic understanding of laser–plasma interactions that was obtained before the declassification of laser-induced compression in the early 1970s and continues with the compression experiments using infrared lasers in the late 1970s that produced thermonuclear neutrons. The problem of suprathermal electrons and the target preheat that they caused, associated with the infrared laser wavelength, led to lasers being built after 1980 to operate at shorter wavelengths, especially 0.35 μm—the third harmonic of the Nd:glass laser—and 0.248 μm (the KrF gas laser). The main physics areas relevant to direct drive are reviewed. The primary absorption mechanism at short wavelengths is classical inverse bremsstrahlung. Nonuniformities imprinted on the target by laser irradiation have been addressed by the development of a number of beam-smoothing techniques and imprint-mitigation strategies. The effects of hydrodynamic instabilities are mitigated by a combination of imprint reduction and target designs that minimize the instability growth rates. Several coronal plasma physics processes are reviewed. The two-plasmon–decay instability, stimulated Brillouin scattering (together with cross-beam energy transfer), and (possibly) stimulated Raman scattering are identified as potential concerns, placing constraints on the laser intensities used in target designs, while other processes (self-focusing and filamentation, the parametric decay instability, and magnetic fields), once considered important, are now of lesser concern for mainline

  19. Direct-drive inertial confinement fusion: A review

    Energy Technology Data Exchange (ETDEWEB)

    Craxton, R. S.; Anderson, K. S.; Boehly, T. R.; Goncharov, V. N.; Harding, D. R.; Knauer, J. P.; McKenty, P. W.; Myatt, J. F.; Short, R. W.; Skupsky, S.; Theobald, W.; Collins, T. J. B.; Delettrez, J. A.; Hu, S. X.; Marozas, J. A.; Maximov, A. V.; Michel, D. T.; Radha, P. B.; Regan, S. P.; Sangster, T. C. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623-1299 (United States); and others

    2015-11-15

    The direct-drive, laser-based approach to inertial confinement fusion (ICF) is reviewed from its inception following the demonstration of the first laser to its implementation on the present generation of high-power lasers. The review focuses on the evolution of scientific understanding gained from target-physics experiments in many areas, identifying problems that were demonstrated and the solutions implemented. The review starts with the basic understanding of laser–plasma interactions that was obtained before the declassification of laser-induced compression in the early 1970s and continues with the compression experiments using infrared lasers in the late 1970s that produced thermonuclear neutrons. The problem of suprathermal electrons and the target preheat that they caused, associated with the infrared laser wavelength, led to lasers being built after 1980 to operate at shorter wavelengths, especially 0.35 μm—the third harmonic of the Nd:glass laser—and 0.248 μm (the KrF gas laser). The main physics areas relevant to direct drive are reviewed. The primary absorption mechanism at short wavelengths is classical inverse bremsstrahlung. Nonuniformities imprinted on the target by laser irradiation have been addressed by the development of a number of beam-smoothing techniques and imprint-mitigation strategies. The effects of hydrodynamic instabilities are mitigated by a combination of imprint reduction and target designs that minimize the instability growth rates. Several coronal plasma physics processes are reviewed. The two-plasmon–decay instability, stimulated Brillouin scattering (together with cross-beam energy transfer), and (possibly) stimulated Raman scattering are identified as potential concerns, placing constraints on the laser intensities used in target designs, while other processes (self-focusing and filamentation, the parametric decay instability, and magnetic fields), once considered important, are now of lesser concern for mainline

  20. Breeder control fusion reactor. Topical interview

    Energy Technology Data Exchange (ETDEWEB)

    Schlueter, A [Max-Planck-Institut fuer Plasmaphysik, Garching/Muenchen (Germany, F.R.)

    1977-09-01

    The energy sources of the future are extremely controversial. The consumption of fossil fuel shall decrease during the next decades, because exhaustion of the resources, pollution, increase of CO/sub 2/ in the atmosphere and other reasons. But at present the question it not yet settled which alternative energy system should replace the fossil fuel. First of all nuclear energy in the form of fission reactions seems to come into operation to a larger extent. The next step may be the controlled thermonuclear fusion reaction. Furthermore, a comparison between fusion and fission is given which shows that fusion would bring about less risks than the breeders. An advantage of the fusion reactor would be the fact that the fuel cycle is closed. Unfortunately, the physical questions are not as yet satisfactorily clarified so that one cannot be sure whether a fusion reactor can really be built.

  1. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

    The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

  2. A look at the fusion reactor technology

    International Nuclear Information System (INIS)

    Rohatgi, V.K.

    1985-01-01

    The prospects of fusion energy have been summarised in this paper. The rapid progress in the field in recent years can be attributed to the advances in various technologies. The commercial fusion energy depends more heavily on the evolution and improvement in these technologies. With better understanding of plasma physics, the fusion reactor designs have become more realistic and comprehensive. It is now possible to make intercomparison between various concepts within the frame work of the established technologies. Assuming certain growth rate of the technological development, it is estimated that fusion energy can become available during the early part of the next century. (author)

  3. Extrap conceptual fusion reactor design study

    International Nuclear Information System (INIS)

    Eninger, J.E; Lehnert, B.

    1987-12-01

    A study has recently been initiated to asses the fusion reactor potential of the Extrap concept. A reactor model is defined that fulfills certain economic and environmental criteria. This model is applied to Extrap and a reference reactor is outlined. The design is optimized by varying parameters subject to both physics and engineering constraints. Several design options are examined and key engineering issues are identified and addressed. Some preliminary results and conclusions of this work are summarized. (authors)

  4. Space Propulsion via Spherical Torus Fusion Reactor

    International Nuclear Information System (INIS)

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

    2003-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 204 days, with an initial mass in low Earth orbit of 1630 mt. Engineering conceptual design, analysis, and assessment were performed on all major systems including nuclear fusion reactor, magnetic nozzle, power conversion, fast wave plasma heating, fuel pellet injector, startup/re-start fission reactor and battery, and other systems. Detailed fusion reactor design included analysis of plasma characteristics, power balance and utilization, first wall, toroidal field coils, heat transfer, and neutron/X-ray radiation

  5. Direct energy conversion of radiation energy in fusion reactor

    International Nuclear Information System (INIS)

    Yamaguchi, S.; Iiyoshi, A.; Motojima, O.; Okamoto, M.; Sudo, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    Direct energy conversion from plasma heat flux has been studied. Since major parts of fusion energy in the advanced fusion reactor are radiation and charged particle energies, the flexible design of the blanket is possible. We discuss the potentiality of the thermoelectric element that generates electricity by temperature gradient in conductors. A strong magnetic field is used to confine the fusion plasma, therefore, it is appropriate to consider the effect of the magnetic field. We propose a new element which is called Nernst element. The new element needs the magnetic field and the temperature gradient. We compare the efficiency of these two elements in a semiconductor model. Finally, a direct energy conversion are mentioned. (author)

  6. Direct energy conversion of radiation energy in fusion reactor

    Science.gov (United States)

    Yamaguchi, S.; Iiyoshi, A.; Motojima, O.; Okamoto, M.; Sudo, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    Direct energy conversion from plasma heat flux has been studied. Since major parts of fusion energy in the advanced fusion reactor are radiation and charged particle energies, the flexible design of the blanket is possible. We discuss the potentiality of the thermoelectric element that generates electricity by temperature gradient in conductors. A strong magnetic field is used to confine the fusion plasma, therefore, it is appropriate to consider the effect of the magnetic field. We propose a new element which is called Nernst element. The new element needs the magnetic field and the temperature gradient. We compare the efficiency of these two elements in a semiconductor model. Finally, a direct energy conversion are mentioned.

  7. Direct energy conversion of radiation energy in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, S.; Iiyoshi, A.; Motojima, O.; Okamoto, M.; Sudo, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-11-01

    Direct energy conversion from plasma heat flux has been studied. Since major parts of fusion energy in the advanced fusion reactor are radiation and charged particle energies, the flexible design of the blanket is possible. We discuss the potentiality of the thermoelectric element that generates electricity by temperature gradient in conductors. A strong magnetic field is used to confine the fusion plasma, therefore, it is appropriate to consider the effect of the magnetic field. We propose a new element which is called Nernst element. The new element needs the magnetic field and the temperature gradient. We compare the efficiency of these two elements in a semiconductor model. Finally, a direct energy conversion are mentioned. (author).

  8. Direct energy conversion of radiation energy in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, S.; Iiyoshi, A.; Motojima, O.; Okamoto, M.; Sudo, S. [National Inst. for Fusion Science, Nagoya (Japan); Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1994-12-31

    Direct energy conversion from plasma heat flux has been studied. Since major parts of fusion energy in the advanced fusion reactor are radiation and charged particle energies, the flexible design of the blanket is possible. We discuss the potentiality of the thermoelectric element that generate electricity by temperature gradient in conductors. A Strong magnetic field is used to confine the fusion plasma, therefore, it is appropriate to consider the effect of the magnetic field. We propose a new element which is called Nernst element. The new element needs the magnetic field and the temperature gradient. We compare the efficiency of these two elements in a semiconductor model. Finally, a direct energy converter are mentioned. (author).

  9. Direct energy conversion of radiation energy in fusion reactor

    International Nuclear Information System (INIS)

    Yamaguchi, S.; Iiyoshi, A.; Motojima, O.; Okamoto, M.; Sudo, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1994-01-01

    Direct energy conversion from plasma heat flux has been studied. Since major parts of fusion energy in the advanced fusion reactor are radiation and charged particle energies, the flexible design of the blanket is possible. We discuss the potentiality of the thermoelectric element that generate electricity by temperature gradient in conductors. A Strong magnetic field is used to confine the fusion plasma, therefore, it is appropriate to consider the effect of the magnetic field. We propose a new element which is called Nernst element. The new element needs the magnetic field and the temperature gradient. We compare the efficiency of these two elements in a semiconductor model. Finally, a direct energy converter are mentioned. (author)

  10. The role of alpha particles in magnetically confined fusion plasmas

    International Nuclear Information System (INIS)

    Lisak, M.; Wilhelmsson, H.

    1986-01-01

    Recent progress in the confinement of hot plasmas in magnetic fusion experiments throughout the world has intensified interest and research in the physics of D-T burning plasmas especially in the wide range of unresolved theoretical as well as experimental questions associated with the role of alpha particles in such devices. In order to review the state-of-the- art in this field, and to identify new issues and problems for further research, the Symposium on the Role of Alpha Particles in Magnetically Confined Fusion Plasmas was held from 24 to 26 June 1986 at Aspenaesgaarden near Goeteborg, Sweden. About 25 leading experts from nine countries attended the Symposium and gave invited talks. The major part of the programme was devoted to alpha-particle effects in tokamaks but some aspects of open systems were also discussed. The possibilities of obtaining ignition in JET and TFTR as well as physics issues for the compact ignition experiments were considered in particular. A special session was devoted to the diagnostics of alpha particles and other fusion products. In this report are summarised some of the highlights of the symposium. (authors)

  11. New materials in nuclear fusion reactors

    International Nuclear Information System (INIS)

    Iwata, Shuichi

    1988-01-01

    In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

  12. Response of a lithium fall to an inertially confined fusion microexplosion

    International Nuclear Information System (INIS)

    Hovingh, J.; Blink, J.; Glenn, L.

    1978-01-01

    One of the most difficult technology problems in an inertially confined fusion reactor is the survival of the structure from the repeated stresses caused by the microexplosion products. To mitigate the damage from the microexplosion products, a thick lithium fall can be circulated in front of the structure. This fall will absorb the short-ranged products and moderate and attenuate the neutrons. This paper discusses the response of the fall to the microexplosion products, and estimates the resulting loading and stresses in the first structural wall

  13. Inertial confinement fusion systems using heavy ion accelerators as drivers

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.; Godlove, T.F.; Keefe, D.

    1980-01-01

    Heavy ion accelerators are the most recent entrants in the effort to identify a practical driver for inertial confinement fusion. They are of interest because of the expected efficient coupling of ion kinetic energy to the thermal energy needed to implode the pellet and because of the good electrical efficiency of high intensity particle accelerators. The beam intensities required, while formidable, lie within the range that can be studied by extensions of the theories and the technology of modern high energy accelerators. (orig.) [de

  14. Open-ended magnetic confinement systems for fusion

    International Nuclear Information System (INIS)

    Post, R.F.; Ryutov, D.D.

    1995-05-01

    Magnetic confinement systems that use externally generated magnetic fields can be divided topologically into two classes: ''closed'' and 'open''. The tokamak, the stellarator, and the reversed-field-pinch approaches are representatives of the first category, while mirror-based systems and their variants are of the second category. While the recent thrust of magnetic fusion research, with its emphasis on the tokamak, has been concentrated on closed geometry, there are significant reasons for the continued pursuit of research into open-ended systems. The paper discusses these reasons, reviews the history and the present status of open-ended systems, and suggests some future directions for the research

  15. Conceptual design studies of experimental and demonstration fusion reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

  16. Impact of confinement physics on reactor design and economics

    International Nuclear Information System (INIS)

    DeFreece, D.A.; Campbell, R.B.; Waganer, L.M.

    1977-01-01

    A variety of confinement laws were employed in a transient, zero dimensional plasma code, which was coupled to the TOCOMO systems code. The purpose was to determine the impact of the confinement laws on reactor design, power costs and changes in the utility interface. A satisfactory reactor and power plant has been defined for the large majority of combinations of confinement law, power plant size and plasma shape. Trapped ion mode (TIM) has been the easiest to work with, since the plasma is thermally stable with a good power density and minimal alpha particle build up. Neoclassical and pseudoclassical along with TEMII result in satisfactory reactor performance, but require active feedback control (by injecting impurities) to prevent plasma temperature excursions. These laws also require some form and degree of confinement time spoiling to allow long burn times, otherwise, alpha particles build up to an unacceptable level. TEM I results in thermal equilibrium at 5 keV and must be driven to provide a reactor quality plasma. The continuous injected power required for a 4300 MW thermal reactor is 540 MW. This added to the other circulating loads results in a net power output of 600 MWe at a very high relative cost. Daughney (empirical) confinement results in a satisfactory, competitive reactor

  17. Mirror hybrid (fusion--fission) reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Neef, W.S.; Devoto, R.S.; Galloway, T.R.; Fink, J.H.; Schultz, K.R.; Culver, D.; Rao, S.

    1977-10-01

    The reference mirror hybrid reactor design performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. The resulting reactor may thus be viewed as a comparatively near-term goal of the fusion program, and we project improved performance for the hybrid in the future as more advanced technology becomes available

  18. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-02-01

    This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

  19. Fusion reactor remote maintenance study. Final report

    International Nuclear Information System (INIS)

    Sniderman, M.

    1979-04-01

    An analysis of a major maintenance operation, the remote replacement of a modular sector of a tokamak reactor, was performed in substantial detail. Specific assumptions were developed which included concepts from various existing designs so that the operation which was studied includes some design features generic to any fusion reactor design. Based on the work performed in this study, the principal conclusions are: (1) It appears feasible to design a tokamak fusion reactor plant with availability comparable to existing fossil and fission plants, but this will require diligence and comprehensive planning during the complete design phase. (2) Since the total fusion program is paced by the success of each device, maintenance considerations must be incorporated into each device during design, even if the device is an experimental unit. (3) Innovative approaches, such as automatic computer controlled operations, should be developed so that large step reductions in planned maintenance times can be achieved

  20. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-09-01

    New options for the development of the nuclear energy economy which might become available by a successful development of fusion-breeders or fusion-fission hybrid power reactors, identified and their nonproliferative attributes are discussed. The more promising proliferation-resistance ettributes identified include: (1) Justification for a significant delay in the initiation of fuel processing, (2) Denaturing the plutonium with 238 Pu before its use in power reactors of any kind, and (3) Making practical the development of denatured uranium fuel cycles and, in particular, denaturing the uranium with 232 U. Fuel resource utilization, time-table and economic considerations associated with the use of fusion-breeders are also discussed. It is concluded that hybrid reactors may enable developing a nuclear energy economy which is more proliferation resistant than possible otherwise, whileat the same time, assuring high utilization of t he uranium and thorium resources in an economically acceptable way. (author)

  1. Remote assembly and maintenance of fusion reactors

    International Nuclear Information System (INIS)

    Becquet, M.C.; Farfaletti-Casali, F.

    1991-01-01

    This paper intend to present the state of the art in the field of remote assembly and maintenance, including system analysis design and operation for controlled fusion device such as JET, and the next NET and ITER reactors. The operational constraints of fusion reactors with respect to temperature, radiations dose rates and cumulated doses are considered with the resulting design requirements. Concepts like articulated boom, in-vessel vehicle and blanket handling device are presented. The close relations between computer simulations and experimental validation of those concepts are emphasized to ensure reliability of the operational behavior. Mockups and prototypes in reduced and full scale, as operating machines are described to illustrate the progress in remote operations for fusion reactors. The developments achieved at the Institute for System Engineering and Informatics of the Joint Research Center, in the field of remote blanket maintenance, reliability assessment of RH systems and remote cut and welding of lips joints are considered. (author)

  2. Nuclear data needs for fusion reactors

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-01-01

    The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

  3. Nuclear data for fusion reactor technology

    International Nuclear Information System (INIS)

    1988-06-01

    The meeting was organized in four sessions and four working groups devoted to the following topics: Requirements of nuclear data for fusion reactor technology (6 papers); Status of experimental and theoretical investigations of microscopic nuclear data (10 papers); Status of existing libraries for fusion neutronic calculations (5 papers); and Status of integral experiments and benchmark tests (6 papers). A separate abstract was prepared for each of these papers

  4. Health physics in fusion reactor design

    International Nuclear Information System (INIS)

    Wong, K.Y.; Dinner, P.J.

    1984-06-01

    Experience in the control of tritium exposures to workers and the public gained through the design and operation of Ontario Hydro's nuclear stations has been applied to fusion projects and to design studies on emerging fusion reactor concepts. Ontario Hydro performance in occupational tritium exposure control and environmental impact is reviewed. Application of tritium control technologies and dose management methodology during facility design is highlighted

  5. TORFA - toroidal reactor for fusion applications

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1980-09-01

    The near-term goal of the US controlled fusion program should be the development, for practical applications, of an intense, quasi-steady, reliable 14-MeV neutron source with an electrical utilization efficiency at least 10 times larger than the value characterizing beam/solid-target neutron generators. This report outlines a method for implementing that goal, based on tokamak fusion reactors featuring resistive toroidal-field coils designed for ease of demountability

  6. Compact approach to fusion power reactors

    International Nuclear Information System (INIS)

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

    1984-01-01

    The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

  7. First wall costs of an ion-beam fusion reactor

    International Nuclear Information System (INIS)

    Hovingh, J.

    1977-08-01

    This paper parametrically investigates the effects of microexplosion energy on the first wall costs of a 4000 MW/sub t/ ion-beam initiated, inertially confined fusion reactor for several first wall materials. The thermodynamic models and the results for microexplosion energies between 400 and 4000 MJ are presented. A solid stainless steel or a composite isotropic graphite over stainless steel first wall can operate for a year at a cost of 0.6 mills per kWh gross electric power output

  8. Transients and burn dynamics in advanced tokamak fusion reactors

    International Nuclear Information System (INIS)

    Mantsinen, M.J.; Salomaa, R.R.E.

    1994-01-01

    Transient behavior of D 3 He-tokamak reactors is investigated numerically using a zero-dimensional code with prescribed profiles. Pure D 3 He start-up is compared to DT-assisted and DT-ignited start-ups. We have considered two categories of transients which could extinguish steady fusion burn: fuelling interruptions and sudden confinement changes similar to the L → H transients occurring in present-day tokamaks. Shutdown with various current and density ramp-down scenarios are studied, too. (author)

  9. Prospects for improved fusion reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

    Ideally, a new energy source must be capable of displacing old energy sources while providing both economic opportunities and enhanced environmental benefits. The attraction of an essentially unlimited fuel supply has generated a strong impetus to develop advanced fission breeders and, even more strongly, the exploitation of nuclear fusion. Both fission and fusion systems trade a reduced fuel charge for a more capital-intensive plant needed to utilize a cheaper and more abundant fuel. Results from early conceptual designs of fusion power plants, however, indicated a capital intensiveness that could override cost savings promised by an inexpensive fuel cycle. Early warnings of these problems appeared, and generalized routes to more economically attractive systems have been suggested; specific examples have also recently been given. Although a direct reduction in the cost (and mass) of the fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils, and primary structure) most directly reduces the overall cost of fusion power, with the mass power density (MPD, ratio of net electric power to FPC mass, kWe/tonne) being suggested as a figure-of-merit in this respect, other technical, safety/environmental, and institutional issues also enter into the definition of and direction for improved fusion concepts. These latter issues and related tradeoffs are discussed

  10. Fuel pellets and optical systems for inertially confined fusion

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1979-01-01

    Current laser-driven ICF targets are complex sets of concentric spherical shells made from a variety of materials including the fuel (e.g., deuterium-tritium), glass, beryllium, gold, polymeric materials, organo-metallics, and several additional organic and inorganic materials depending on the particular experiments to be done. While it is not yet known what the reactor targets will be exactly, there is little reason to believe they will be just simple, low quality glass shells containing DT gas or simple spheres of deuterated polyethylene or other fuel. Consequently, many of the current targets, materials, and fabrication techniques are considered to be applicable to the long range problems of ICF reactor target fabrication. Many current material problems and fabrication techniques are discussed and various quality factors are presented in an attempt to bring an awareness of the possible fusion reactor target materials problems to the scientific and technical community

  11. Light ion driven inertial fusion reactor concepts

    International Nuclear Information System (INIS)

    Cook, D.L.; Sweeney, M.A.; Buttram, M.T.; Prestwich, K.R.; Moses, G.A.; peterson, R.R.; Lovell, E.G.; Englestad, R.L.

    1980-01-01

    The possibility of designing fusion reactor systems using intense beams of light ions has been investigated. concepts for beam production, transport, and focusing on target have been analyzed in light of more conservative target performance estimates. Analyses of the major criteria which govern the design of the beam-target-cavity tried indicate the feasibility of designing power systems at the few hundred megawatt (electric) level. This paper discusses light ion fusion reactor (LIFR) concepts and presents an assessment of the design limitations through quantitative examples

  12. Progress of electromagnetic analysis for fusion reactors

    International Nuclear Information System (INIS)

    Takagi, T.; Ruatto, P.; Boccaccini, L.V.

    1998-01-01

    This paper describes the recent progress of electromagnetic analysis research for fusion reactors including methods, codes, verification tests and some applications. Due to the necessity of the research effort for the structural design of large tokamak devices since the 1970's with the help of the introduction of new numerical methods and the advancement of computer technologies, three-dimensional analysis methods have become as practical as shell approximation methods. The electromagnetic analysis is now applied to the structural design of new fusion reactors. Some more modeling and verification tests are necessary when the codes are applied to new materials with nonlinear material properties. (orig.)

  13. A Parallel Boltzmann Simulation for Multi-grid Inertial Electrostatic Confinement Fusion

    Data.gov (United States)

    National Aeronautics and Space Administration — Inertial electrostatic confinement (IEC) is a means of confining a non-neutral, non-Maxwellian plasma with an electric field, with the goal of creating fusion...

  14. Utilization of a Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research. Report of a Coordinated Research Project 2011–2016

    International Nuclear Information System (INIS)

    2016-12-01

    The IAEA actively promotes the development of controlled fusion as a source of energy. Through its coordinated research activities, the IAEA helps Member States to exchange and establish scientific and technical knowledge required for the design, construction and operation of a fusion reactor. Due to their compactness, flexibility and low operation costs, small fusion devices are a great resource for supporting and accelerating the development of mainstream fusion research on large fusion devices such as the International Thermonuclear Experimental Reactor. They play an important role in investigating the physics of controlled fusion, developing innovative technologies and diagnostics, testing new materials, training highly qualified personnel for larger fusion facilities, and supporting educational programmes for young scientists. This publication reports on the research work accomplished within the framework of the Coordinated Research Project (CRP) on Utilization of the Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research, organized and conducted by the IAEA in 2011–2016. The CRP has contributed to the coordination of a network of research institutions, thereby enhancing international collaboration through scientific visits, joint experiments and the exchange of information and equipment. A total of 16 institutions and 14 devices from 13 Member States participated in this CRP (Belgium, Bulgaria, Canada, China, Costa Rica, the Czech Republic, the Islamic Republic of Iran, Kazakhstan, Pakistan, Portugal, the Russian Federation, Ukraine and the United Kingdom).

  15. Steady-state operation requirements of tokamak fusion reactor concepts

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1991-06-01

    In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

  16. Present status of fusion reactor materials, 4

    International Nuclear Information System (INIS)

    Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

    1982-01-01

    Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

  17. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

  18. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohnishi, Masami, E-mail: onishi@kansai-u.ac.jp [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Yamamoto, Yasushi; Osawa, Hodaka [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Hatano, Yuji; Torikai, Yuji [Hydrogen Isotope Science Center, University of Toyama, Gofuku, Toyama 930-8555 (Japan); Murata, Isao [Faculty of Engineering Environment and Energy Department, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Masuda, Kai [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hotta, Eiki [Interdisciplinary Graduate School of Science and Engineering, Tokyo Institute of Technology, 4259 Nagatsuda-cho, Midori-ku, Yokohama 226-8503 (Japan)

    2016-11-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10{sup 8}(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  19. Tritium burning in inertial electrostatic confinement fusion facility

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Osawa, Hodaka; Hatano, Yuji; Torikai, Yuji; Murata, Isao; Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki; Masuda, Kai; Hotta, Eiki

    2016-01-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10"8(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  20. The VISTA spacecraft: Advantages of ICF [Inertial Confinement Fusion] for interplanetary fusion propulsion applications

    International Nuclear Information System (INIS)

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

    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

  1. Occupational health physics at a fusion reactor

    International Nuclear Information System (INIS)

    Shank, K.E.; Easterly, C.E.; Shoup, R.L.

    1975-01-01

    Future generation of electrical power using controlled thermonuclear reactors will involve both traditional and new concerns for health protection. A review of the problems associated with exposures to tritium and magnetic fields is presented with emphasis on the occupational worker. The radiological aspects of tritium, inventories and loss rates of tritium for fusion reactors, and protection of the occupational worker are discussed. Magnetic fields in which workers may be exposed routinely and possible biological effects are also discussed

  2. Progress in the pulsed power Inertial Confinement Fusion program

    International Nuclear Information System (INIS)

    Quintenz, J.P.; Matzen, M.K.; Mehlhorn, T.A.

    1996-01-01

    Pulsed power accelerators are being used in Inertial Confinement Fusion (ICF) research. In order to achieve our goal of a fusion yield in the range of 200 - 1000 MJ from radiation-driven fusion capsules, it is generally believed that ∼10 MJ of driver energy must be deposited within the ICF target in order to deposit ∼1 MJ of radiation energy in the fusion capsule. Pulsed power represents an efficient technology for producing both these energies and these radiation environments in the required short pulses (few tens of ns). Two possible approaches are being developed to utilize pulsed power accelerators in this effort: intense beams of light ions and z- pinches. This paper describes recent progress in both approaches. Over the past several years, experiments have successfully answered many questions critical to ion target design. Increasing the ion beam power and intensity are our next objectives. Last year, the Particle Beam Fusion Accelerator H (PBFA II) was modified to generate ion beams in a geometry that will be required for high yield applications. This 2048 modification has resulted in the production of the highest power ion beam to be accelerated from an extraction ion diode. We are also evaluating fast magnetically-driven implosions (z-pinches) as platforms for ICF ablator physics and EOS experiments. Z-pinch implosions driven by the 20 TW Saturn accelerator have efficiently produced high x- ray power (> 75 TW) and energy (> 400 kJ). Containing these x-ray sources within a hohlraum produces a unique large volume (> 6000 mm 3 ), long lived (>20 ns) radiation environment. In addition to studying fundamental ICF capsule physics, there are several concepts for driving ICF capsules with these x-ray sources. Progress in increasing the x-ray power on the Saturn accelerator and promise of further increases on the higher power PBFA II accelerator will be described

  3. Ignition and Inertial Confinement Fusion at The National Ignition Facility

    International Nuclear Information System (INIS)

    Moses, E.

    2009-01-01

    The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and for studying high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). The NIF is now conducting experiments to commission the laser drive, the hohlraum and the capsule and to develop the infrastructure needed to begin the first ignition experiments in FY 2010. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. NIF will achieve this by concentrating the energy from the 192 beams into a mm 3 -sized target and igniting a deuterium-tritium mix, liberating more energy than is required to initiate the fusion reaction. NIF's ignition program is a national effort managed via the National Ignition Campaign (NIC). The NIC has two major goals: execution of DT ignition experiments starting in FY2010 with the goal of demonstrating ignition and a reliable, repeatable ignition platform by the conclusion of the NIC at the end of FY2012. The NIC will also develop the infrastructure and the processes required to operate NIF as a national user facility. The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on laser fusion as a viable energy option. A laser fusion-based energy concept that builds on NIF, known as LIFE (Laser Inertial Fusion Energy), is currently under development. LIFE is inherently safe and can provide a global carbon-free energy generation solution in the 21st century. This paper describes recent progress on NIF, NIC, and the LIFE concept.

  4. Tritium resources available for fusion reactors

    Science.gov (United States)

    Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

    2018-02-01

    The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

  5. Osiris and SOMBRERO inertial confinement fusion power plant designs

    International Nuclear Information System (INIS)

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

    1992-03-01

    The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs

  6. Non-LTE effects in inertial confinement fusion target chambers

    International Nuclear Information System (INIS)

    MacFarlane, J.J.; Moses, G.A.; Peterson, R.R.

    1989-01-01

    In previous studies of transport processes in inertial confinement fusion target chambers, the radiative properties of the background plasma were calculated under the assumption of local thermodynamic equilibrium (LTE). In this paper, the authors present a study of the equation of state and the radiative properties of high temperature, low-to-moderate density ( 21 cm -3 ) plasmas for the determination of the conditions under which non-LTE effects become important and for an assessment of the importance of non-LTE processes in target chambers during high yield inertial fusion target explosions. For this purpose, two-body (radiative and dielectronic) and three-body (collisional) recombination and de-excitation processes are considered in calculating the steady state ionization and excitation populations. The results of this study indicate that non-LTE processes generally become important at temperatures of > or approx. 1, 10 and 100 eV for plasma densities of 10 18 , 10 19 and 10 21 cm -3 , respectively. Radiation hydrodynamic simulations utilizing the equation of state and the opacities for a non-LTE argon plasma were performed to study the response of a background gas to an inertial fusion target explosion. These calculations indicate that non-LTE processes are often the dominant atomic processes in the background plasma and that they can strongly affect the radiative and shock properties as energy is transported away from the point of the target explosion. (author). 22 refs, 10 figs, 1 tab

  7. Control of Internal Transport Barriers in Magnetically Confined Fusion Plasmas

    Science.gov (United States)

    Panta, Soma; Newman, David; Sanchez, Raul; Terry, Paul

    2016-10-01

    In magnetic confinement fusion devices the best performance often involves some sort of transport barriers to reduce the energy and particle flow from core to edge. Those barriers create gradients in the temperature and density profiles. If gradients in the profiles are too steep that can lead to instabilities and the system collapses. Control of these barriers is therefore an important challenge for fusion devices (burning plasmas). In this work we focus on the dynamics of internal transport barriers. Using a simple 7 field transport model, extensively used for barrier dynamics and control studies, we explore the use of RF heating to control the local gradients and therefore the growth rates and shearing rates for barrier initiation and control in self-heated fusion plasmas. Ion channel barriers can be formed in self-heated plasmas with some NBI heating but electron channel barriers are very sensitive. They can be formed in self-heated plasmas with additional auxiliary heating i.e. NBI and radio-frequency(RF). Using RF heating on both electrons and ions at proper locations, electron channel barriers along with ion channel barriers can be formed and removed demonstrating a control technique. Investigating the role of pellet injection in controlling the barriers is our next goal. Work supported by DOE Grant DE-FG02-04ER54741.

  8. Plasma fluctuations and confinement of fusion reaction products

    International Nuclear Information System (INIS)

    Coppi, B.; Pegoraro, F.

    1981-01-01

    The interaction between the fluctuations that can be excited in a magnetically confined plasma and the high-energy-particle population produced by fusion reactions is analyzed in view of its relevance to the process of thermonuclear ignition. The spectrum of the perturbations that, in the absence of fusion reaction products, would be described by the incompressible ideal magnetohydrodynamic approximation is studied considering finite value of the plasma pressure relative ot the magnetic pressure. The combined effects of the magnetic field curvature and shear are taken into account and the relevant spectrum is shown to consist of a continuous portion, that could be identified as a mixture of shear-Alfven and interchange oscillations, and a discrete unstable part corresponding to the so-called ballooning modes. The rate of diffusion of the fusion reaction products induced by oscillations in the continuous part of the spectrum, as estimated from the appropriate quasi-linear theory, is found to be significantly smaller than could be expected if normal modes (i.e., nonconvective solutions) were excited. However, a relatively wide intermediate region is identified where opalescent fluctuations, capable of achieving significant amplitudes and corresponding to a quasi-discrete spectrum, can be excited

  9. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Gerstl, S.A.; Houck, D.L.; Jalbert, R.A.; Krakowski, R.A.; Linford, R.K.; McDonald, T.E.; Rogers, J.D.; Thomassen, K.I.

    1975-01-01

    A general design of the system is given. The implosion heating and compression systems (METS) are described. Tritium handling, shielding and activation of the reactor, and safety and environmental aspects are discussed

  10. Tritium breeding in fusion reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1982-10-01

    Key technological problems that influence tritium breeding in fusion blankets are reviewed. The breeding potential of candidate materials is evaluated and compared to the tritium breeding requirements. The sensitivity of tritium breeding to design and nuclear data parameters is reviewed. A framework for an integrated approach to improve tritium breeding prediction is discussed with emphasis on nuclear data requirements

  11. Progress on the conceptual design of a mirror hybrid fusion--fission reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1975-01-01

    A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

  12. Materials problems associated with fusion reactor technology

    International Nuclear Information System (INIS)

    Dutton, R.

    This paper outlines the principles of design and operation of conceptual fusion reactors, indicates the level of research funding and activity being proposed at major centres and reviews the major materials problems which have been identified, together with an outline of the experimental techniques which have been suggested for investigating these problems. (author)

  13. Physics considerations of the Reversed-Field Pinch fusion reactor

    International Nuclear Information System (INIS)

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

    1980-01-01

    A conceptual engineering design of a fusion reactor based on plasma confinement in a toroidal Reversed-Field Pinch (RFP) configuration is described. The plasma is ohmically ignited by toroidal plasma currents which also inherently provide the confining magnetic fields in a toroidal chamber having major and minor radii of 12.7 and 1.5 m, respectively. The DT plasma ignites in 2 to 3 s and undergoes a transient, unrefueled burn at 10 to 20 keV for approx. 20 s to give a DT burnup of approx. 50%. Accounting for all major energy sinks yields a cost-optimized system with a recirculating power fraction of 0.17; the power output is 750 MWe

  14. Waste management for JAERI fusion reactors

    International Nuclear Information System (INIS)

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

    2004-01-01

    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)

  15. Computer simulation of multi-elemental fusion reactor materials

    International Nuclear Information System (INIS)

    Voertler, K.

    2011-01-01

    Thermonuclear fusion is a sustainable energy solution, in which energy is produced using similar processes as in the sun. In this technology hydrogen isotopes are fused to gain energy and consequently to produce electricity. In a fusion reactor hydrogen isotopes are confined by magnetic fields as ionized gas, the plasma. Since the core plasma is millions of degrees hot, there are special needs for the plasma-facing materials. Moreover, in the plasma the fusion of hydrogen isotopes leads to the production of high energetic neutrons which sets demanding abilities for the structural materials of the reactor. This thesis investigates the irradiation response of materials to be used in future fusion reactors. Interactions of the plasma with the reactor wall leads to the removal of surface atoms, migration of them, and formation of co-deposited layers such as tungsten carbide. Sputtering of tungsten carbide and deuterium trapping in tungsten carbide was investigated in this thesis. As the second topic the primary interaction of the neutrons in the structural material steel was examined. As model materials for steel iron chromium and iron nickel were used. This study was performed theoretically by the means of computer simulations on the atomic level. In contrast to previous studies in the field, in which simulations were limited to pure elements, in this work more complex materials were used, i.e. they were multi-elemental including two or more atom species. The results of this thesis are in the microscale. One of the results is a catalogue of atom species, which were removed from tungsten carbide by the plasma. Another result is e.g. the atomic distributions of defects in iron chromium caused by the energetic neutrons. These microscopic results are used in data bases for multiscale modelling of fusion reactor materials, which has the aim to explain the macroscopic degradation in the materials. This thesis is therefore a relevant contribution to investigate the

  16. Studies on the feasibility of heavy-ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    1982-04-01

    This annual report summarizes the scientific results of work carried out in 1981 in the framework of a feasibility study for inertial confinement fusion (ICF) with heavy ion beams. This program, established in autumn 1979 and funded by the German Ministry for Science and Technology, is devoted in a first period until 1984 to the study of fundamental aspects of the field. Its principal aims are the investigation of key problems and the identification of critical issues of the heavy ion ICF concept in the fields of accelerator research, pellet physics, atomic physics, and reactor design. The research is carried out by about ten working groups at various German research centers and universities. In addition, together with a group of the University of Wisconsin a conceptual design study for a reactor plant (HIBALL) has been started in 1980 and was continued 1981. (orig.) [de

  17. Materials design data for fusion reactors

    International Nuclear Information System (INIS)

    Tavassoli, A.A.F.

    1998-01-01

    Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.)

  18. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  19. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

  20. Materials design data for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.A.F. [CEA Commissariat a l`Energie Atomique, Gif sur Yvette (France). CEREM

    1998-10-01

    Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.) 19 refs.

  1. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

    Purcell, J.R.; Chen, W.; Thomas, R.

    1976-01-01

    Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

  2. Tritium management for fusion reactors

    International Nuclear Information System (INIS)

    Rouyer, J.L.; Djerassi, H.

    1985-01-01

    To determine a waste management strategy, one has to identify first the wastes (quantities, activities, etc.), then to define options, and to compare these options by appropriate criteria and evaluations. Two European Associations are working together, i.e., Studsvik and CEA, on waste treatment and tritium problems. A contribution to fusion specific tritiated waste management strategy is presented. It is demonstrated that the best strategy is to retain tritium (outgas and recover, or immobilize it) so that residual tritium releases are kept to a minimum. For that, wastes are identified, actual regulations are described and judged inadequate without amendments for fusion problems. Appropriate criteria are defined. Options for treatment and disposal of tritiated wastes are proposed and evaluated. A tritium recovery solution is described

  3. Radiation environment of fusion experimental reactor

    International Nuclear Information System (INIS)

    Mori, Seiji; Seki, Yasushi

    1988-01-01

    Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

  4. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-01-01

    Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

  5. A conceptual fusion reactor based on the high-plasma-density Z-pinch

    International Nuclear Information System (INIS)

    Hartman, C.W.; Carlson, G.; Hoffman, M.; Werner, R.

    1977-01-01

    Conceptual DT and DD fusion reactors are discussed based on magnetic confinement with the high-plasma-density Z-pinch. The reactor concepts have no ''first wall'', the fusion neutrons and plasma energy being absorbed directly into a surrounding lithium vortex blanket. Efficient systems with low re-circulated power are projected, based on a flow-through pinch cycle for which overall Q values can approach 10. The conceptual reactors are characterized by simplicity, small minimum size (100MW(e)) and by the potential for minimal radioactivity hazards. (author)

  6. Inertial confinement fusion reaction chamber and power conversion system study. Final report

    International Nuclear Information System (INIS)

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

    1985-10-01

    This report summarizes the results of the second year of a two-year study on the design and evaluation of the Cascade concept as a commercial inertial confinement fusion (ICF) reactor. We developed a reactor design based on the Cascade reaction chamber concept that would be competitive in terms of both capital and operating costs, safe and environmentally acceptable in terms of hazard to the public, occupational exposure and radioactive waste production, and highly efficient. The Cascade reaction chamber is a double-cone-shaped rotating drum. The granulated solid blanket materials inside the rotating chamber are held against the walls by centrifugal force. The fusion energy is captured in a blanket of solid carbon, BeO, and LiAlO 2 granules. These granules are circulated to the primary side of a ceramic heat exchanger. Primary-side granule temperatures range from 1285 K at the LiAlO 2 granule heat exchanger outlet to 1600 K at the carbon granule heat exchanger inlet. The secondary side consists of a closed-cycle gas turbine power conversion system with helium working fluid, operating at 1300 K peak outlet temperature and achieving a thermal power conversion efficiency of 55%. The net plant efficiency is 49%. The reference design is a plant producing 1500 MW of D-T fusion power and delivering 815 MW of electrical power for sale to the utility grid. 88 refs., 44 figs., 47 tabs

  7. Polyvinyl alcohol coating of polystyrene inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Annamalai, P.; Lee, M.C.; Crawley, R.L.; Downs, R.L.

    1985-01-01

    An inertial confinement fusion (ICF) target made of polystyrene is first levitated in an acoustic field. The surface of the target is then etched using an appropriate solution (e.g., cyclohexane) to enhance the wetting characteristics. A specially prepared polyvinyl alcohol solution is atomized using an acoustic atomizer and deposited on the surface of the target. The solution is air dried to form a thin coating (2 μm) on the target (outside diameter approx.350--850 μm). Thicker coatings are obtained by repeated applications of the coating solution. Preliminary results indicate that uniform coatings may be achievable on the targets with a background surface smoothness in the order of 1000 A

  8. Tertiary proton diagnostics in future inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Cremer, S.; Verdon, C.P.; Petrasso, R.D.

    1998-01-01

    Recently, it was proposed to use energetic (up to 31 MeV) tertiary protons produced during the final stage of inertial confinement fusion implosions to measure the fuel areal density of compressed deuterium endash tritium (DT). The method is based on seeding the fuel with 3 He. The reaction of 3 He ions with the energetic knock-on deuterons, produced via the elastic scattering of 14.1 MeV neutrons, is a source of very energetic protons capable of escaping from very large areal density targets. This work presents results of detailed time-dependent Monte Carlo simulations of the nuclear processes involved in producing and transporting these protons through imploding targets proposed for direct-drive experiments on OMEGA [D. K. Bradley et al., Phys. Plasmas 5, 1870 (1998)] and the National Ignition Facility [S. W. Haan et al., Phys. Plasmas 2, 2480 (1995)]. copyright 1998 American Institute of Physics

  9. Fabrication and characterization of cryogenic targets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Rieger, H.; Kim, K.

    1979-08-01

    A new technique has been developed which is capable of fabricating uniform cryogenic targets for use in inertial confinement fusion. The essence of the technique is to directly wet a target with a cold helium gas jet, which results in freezing of the DT mixture contained in the target. A controlled amount of current is pulsed through a heater wire surrounding the target, giving rise to fast evaporation and refreezing of the DT-condensate into a uniform layer. Experiments, which have been performed with D 2 -filled glass microshells, successfully produce uniform layers of both liquid and solid D 2 inside the glass shells. A set of data illustrating the technique is presented and analyzed

  10. Passive Spectroscopic Diagnostics for Magnetically-confined Fusion Plasmas

    International Nuclear Information System (INIS)

    Stratton, B.C.; Bitter, M.; Hill, K.W.; Hillis, D.L.; Hogan, J.T.

    2007-01-01

    Spectroscopy of radiation emitted by impurities and hydrogen isotopes plays an important role in the study of magnetically-confined fusion plasmas, both in determining the effects of impurities on plasma behavior and in measurements of plasma parameters such as electron and ion temperatures and densities, particle transport, and particle influx rates. This paper reviews spectroscopic diagnostics of plasma radiation that are excited by collisional processes in the plasma, which are termed 'passive' spectroscopic diagnostics to distinguish them from 'active' spectroscopic diagnostics involving injected particle and laser beams. A brief overview of the ionization balance in hot plasmas and the relevant line and continuum radiation excitation mechanisms is given. Instrumentation in the soft X-ray, vacuum ultraviolet, ultraviolet, visible, and near-infrared regions of the spectrum is described and examples of measurements are given. Paths for further development of these measurements and issues for their implementation in a burning plasma environment are discussed.

  11. Passive Spectroscopic Diagnostics for Magnetically-confined Fusion Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Stratton, B. C.; Biter, M.; Hill, K. W.; Hillis, D. L.; Hogan, J. T.

    2007-07-18

    Spectroscopy of radiation emitted by impurities and hydrogen isotopes plays an important role in the study of magnetically-confined fusion plasmas, both in determining the effects of impurities on plasma behavior and in measurements of plasma parameters such as electron and ion temperatures and densities, particle transport, and particle influx rates. This paper reviews spectroscopic diagnostics of plasma radiation that are excited by collisional processes in the plasma, which are termed 'passive' spectroscopic diagnostics to distinguish them from 'active' spectroscopic diagnostics involving injected particle and laser beams. A brief overview of the ionization balance in hot plasmas and the relevant line and continuum radiation excitation mechanisms is given. Instrumentation in the soft X-ray, vacuum ultraviolet, ultraviolet, visible, and near-infrared regions of the spectrum is described and examples of measurements are given. Paths for further development of these measurements and issues for their implementation in a burning plasma environment are discussed.

  12. Free Electron Laser as Energy Driver for Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Saldin, E.L.; Shnejdmiller, E.A.; Ul'yanov, Yu.N.; Sarantsev, V.P.; Yurkov, M.V.

    1994-01-01

    A FEL based energy driver for Inertial Confinement Fusion (ICF) is proposed. The key element of the scheme is free electron laser system. Novel technical solutions reveal a possibility to construct the FEL system operating at radiation wavelength λ = 0.5 μm and providing flash energy E = 1 MJ and brightness 4 x 10 22 W cm -2 sr -1 within steering pulse duration 0.1-2 ns. Total energy efficiency of the proposed ICF energy driver is about of 11% and repetition rate is 40 Hz. Dimensions of such an ICF driver are comparable with those of heavy-ion ICF driver, while the problem of technical realization seems to be more realistic. It is shown that the FEL based ICF energy driver may be constructed at the present level of accelerator technique R and D. 27 refs., 10 figs., 3 tabs

  13. Application of inertial confinement fusion to weapon technology

    International Nuclear Information System (INIS)

    Toepfer, A.J.; Posey, L.D.

    1978-12-01

    This report reviews aspects of the military applications of the inertial confinement fusion (ICF) program at Sandia Laboratories. These applications exist in the areas of: (1) weapon physics research, and (2) weapon effects simulation. In the area of weapon physics research, ICF source technology can be used to study: (1) equations-of-state at high energy densities, (2) implosion dynamics, and (3) laboratory simulation of exoatmospheric burst phenomena. In the area of weapon effects simulation, ICF technology and facilities have direct near, intermediate, and far term applications. In the near term, short pulse x-ray simulation capabilities exist for electronic component effects testing. In the intermediate term, capabilities can be developed for high energy neutron exposures and bremsstrahlung x-ray exposures of components. In the far term, system level exposures of full reentry vehicles will be possible if sufficiently high pellet gains are achieved

  14. Computer modeling and simulation in inertial confinement fusion

    International Nuclear Information System (INIS)

    McCrory, R.L.; Verdon, C.P.

    1989-03-01

    The complex hydrodynamic and transport processes associated with the implosion of an inertial confinement fusion (ICF) pellet place considerable demands on numerical simulation programs. Processes associated with implosion can usually be described using relatively simple models, but their complex interplay requires that programs model most of the relevant physical phenomena accurately. Most hydrodynamic codes used in ICF incorporate a one-fluid, two-temperature model. Electrons and ions are assumed to flow as one fluid (no charge separation). Due to the relatively weak coupling between the ions and electrons, each species is treated separately in terms of its temperature. In this paper we describe some of the major components associated with an ICF hydrodynamics simulation code. To serve as an example we draw heavily on a two-dimensional Lagrangian hydrodynamic code (ORCHID) written at the University of Rochester's Laboratory for Laser Energetics. 46 refs., 19 figs., 1 tab

  15. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Linford, R.K.; Oliphant, T.A.; Thomassen, K.I.

    1976-01-01

    The SFTR is a proposed 80-m diameter D-T burning toroidal theta pinch. The system is designed to achieve Q = 1 where Q is the ratio of the total thermonuclear energy output to the maximum stored energy in the plasma. SFTR design studies [1] will provide valuable guidance to the Scyllac related research and to the needed technological development. The portion of the system directly related to the plasma confinement, stability, and heating, is described, and the approach used to obtain an operating point consistent with Q = 1, m = 1 stability, and technological limitations is outlined. (U.K.)

  16. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Linford, R.K.; Oliphant, T.A.; Thomassen, K.I.

    1975-01-01

    The SFTR is a proposed 80-m diameter D--T burning toroidal theta pinch. The system is designed to achieve Q = 1 where Q is the ratio of the total thermonuclear energy output to the maximum stored energy in the plasma. SFTR design studies will provide valuable guidance to the Scyllac related research and to the needed technological development. This paper describes the portion of the system directly related to the plasma confinement, stability, and heating, and outlines the approach used to obtain an operating point consistent with Q = 1, m = 1 stability, and technological limitations. (auth)

  17. Inertial Confinement Fusion Target Component Fabrication and Technology Development report

    International Nuclear Information System (INIS)

    Steinman, D.

    1994-03-01

    On December 30, 1990, the US Department of Energy entered into a contract with General Atomics (GA) to be the Inertial Confinement Fusion Target Component Fabrication and Technology Development Support contractor. This report documents the technical activities which took place under this contract during the period of October 1, 1992 through September 30, 1993. During this period, GA was assigned 18 tasks in support of the Inertial Confinement Fusion program and its laboratories. These tasks included ''Capabilities Activation'' and ''Capabilities Demonstration'' to enable us to begin production of glass and composite polymer capsules. Capsule delivery tasks included ''Small Glass Shell Deliveries'' and ''Composite Polymer Capsules'' for Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). We also were asked to provide direct ''Onsite Support'' at LLNL and LANL. We continued planning for the transfer of ''Micromachining Equipment from Rocky Flats'' and established ''Target Component Micromachining and Electroplating Facilities'' at GA. We fabricated over 1100 films and filters of 11 types for Sandia National Laboratory and provided full-time onsite engineering support for target fabrication and characterization. We initiated development of methods to make targets for the Naval Research Laboratory. We investigated spherical interferometry, built an automated capsule sorter, and developed an apparatus for calorimetric measurement of fuel fill for LLNL. We assisted LANL in the ''Characterization of Opaque b-Layered Targets.'' We developed deuterated and UV-opaque polymers for use by the University of Rochester's Laboratory for Laser Energetics (UR/LLE) and devised a triple-orifice droplet generator to demonstrate the controlled-mass nature of the microencapsulation process

  18. Tritium behavior in the Caisson, a simulated fusion reactor room

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Yamada, Masayuki; Suzuki, Takumi; O'hira, Shigeru; Nakamura, Hirofumi; Shu, Weimin; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Konishi, Satoshi; Nishi, Masataka

    2000-01-01

    In order to confirm tritium confinement ability in the deuterium-tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called 'Caisson', at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak-tight vessel of 12 m 3 , simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code

  19. Neutron personnel dosimetry considerations for fusion reactors

    International Nuclear Information System (INIS)

    Barton, T.P.; Easterly, C.E.

    1979-07-01

    The increasing development of fusion reactor technology warrants an evaluation of personnel neutron dosimetry systems to aid in the concurrent development of a radiation protection program. For this reason, current state of knowledge neutron dosimeters have been reviewed with emphasis placed on practical utilization and the problems inherent in each type of dosimetry system. Evaluations of salient parameters such as energy response, latent image instability, and minimum detectable dose equivalent are presented for nuclear emulsion films, track etch techniques, albedo and other thermoluminescent dosimetry techniques, electrical conductivity damage effects, lyoluminescence, thermocurrent, and thermally stimulated exoelectron emission. Brief summaries of dosimetry regulatory requirements and intercomparison study results help to establish compliance and recent trends, respectively. Spectrum modeling data generated by the Neutron Physics Division of Oak Ridge National Laboratory for the Princeton Tokamak Fusion Test Reactor (TFTR) Facility have been analyzed by both International Commission on Radiological Protection fluence to dose conversion factors and an adjoint technique of radiation dosimetry, in an attempt to determine the applicability of current neutron dosimetry systems to deuterium and tritium fusion reactor leakage spectra. Based on the modeling data, a wide range of neutron energies will probably be present in the leakage spectra of the TFTR facility, and no appreciable risk of somatic injury to occupationally exposed workers is expected. The relative dose contributions due to high energy and thermal neutrons indicate that neutron dosimetry will probably not be a serious limitation in the development of fusion power

  20. Control of a laser inertial confinement fusion-fission power plant

    Science.gov (United States)

    Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

    2015-10-27

    A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

  1. Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

  2. Tritium monitor for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jalbert, R.A.

    1982-08-01

    This report describes the design, operation, and performance of a flow-through ion-chamber instrument designed to measure tritium concentrations in air containing /sup 13/N, /sup 16/N, and /sup 41/Ar produced by neutrons generated by D-T fusion devices. The instrument employs a chamber assembly consisting of two coaxial ionization chambers. The inner chamber is the flow-through measuring chamber and the outer chamber is used for current subtraction. A thin wall common to both chambers is opaque to the tritium betas. Currents produced in the two chambers by higher energy radiation are automatically subtracted, leaving only the current due to tritium.

  3. Addressing Common Technical challenges in Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Haynes, Donald A.

    2016-01-01

    The implosion phase for Inertial Confinement Fusion (ICF) occurs from initiation of the drive until just before stagnation. Evolution of the shell and fusion fuel during the implosion phase is affected by the initial conditions of the target, the drive history. Poor performing implosions are a result of the behavior that occurs during the implosion phase such as low mode asymmetries, mixing of the ablator into the fuel, and the hydrodynamic evolution of initial target features and defects such as the shell mounting hardware. The ultimate results of these effects can only be measured at stagnation. However, studying the implosion phase can be effective for understanding and mitigating these effects and for of ultimately improving the performance of ICF implosions. As the ICF program moves towards the 2020 milestone to ''determine the efficacy of ignition'', it will be important to understand the physics that occurs during the implosion phase. This will require both focused and integrated experiments. Focused experiments will provide the understanding and the evidence needed to support any determination concerning the efficacy of ignition.

  4. Addressing Common Technical challenges in Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Haynes, Donald A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-22

    The implosion phase for Inertial Confinement Fusion (ICF) occurs from initiation of the drive until just before stagnation. Evolution of the shell and fusion fuel during the implosion phase is affected by the initial conditions of the target, the drive history. Poor performing implosions are a result of the behavior that occurs during the implosion phase such as low mode asymmetries, mixing of the ablator into the fuel, and the hydrodynamic evolution of initial target features and defects such as the shell mounting hardware. The ultimate results of these effects can only be measured at stagnation. However, studying the implosion phase can be effective for understanding and mitigating these effects and for of ultimately improving the performance of ICF implosions. As the ICF program moves towards the 2020 milestone to “determine the efficacy of ignition”, it will be important to understand the physics that occurs during the implosion phase. This will require both focused and integrated experiments. Focused experiments will provide the understanding and the evidence needed to support any determination concerning the efficacy of ignition.

  5. Conceptual design study on inertial confinement reactor ''SENRI-II''

    International Nuclear Information System (INIS)

    Nakamura, N.; Ouura, H.

    1983-01-01

    Design features of a laser fusion reactor concept SENRI-II are reviewed and discussed. A conceptual design study of the ICF reactor SENRI-II (an advanced design of SENRI-I) has been carried out over 2 years in the Research Committee of ICF Reactors, Institute of Laser Engineering, Osaka University. While the ICF reactor SENRI-I utilized a magnetic field to guide and control an inner liquid lithium flow, SENRI-II is designed to use porous metal as the liquid lithium flow guide. In the design of SENRI-II, a metal porous lithium blanket serves as the protection of a wall against fusion products and as wall per se. Because of the separation of these two functions, a high power density can be attained

  6. Fusion reactors for hydrogen production via electrolysis

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.R.; Steinberg, M.

    1979-01-01

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approx. 40 to 60% and hydrogen production efficiencies by high temperature electrolysis of approx. 50 to 70% are projected for fusion reactors using high temperature blankets

  7. ITER: the first experimental fusion reactor

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1995-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a multiphased project, at present proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the USA and the Government of Russia (''the parties''). The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world.The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test technologies for a demonstration fusion power plant. During the extended performance phase of ITER, it will demonstrate the characteristics of a fusion power plant, producing more than 1500MW of fusion power.The objective of the engineering design activity (EDA) phase is to produce a detailed, complete and fully integrated engineering design of ITER and all technical data necessary for the future decision on the construction of ITER.The ITER device will be a major step from present fusion experiments and will encompass all the major elements required for a fusion reactor. It will also require the development and the implementation of major new components and technologies.The inside surface of the plasma containment chamber will be designed to withstand temperature of up to 500 C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures, and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stage of development of testing. (orig.)

  8. Power exhaust by impurity seeding in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bernert, Matthias; Kallenbach, Arne; Dux, Ralph; Wischmeier, Marco [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, IEK, Juelich (Germany); Lipschultz, Bruce [University of York, York Plasma Institute, Heslington, York (United Kingdom); Collaboration: the ASDEX Upgrade team; the EUROfusion MST1 Team

    2016-07-01

    Power exhaust is one of the big challenges for future fusion reactors. The power load at the divertor targets, the primary plasma-wall interaction zone, would exceed material limits and, thus, must be reduced. Therefore, 90% of the exhaust power needs to be dissipated and the divertor is anticipated to be in the detached regime, where the interaction of the plasma with the wall is significantly reduced. Radiation is the dominant dissipation process and is increased by impurity seeding. The radiation distribution can be tailored by using different seed impurities (N for radiation outside, Ne and Ar for radiation at the edge of and Kr for radiation inside the confined region). The tailoring of the radiation profile is required in order to maximize the radiated power and at the same time minimize the impact on the energy confinement. Recent experiments with intense impurity seeding at the ASDEX Upgrade tokamak demonstrate operation at highest heat fluxes and detached divertor targets at radiated power fractions of up to 90%. In these scenarios the radiation originates predominantly from the confined region and leads to an unexpectedly small confinement reduction.

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

    International Nuclear Information System (INIS)

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

    1994-09-01

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

  10. Reactor potential of the magnetically insulated inertial fusion (MICF) system

    International Nuclear Information System (INIS)

    Kammash, T.; Galbraith, D.L.

    1987-01-01

    The Magnetically Insulated Inertial Confinement Fusion (MICF) scheme is examined with regard to its potential as a power-producing reactor. This approach combines the favorable aspects of both magnetic and inertial fusions in that physical containment of the plasma is provided by a metallic shell while thermal insulation of its energy is provided by a strong, self-generated magnetic field. The plasma is created at the core of the target as a result of irradiation of the fuel-coated inner surface by a laser beam that enters through a hole in the spherical shell. The instantaneous magnetic field is generated by the current loops formed by the laser-heated, laser-ablated electrons, and preliminary experimental results at Osaka University have confirmed the presence of such a field. These same experiments have also yielded a Lawson parameter of about 5x10 12 cm -3 sec, and because of these unique properties, the plasma lifetimes in MICF have been shown to be about two orders of magnitude longer than conventional, pusher type inertial fusion schemes. In this paper a quasi one dimensional, time dependent set of particle and energy balance equations for the thermal species, namely, electrons, ions and thermal alphas which also allows for an appropriate set of fast alpha groups is utilized to assess the reactor prospects of a DT-burning MICF system. (author) [pt

  11. Hybrid fission-fusion nuclear reactors

    International Nuclear Information System (INIS)

    Zucchetti, Massimo

    2011-01-01

    A fusion-fission hybrid could contribute to all components of nuclear power - fuel supply, electricity production, and waste management. The idea of the fusion-fission hybrid is many decades old. Several ideas, both new and revisited, have been investigated by hybrid proponents. These ideas appear to have attractive features, but they require various levels of advances in plasma science and fusion and nuclear technology. As a first step towards the development of hybrid reactors, fusion neutron sources can be considered as an option. Compact high-field tokamaks can be a candidate for being the neutron source in a fission-fusion hybrid, essentially due to their design characteristics, such as compact dimensions, high magnetic field, flexibility of operation. This study presents the development of a tokamak neutron source for a material testing facility using an Ignitor-based concept. The computed values show the potential of this neutron-rich device for fusion materials testing. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa. (Author)

  12. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

  13. Reduction of surface erosion in fusion reactors

    International Nuclear Information System (INIS)

    Rossing, T.D.; Das, S.K.; Kaminsky, M.

    1976-01-01

    Some of the major processes leading to surface erosion in fusion reactors are reviewed briefly, including blistering by implanted gas, sputtering by ions, atoms, and neutrons, and vaporization by local heating. Surface erosion affects the structural integrity and limits the lifetime of reactor components exposed to plasma radiation. In addition, some of the processes leading to surface erosion also cause the release of plasma contaminants. Methods proposed to reduce surface erosion have included control of surface temperature, selection of materials with a favorable microstructure, chemical and mechanical treatment of surfaces, and employment of protective surface coatings, wall liners, and divertors. The advantages and disadvantages of some of these methods are discussed

  14. Nuclear data requirements for fusion reactor shielding

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1979-01-01

    The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

  15. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

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

  16. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    Mantsinen, M.J.

    1995-03-01

    The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

  17. Overview of materials research for fusion reactors

    International Nuclear Information System (INIS)

    Muroga, T.; Gasparotto, M.; Zinkle, S.J.

    2002-01-01

    Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale

  18. Environmental aspects of fusion reactors 1985

    International Nuclear Information System (INIS)

    Casini, G.; Ponti, C.; Rocco, P.

    1986-01-01

    The aspects of the environmental impact as expected from future fusion reactors are reviewed. The radioactive inventories consist in tritium and neutron-induced radioactivity in the structures. An analysis is performed of the radioactive releases from the different plant's systems in normal and accident conditions and typical emissions to the ambient are defined. Information is given on the waste management problems. Two appendixes give general information on tritium and safety guidelines

  19. Designs of tandem-mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Barr, W.L.; Boghosian, B.M.

    1981-01-01

    We have completed a comparative evaluation of several end plug configurations for tandem mirror fusion reactors with thermal barriers. The axi-cell configuration has been selected for further study and will be the basis for a detailed conceptual design study to be carried out over the next two years. The axi-cell end plug has a simple mirror cell produced by two circular coils followed by a transition coil and a yin-yang pair, which provides for MHD stability

  20. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  1. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

    It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

  2. Survey of fusion reactor technology

    International Nuclear Information System (INIS)

    Chung, M.K.; Kang, H.D.; Cho, S.W.; Kim, Y.C.; Kim, T.S.; Whang, C.K.; Oh, Y.K.

    1982-01-01

    Design parameters of the Tokamak are as follows: 51 cm of major radius, 14.5 cm minor radius, 2.5 T toroidal magnetic field, 5 x 10 13 /cm 3 electron density, 800 eV electron temperature, 110 KA plasma current and 2 ms energy confinement time. And the Tokamak will have an ironcore transformer with no return legs for ohmic heating. Toroidal field configurations, ripples and pure tension D shape of toroidal field coils were computer-calculated and the rough calculation of poloidal field configuration was carried out manually and then a detailed calculation is underway by utilizing POISSON code. The power supply for toroidal field coils was designed to be enable to store 4 MJ energy by a capacitor bank and the mechanical parts such as a vacuum chamber, observation ports and the supporting mechanism etc. were designed with the aid of a mock-up of actual scale to realize maximum conveniency for operation and maintenance and proper arrangement and dimensions of the constituent parts. The detailed designs of various plasma diagnostics systems were also done. They are magnetic, microwave, X-ray, Laser, electrostatic systems of which the observed data will be analyzed automatically by a CAMAC data acquistion and analysis system in conjunction with Tokamak discharge. Apart from the design works of a compact Tokamak and related facilities, a proto-type theta pinch device was constructed and applied to the turbulent heating experiments. Also, accelerating column and high tension (200 KDC, 100 mA) for intense D-T neutron generator were constructed and now they are under tests. (Author)

  3. Fusion reactors-high temperature electrolysis (HTE)

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

    Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 1800 0 C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400 0 C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) [$1000/KW(E) equivalent], the H 2 energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 10 6 scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen

  4. Innovative energy production in fusion reactors

    International Nuclear Information System (INIS)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

  5. Innovative energy production in fusion reactors

    International Nuclear Information System (INIS)

    Iiyoshi, A.; Momota, H.; Motojima, O.

    1994-01-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

  6. Innovative energy production in fusion reactors

    Science.gov (United States)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

  7. Innovative energy production in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author).

  8. SEBREZ: an inertial-fusion-reactor concept

    International Nuclear Information System (INIS)

    Meier, W.R.

    1982-01-01

    The neutronic aspects of an inertial fusion reactor concept that relies on asymmetrical neutronic effects to enhance the tritium production in the breeding zones have been studied. We find that it is possible to obtain a tritium breeding ratio greater than 1.0 with a chamber configuration in which the breeding zones subtend only a fraction of the total solid angle. This is the origin of the name SEBREZ which stands for SEgregated BREeding Zones. It should be emphasized that this is not a reactor design study; rather this study illustrates certain neutronic effects in the context of a particular reactor concept. An understanding of these effects forms the basis of a design technique which has broader application than just the SEBREZ concept

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

    International Nuclear Information System (INIS)

    Lasche, G.P.

    1988-01-01

    A method for recovering energy in an inertial confinement fusion reactor having a reactor chamber and a sphere forming means positioned above an opening in the reactor chamber is described, comprising: embedding a fusion target fuel capsule having a predetermined yield in the center of a hollow solid lithium tube and subsequently embedding the hollow solid lithium tube in a liquid lithium medium; using the sphere forming means for forming the liquid lithium into a spherical shaped liquid lithium mass having a diameter smaller than the length of the hollow solid lithium tube with the hollow solid lithium tube being positioned along a diameter of the spherical shaped mass, providing the spherical shaped liquid lithium mass with the fusion fuel target capsule and hollow solid lithium tube therein as a freestanding liquid lithium shaped spherical shaped mass without any external means for maintaining the spherical shape by dropping the liquid lithium spherical shaped mass from the sphere forming means into the reactor chamber; producing a magnetic field in the reactor chamber; imploding the target capsule in the reactor chamber to produce fusion energy; absorbing fusion energy in the liquid lithium spherical shaped mass to convert substantially all the fusion energy to shock induced kinetic energy of the liquid lithium spherical shaped mass which expands the liquid lithium spherical shaped mass; and compressing the magnetic field by expansion of the liquid lithium spherical shaped mass and recovering useful energy

  10. Interatomic potentials for fusion reactor material simulations

    International Nuclear Information System (INIS)

    Bjoerkas, C.

    2009-01-01

    In this thesis, the behaviour of a material situated in a fusion reactor was studied using molecular dynamics simulations. Simulations of processes in the next generation fusion reactor ITER include the reactor materials beryllium, carbon and tungsten as well as the plasma hydrogen isotopes. This means that interaction models, i.e. interatomic potentials, for this complicated quaternary system are needed. The task of finding such potentials is nonetheless nearly at its end, since models for the beryllium-carbon-hydrogen interactions were constructed in this thesis and as a continuation of that work, a beryllium-tungsten model is under development. These potentials are combinable with the earlier tungsten-carbon-hydrogen ones. The potentials were used to explain the chemical sputtering of beryllium due to deuterium plasma exposure. During experiments, a large fraction of the sputtered beryllium atoms were observed to be released as BeD molecules, and the simulations identified the swift chemical sputtering mechanism, previously not believed to be important in metals, as the underlying mechanism. Radiation damage in the reactor structural materials vanadium, iron and iron chromium, as well as in the wall material tungsten and the mixed alloy tungsten carbide, was also studied in this thesis. Interatomic potentials for vanadium, tungsten and iron were modified to be better suited for simulating collision cascades that are formed during particle irradiation, and the potential features affecting the resulting primary damage were identified. Including the often neglected electronic effects in the simulations was also shown to have an impact on the damage. With proper tuning of the electronphonon interaction strength, experimentally measured quantities related to ion-beam mixing in iron could be reproduced. The damage in tungsten carbide alloys showed elemental asymmetry, as the major part of the damage consisted of carbon defects. On the other hand, modelling the damage

  11. Achieving competitive excellence in nuclear energy: The threat of proliferation; the challenge of inertial confinement fusion

    International Nuclear Information System (INIS)

    Nuckolls, J.H.

    1994-06-01

    Nuclear energy will have an expanding role in meeting the twenty-first-century challenges of population and economic growth, energy demand, and global warming. These great challenges are non-linearly coupled and incompletely understood. In the complex global system, achieving competitive excellence for nuclear energy is a multi-dimensional challenge. The growth of nuclear energy will be driven by its margin of economic advantage, as well as by threats to energy security and by growing evidence of global warming. At the same time, the deployment of nuclear energy will be inhibited by concerns about nuclear weapons proliferation, nuclear waste and nuclear reactor safety. These drivers and inhibitors are coupled: for example, in the foreseeable future, proliferation in the Middle East may undermine energy security and increase demand for nuclear energy. The Department of Energy's nuclear weapons laboratories are addressing many of these challenges, including nuclear weapons builddown and nonproliferation, nuclear waste storage and burnup, reactor safety and fuel enrichment, global warming, and the long-range development of fusion energy. Today I will focus on two major program areas at the Lawrence Livermore National Laboratory (LLNL): the proliferation of nuclear weapons and the development of inertial confinement fusion (ICF) energy

  12. Inertial electrostatic confinement fusion neutron source R ampersand D and issues

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Hasegawa, Mitsunori

    1997-01-01

    An inertial electrostatic confinement (IEC) fusion is the scheme of injecting the ions and electrons toward the spherical center, trapping both species in the electrostatic self-field and giving rise to fusion reactions in the dense core. An IEC is expected to have wide application from a small neutron source to a D- 3 He fusion reactor. Hirsch reported 10 9 n/s deuterium-tritium (D-T) neutron production in the device equipped with ion guns. Recently, Gu et al. measured 10 6 n/s using a D 2 gas discharge between the spherical wire cathode and the anode vacuum vessel, where the applied voltage is 60 kV and the current is 15 mA. We have also obtained similar neutron production at a lower voltage, ∼45 kV in a single-grid IEC device. Fusion reaction rates obtained by IEC experiments so far cannot be explained by a model of a simple potential well structure because the electrical potential peaked at the center prevents making a dense core. Hirsch proposed a multiwell structure called open-quotes poissorsclose quotes to explain the experiments. It is generally believed that there may be some correlation between the potential well structure and the neutron production rate. The scaling of neutron production on the injected ion current is a most important aspect of the problem for the prospect of utilizing IEC for fusion energy. The potential structure and its behavior are keys to the physics in understanding the principle of an IEC

  13. Fabrication of thin-wall, freestanding inertial confinement fusion targets by chemical vapor deposition

    International Nuclear Information System (INIS)

    Carroll, D.W.; McCreary, W.J.

    1982-01-01

    To meet the requirements for plasma physics experiments in the inertial confinement fusion (ICF) program, chemical vapor deposition (CVD) in fluid beds was used to fabricate freestanding tungsten spheres and cylinders with wall thicknesses less than 5.0 μm. Molybdenum and molybdenum alloy (TZM) mandrels of the desired geometry were suspended in a carrier bed of dense microspheres contained in an induction-heated fluid-bed reactor. The mandrels were free to float randomly through the bed, and using the reaction WF 6 +3H 2 →/sub /KW +6HF, very fine-grained tungsten was deposited onto the surface at a rate and in a grain size determined by temperature, gas flow rate, system pressure, and duration of the reaction. After coating, a portion of each mandrel was exposed by hole drilling or grinding. The mandrel was then removed by acid leaching, leaving a freestanding tungsten shape. Experimental procedures, mandrel preparation, and results obtained are discussed

  14. Repetitive pulse accelerator technology for light ion inertial confinement fusion

    International Nuclear Information System (INIS)

    Buttram, M.T.

    1985-01-01

    Successful ignition of an inertial confinement fusion (ICF) pellet is calculated to require that several megajoules of energy be deposited in the pellet's centimeter-sized shell within 10 ns. This implies a driver power of several hundreds of terawatts and power density around 100 TW/cm 2 . The Sandia ICF approach is to deposit the energy with beams of 30 MV lithium ions. The first accelerator capable of producing these beams (PBFA II, 100 TW) will be used to study beam formation and target physics on a single pulse basis. To utilize this technology for power production, repetitive pulsing at rates that may be as high as 10 Hz will be required. This paper will overview the technologies being studied for a repetitively pulsed ICF accelerator. As presently conceived, power is supplied by rotating machinery providing 16 MJ in 1 ms. The generator output is transformed to 3 MV, then switched into a pulse compression system using laser triggered spark gaps. These must be synchronized to about 1 ns. Pulse compression is performed with saturable inductor switches, the output being 40 ns, 1.5 MV pulses. These are transformed to 30 MV in a self-magnetically insulated cavity adder structure. Space charge limited ion beams are drawn from anode plasmas with electron counter streaming being magnetically inhibited. The ions are ballistically focused into the entrances of guiding discharge channels for transport to the pellet. The status of component development from the prime power to the ion source will be reviewed

  15. Production of hollow microspheres for inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Cook, R.

    1994-12-01

    The targets used in inertial confinement fusion (ICF) experiments at the Lawrence Livermore National Laboratory are plastic capsules roughly 0.5 mm in diameter. The capsules, which typically have wall thicknesses from 20 to 60 μm, must possess extraordinary symmetry and concentricity and must have surface finishes of less than 1000 Angstrom peak-to-valley variation over surface contours of from 10 to 100's of μm. This paper reviews the fabrication of these capsules, focusing on the production of the thin-walled polystyrene microshell mandrel around which the capsule is built. The relationship between the capsule characteristics, especially surface finish, and capsule performance is discussed, as are the methods of surface characterization and modification necessary for experiments designed to study the effects of surface roughness on implosion dynamics. Targets for the next generation of ICF facilities using more powerful laser drivers will have to be larger while meeting the same or even more stringent symmetry and surface finish requirements. Some of the technologies for meeting these needs are discussed briefly

  16. Neutron imaging for inertial confinement fusion and molecular optic imaging

    International Nuclear Information System (INIS)

    Delage, O.

    2010-01-01

    Scientific domains that require imaging of micrometric/nano-metric objects are dramatically increasing (Plasma Physics, Astrophysics, Biotechnology, Earth Sciences...). Difficulties encountered in imaging smaller and smaller objects make this research area more and more challenging and in constant evolution. The two scientific domains, through which this study has been led, are the neutron imaging in the context of the inertial confinement fusion and the fluorescence molecular imaging. Work presented in this thesis has two main objectives. The first one is to describe the instrumentation characteristics that require such imagery and, relatively to the scientific domains considered, identify parameters likely to optimize the imaging system accuracy. The second one is to present the developed data analysis and reconstruction methods able to provide spatial resolution adapted to the size of the observed object. Similarities of numerical algorithms used in these two scientific domains, which goals are quiet different, show how micrometric/nano-metric object imaging is a research area at the border of a large number of scientific disciplines. (author)

  17. Modeling Small-Amplitude Perturbations in Inertial Confinement Fusion Pellets

    Science.gov (United States)

    Zalesak, Steven; Metzler, N.; Velikovich, A. L.; Gardner, J. H.; Manheimer, W.

    2005-10-01

    Recent advances in inertial confinement fusion (ICF) technology serve to ensure that imploding laser-driven ICF pellets will spend a significantly larger portion of their time in what is regarded as the ``linear'' portion of their perturbation evolution, i.e., in the presence of small-amplitude but nonetheless evolving perturbations. Since the evolution of these linear perturbations collectively form the initial conditions for the subsequent nonlinear evolution of the pellet, which in turn determines the energy yield of the pellet, the accurate numerical modeling of these small-amplitude perturbations has taken on an increased importance. This modeling is difficult despite the expected linear evolution of the perturbations themselves, because these perturbations are embedded in a highly nonlinear, strongly-shocked, and highly complex flow field which in and of itself stresses numerical computation capabilities, and whose simulation often employs numerical techniques which were not designed with the proper treatment of small-amplitude perturbations in mind. In this paper we will review some of the techniques that we have recently found to be of use toward this end.

  18. Statistical Relations for Yield Degradation in Inertial Confinement Fusion

    Science.gov (United States)

    Woo, K. M.; Betti, R.; Patel, D.; Gopalaswamy, V.

    2017-10-01

    In inertial confinement fusion (ICF), the yield-over-clean (YOC) is a quantity commonly used to assess the performance of an implosion with respect to the degradation caused by asymmetries. The YOC also determines the Lawson parameter used to identify the onset of ignition and the level of alpha heating in ICF implosions. In this work, we show that the YOC is a unique function of the residual kinetic energy in the compressed shell (with respect to the 1-D case) regardless of the asymmetry spectrum. This result is derived using a simple model of the deceleration phase as well as through an extensive set of 3-D radiation-hydrodynamics simulations using the code DEC3D. The latter has been recently upgraded to include a 3-D spherical moving mesh, the HYPRE solver for 3-D radiation transport and piecewise-parabolic method for robust shock-capturing hydrodynamic simulations. DEC3D is used to build a synthetic single-mode database to study the behavior of yield degradation caused by Rayleigh-Taylor instabilities in the deceleration phase. The relation between YOC and residual kinetic energy is compared with the result in an adiabatic implosion model. The statistical expression of YOC is also applied to the ignition criterion in the presence of multidimensional nonuniformities. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  19. Shock-timing experiments for Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Debras, G.

    2012-01-01

    The Laser Megajoule (LMJ), which should achieve energy gain in an indirect drive inertial confinement fusion configuration, is being built in France by the CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives). To achieve thermonuclear ignition, the compression of a spherical target will have to be controlled by a series of accurately timed centripetal shocks, with a finely tuned level. A first experiment, performed in 2010 on the LIL (Ligne d'Integration Laser) facility at CEA, has allowed us to study the coalescence of two planar shocks in an indirectly-driven sample of polystyrene, within the framework of shock timing. The main objectives were to validate the experimental concept and the numerical simulations, as a proof-of-principle for future shock-timing campaigns. The main diagnostics used for this study are VISAR (Velocity Interferometer System for Any Reflection) and an optical shock breakout diagnostic, taking into account optical perturbations caused by X-rays. In another experiment, conducted on the LULI (Laboratoire pour l'Utilisation des Lasers Intenses) laser facility in 2010, we studied the timing of two planar directly-driven shocks using the same diagnostics. This latter study is related to the shock ignition concept, with the long-term perspective of energy production. This thesis presents these two experiments and their results. (author) [fr

  20. First wall studies of a laser-fusion hybrid reactor design

    International Nuclear Information System (INIS)

    Hovingh, J.

    1976-09-01

    The design of a first wall for a 20 MW thermonuclear power laser fusion hybrid reactor is presented. The 20 mm thick graphite first wall is located 3.5 m from the DT microexplosion with a thermonuclear yield of 10 MJ. Estimates of the energy deposition, temperature, stresses, and material vaporized from the first wall due to the interaction of the x-rays, charged particle debris, and reflected laser light with the graphite are presented, along with a brief description of the analytical methods used for these estimations. Graphite is a viable first wall material for inertially-confined fusion reactors, with lifetimes of a year possible

  1. Power balance in an Ohmically heated fusion reactor

    International Nuclear Information System (INIS)

    Christiansen, J.P.; Roberts, K.V.

    1982-01-01

    A simplified power-balance equation (zero-dimensional model) is used to study the performance of an Ohmically heated fusion reactor with emphasis on a pulsed reversed-field pinch concept (RFP). The energy confinement time tausub(E) is treated as an adjustable function, and empirical tokamak scaling laws are employed in the numerical estimates, which are supplemented by 1-D ATHENE code calculations. The known heating rates and energy losses are represented by the net energy replacement time tausub(W), which is exhibited as a surface in density (n) and temperature (T) space with a saddle point (nsub(*), Tsub(*)), the optimum ignition point. It is concluded that i) ignition by Ohmic heating is more practicable for the RFP reactor than for a tokamak reactor with the same tausub(E), (ii) if at fixed current the minor radius can be reduced or at fixed minor radius the current can be increased, then it is found that Ohmic ignition becomes more likely when present tokamak scaling laws are used. More definitive estimates require, however, a knowledge of tausub(E), which can only be obtained by establishing a reliable set of experimental RFP scaling laws and, in particular, by extending RFP experiments closer to the reactor regime. (author)

  2. Focused proton beams propagating in reactor of fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Niu, K [Teikyo Heisei Univ., Uruido, Ichihara, Chiba (Japan)

    1997-12-31

    One of the difficult tasks of light ion beam fusion is to propagate the beam in the reactor cavity and to focus the beam on the target. The light ion beam has a certain local divergence angle because there are several causes for divergence at the diode. The electrostatic force induced at the leading edge causes beam divergence during propagation. To confine the beam within a small radius during propagation, the magnetic field must be employed. Here the electron beam is proposed to be launched simultaneously with the launching of the proton beam. If the electron beam has the excess current, the beam induces a magnetic field in the negative azimuthal direction, which confines the ion beam within a small radius by the electrostatic field as well as the electron beam by the Lorentz force. The metal guide around the beam path helps the beam confinement and reduces the total amount of magnetic field energy induced by the electron current. (author). 2 figs., 15 refs.

  3. Survey of the laser-solenoid fusion reactor

    International Nuclear Information System (INIS)

    Amherd, N.A.

    1975-09-01

    This report surveys the prospects for a laser-solenoid fusion reactor. A sample reactor and scaling laws are used to describe the concept's characteristics. Experimental results are reviewed, and the laser and magnet technologies that undergird the laser-solenoid concept are analyzed. Finally, a systems analysis of fusion power reactors is given, including a discussion of direct conversion and fusion-fission effects, to ascertain the system attributes of the laser-solenoid configuration

  4. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.; Masson, M.; Briec, M.

    1986-09-01

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

  5. Investigation of materials for fusion power reactors

    Science.gov (United States)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  6. Series lecture on advanced fusion reactors

    International Nuclear Information System (INIS)

    Dawson, J.M.

    1983-01-01

    The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

  7. Conceptual design of Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Tone, T.; Fujisawa, N.

    1983-01-01

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been performed. The FER has an objective of achieving selfignition and demonstrating engineering feasibility as a next generation tokamak to JT-60. Various concepts of the FER have been considered. The reference design is based on a double-null divertor. Optional design studies with some attractive features based on advanced concepts such as pumped limiter and RF current drive have been carried out. Key design parameters are; fusion power of 440 MW, average neutron wall loading of 1MW/m 2 , major radius of 5.5m, plasma minor radius of 1.1m, plasma elongation of 1.5, plasma current of 5.3MA, toroidal beta of 4%, toroidal field on plasma axis of 5.7T and tritium breeding ratio of above unity

  8. Assessment of materials needs for fusion reactors

    International Nuclear Information System (INIS)

    Allison, G.S.

    1976-07-01

    This report has the goal of presenting for the CTR designer and material supplier potentially significant problem areas in materials manufacturing and in structural material resources projected for potential application in fusion power reactor construction. The projected material requirements are based on presently available bills-of-materials for conceptual CTR designs used for constructing a hypothetical fusion power generating capacity of 10 6 MW(e) maturing exponentially over a 20-year period. The projected elemental requirements, the ratio of these requirements to the projected total U.S. demand, and the salient problems currently identified with the CTR use of these elements are summarized. The projected requirements are based upon a ''model'' industry, which is described, and the estimated potential use of molybdenum, niobium, vanadium, and tantalum as blanket structural materials

  9. Assessment of materials needs for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Allison, G.S. (comp.)

    1976-07-01

    This report has the goal of presenting for the CTR designer and material supplier potentially significant problem areas in materials manufacturing and in structural material resources projected for potential application in fusion power reactor construction. The projected material requirements are based on presently available bills-of-materials for conceptual CTR designs used for constructing a hypothetical fusion power generating capacity of 10/sup 6/ MW(e) maturing exponentially over a 20-year period. The projected elemental requirements, the ratio of these requirements to the projected total U.S. demand, and the salient problems currently identified with the CTR use of these elements are summarized. The projected requirements are based upon a ''model'' industry, which is described, and the estimated potential use of molybdenum, niobium, vanadium, and tantalum as blanket structural materials.

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

    International Nuclear Information System (INIS)

    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

  11. The scientific benefits of inertially confined fusion research

    International Nuclear Information System (INIS)

    Key, M

    1999-01-01

    A striking feature of 25 years of research into inertially confined fusion (ICF) and inertial fusion energy (IFE) has been its significant impact in other fields of science. Most ICF facilities worldwide are now being used in part to support a wider portfolio of research than simply ICF. Reasons for this trend include the high intrinsic interest of the new science coupled with the relative ease and low marginal cost of adapting the facilities particularly lasers, to carry out experiments with goals other than ICF. The availability at ICF laboratories of sophisticated theory and modeling capability and advanced diagnostics has given added impetus. The expertise of ICF specialists has also triggered more lateral scientific spin-offs leading for example to new types of lasers and to related developments in basic science. In a generic sense, the facilities developed for ICF have made possible study of new regimes of the properties of matter at extremely high-energy density and the interaction of ultraintense light with matter. This general opportunity has been exploited in numerous and diverse specific lines of research. Examples elaborated below include laboratory simulation of astrophysical phenomena; studies of the equation of state (EOS) of matter under conditions relevant to the interior of planets and stars; development of uniquely intense sources of extreme ultraviolet (EUV) to hard x-ray emission, notably the x-ray laser; understanding of the physics of strong field interaction of light and matter; and related new phenomena such as laser-induced nuclear processes and high-field-electron accelerators. Some of these developments have potential themselves for further scientific exploitation such as the scientific use of advanced light sources. There are also avenues for commercial exploitation, for example the use of laser plasma sources in EUV lithography. Past scientific progress is summarized here and projections are made for new science that may flow from the

  12. Afterheat assessment of a conceptual fusion reactor

    International Nuclear Information System (INIS)

    Jayatissa, S.P.; Goddard, A.J.H.

    1987-01-01

    Structural activation and decay heat deposition calculations have been undertaken for the DEMO fusion reactor design. The DEMO design was based on an earlier conceptual design of a blanket sector which could breed tritium and generate electricity. These calculations have taken account of the redistribution of energy by the transport of γ radiation. Calculated heat deposition patterns have been used as data for simplified heat transfer calculations to judge temperature rises in relation to materials limits in a severe accident involving complete coolant flow failure. (author)

  13. Dust removal system for fusion experimental reactors

    International Nuclear Information System (INIS)

    Onozuka, M.; Ueda, Y.; Takahashi, K.; Oda, Y.; Seki, Y.; Ueda, S.; Aoki, I.

    1995-01-01

    Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by the DC electric field. The dust is then transported by the electric curtain formed by the three-phase AC electric field. Experimental investigation has been conducted to examine the characteristics of the system. Current research results indicate that the dust removal system using static electricity can be used for fusion experimental reactors

  14. Dust removal system for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M.; Ueda, Y.; Takahashi, K.; Oda, Y. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Seki, Y.; Ueda, S.; Aoki, I. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1995-12-31

    Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by the DC electric field. The dust is then transported by the electric curtain formed by the three-phase AC electric field. Experimental investigation has been conducted to examine the characteristics of the system. Current research results indicate that the dust removal system using static electricity can be used for fusion experimental reactors.

  15. Organic materials for fusion-reactor applications

    International Nuclear Information System (INIS)

    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

  16. Vanadium alloys for fusion reactor applications

    International Nuclear Information System (INIS)

    Mattas, R.F.; Loomis, B.A.; Smith, D.L.

    1992-01-01

    This paper reports that fusion reactors will produce a severe operating environment for structural materials. The material should have good mechanical strength and ductility to high temperature, be corrosion resistant to the local environment, have attractive thermophysical properties to accommodate high heat loads, and be resistant to neutron damage. Vanadium alloys are being developed for such applications, and they exhibit desirable properties in many areas Recent progress in vanadium alloy development indicates good strength and ductility to 700 degrees C, minimal degradation by neutron irradiation, and reduced radioactivity compared with other candidate alloy systems

  17. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Young, J.R.

    1975-01-01

    Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

  18. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    2003-01-01

    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 and 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 and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and 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

  19. Tritium permeation in fusion reactors: INTOR

    International Nuclear Information System (INIS)

    Baskes, M.I.; Bauer, W.; Kerst, R.A.; Swansiger, W.A.; Wilson, K.L.

    1981-12-01

    Tritium permeation through the first wall of advanced fusion reactors is examined. A fraction of the D-T which bombards the first wall as charge exchange neutral particles will permeate through the first wall and enter the coolant. Calculations of the steady state permeation rate for the US INTOR Tokamak design result in values of less than or equal to 0.002 grams of tritium per day under the most favorable conditions. For unfavorable surface conditions the rate is greater than or equal to 0.1 g/day. The magnitude of these permeation rates is critically dependent on the temperatures and surface conditions of the wall. The introduction of permeation barriers at the wall-coolant interface can significantly reduce permeation rates and hence may be desirable for reactor applications

  20. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1985-01-01

    The Fusion Experimental Reactor (FER) being developed at JAERI as a next generation tokamak to JT-60 has a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. During FY82 and FY83 a comprehensive and intensive conceptual design study has been conducted for a pulsed operation FER as a reference option which employs a conventional inductive current drive and a double-null divertor. In parallel with the reference design, studies have been carried out to evaluate advanced reactor concepts such as quasi-steady state operation and steady state operation based on RF current drive and pumped limiter, and comparative studies for single-null divertor/pumped limiter. This report presents major results obtained primarily from FY83 design studies, while the results of FY82 design studies are described in previous references (JAERI-M 83-213--216). (author)

  1. Contribution to the numerical modeling of inertial confinement fusion

    International Nuclear Information System (INIS)

    Maire, P.H.

    2011-02-01

    This work was realized by writing the CHIC code, which is a software for designing and restoring experience in the field of inertial confinement fusion. The theoretical model describing the implosion of a laser target is a system of partial differential equations in the center of which is the Euler equations written in Lagrangian formalism, coupled with diffusion equations modeling the nonlinear transport of energy by electrons and photons. After a brief review of the physical context, we describe two novel methods which constitute the backbone of the CHIC code. These are 2 high-order finite volume schemes respectively dedicated to solving the equations of Lagrangian hydrodynamics and the anisotropic diffusion equations on bi-dimensional unstructured grids. The first scheme, called EUCCLHYD (Explicit Unstructured Lagrangian Hydrodynamics), solves the equations of gas dynamics on a moving mesh that moves at the speed of light. It is obtained from a general formalism based on the concept of sub-cell forces. In this context, the numerical fluxes are expressed in terms of the sub-cell force and the nodal velocity. Their determination is based on 3 basic principles: geometric compatibility between the movement of nodes and the volume change of mesh (geometric conservation law), compatibility with the second law of thermodynamics and conservation of total energy and momentum. The high-order extension is performed using a method based on solving a generalized Riemann problem in the acoustic approximation. The second scheme, called CCLAD (Cell-Centered Lagrangian Diffusion), solves the anisotropic heat equation. The corresponding discretization relies on a discrete variational formulation based on the sub-cell that allows to build a multipoint approximation of heat flux. This high-order discretization makes possible the resolution of the equations of anisotropic diffusion with satisfactory accuracy on highly distorted Lagrangian meshes. (author)

  2. Mixing with applications to inertial-confinement-fusion implosions

    Science.gov (United States)

    Rana, V.; Lim, H.; Melvin, J.; Glimm, J.; Cheng, B.; Sharp, D. H.

    2017-01-01

    Approximate one-dimensional (1D) as well as 2D and 3D simulations are playing an important supporting role in the design and analysis of future experiments at National Ignition Facility. This paper is mainly concerned with 1D simulations, used extensively in design and optimization. We couple a 1D buoyancy-drag mix model for the mixing zone edges with a 1D inertial confinement fusion simulation code. This analysis predicts that National Ignition Campaign (NIC) designs are located close to a performance cliff, so modeling errors, design features (fill tube and tent) and additional, unmodeled instabilities could lead to significant levels of mix. The performance cliff we identify is associated with multimode plastic ablator (CH) mix into the hot-spot deuterium and tritium (DT). The buoyancy-drag mix model is mode number independent and selects implicitly a range of maximum growth modes. Our main conclusion is that single effect instabilities are predicted not to lead to hot-spot mix, while combined mode mixing effects are predicted to affect hot-spot thermodynamics and possibly hot-spot mix. Combined with the stagnation Rayleigh-Taylor instability, we find the potential for mix effects in combination with the ice-to-gas DT boundary, numerical effects of Eulerian species CH concentration diffusion, and ablation-driven instabilities. With the help of a convenient package of plasma transport parameters developed here, we give an approximate determination of these quantities in the regime relevant to the NIC experiments, while ruling out a variety of mix possibilities. Plasma transport parameters affect the 1D buoyancy-drag mix model primarily through its phenomenological drag coefficient as well as the 1D hydro model to which the buoyancy-drag equation is coupled.

  3. Rayleigh-Taylor instability in inertial confinement fusion

    International Nuclear Information System (INIS)

    Gupta, N.K.

    1987-01-01

    This report summarises the main results of theoretical analysis on the problem of Rayleigh-Tylor instability in inertial confinement fusion (ICF). Work presented in this report essentially covers four basic problems. Firstly, an analytical formulation to analyse the effects of plasma density inhomogeneities on the growth of the instability in plane geometry is presented. As a result of this analysis it is concluded that, for minimizing the growth rate of the instability, it may be advantageous to use the driver laser beams of higher irradiance and an optimum wave length in an ICF experiment. Secondly, a new formulation for the analysis of the instability in curved (cylindrical and spherical) geometries is presented. A general eigenvalue equation for the growth rate of the instability which is applicable for both plane and curved geometries is derived. A comparative study is made between the plane, cylindrical and spherical geometries. Also analytical expressions for the growth rates are obtained in the cases of spherical and cylindrical shell targets and their variations with respect to the aspect ratios of the shells are discussed. Thirdly, a semi-analytical analysis of the instability where the growth rate is obtained by solving numerically a (2N-1)x(2N-1) determinantal equation is presented. The semi-analytical analysis developed is applicable for the study of the growth of the instability in the present day multi-structured spherical shell targets. Finally, a dynamic analysis of the growth of the instability for a representative spherical solid target driven by laser beams symmetrically from all the sides is carried out numerically using a computer code developed for this purpose. This study confirms analytical predictions. Further, it is observed that an approximate analytical analysis with time independent density profile gives conservative estimates for the growth rate. In passing, the computer code is also used to estimate the pellet gain for spin

  4. Study of the shock ignition scheme in inertial confinement fusion

    International Nuclear Information System (INIS)

    Lafon, M.

    2011-01-01

    The Shock Ignition (SI) scheme is an alternative to classical ignition schemes in Inertial Confinement Fusion. Its singularity relies on the relaxation of constraints during the compression phase and fulfilment of ignition conditions by launching a short and intense laser pulse (∼500 ps, ∼300 TW) on the pre-assembled fuel at the end of the implosion.In this thesis, it has been established that the SI process leads to a non-isobaric fuel configuration at the ignition time thus modifying the ignition criteria of Deuterium-Tritium (DT) against the conventional schemes. A gain model has been developed and gain curves have been inferred and numerically validated. This hydrodynamical modeling has demonstrated that the SI process allows higher gain and lower ignition energy threshold than conventional ignition due to the high hot spot pressure at ignition time resulting from the ignitor shock propagation.The radiative hydrodynamic CHIC code developed at the CELIA laboratory has been used to determine parametric dependences describing the optimal conditions for target design leading to ignition. These numerical studies have enlightened the potential of SI with regards to saving up laser energy, obtain high gains but also to safety margins and ignition robustness.Finally, the results of the first SI experiments performed in spherical geometry on the OMEGA laser facility (NY, USA) are presented. An interpretation of the experimental data is proposed from mono and bidimensional hydrodynamic simulations. Then, different trails are explored to account for the differences observed between experimental and numerical data and alternative solutions to improve performances are suggested. (author) [fr

  5. X-ray sources by Z-pinch for inertial confinement fusion

    International Nuclear Information System (INIS)

    Akiyama, Hidenori; Katsuki, Sunao; Lisitsyn, Igor

    1999-01-01

    Inertial confinement nuclear fusion driven by X-ray from Z-pinch plasmas has been developed. Recently, extremely high X-ray power (290 TW) and energy (1.8 MJ) were produced in fast Z-pinch implosions on the Z accelerator (Sandia National Laboratories). Wire arrays are used to produce the initial plasma. The X-ray from Z-pinch plasmas produced by pulsed power has great potential as a driver of inertial confinement nuclear fusion. (author)

  6. Inertial Confinement Fusion quarterly report, January--March 1995. Volume 5, No. 2

    International Nuclear Information System (INIS)

    1995-01-01

    The ICF quarterly report is published by the Inertial Confinement Fusion Program at the Lawrence Livermore National Laboratory. Topics included this quarter include: the role of the National Ignition Facility in the development of Inertial Confinement Fusion, laser-plasma interactions in large gas-filled hohlraums, evolution of solid-state induction modulators for a heavy-ion recirculator, the National Ignition Facility project, and terminal-level relaxation in Nd-doped laser material

  7. Structural properties of hydrogen isotopes in solid phase in the context of inertial confinement fusion

    Directory of Open Access Journals (Sweden)

    Guerrero Carlo

    2013-11-01

    Full Text Available Quality of Deuterium-Tritium capsules is a critical aspect in Inertial Confinement Fusion. In this work, we present a Quantum Molecular Dynamics methodology able to model hydrogen isotopes and their structural molecular organisation at extreme pressures and cryogenic temperatures (< 15 K. Our study sets up the basis for a future analysis on the mechanical and structural properties of DT-ice in inertial confinement fusion (ICF target manufacturing conditions.

  8. Activation product transport in fusion reactors

    International Nuclear Information System (INIS)

    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

  9. Pulsed power ion accelerators for inertially confined fusion

    International Nuclear Information System (INIS)

    Olson, C.L.

    1976-01-01

    Current research is described on pulsed power ion accelerators for inertial fusion, i.e., ion diodes and collective accelerators. Particle beam energy and power requirements for fusion, and basic deposition characteristics of charged particle beams are discussed. Ion diodes and collective accelerators for fusion are compared with existing conventional accelerators

  10. Robust stabilization of burn conditions in subignited fusion reactors using artificial neural networks

    International Nuclear Information System (INIS)

    Vitela, E. Javier; Martinell, J. Julio

    2000-01-01

    In this work it is shown that robust burn control in long pulse operations of thermonuclear reactors can be successfully achieved with artificial neural networks. The results reported here correspond to a volume averaged zero-dimensional nonlinear model of a subignited fusion device using the design parameters of the tokamak EDA-ITER group. A Radial Basis Neural Network (RBNN) was trained to provide feedback stabilization at a fixed operating point independently of any particular scaling law that the reactor confinement time may follow. A numerically simulated transient is used to illustrate the stabilization capabilities of the resulting RBNN when the reactor follows an ELMy scaling law corrupted with Gaussian noise. (author)

  11. SOLASE conceptual laser fusion reactor study

    International Nuclear Information System (INIS)

    Moses, G.A.; Conn, R.W.; Abdel-Khalik, S.I.; Cooper, G.W.; Howard, J.; Magelssen, G.R.

    1978-01-01

    A conceptual laser fusion reactor for electric power, SOLASE, has been designed. The SOLASE design utilizes a 1 MJ, 6.7% efficient laser to implode 20 fusion targets per second. The target gain is 150 and produces a net electrical power of 1000 MW. The reactor cavity is spherical with a 6 m radius. The first wall is graphite and has a neutron wall loading of 5 MW/m 2 . It is protected from the target debris by low pressure xenon gas that is introduced into the cavity. The blanket structure is a honeycombed graphite composite. The tritium breeding and heat transport medium is Li 2 O in the form of pellets that flow through the blanket. The tritium breeding ration is 1.34. Temperature decoupling of the graphite structure and the Li 2 O coolant enables the structure to operate at temperatures that minimize radiation damage effects. The graphite blanket is replaced every year but exhibits low levels of radioactivity so that limited hands on maintenance is possible two weeks after shutdown, thus facilitating rapid replacement

  12. Process for manufacture of inertial confinement fusion targets and resulting product

    International Nuclear Information System (INIS)

    Solomon, D.E.; Wise, K.D.; Wuttke, G.H.; Masnari, N.A.; Rensel, W.B.; Robinson, M.G.

    1980-01-01

    A method of manufacturing inertial confinement fusion targets is described which is adaptable for high volume production of low cost targets in a wide variety of sizes. The targets include a spherical pellet of fusion fuel surrounded by a protective concentric shell. (UK)

  13. Development of heavy-ion accelerators as drivers for inertially confined fusion

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.

    1979-06-01

    The commercialization of inertial confinement fusion is discussed in terms of power costs. A chapter on heavy ion accelerators covers the prinicpal components, beam loss mechanisms, and theoretical considerations. Other tyopics discussed include the following: (1) heavy ion fusion implementation plan, (2) driver with accumulator rings fed by an rf LINAC, (3) single pass driver with an induction LINAC, and (4) implementation scenarios

  14. Inertial Confinement Fusion quarterly report, April--June 1995. Volume 5, No. 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The ICF Quarterly Reports is published four times each fiscal year by the Inertial Confinement Fusion Program at the Lawrence Livermore National Laboratory. The journal reports selected current research within the ICF Program. Major areas of investigation presented here include fusion target theory and design, target fabrication, target experiments, and laser and optical science and technology.

  15. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

    The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

  16. Hydrogen production from high temperature electrolysis and fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, J.F.; Issacs, H.S.; Lazareth, O.; Powell, J.R.; Salzano, F.J.

    1978-01-01

    Production of hydrogen from high temperature electrolysis of steam coupled with a fusion reactor is studied. The process includes three major components: the fusion reactor, the high temperature electrolyzer and the power conversion cycle each of which is discussed in the paper. Detailed process design and analysis of the system is examined. A parametric study on the effect of process efficiency is presented

  17. Graphs of neutron cross section data for fusion reactor development

    International Nuclear Information System (INIS)

    Asami, Tetsuo; Tanaka, Shigeya

    1979-03-01

    Graphs of neutron cross section data relevant to fusion reactor development are presented. Nuclides and reaction types in the present compilation are based on a WRENDA request list from Japan for fusion reactor development. The compilation contains various partial cross sections for 55 nuclides from 6 Li to 237 Np in the energy range up to 20 MeV. (author)

  18. Burn Control in Fusion Reactors via Nonlinear Stabilization Techniques

    International Nuclear Information System (INIS)

    Schuster, Eugenio; Krstic, Miroslav; Tynan, George

    2003-01-01

    Control of plasma density and temperature magnitudes, as well as their profiles, are among the most fundamental problems in fusion reactors. Existing efforts on model-based control use control techniques for linear models. In this work, a zero-dimensional nonlinear model involving approximate conservation equations for the energy and the densities of the species was used to synthesize a nonlinear feedback controller for stabilizing the burn condition of a fusion reactor. The subignition case, where the modulation of auxiliary power and fueling rate are considered as control forces, and the ignition case, where the controlled injection of impurities is considered as an additional actuator, are treated separately.The model addresses the issue of the lag due to the finite time for the fresh fuel to diffuse into the plasma center. In this way we make our control system independent of the fueling system and the reactor can be fed either by pellet injection or by puffing. This imposed lag is treated using nonlinear backstepping.The nonlinear controller proposed guarantees a much larger region of attraction than the previous linear controllers. In addition, it is capable of rejecting perturbations in initial conditions leading to both thermal excursion and quenching, and its effectiveness does not depend on whether the operating point is an ignition or a subignition point.The controller designed ensures setpoint regulation for the energy and plasma parameter β with robustness against uncertainties in the confinement times for different species. Hence, the controller can increase or decrease β, modify the power, the temperature or the density, and go from a subignition to an ignition point and vice versa

  19. The ARIES tokamak fusion reactor study

    International Nuclear Information System (INIS)

    Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

    1989-01-01

    The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

  20. The need and prospects for improved fusion reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  1. HYLIFE-II inertial confinement: Fusion power plant design

    International Nuclear Information System (INIS)

    Moir, R.W.

    1990-01-01

    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 2 BeF 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·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. 16 refs., 6 figs., 2 tabs

  2. High density plasmas formation in Inertial Confinement Fusion and Astrophysics

    International Nuclear Information System (INIS)

    Martinez-Val, J. M.; Minguez, E.; Velarde, P.; Perlado, J. M.; Velarde, G.; Bravo, E.; Eliezer, S.; Florido, R.; Garcia Rubiano, J.; Garcia-Senz, D.; Gil de la Fe, J. M.; Leon, P. T.; Martel, P.; Ogando, F.; Piera, M.; Relano, A.; Rodriguez, R.; Garcia, C.; Gonzalez, E.; Lachaise, M.; Oliva, E.

    2005-01-01

    In inertially confined fusion (ICF), high densities are required to obtain high gains. In Fast Ignition, a high density, low temperature plasma can be obtained during the compression. If the final temperature reached is low enough, the electrons of the plasma can be degenerate. In degenerate plasmas. Bremsstrahlung emission is strongly suppressed an ignition temperature becomes lower than in classical plasmas, which offers a new design window for ICF. The main difficulty of degenerate plasmas in the compression energy needed for high densities. Besides that, the low specific heat of degenerate electrons (as compared to classical values) is also a problem because of the rapid heating of the plasma. Fluid dynamic evolution of supernovae remnants is a very interesting problem in order to predict the thermodynamical conditions achieved in their collision regions. Those conditions have a strong influence in the emission of light and therefore the detection of such events. A laboratory scale system has been designed reproducing the fluid dynamic field in high energy experiments. The evolution of the laboratory system has been calculated with ARWEN code, 2D Radiation CFD that works with Adaptive Mesh Refinement. Results are compared with simulations on the original system obtained with a 3D SPH astrophysical code. New phenomena at the collision plane and scaling of the laboratory magnitudes will be described. Atomic physics for high density plasmas has been studied with participation in experiments to obtain laser produced high density plasmas under NLTE conditions, carried out at LULI. A code, ATOM3R, has been developed which solves rate equations for optically thin plasmas as well as for homogeneous optically thick plasmas making use of escape factors. New improvements in ATOM3R are been done to calculate level populations and opacities for non homogeneous thick plasmas in NLTE, with emphasis in He and H lines for high density plasma diagnosis. Analytical expression

  3. Nonthermal fusion reactor concept based on Hall-effect magnetohydrodynamics plasma theory

    International Nuclear Information System (INIS)

    Witalis, E.A.

    1988-01-01

    The failure of magnetic confinement controlled thermonuclear fusion research to achieve its goal is attributed to its foundation on the incomplete MHD plasma description instead of the more general HMHD (Hall-effect magnetohydrodynamics) theory. The latter allows for a certain magnetic plasma self-confinement under described stringent conditions. A reactor concept based on the formation, acceleration, and forced disintegration of magnetized whirl structures, plasmoids, is proposed. The four conventional MHD theory objections, i.e., absence of dynamo action, fast decay caused by resistivity, non-existence of magnetic self-confinement, and negligible non-thermal fusion yield, are shown not to apply. Support for the scheme from dense plasma focus research is pointed out. (orig.) [de

  4. Symbiosis of near breeder HTR's with hybrid fusion reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1978-07-01

    In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

  5. Simulation of transition dynamics to high confinement in fusion plasmas

    DEFF Research Database (Denmark)

    Nielsen, Anders Henry; Xu, G. S.; Madsen, Jens

    2015-01-01

    The transition dynamics from the low (L) to the high (H) confinement mode in magnetically confined plasmas is investigated using a first-principles four-field fluid model. Numerical results are in agreement with measurements from the Experimental Advanced Superconducting Tokamak - EAST...

  6. Proceedings of the US-Japan workshop and the satellite meeting of ITC-9 on physics of high beta plasma confinement in innovative fusion system

    International Nuclear Information System (INIS)

    Goto, Seiichi; Yoshimura, Satoru

    1999-04-01

    The US-Japan Workshop on Physics of High Beta Plasma Confinement in Innovative Fusion System was held jointly with the Satellite Meeting of ITC-9 at National Institute for Fusion Science (NIFS), Toki-city during December 14-15, 1998. This proceedings book includes the papers of the talks given at the workshop. These include: Theoretical analysis on the stability of field reversed configuration (FRC) plasmas; Theory and Modeling of high β plasmas; Recent progressive experiments in high β systems; Formation of high β plasmas using merging phenomenon; Theory and Modeling of a FRC Fusion Reactor. The 15 papers are indexed individually. (J.P.N.)

  7. Proceedings of the US-Japan workshop and the satellite meeting of ITC-9 on physics of high beta plasma confinement in innovative fusion system

    Energy Technology Data Exchange (ETDEWEB)

    Goto, Seiichi; Yoshimura, Satoru [eds.

    1999-04-01

    The US-Japan Workshop on Physics of High Beta Plasma Confinement in Innovative Fusion System was held jointly with the Satellite Meeting of ITC-9 at National Institute for Fusion Science (NIFS), Toki-city during December 14-15, 1998. This proceedings book includes the papers of the talks given at the workshop. These include: Theoretical analysis on the stability of field reversed configuration (FRC) plasmas; Theory and Modeling of high {beta} plasmas; Recent progressive experiments in high {beta} systems; Formation of high {beta} plasmas using merging phenomenon; Theory and Modeling of a FRC Fusion Reactor. The 15 papers are indexed individually. (J.P.N.)

  8. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  9. Potential mirror concepts for radiation testing of fusion reactor materials

    International Nuclear Information System (INIS)

    Miley, G.H.

    1977-01-01

    Studies under the University of Illinois PROMETHEUS (Plasma Reactor Optimized for Materials Experimentation for Thermonuclear Energy Usage) project are described that started in 1971 with the realization that a practical fusion-plasma neutron source was feasible with a net-power input (rather than production). The basic objectives were similar to those in later FERF (Fusion Engineering Research Facility) studies: namely, to maximize the neutron flux and usable experimental volume; to include the flexibility to handle a variety of both materials and engineering experiments; to minimize capital and operating costs; and to utilize near- term technology. The PROMETHEUS design provides a neutron flux of approximately 5x10 14 n/cm 2 s by injection of approximately 30 MW of neutral-beams into a 20 cm radius mirror-confined plasma. Charge-exchange bombardment of the first wall is viewed as a key problem in the design and is discussed in some detail. To gain yet higher neutron fluxes for accelerated testing, two alternate designs have been studied: a 'Twin-beam' injection device and a field reversed mirror concept. The latter potentially offers fluxes approaching 10 16 n/cm 2 s but involves more speculative technology. (Auth.)

  10. Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

  11. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  12. Advanced real-time control systems for magnetically confined fusion plasmas

    International Nuclear Information System (INIS)

    Goncalves, B.; Sousa, J.; Fernandes, H.; Rodrigues, A.P.; Carvalho, B.B.; Neto, A.; Varandas, C.A.F.

    2008-01-01

    Real-time control of magnetically confined plasmas is a critical issue for the safety, operation and high performance scientific exploitation of the experimental devices on regimes beyond the current operation frontiers. The number of parameters and the data volumes used for the plasma properties identification scale normally not only with the machine size but also with the technology improvements, leading to a great complexity of the plant system. A strong computational power and fast communication infrastructure are needed to handle in real-time this information, allowing just-in-time decisions to achieve the fusion critical plasma conditions. These advanced control systems require a tiered infrastructure including the hardware layer, the signal-processing middleware, real-time timing and data transport, the real-time operating system tools and drivers, the framework for code development, simulation, deployment and experiment parameterization and the human real-time plasma condition monitoring and management. This approach is being implemented at CFN by offering a vertical solution for the forthcoming challenges, including ITER, the first experimental fusion reactor. A given set of tools and systems are described on this paper, namely: (i) an ATCA based hardware multiple-input-multiple-output (MIMO) platform, PCI and PCIe acquisition and control modules; (ii) FPGA and DSP parallelized signal processing algorithms; (iii) a signal data and event distribution system over a 2.5/10Gb optical network with sub-microsecond latencies; (iv) RTAI and Linux drivers; and (v) the FireSignal, FusionTalk, SDAS FireCalc application tools. (author)

  13. Correlation between potential well structure and neutron production in inertial electrostatic confinement fusion

    International Nuclear Information System (INIS)

    Ohnishi, M.; Yamamoto, Y.; Yoshikawa, K.; Sato, K.H.

    1997-01-01

    The electrostatic potential well in inertial electrostatic confinement (IEC) is studied using two approaches. First, the equilibrium potential profile is obtained by solving the charge neutrality condition, i.e. n i n e , assuming the appropriate distribution functions for the ions and the electrons. The formation of a double well structure is demonstrated, with a depth depending upon the ratio between the focus radii of the electrons and the ions. The correlations between the well depth and the volume integrated neutron production due to deuterium-deuterium (DD) reactions are obtained. Second, in order to study the stability of the well, the dynamic behaviours of the potential well are calculated by performing time advancing numerical simulations on the basis of the particle in cell method. Single, double and triple wells, depending on the amount of injected ion current, are observed to be formed for ions with a monoenergetic distribution. The well in the centre of the multiwell structure is unstable and oscillates with a periods much longer than the inverse ion plasma frequency. A double well structure can be formed even for ions with a spread out energy distribution when the ion current is larger than the threshold value. The time averaged neutron production by DD fusion events is proportional to a power of the ion current involved in forming the double well structure. The results strongly suggest that the high neutron production rate should be attributed to not only the well depth but also the unstable behaviour of the potential, i.e. the intermittent peaking of the density in the centre region. A numerical simulation reveals that IEC possesses a favourable dependence of fusion reactions on the injected ion current for the application to a neutron source or a fusion reactor. (author). 9 refs, 9 figs

  14. Blanket materials for DT fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1981-01-01

    This paper presents an overview of the critical materials issues that must be considered in the development of a tritium breeding blanket for a tokamak fusion reactor that operates on the D-T-Li fuel cycle. The primary requirements of the blanket system are identified and the important criteria that must be considered in the development of blanket technology are summarized. The candidate materials are listed for the different blanket components, e.g., breeder, coolant, structure and neutron multiplier. Three blanket concepts that appear to offer the most potential are: (1) liquid-metal breeder/coolant, (2) liquid-metal breeder/separate coolant, and (3) solid breeder/separate coolant. The major uncertainties associated with each of the design concepts are discussed and the key materials R and D requirements for each concept are identified

  15. Burning nuclear wastes in fusion reactors

    International Nuclear Information System (INIS)

    Meldner, H.W.; Howard, W.M.

    1979-01-01

    A study was made up of actinide burn-up in ICF reactor pellets; i.e. 14 Mev neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet fuel region burn-up is safety: only milligrams of highly toxic and active material need to be present in the fusion chamber, whereas blanket burn-up requires the continued presence of tons of actinides in a small volume. The actinide data tables required for Monte Carlo calculations of the burn-up of /sup 241/Am and /sup 243/Am are discussed in connection with a study of the sensitivity to cross section uncertainties. More accurate and complete cross sections are required for realistic quantitative calculations. 13 refs

  16. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

  17. Historic overview of inertial confinement fusion: What have we learned

    International Nuclear Information System (INIS)

    Glass, A.J.

    1986-01-01

    Although laser fusion has been the subject of research since the early 1960s, it has only been intensively studied for about 14 years. During that time, substantive advances have been made in our understanding of the complex physics of laser-heated plasmas, in the development of sophisticated diagnostic instrumentation, and in the technology of fusion targets and inertial fusion drivers. These advances will be reviewed. Of equal importance are the lessons learned in the economic and political arenas. These lessons may be of greater significance for scientific endeavors in other fields of research. The economic and political issues surrounding inertial fusion research will be discussed. Possible future directions for inertial fusion development will be presented

  18. Classical impurity ion confinement in a toroidal magnetized fusion plasma.

    Science.gov (United States)

    Kumar, S T A; Den Hartog, D J; Caspary, K J; Magee, R M; Mirnov, V V; Chapman, B E; Craig, D; Fiksel, G; Sarff, J S

    2012-03-23

    High-resolution measurements of impurity ion dynamics provide first-time evidence of classical ion confinement in a toroidal, magnetically confined plasma. The density profile evolution of fully stripped carbon is measured in MST reversed-field pinch plasmas with reduced magnetic turbulence to assess Coulomb-collisional transport without the neoclassical enhancement from particle drift effects. The impurity density profile evolves to a hollow shape, consistent with the temperature screening mechanism of classical transport. Corroborating methane pellet injection experiments expose the sensitivity of the impurity particle confinement time to the residual magnetic fluctuation amplitude.

  19. Intelligible seminar on fusion reactors. (12) Next step toward the realization of fusion reactors. Future vision of fusion energy research and development

    International Nuclear Information System (INIS)

    Okano, Kunihiko; Kurihara, Kenichi; Tobita, Kenji

    2006-01-01

    In the last session of this seminar the progress of research and development for the realization of fusion reactors and future vision of fusion energy research and development are summarized. The some problems to be solved when the commercial fusion reactors would be realized, (1) production of deuterium as the fuel, (2) why need the thermonuclear reactors, (3) environmental problems, and (4) ITER project, are described. (H. Mase)

  20. Early fusion reactor neutronic calculations: A reevaluation

    International Nuclear Information System (INIS)

    Perry, R.T.

    1996-01-01

    Several fusion power plant design studies were made at a number of universities and laboratories in the late 1960s and early 1970s. These studies included such designs as the Princeton Plasma Physics Laboratory Fusion Power Plan and the University of Wisconsin UWMAK-I Reactor Neutronic analyses of the blankets and shields were part of the studies. During this time there were dissertations written on neutronic analysis systems and the results of neutronic analysis on several blanket and shield designs. The results were presented in the literature. Now in the fifth decade of fusion research, investigators often return to the earlier analyses for the neutronic results that are applicable to current blanket and shield designs, with the idea of using the older work as a basis for the new. However, the analyses of the past were made with cross-section data sets that have long been replaced with more modern versions. In addition, approximations were often made to the cross sections used because more exact data were not available. Because these results are used as guides, it is important to know if they are reproducible using more modern data. In this paper, several of the neutronic calculations made in the early studies are repeated using the MATXS-11 data library. This library is the ENDF/B-VI version of the MATXS-5 library. The library has 80 neutron groups. Tritium breeding ratios, heating rates, and fluxes are calculated and compared. This transport code used here is the one- dimensional S n code, ONEDANT. It is important to note that the calculations here are not to be considered as benchmarks because parameter and sensitivity studies were not made. They are used only to see if the results of older calculations are in reasonable agreement with a more modern library

  1. The LOFA analysis of fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Z.-C.; Xie, H.

    2014-01-01

    The fusion-fission hybrid energy reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc, with the fusion neutron source striking the subcritical blanket. The passive safety system, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (author)

  2. The technology benefits of inertial confinement fusion research

    International Nuclear Information System (INIS)

    Powell, H.T.

    1999-01-01

    The development and demonstration of inertial fusion is incredibly challenging because it requires simultaneously controlling and precisely measuring parameters at extreme values in energy, space, and time. The challenges range from building megajoule (10 6 J) drivers that perform with percent-level precision to fabricating targets with submicron specifications to measuring target performance at micron scale (10 -6 m) with picosecond (10 -12 s) time resolution. Over the past 30 years in attempting to meet this challenge, the inertial fusion community around the world has invented new technologies in lasers, particle beams, pulse power drivers, diagnostics, target fabrication, and other areas. These technologies have found applications in diverse fields of industry and science. Moreover, simply assembling the teams with the background, experience, and personal drive to meet the challenging requirements of inertial fusion has led to spin-offs in unexpected directions, for example, in laser isotope separation, extreme ultraviolet lithography for microelectronics, compact and inexpensive radars, advanced laser materials processing, and medical technology. The experience of inertial fusion research and development of spinning off technologies has not been unique to any one laboratory or country but has been similar in main research centers in the US, Europe, and Japan. Strengthening and broadening the inertial fusion effort to focus on creating a new source of electrical power (inertial fusion energy [IFE]) that is economically competitive and environmentally benign will yield rich rewards in technology spin-offs. The additional challenges presented by IFE are to make drivers affordable, efficient, and long-lived while operating at a repetition rate of a few Hertz; to make fusion targets that perform consistently at high-fusion yield; and to create target chambers that can repetitively handle greater than 100-MJ yields while producing minimal radioactive by

  3. The international thermonuclear experimental reactor and the future of nuclear fusion energy

    International Nuclear Information System (INIS)

    Pan Chuanhong

    2010-01-01

    Energy shortage and environmental problems are now the two largest challenges for human beings. Magnetic confinement nuclear fusion, which has achieved great progress since the 1990's, is anticipated to be a way to realize an ideal source of energy in the future because of its abundance, environmental compatibility, and zero carbon release. Exemplified by the construction of the International Thermonuclear Experimental Reactor (ITER), the development of nuclear fusion energy is now in its engineering phase, and should be realized by the middle of this century if all objectives of the ITER project are met. (author)

  4. The heat transport system and plant design for the HYLIFE-2 fusion reactor

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1990-01-01

    HYLIFE is the name given to a family of self-healing liquid-wall reactor concepts for inertial confinement fusion. This HYLIFE-II concept employs the molten salt, Flibe, for the liquid jets instead of liquid lithium used in the original HYLIFE-I study. A preliminary conceptual design study of the heat transport system and the balance of plant of the HYLIFE-II fusion power plant is described in this paper with special emphasis on a scoping study to determine the best intermediate heat exchanger geometry and flow conditions for minimum cost of electricity. 11 refs., 8 figs

  5. Major features of a mirror fusion--fast fission hybrid reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1974-01-01

    A conceptual design was made of a fusion-fission reactor. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and sustained by hot neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and is cooled by helium. It was shown how the reactor can be built using essentially present day construction technology and how the uranium bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel of which approximately 1200 kg of plutonium are produced each year along with the approximately 750 MW of electricity. (U.S.)

  6. Application of Bondarenko formalism to fusion reactors

    International Nuclear Information System (INIS)

    Soran, P.D.; Dudziak, D.J.

    1975-01-01

    The Bondarenko formalism used to account for resonance self-shielding effects (temperature and composition) in a Reference Theta-Pinch Reactor is reviewed. A material of interest in the RTPR blanket is 93 Nb, which exhibits a large number of capture resonance in the energy region below 800 keV. Although Nb constitutes a small volume fraction of the blanket, its presence significantly affects the nucleonic properties of the RTPR blanket. The effects of self-shielding in 93 Nb on blanket parameters such as breeding ratio, total afterheat, radioactivity, magnet-coil heating and total energy depositions have been studied. Resonance self-shielding of 93 Nb, as compared to unshielded cross sections, will increase tritium breeding by approximately 7 percent in the RTPR blanket and will decrease blanket radioactivity, total recoverable energy, and magnet-coil heating. Temperature effects change these parameters by less than 2 percent. The method is not restricted to the RTPR, as a single set of Bondarenko f-factors is suitable for application to a variety of fusion reactor designs

  7. Conceptual design of imploding liner fusion reactors

    International Nuclear Information System (INIS)

    Turchi, P.J.; Robson, A.E.

    1976-01-01

    The basic new ingredient is the concept of rotationally stabilized liquid metal liners accelerated with free pistons. The liner motion is constrained on its outer surface by the pistons, laterally by channel walls, during acceleration, and on its inner surface, where megagauss field levels are attained by the centrifugal motion of the liner material. In this way, stable, reversible motion of the liner should be possible, permitting repetitive, pulsed operation at interior pressures far greater than can be allowed in static conductor systems. Such higher operating pressures permit the use of simple plasma geometries, such as theta pinches, with greatly reduced dimensions. Furthermore, the implosion of thick, lithium-bearing liners with large radial compression ratios inherently provides the plasma with a surrounding blanket of neutron absorbing liquid metal, thereby substantially reducing the problems of induced radioactivity and first wall damage that haunt conventional fusion reactor designs. The following article discusses the basic operation of liner reactors and several important features influencing their design

  8. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  9. On the possibility of D-3He fusion based on fast - ignition inertial confinement scheme

    International Nuclear Information System (INIS)

    Nakao, Y.; Hegi, K.; Ohmura, T.; Katsube, M.; Kudo, K.; Johzaki, T.; Ohta, M.

    2007-01-01

    Although nuclear fusion reactors adopting D 3 He fuel could provide many advantages, such as low neutron generation and efficient conversion of output fusion energy, the achievement of ignition is a difficult problem. It is therefore of particular importance to find some methods or schemes that relax the ignition requirements. In inertial confinement scheme, the use of pure D 3 He fuel is impractical because of the excessive requirement on driver energy. A small amount of DT fuel as 'igniter' is hence indispensable [1]. Our previous burn simulation [1] for DT/D 3 He fuels compressed to 5000 times the liquid density showed that substantial fuel gains (∼500) are obtained from fuels having parameters ρ R D T = 3 g/cm 2 , ρ R t otal 14 g/cm 2 and a central spark temperature of 5 keV. The driver energy needed to achieve these gains is estimated to be ∼30 MJ when the coupling efficiency is 10%; in this case the target gain is ∼50. Subsequent implosion simulation [2], however, showed that after void closure the central DT fuel is ignited while the bulk of the main D 3 He fuel is still imploding with high velocities. This pre-ignition of DT fuel leads to a low compression of the main fuel and prevents the DT/D 3 He fuel from obtaining required gain. These difficulties associated with the pre-ignition of DT fuel could be resolved or mitigated if other ignition schemes such as fast-ignition [3] and/or impact-ignition [4] are adopted, because in these schemes compression and ignition phases are separated. Furthermore, the reduction of driver energy can be expected. In the present study, we examine the possibility of D 3 He fusion in the fast-ignition scheme. Simulations until now have been made for a DT/D 3 He fuel compressed to 5000 times the liquid density by using FIBMET (2D fusion ignition and burning code) [5] and a newly developed neutron diffusion code. DT igniter was assumed to be placed at a corner of the compressed fuel. The ρ R values and temperature of

  10. Controlled Nuclear Fusion by Magnetic Confinement and ITER

    CERN Document Server

    CERN. Geneva. Audiovisual Unit; Alvarez-Gaumé, Luís

    2005-01-01

    For may years harnessing fusion energy was considered the final solution to the world's energy crisis. ITER is the last step in the elusive quest. This presentation will provide in its various acientific, technological and political aspects.

  11. Fusion reactor start-up without an external tritium source

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, S., E-mail: Shanliang.Zheng@ccfe.ac.uk; King, D.B.; Garzotti, L.; Surrey, E.; Todd, T.N.

    2016-02-15

    Highlights: • Investigated the feasibility (including plasma physics, neutronics and economics) of starting a fusion reactor from running pure D–D fusion reactor to gradually move towards the D–T operation. • Proposed building up tritium from making use of neutrons generated by D–D fusion reactions. • Studied plasma physics feasibility for pure D–D operation and provided consistent fusion power and neutron yield in the plasma with different mixture of deuterium and tritium. • Discussed the economics aspect for operating a pure D–D fusion reactor towards a full-power D–T fusion reactor. - Abstract: It has long been recognised that the shortage of external tritium sources for fusion reactors using D–T, the most promising fusion fuel, requires all such fusion power plants (FPP) to breed their own tritium. It is also recognised that the initial start-up of a fusion reactor will require several kilograms of tritium within a scenario in which radioactive decay, ITER and subsequent demonstrator reactors are expected to have consumed most of the known tritium stockpile. To circumvent this tritium fuel shortage and ultimately achieve steady-state operation for a FPP, it is essential to first accumulate sufficient tritium to compensate for loss due to decay and significant retention in the materials in order to start a new FPP. In this work, we propose to accumulate tritium starting from D–D fusion reactions, since D exists naturally in water, and to gradually build up the D–T plasma targeted in fusion reactor designs. There are two likely D–D fusion reaction channels, (1) D + D → T + p, and (2) D + D → He3 + n. The tritium can be generated via the reaction channel ‘(1)’ and the 2.45 MeV neutrons from ‘(2)’ react with lithium-6 in the breeding blanket to produce more tritium to be fed back into plasma fuel. Quantitative evaluations are conducted for two blanket concepts to assess the feasibility and suitability of this approach to FPP

  12. Self-sustaining nuclear pumped laser-fusion reactor experiment

    International Nuclear Information System (INIS)

    Boody, F.P.; Choi, C.K.; Miley, G.H.

    1977-01-01

    The features of a neutron feedback nuclear pumped (NFNP) laser-fusion reactor equipment were studied with the intention of establishing the feasibility of the concept. The NFNP laser-fusion concept is compared schematically to electrically pumped laser fusion. The study showed that, once a method of energy storage has been demonstrated, a self-sustaining fusion-fission hybrid reactor with a ''blanket multiplication'' of two would be feasible using nuclear pumped Xe F* excimer lasers having efficiencies of 1 to 2 percent and D-D-T pellets with gains of 50 to 100

  13. Power plant production of inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.; Johnson, W.L.

    1979-01-01

    Many of the current techniques for fabricating experimental targets appear to be directly extendable to the high-rate, low-cost production of reactor targets. This report describes several new techniques that, in conjunction with the expansion of existing techniques, can constitute a target factory. We have evaluated this concept on the basis of a generalized reactor target design and the projected specifications of reactor-grade targets

  14. Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

    International Nuclear Information System (INIS)

    Demachi, K.; Yoshida, Y.; Miya, K.

    1994-01-01

    For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

  15. HIBALL - a conceptual heavy ion beam driven fusion reactor study. Vol. 1

    International Nuclear Information System (INIS)

    Badger, B.; El-Guebaly, L.; Engelstad, R.; Hassanein, A.; Klein, A.; Kulcinski, G.; Larsen, E.; Lee, K.; Lovell, E.; Moses, G.

    1981-12-01

    A preliminary concept for a heavy-ion beam driven inertial confinement fusion power plant is presented. The high repetition rate of the RF accelerator driver is utilized to serve four reactor chambers alternatingly. In the chambers a novel first-wall protection scheme is used. At a target gain of 83 the total net electrical output is 3.8 GW. The recirculating power fraction is below 15%. The main goal of the comprehensive HIBALL study (which is continuing) is to demonstrate the compatibility of the design of the driver, the target and the reactor chambers. Though preliminary, the present dessign is essentially self-consistent. Tentative cost estimates are given. The costs compare well with those found in similar studies on other types of fusion reactors. (orig.) [de

  16. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

    A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  17. A probabilistic methodology for the design of radiological confinement of tokamak reactors

    International Nuclear Information System (INIS)

    Golinescu, Ruxandra P.; Morosan, Florinel; Kazimi, Mujid S.

    1997-01-01

    A methodology using probabilistic risk assessment techniques is proposed for evaluating the design of multiple confinement barriers for a fusion plant within the context of a limited allowable risk. The methodology was applied to the reference design of the International Thermonuclear Experimental Reactor (ITER). Accident sequence models were developed to determine the probability of radioactive releases from each confinement barrier. The current ITER design requirements, that set environmental radioactive release limits for individual event sequences grouped in categories by frequency, is extended to derive a limit on the plant overall risk. This avoids detailed accounting for event uncertainties in both frequency and consequence. Thus, an analytical form for a limit line is derived as a complementary cumulative frequency of permissible radioactive releases to the environment. The line can be derived using risk aversion of the designer's own choice. By comparing the releases from each confinement barrier against this limit line, a decision can be made about the number of barriers required to comply with the design requirements. A decision model using multi-attribute utility function theory was constructed to help the designer in choosing the type of the tokamak building while considering preferences for attributes such as construction cost, project completion time, technical feasibility and public attitude. Sensitivity analysis on some of the relevant parameters in the model was performed

  18. Engineering the fusion reactor first wall

    International Nuclear Information System (INIS)

    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. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    Jung, J.

    1983-12-01

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

  20. Control of tritium permeation through fusion reactor strucural materials

    International Nuclear Information System (INIS)

    Maroni, V.A.

    1978-01-01

    The intention of this paper is to provide a brief synopsis of the status of understanding and technology pertaining to the dissolution and permeation of tritium in fusion reactor materials. The following sections of this paper attempt to develop a simple perspective for understanding the consequences of these phenomena and the nature of the technical methodology being contemplated to control their impact on fusion reactor operation. Considered in order are: (1) the occurrence of tritium in the fusion fuel cycle, (2) a set of tentative criteria to guide the analysis of tritium containment and control strategies, (3) the basic mechanisms by which tritium may be released from a fusion plant, and (4) the methods currently under development to control the permeation-related release mechanisms. To provide background and support for these considerations, existing solubility and permeation data for the hydrogen isotopes are compared and correlated under conditions to be expected in fusion reactor systems

  1. Plutonium-239 production rate study using a typical fusion reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Havasi, H.; Amin-Mozafari, M.

    2008-01-01

    The purpose of the present paper is to compute fissile 239 Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m 2 ) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate 239 Pu production rate. Produced 239 Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, 239 Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type

  2. Plutonium-239 production rate study using a typical fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Havasi, H.; Amin-Mozafari, M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

    2008-05-15

    The purpose of the present paper is to compute fissile {sup 239}Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m{sup 2}) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate {sup 239}Pu production rate. Produced {sup 239}Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, {sup 239}Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type.

  3. Alternative fusion concepts and the prospects for improved reactors

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1985-01-01

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

  4. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  5. Health physics aspects of activation products from fusion reactors

    International Nuclear Information System (INIS)

    Shoup, R.L.; Poston, J.W.; Easterly, C.E.; Jacobs, D.G.

    1975-01-01

    A review of the activation products from fusion reactors and their attendant impacts is discussed. This includes a discussion on their production, expected inventories, and the status of metabolic data on these products

  6. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  7. Studies in the evolution of hydrodynamic instabilities and their role in inertial confinement fusion

    International Nuclear Information System (INIS)

    Shvarts, D.; Oron, D.; Sadot, O.

    2001-01-01

    Hydrodynamic instabilities, such as the Rayleigh-Taylor and Richtmyer-Meshkov instabilities, have a central role when trying to achieve net thermonuclear fusion energy via the method of Inertial Confinement Fusion. We shall review recent theoretical, numerical and experimental work that describes the evolution of two- and three-dimensional perturbations. Finally, the effects of these perturbation on the ignition conditions, using new self-similar solutions for perturbed burn wave propagation will be discussed. (author)

  8. Present status of inertial confinement fusion in Japan

    International Nuclear Information System (INIS)

    Yamanaka, Chiyoe

    1984-01-01

    The Japanese inertial fusion program has made important progress towards implosion fusion process and the technical development required for realizing the breakeven of inertial fusion energy. The key issues for the ICF research are the development of a high power driver, the pertinent pellet design for implosion by a super computer code, and the diagnostics of implosion process with high space and time resolution. The Institute of Laser Engineering (ILE), Osaka University, is the central laboratory for ICF research in Japan. The ILE Osaka has advanced the Kongo Project aiming at the breakeven of inertial fusion since 1980, and as the first phase, the Gekko 12 Nd glass laser of 20 kJ having 12 beams was constructed. The ILE has also the Lekko 8 CO 2 laser and the Reiden 4 light ion beam machine. In the second phase, a 100 kJ class driver will be provided. At the ILE, rare gas halide lasers such as KrF and ArF have been investigated. Laser plasma coupling, the scaling law for implosion pressure, the invention of a new type target ''Cannonball'', and the development of computer codes are described. Also the activities in universities, government laboratories and industrial companies are reported. (Kako, I.)

  9. Studies on the feasibility of heavy ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    1983-05-01

    This Annual Report summarizes the scientific results of work carried out in 1982 in the framework of a feasibility study for inertial confinement fusion with heavy ion beams funded by the German Ministry of Research and Technology. The principal aim of this basic research program is the investigation of key problems and the identification of critical issues of the heavy ion ICF concept in the fields of accelerator research, atomic physics, target physics, and reactor design. The research is carried out by about ten working groups at various research centers and universities. One of the highlights in 1982 was a symposium held end of March at GSI which focussed on a critical analysis of the HIBALL accelerator concept. Whereas technical issues and hardware parameters were found feasible the beam dynamics in the storage rings turned out to be beyond the so far believed stability limits. As a consequence a revised accelerator scenario based on a lower charge state and a higher linac current has been investigated during the last year. First considerations were made on an experimental facility necessary for the study of high-intensity beam dynamics and of beam target interaction. Experimental studies of this kind will be of increasing importance for the future of the project. (orig.) [de

  10. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  11. Influence of Impurities on the Fuel Retention in Fusion Reactors

    OpenAIRE

    Reinhart, Michael

    2015-01-01

    The topic of this thesis is the influence of plasma impurities on the hydrogen retentionin metals, in the scope of plasma-wall-interaction research for fusion reactors.This is addressed experimentally and by modelling. The mechanisms of the hydrogenretention are influenced by various parameters like the wall temperature, ionenergy, flux and fluence as well as the plasma composition. The plasma compositionis a relevant factor for hydrogen retention in fusion reactors, as their plasma willalso ...

  12. Inertial confinement fusion with direct electric generation by magnetic flux comparession

    International Nuclear Information System (INIS)

    Lasche, G.P.

    1983-01-01

    A high-power-density laser-fusion-reactor concept in investigated in which directed kinetic enery imparted to a large mass of liquid lithium--in which the fusion target is centrally located--is maximized. In turn, this kinetic energy is converted directly to electricity with, potentially, very high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the concept maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall can be many orders of magnitude less than is typical of D-T fusion reactor concepts

  13. An analysis of the estimated capital cost of a fusion reactor

    International Nuclear Information System (INIS)

    Hollis, A.A.

    1981-06-01

    The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

  14. An analysis of the estimated capital cost of a fusion reactor

    International Nuclear Information System (INIS)

    Hollis, A.A.; Evans, L.S.

    1981-01-01

    The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

  15. Tritium-related materials problems in fusion reactors

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1976-01-01

    Pressing materials problems that must be solved before tritium can be used to produce energy economically in fusion reactors are discussed. The following topics are discussed: (1) breeding tritium, (2) recovering bred tritium, (3) containing tritium, (4) fuel recycling, and (5) laser-fusion fueling

  16. Thermal aspects of a superconducting coil for fusion reactor

    International Nuclear Information System (INIS)

    Yeh, H.T.

    1975-01-01

    Computer models are used to simulate both localized and extensive thermal excursions in a large superconducting magnet for fusion reactor. Conditions for the failure of fusion magnet due to thermal excursion are delineated. Designs to protect the magnet against such thermal excursion are evaluated

  17. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

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

    1982-12-01

    Results of a conceptual design study of a 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

  18. Communication links for fusion reactor maintenance operations

    International Nuclear Information System (INIS)

    Van Uffelen, M.

    2005-01-01

    Different architectures are envisaged for data transmission with fibre optic links in a radiation environment, as proposed in literature for both space and high energy physics applications. Their needs and constraints differ from those encountered for maintenance tasks in the future ITER environment, not only in terms of temperature and radiation levels, but also with respect to transmission speed requirements. Our approach attempts to limit the use of radiation-sensitive electronics for transmission of both digital and/or analogue data to the control room, using glass fibres as transport medium. We therefore assessed the radiation behaviour of a cost-effective fibre optic transmitter at 850 nm, consisting of a PWM (pulse width modulator), a radiation tolerant current driver (previously developed at SCK-CEN) and a VCSEL (Vertical-Cavity Surface Emitting Laser assembly, up to 10 MGy at 60 degrees Celsius. The PWM enables to transform an analogue sensor signal into a pseudo numerical signal, with a pulse width proportional to the incoming signal. The main objective of this task is to contribute to the major design of the maintenance equipment and strategy needed for the remote replacement of the divertor system in the future ITER fusion reactor, with particular attention to the implications of radiation hardening rules and recommendations. Next to the radiation assessment studies of remote handling tools, including actuators and sensors, we also develop radiation tolerant communication links with multiplexing capabilities

  19. Safety vessels for explosive fusion reactor

    International Nuclear Information System (INIS)

    Mineev, V.

    1994-01-01

    The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

  20. Knudsen and inverse Knudsen layer effect on tail ion distribution and fusion reactivity in inertial confinement fusion targets

    Science.gov (United States)

    McDevitt, C. J.; Tang, X.-Z.; Guo, Z.; Berk, H. L.

    2014-10-01

    A series of reduced models are used to study the fast ion tail in the vicinity of a transition layer between plasmas at disparate temperatures and densities, which is typical of the gas-pusher interface in inertial confinement fusion targets. Emphasis is placed on utilizing progressively more comprehensive models in order to identify the essential physics for computing the fast ion tail at energies comparable to the Gamow peak. The resulting fast ion tail distribution is subsequently used to compute the fusion reactivity as a function of collisionality and temperature. It is found that while the fast ion distribution can be significantly depleted in the hot spot, leading to a reduction of the fusion reactivity in this region, a surplus of fast ions is present in the neighboring cold region. The presence of this fast ion surplus in the neighboring cold region is shown to lead to a partial recovery of the fusion yield lost in the hot spot.