WorldWideScience

Sample records for charging fusion reactor

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

    International Nuclear Information System (INIS)

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

  2. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  3. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  4. The fusion reactor

    International Nuclear Information System (INIS)

    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.)

  5. Mirror fusion reactors

    International Nuclear Information System (INIS)

    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

  6. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    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

  7. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    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

  8. Fusion reactor materials

    International Nuclear Information System (INIS)

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  9. Fusion reactor safety

    International Nuclear Information System (INIS)

    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

  10. Fusion reactor materials

    International Nuclear Information System (INIS)

    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

  11. Compact fusion reactors

    CERN Document Server

    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.

  12. 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.

  13. Colliding Beam Fusion Reactors

    Science.gov (United States)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

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

  14. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    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

  15. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    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

  16. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. Fusion reactor materials

    International Nuclear Information System (INIS)

    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

  19. Spherical torus fusion reactor

    Science.gov (United States)

    Peng, Yueng-Kay M.

    1989-01-01

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

  20. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    In the middle of 21st century, the population on the earth is expected to double, and the energy that mankind consumes to triple. The nuclear fusion which is said the ultimate energy source for mankind is expected to solve this energy problem. As for fusion reactors, fuel materials exist inexhaustibly, distributing evenly, they have high safety in principle, the product of burning is harmless nonradioactive substance that does not require the treatment and disposal, and the attenuation of induced radioactivity due to neutrons is quick and the effect to global environment is little. The basic plan of second stage nuclear fusion research and development was decided in 1975, aiming at attaining the critical plasma condition. JT-60 has attained it in 1987. The project of international thermonuclear fusion experimental reactor (ITER) was started, and the conceptual design was carried out. Under such background, the third stage basic plan was decided in 1992, and its objective is self ignition condition, long time burning and the basis of the reactor engineering technology. The engineering design of the ITER is investigated. (K.I.)

  1. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

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

  2. 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)

  3. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    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

  4. Migma fusion reactor

    International Nuclear Information System (INIS)

    Collisions of atomic and molecular ions of like charge are produced in a device including a magnetic field which decreases with the radial distance from its central axis and increases with the distance along the central axis from its center plane. Injected accelerated ion beams are mixed in an organized manner in precessing orbits designed to make them collide head-on or nearly so in the central region of the device continuously and automatically. Ions that have not undergone fusion are continuously and automatically returned by the field to the collision region. The collision probability is further increased by accelerating (rather than heating) the ions to an energy at which the reaction parameter (the product of the fusion cross-secton and the relative ion velocity) is maximized. The atomic nuclei are confined in the device by 'self-trapping' processes. By limiting the injection energy of deuterons to a particular range, it is possible to achieve a breeding effect. Means are presented to maintain the density of the organized ion mixture along with a geometrical configuration of the magnetic field-producing coils and the external electrical fields in such a manner that the charged nuclei resulting from the fusion reactions may have their energy directly converted into electric energy by a decelerating electric potential outside the magnetic field. (LL)

  5. Coatings for fusion reactor environments

    International Nuclear Information System (INIS)

    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

  6. Radioactivity production around the surface of a cooling water pipe in a D-T fusion reactor by sequential charged particle reactions

    International Nuclear Information System (INIS)

    Around the surface of a cooling pipe in a D-T fusion reactor, it is expected that the radioactivity production via what is known as 'Sequential Charged Particle Reaction (SCPR)' would be enhanced by recoiled proton from hydrogen in cooling water. In order to simulate the circumstances, several sheets of foil with a thickness of 50-250 μm were laminated on a polyethylene board for six fusion materials (Fe, Cu, V, Ti, W, Pb). The laminated samples were irradiated with intense D-T neutrons at the fusion neutronics source facility in JAERI. After irradiation, the decay gamma rays emitted from the sequential reaction products (56Co, 65Zn, 51Cr, 48V, 184Re, 206Bi) were measured and the effective cross-sections for producing those were obtained at several positions. The present results indicated that the sequential reaction rate increases prominently as the location becomes closer to hydrogen compounds

  7. Design of a Fast Neutral He Beam System for Feasibility Study of Charge-Exchange Alpha-Particle Diagnostics in a Thermonuclear Fusion Reactor

    CERN Document Server

    Shinto, Katsuhiro; Kitajima, Sumio; Kiyama, Satoru; Nishiura, Masaki; Sasao, Mamiko; Sugawara, Hiroshi; Takenaga, Mahoko; Takeuchi, Shu; Wada, Motoi

    2005-01-01

    For alpha-particle diagnostics in a thermonuclear fusion reactor, neutralization using a fast (~2 MeV) neutral He beam produced by the spontaneous electron detachment of a He- is considered most promising. However, the beam transport of produced fast neutral He has not been studied, because of difficulty for producing high-brightness He- beam. Double-charge-exchange He- sources and simple beam transport systems were developed and their results were reported in the PAC99* and other papers.** To accelerate an intense He- beam and verify the production of the fast neutral He beam, a new test stand has been designed. It consists of a multi-cusp He+

  8. Hydrogen Production in Fusion Reactors

    OpenAIRE

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

    1993-01-01

    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.

  9. Secondary charged particle activation method for measuring the tritium production rate in the breeding blankets of a fusion reactor

    International Nuclear Information System (INIS)

    In this work, a new passive technique has been developed for measuring the tritium production rate in ITER (International Thermonuclear Experimental Reactor) test blanket modules. This method is based on the secondary charged particle activation, in which the irradiated sample contains two main components: a tritium producing target (6Li or 7Li) and an indicator nuclide, which has a relatively high cross-section for an incoming tritium particle (triton). During the neutron irradiation, the target produces a triton, which has sufficiently high energy to cause the so-called secondary charged particle activation on an indicator nuclide. If the product of this reaction is a radioactive nuclide, its activity must be proportional to the amount of generated tritium. A comprehensive set of irradiations were performed at the Training Reactor of the Budapest University of Technology and Economics. The following charged particle reactions were observed and investigated: 27Al(t,p)29Al; 26Mg(t,p)28Mg; 26Mg(t,n)28Al; 32S(t,n)34mCl; 16O(t,n)18F; and 18O(t,α)17N. The optimal atomic ratio of the indicator elements and 6Li was also investigated. The reaction rates were estimated using calculations with the MCNPX Monte Carlo particle transport code. The trend of the measured and the simulated data are in good agreement, although accurate data for triton induced reaction cross-sections cannot be found in the literature. Once the technique is calibrated with a reference LSC (Liquid Scintillation Counting) measurement, a new passive method becomes available for tritium production rate measurements.

  10. Proton Collimators for Fusion Reactors

    Science.gov (United States)

    Miley, George H.; Momota, Hiromu

    2003-01-01

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

  11. Polymer materials for fusion reactors

    International Nuclear Information System (INIS)

    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.)

  12. Magnetic fusion reactor economics

    International Nuclear Information System (INIS)

    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

  13. Charged fusion products in a Tokamak

    International Nuclear Information System (INIS)

    In fusion reactions, charged particules are generated; they are more or less confined in the magnetic fields. Results reachable by charged fusion products analysis justify the work, especially on the large experiments like Tore-Supra or J.E.T., where the power produced may reach a few MW. They can be used as diagnostic for the plasma, and as experimental prediction of the confinement of alpha particules, which is necessary for the reactor. Practical use of a semi-conductor detector on a Tokamak is technically difficult: the problems have been studied on T.F.R. Encouraging results have been obtained, with 3 MeV (D-D) and 15 MeV (D-He3) proton spectra. Calculations on particle trajectories, damping and scattering in Tore Supra are also presented

  14. Materials requirements for fusion reactors

    International Nuclear Information System (INIS)

    Once the physics of fusion devices is understood, one or more experimental power reactors (EPR) are planned which will produce net electrical power. The structural material for the device will probably be a modification of an austenitic stainless steel. Unlike fission reactors, whose pressure boundaries are subjected to no or only light irradiation, the pressure boundary of a fusion reactor is subjected to high atomic displacement-damage and high production rates of transmutation products, e.g., helium and hydrogen. The design data base must include irradiated materials. Since in situ testing to obtain tensile, fatigue, creep, crack-growth, stress-rupture, and swelling data is currently impossible for fusion reactor conditions, a program of service-temperature irradiations in fission reactors followed by postirradiation testing, simulation of fusion conditions, and low-fluence 14 MeV neutron-irradiation tests are planned. For the Demonstration Reactor (DEMO) expected to be built within ten years after theEPR, higher heat fluxes may require the use of refractory metals, at least for the first 20 cm. A partial data base may be provided by high-flux 14 MeV neutron sources being planned. Many materials other than those for structural components will be required in the EPR and DEMO. These include superconducting magnets, insulators, neutron reflectors and shields, and breeding materials. The rest of the device should utilize conventional materials except that portion involved in tritium confinement and recovery

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

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

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

  16. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  17. Compact fusion reactors

    International Nuclear Information System (INIS)

    Compact, high-power-density approaches to fusion power are proposed to improve economic viability through the use of less-advanced technology in systems of considerably reduced scale. The rationale for and the means by which these systems can be achieved are discussed, as are unique technological problems

  18. Fusion reactor critical issues

    International Nuclear Information System (INIS)

    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

  19. Prospects for toroidal fusion reactors

    International Nuclear Information System (INIS)

    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

  20. Swedish thermometer for fusion reactors

    International Nuclear Information System (INIS)

    A neutron spectrometer called Tansy, which can measure temperatures in a fusion reactor, has been developed in the Department of Reactor Physics at Chalmers University of Technology in Gothenburg, Sweden. The instrument has been designed, constructed and tested over the past eight years and it has become one of Sweden's contributions of scientific know-how to the JET (Joint European Torus) fusion centre in Britain. A thesis by Dr. D. Aronsson entitled 'The development of a spectrometer for 14MeV neutrons from fusion' describes his part in the development of the instrument. Hydrogen fusion could become an important future source of energy provided we learn to use it properly, he says. As the release of energy during fusion takes place at such high temperatures, the plasma (fuel) can only be kept in place by strong magnetic fields. One way to measure temperatures of such magnitude is to study the neutrons scattered by the process. With the aid of Tansy, it is possible to study the variations in speed between different neutrons emerging from the fusion process at an average speed of 50,000 km/sec. The basic principle is quite simple; the released neutrons hit a thin polyethylene foil and some of them collide with the hydrogen atom nuclei. After the collision, the particles continue in different directions. Tansy has a system of detectors which can register and identify the two types of particles at the same time. A computer can then use this information together with knowledge about the effects of the collision to calculate the speed of the neutrons released by the fusion, and from this determine the temperature at combustion. In its present form, Tansy is a one-off and will probably not be produced again, Dr. Aronsson says, but the principle is likely to be used again. If fusion becomes an energy producing method of the future, instruments like Tansy will be needed to control the process

  1. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    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. Innovative energy production in fusion reactors

    International Nuclear Information System (INIS)

    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)

  3. (Meeting on fusion reactor materials)

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.H. (Pacific Northwest Lab., Richland, WA (USA)); Klueh, R.L.; Rowcliffe, A.F.; Wiffen, F.W. (Oak Ridge National Lab., TN (USA)); Loomis, B.A. (Argonne National Lab., IL (USA))

    1990-11-01

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

  4. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    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

  5. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    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.)

  6. Prospects for improved fusion reactors

    International Nuclear Information System (INIS)

    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

  7. Assessment of fusion reactor development. Proceedings

    International Nuclear Information System (INIS)

    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 (D3He). (4) Critical issues of fusion reactor development such as Li extraction from the sea water, structural material and safety. (author)

  8. Nuclear data requirements for fusion reactor nucleonics

    International Nuclear Information System (INIS)

    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

  9. Generic magnetic fusion reactor cost assessment

    International Nuclear Information System (INIS)

    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

  10. The Extrap fusion reactor concept

    International Nuclear Information System (INIS)

    ABSTRACT A study has recently been initiated to assess the fusion reactor potential of the Extrap high-beta toroidal z-pinch concept. A reactor model is defined that fullfills certain economic and operational criteria that are characteristic of compact toroidal systems, including moderately large electric power output, high power density, high first wall loading, and simple construction. This model is applied to Extrap, and a 1000 MWe reference reactor having a first wall neutron loading of 10 MW/m2 is outlined. The minor plasma radius is 1.5 m, the major radius 4.5 m and the pinch current 10 MA. A 0.7 m thick blanket/refletor/shield is chosen to achieve sufficient breeding of tritium, good energy multiplication, and shielding of normal copper coils. (author). 12 refs.; 3 figs.; 1 tab

  11. Stellarator fusion reactors - an overview

    International Nuclear Information System (INIS)

    The stellarator system offers a distinct alternative to the mainline approaches to magnetic fusion power and has several potentially major advantages. Since the first proposal of the stellarator concept many reactor studies have been published and these studies reflect the large variety of stellarator configurations. The main representatives are the continuous-coil configurations and the modular-coil configurations. As a continuation of the LHD experiment two reactor configurations, FFHR1 and FFHR2, have been investigated, which use continuous helical windings for providing the magnetic field. The modular coil concept has been realized in the MHH-reactor study (USA 1997) and in the Helias reactor. The Helias reactor combines the principle of plasma optimisation with a modular coil system. The paper also discusses the issues associated with the blanket and the maintenance process. Stellarator configurations with continuous coils such as LHD possess a natural helical divertor, which can be used favourably for impurity control. In advanced stellarators with modular coils the same goal can be achieved by the island divertor. Plasma parameters in the various stellarator reactors are computed on the basis of presently known scaling laws showing that confinement is sufficiently good to provide ignition and self-sustained burn. (author)

  12. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    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

  13. The concept of a research fusion reactor

    International Nuclear Information System (INIS)

    Thus,for advancement towards a commercial fusion reactor,we have proposed here as a next step a steady state operated research fusion reactor with an increased plasma-wall detachment so as to further guarantee not only the production but also a long-term (for many years) confinement of a self-sustained plasma at the existing technology level. We consider the primary goal of the research fusion reactor is the provision of full-scale conditions for carrying out materials science experiments to create and test 1 st wall materials for the commercial fusion reactor

  14. Modular Stellarator Fusion Reactor concept

    International Nuclear Information System (INIS)

    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. Alternate fusion concepts as reactors

    International Nuclear Information System (INIS)

    The recent successes of the tokamak concept of controlled fusion have not quenched interest in possible alternatives. This report summarizes a recent study sponsored by the Electric Power Research Institute, which tried to quantify which hoped-for advantages persist when a serious attempt is made to design reactor plants around eight specific alternative concepts (Electron Beam-Heated Solenoid, Elmo Bumpy Torus, Fast Liner, Laser-Heated Solenoid, Linear Theta-Pinch, LINUS, Reversed-Field-Pinch, and Shock-Heated Annulus) addressing key technological issues and economic issues for each concept. The study aimed to isolate the cost drivers for the reactor plant and to compare their capital cost per kilowatt of electricity as well as address the impact of technological difficulty. Results of the study indicated that reactor block costs for the eight plants studied represent a substantially larger fraction of total plant costs than the corresponding fraction for light water reactors; bottom line costs of $ /kWe range over a factor of about two with cost drivers being the physical size of the power producing plasma and the relative magnitudes of the circulating power fraction and the nature of the power circulation. Other cost considerations are also enumerated and the author concludes by noting that one value of the engineering study and cost estimate has been to quantify the relation between physics uncertainty and cost uncertainty

  16. Spherical tokamak research for fusion reactor

    International Nuclear Information System (INIS)

    Between ITER and the commercial fusion reactor, there are many technological problems to be solved such as cost, neutron and steady-state operation. In the conceptual design of VECTOR and Slim CS reactors it was shown that the key is 'low aspect ratio'. The spherical tokamak (ST) has been expected as the base for fusion reactors. In US, ST is considered as a non-superconducting reactor for use in the neutron irradiation facility. Conceptual design of the superconducting ST reactor is conducted in Japan and Korea independently. In the present article, the prospect of the ST reactor design is discussed. (author)

  17. Plasma facing materials for fusion reactor applications

    OpenAIRE

    Gonzalez Arrabal, Raquel; Gordillo Garcia, Nuria; Rivera de Mena, Antonio; Alvarez Ruiz, Jesus; Garoz, D.; Perlado Martin, Jose Manuel

    2012-01-01

    The lack of plasma facing materials (PFM) able to withstand the severe magnetiicffusiion radiation conditions expected in fusion reactors is the actual bottle In both fusions approaches energy is released in the form of kinetic energy of neck for fusion to becomes a reality.

  18. Tritium Behaviour in the Fusion Reactor Materials

    OpenAIRE

    Pajuste, Elīna

    2012-01-01

    ABSTRACT Doctoral thesis is devoted to the development of future energy source nuclear fusion. The objective of this research is to evaluate fusion reactor material suitability regarding their behaviour and tritium retention in the fusion reactor relevant conditions. Methods and technique developed in the UL Institute of Chemical Physics Laboratory of Radiation Chemistry of Solid State has been used in this study. Synergetic facilitating effect of accelerated electrons and high magnetic fi...

  19. Status of Fusion Experimental Reactor (FER) design

    International Nuclear Information System (INIS)

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been conducted at JAERI in line with a long-range plan for fusion reactor development laid out in the long-term program of the Atomic Energy Commission issued in 1982. The FER succeeding the tokamak device JT-60 is a tokamak reactor with a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. The paper describes recent developments of the FER design concept

  20. Cold nuclear fusion reactor and nuclear fusion rocket

    OpenAIRE

    Huang Zhenqiang

    2013-01-01

    "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nucle...

  1. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    Full text: A long term solution to problems of energy production, green house gas generation, and pollution control may rest with controlled nuclear fusion reactors. Candidate structural materials for such reactors include low activation ferritic steels. Understanding and eliminating deleterious irradiation effects in these materials is the goal of these experiments in this collaboration using the ANL facility. In recent experiments on ferritic alloys we have recently found a significant difference with alloy composition in the microstructural response to irradiation, which corresponds to a bulk mechanical property change at a similar composition. In a collaboration between the Department of Materials at the University of Oxford and the Materials Science Division at Argonne National Laboratory, experiments which employ the unique transmission electron microscope and in situ ion irradiation user facility at ANL were performed on a series of Fe-Cr alloys. Enhanced nanometer-sized defect formation with Cr concentrations up to 11 % have been found and correlated with a decrease in mechanical hardening and embrittlement in similar alloys. (author)

  2. Compound cryopump for fusion reactors

    CERN Document Server

    Kovari, M; Shephard, T

    2013-01-01

    We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

  3. Feasibility of a laser or charged-particle-beam fusion-reactor concept with direct electric generation by magnetic-flux compression

    International Nuclear Information System (INIS)

    A new concept for an inertial-confinement fusion reactor is described which, because of its fundamentally different approach to blanket geometry and energy conversion, makes possible a unique combination of high efficiency, high power density, and low radioactivity. The conventional blanket is replaced with a liquid-density mass of lithium contiguously surrounding the fusion yield. This compact blanket configuration produces the maximum shock-induced kinetic energy in liquid metal and the maximum neutron absorption per unit mass. The shock-induced kinetic energy of the liquid lithium is converted directly to electricity with high efficiency by work done against a pulsed normal-conducting magnetic field applied to the exterior of the lithium

  4. Alternative fusion reactors as future commercial power plants

    International Nuclear Information System (INIS)

    Alternative reactor based on a field-reversed configuration (FRC) has advantages of the cylindrical geometry, the open field line geometry (direct energy conversion (DEC) of the charged-particle flow), and high β (plasma pressure/magnetic-field pressure). This paper aims to evaluate the attractiveness of a low radioactive FRC fusion core. Analysis of a conceptual deuterium - helium-3 (D-3He) fusion power reactor is presented and reference point is defined. Principal parameters of the D-3He plasma reference case (RC) and comparison with conceptual D-3He tokamak and FRC power plants are shown. (author)

  5. Distance to realization of nuclear fusion reactors

    International Nuclear Information System (INIS)

    Recently the research and development of nuclear fusion have progressed conspicuously, and reached the point of attaining the critical condition. In this paper, it is attempted to forecast how long does it take to realize a final nuclear fusion power reactor (a demonstration reactor). The research and development of nuclear fusion have two important meanings. One is it is a promising means for ensuring an energy source for the future in Japan. Another is it has been brought up to the present status as the large scale project research maintaining the creativity and originality without requiring the introduction of technology from foreign countries. Hereafter, it is necessary to bring it up large as the Japanese basic technology. The research and development of nuclear fusion has advanced steadily, producing many physical knowledges and technical development. The principle and present status of nuclear fusion are explained. Now, an experimental fusion reactor is investigated as the next step. A large helical system project was started as 7-year project from 1990. The start of the operation of a prototype nuclear fusion power reactor is assumed in 2026, and that of a demonstration reactor is assumed in 2040. The investment for nuclear fusion and the extending effect are discussed. (K.I.)

  6. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    CERN Document Server

    Lee, Seung Kyu; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

  7. Designing the Cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    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

  8. Inertial fusion reactors and magnetic fields

    International Nuclear Information System (INIS)

    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

  9. Safety and environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    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)

  10. A Compact Nuclear Fusion Reactor for Space Flights

    International Nuclear Information System (INIS)

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

  11. Fusion reactor safety studies, FY 1977

    International Nuclear Information System (INIS)

    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

  12. Generic Magnetic Fusion Reactor Revisited

    Science.gov (United States)

    Sheffield, John; Milora, Stanley

    2015-11-01

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

  13. Technology of compact fusion-reactor concepts

    International Nuclear Information System (INIS)

    An identification of future engineering needs of compact, high-power-density approaches to fusion power is presented. After describing a rationale for the compact approach and a number of compact fusion reactors, key technology needs are assessed relative to the similar needs of the conventional tokamak in order to emphasize differences in required technology with respect to the well-documented mainline approaches

  14. Safeguarding fusion reactors. Plea for a proliferation resistant design of nuclear fusion

    International Nuclear Information System (INIS)

    The contribution on safeguarding of thermonuclear fusion reactors covers the following topics: The development of the technology during and after the Cold War, the production of fissile material, the non-civil use of nuclear fusion - hypothetical military use of fusion reactors, options for regulations and design for a peaceful use of nuclear fusion, safeguards of nuclear fusion reactors.

  15. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

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

  16. Concept of multi-function fusion reactor

    International Nuclear Information System (INIS)

    To really use the fusion energy and make the fusion energy as a main energy in the world could need more than 50 years. The construction of ITER starts the real course to realize the peaceful use of the fusion energy. It means the technologies developed in the world are feasible to built fusion facilities or reactors with fusion core plasmas. Based on the technologies nowadays, a concept of multi-function fusion reactor (MFFR) is proposed. MFFR has following functions: fission waste disposal, plutonium 239 breeding from uranium 238, hydrogen producing, tritium producing, components test for fusion reactors, or even electricity power plant demonstration. The preliminary considerations of MFFR are: (a) reasonable configuration and changeable invessel function blanket modules, (b) enough flexibility to realize multi-functions separately or at the same time in the facility, (c) suitable plasma core parameters and blanket concept, (d) fully superconducting toroidal and poloidal magnets. Two major types of functional blankets, sub-critical blanket and energy exchange blanket, are defined in MFFR for transmutation of fission waste/fuel and fusion energy transfer. In this paper, the concept of MFFR and the functional blanket based on the results gained in the past years are introduced. (author)

  17. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    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

  18. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    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

  19. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    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

  20. Charge density path in cold fusion reactions

    International Nuclear Information System (INIS)

    Cold fusion reactions are very frequently employed to produce compound nuclei with a relatively low excitation energy, which is extremely important for a successful synthesis method, particularly in the region of superheavy nuclei. Usually the charge densities of the projectile, target, and compound nucleus are different. We present a method allowing to take into consideration this difference continuously during the fusion process. Applications are given both in the intermediate mass and the superheavy region. Different cold fusion paths are studied with respect to the change of the charge density within the overlapping region. A transition formula from separated fusion partners up to the compound nucleus is obtained as depending on the geometrical changes. Macroscopic-microscopic approach is used to compute the total deformation energy. Shell corrections are obtained with Strutinsky method, having the new deformed two-center single particle energy levels as an input. Yukawa-plus-exponential model is employed to compute the macroscopic part. Spheroidal deformations are taken into account. By changing the absolute value of semiaxes as well as their ratio, the charge densities of the partners are modified during fusion. As a result of minimization against different paths of the semiaxes ratios from projectile and target values to synthesized nucleus, charge density variation can lower the cold fusion deformation energy. This kind of influence is especially active in the last part of the fusion process, when the projectile is already at least half embedded in the target. For cold fusion of light and intermediate nuclei, the energy variation in the last part of the deformation path reaches 4 MeV for 102 Ru and 3.7 MeV for 152 Dy synthesis. For a possible superheavy production the influence of charge density changes are quantitatively more important. The energy difference in the cold fusion channel barrier of 292 116 reaches about 8 MeV in the last part of the

  1. A look at the fusion reactor technology

    International Nuclear Information System (INIS)

    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)

  2. Extrap conceptual fusion reactor design study

    International Nuclear Information System (INIS)

    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)

  3. Cold nuclear fusion reactor and nuclear fusion rocket

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang

    2013-10-01

    Full Text Available "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nuclei to be bound fusion on the same line (track of. Re-use nuclear spin angular momentum vector inherent nearly the speed of light to form a super strong spin rotation gyro inertial guidance features, to overcome the Coulomb repulsion strong bias barrier to achieve fusion directly hit. Similar constraints apply nuclear inertial guidance mode for different speeds and energy ion beam mixing speed, the design of ion speed dc transformer is cold fusion reactors, nuclear fusion engines and such nuclear power plants and power delivery systems start important supporting equipment, so apply for a patent merger

  4. Nuclear data needs for fusion reactors

    International Nuclear Information System (INIS)

    The nuclear design of fusion reactor 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

  5. Generic Stellarator-like Magnetic Fusion Reactor

    Science.gov (United States)

    Sheffield, John; Spong, Donald

    2015-11-01

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

  6. Overview of FER (Fusion Experimental Reactor)

    International Nuclear Information System (INIS)

    The FER (Fusion Experimental Reactor) project is proposed to construct a next generation tokamak machine in Japan, in order to take a leadership in realizing a fusion reactor under international cooperation. The FER is the machine, which comes in between the present large tokamak machines like the JT-60 and the DEMO reactor for power generation. The mission of the FER is to realize a long controlled burn and to develop and test major fusion component technologies, super conducting magnet and breeding blanket and so on, that is, to demonstrate the engineering feasibility of a fusion reactor. The conceptual design of the FER was started in 1980. In April 1988, a new organization (Fusion Experimental Reactor Team) was constructed to support the ITER activities and also to design the FER. In order to make the FER and the ITER complementary, the FER concept was reconsidered. The FER described in this report is a new version, and the conceptual design will be finished in December, 1990. (author)

  7. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    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)

  8. Nuclear data needs for fusion reactors

    International Nuclear Information System (INIS)

    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

  9. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

    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 238Pu 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 232U. 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)

  10. Computational mathematics and physics of fusion reactors

    OpenAIRE

    Garabedian, Paul R.

    2003-01-01

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

  11. Health physics in fusion reactor design

    International Nuclear Information System (INIS)

    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

  12. Compact approach to fusion power reactors

    International Nuclear Information System (INIS)

    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

  13. Nuclear data for fusion reactor technology

    International Nuclear Information System (INIS)

    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

  14. Recycling of vanadium alloys in fusion reactors

    International Nuclear Information System (INIS)

    The feasibility of reprocessing a vanadium alloy after its use as a structural material in a fusion reactor, in order to enable its subsequent hands-on recycling within the nuclear industry, has been determined. For less neutron-exposed components, clearance of materials has also been considered. A conceptual model for the radiochemical processing of the alloy has been developed and tested experimentally. Using di-2-ethyl-hexyl-phosphoric acid it is possible to purify the components of the V-Cr-Ti alloy after its exposure in a fusion reactor down to the required level of activation product concentrations

  15. Safety scenario for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A scenario to ensure the safety of the Fusion Experimental Reactor (FER) is proposed. The safety features of a fusion reactor are given and their impacts on the safety design are shown. The requirements in the design of major components of FER to achieve safety and the safety evaluation process are described. The results of the evaluation showed that even in the event of the maximum credible accidents, the radiological consequence to the public can be held at an acceptable level. The applicability to FER of various aspects of the regulations for facilities treating large amounts of radioisotopes is discussed with a positive conclusion. (author). 11 refs, 1 fig

  16. Small fusion reactors: problems, promise, and pathways

    International Nuclear Information System (INIS)

    The prevalent vision of magnetic fusion as a central-station power plant projects a high-technology, low-power-density nuclear boiler that may require high energy costs to be economic. Smaller, higher-power-density approaches can reduce the impact of the fusion power core and associated support equipment on the overall cost equation for fusion. In the course of attaining sizes, power capacity, and costs that are more in line with alternative energy sources, a range of problems, promise, and pathways can be identified. The issues related to these more compact systems are addressed on the basis of generic reactor models

  17. A charged fusion product diagnostic for a spherical tokamak

    Science.gov (United States)

    Perez, Ramona Leticia Valenzuela

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

  18. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    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)

  19. Impurity control in Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Poloidal divertor system is employed as the impurity control measure in Fusion Experimental Reactor (FER). The authors will report the overall survey of impurity control physics in FER. The results obtained are as follows. (1) The triple-valued solutions of divertor plasma equilibrium are obtained as a function of incoming ion flux. Engineering design is carried out based on the stable dense and cold divertor plasma. (2) Low density and high temperature solution disappears when the geometry is extremely closed (chamber length=50cm and void width=O). (3) Plasma temperature can become slightly high on the side of exhaust duct near the plate. (4) Erosion rate on the first wall by charge-exchange neutrals is recognized to be about 1cm/year by DEGAS code that is also obtained by even our divertor code and simple order estimation. (5) Cold and dense divertor plasma could be formed during noninductive current drive phase either LHRF or NBI, if the current drive efficiency is improved

  20. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Safety

    International Nuclear Information System (INIS)

    A number of safety features of fusion reactors as compared with fission reactors have been clarified. Utilizing these safety features, the measures to achieve a significantly safe fusion reactor emphasizing the radioactive material confinement, are described. The results of the safety analysis of ITER, which is the most well defined fusion reactor, show that the safety of ITER is clearly secured. Based on the results, the way to further improve the safety in fusion power reactors is shown. (author)

  1. Occupational health physics at a fusion reactor

    International Nuclear Information System (INIS)

    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. Tritium breeding in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    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

  4. Nuclear fusion reactor material data base

    International Nuclear Information System (INIS)

    The working conditions for the materials to be used for nuclear fusion reactors are many sided, complicated and harsh. The existing experimental results can not be employed directly for reactor design. In such a case, it is insufficient to simply accumulate the experimental data on the specific properties of specific materials, and it is necessary to predict the material behaviour in the reactor system by rearranging those data in accordance with the purpose. When extreme characteristics are frequently pursued, wide insight is necessary regarding from the fundamental theory to the testing of practical equipment. In the development of nuclear fusion reactor materials, it is especially important to satisfy the condition that the design purpose of the system for selecting the optimum materials should be fully understood. A new material engineering approach has become necessary, in which a barrier existing so far between materials and the design is removed. From this viewpoint, the specifications, present status and design and development of material data base presently under development, the use of the data base made for trial, and the interface of material development and nuclear fusion reactor design, are described. In this data base, most of the data handle literature data, and the event data base mainly composed of experimental data is very few, similarly to other fields. Data modification will be necessary to respond the questions of users. (Wakatsuki, Y.)

  5. BR2 Reactor: Irradiation of Fusion Materials

    International Nuclear Information System (INIS)

    In collaboration with the EFDA (European Fusion Development Agreement), SCK-CEN irradiates several materials in the BR2 reactor at different temperatures and up to different doses to study their mechanical and physical properties during and after irradiation. Those materials are candidates for the construction of different parts of the ITER fusion reactor and of the long-term DEMO (DEMOnstration) reactor. The objectives of research performed at SCK-CEN are to irradiate up to 2 dpa RAFM (Reduced Activity Ferritic Martensitic) steels joints and RAFM ODS (Oxide Dispersion Strengthening) at 300 degrees Celsius; to build and test an experimental rig to perform in-situ creep-fatigue tests under neutron irradiation and its out-pile equipment and to design a new irradiation basket to irradiate in BR2 copper/stainless steel joints and RAFM specimens with implanted helium at low dose

  6. Optical design considerations for laser fusion reactors

    International Nuclear Information System (INIS)

    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

  7. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    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)

  8. Direct energy conversion of radiation energy in fusion reactor

    International Nuclear Information System (INIS)

    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)

  9. Radiolytic production of chemical fuels in fusion reactor systems

    International Nuclear Information System (INIS)

    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

  10. Tritium monitor for fusion reactors

    International Nuclear Information System (INIS)

    This report describes the design, operation, and performance of a flow-through ion-chamber instrument designed to measure tritium concentrations in air containing 13N, 16N, and 41Ar 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

  11. A fusion transmutation of waste reactor

    International Nuclear Information System (INIS)

    A design concept and the performance characteristics for a fusion transmutation of waste reactor (FTWR)--a sub-critical fast reactor driven by a tokamak fusion neutron source--are presented. The present design concept is based on nuclear, processing and fusion technologies that either exist or are at an advanced stage of development and on the existing tokamak plasma physics database. A FTWR, operating with keff≤0.95 at a thermal power output of about 3 GW and with a fusion neutron source operating at Qp=1.5-2, could fission the transuranic content of about a hundred metric tons of spent nuclear fuel per full-power-year and would be self-sufficient in both electricity and tritium production. In equilibrium, a nuclear fleet consisting of Light Water Reactors (LWRs) and FTWRs in the electrical power ratio of 3/1 would reduce the actinides discharged from the LWRs in a once-through fuel cycle by 99.4% in the waste stream that must be stored in high-level waste repositories

  12. An overview of inertial fusion reactor design

    International Nuclear Information System (INIS)

    Recent progress in the conceptual design of inertial fusion reaction chambers and power plants is reviewed. A discussion of expected operating parameters and a brief historical perspective are provided to organize the rich array of chamber and driver concepts. The technical feasibility of several reaction chamber concepts is discussed, along with technical issues that require future analysis, experiment, and development. Where these chambers have been integrated into a power plant design, the characteristics are described. Finally, requirements on the future development of inertial fusion reactor technology are discussed

  13. Fusion reactors for hydrogen production via electrolysis

    International Nuclear Information System (INIS)

    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

  14. Testing the reactor charging machine

    International Nuclear Information System (INIS)

    One of the main objective of the R - D technological engineering program devoted to the Fuel Handling System is domestic production of equipment and technology for testing the ends of the reactor charging machine (MID) destined to Cernavoda NPP, beginning with Unit 2. To achieve the objective based on an own design, a bench-scale testing stand of MIDs which can simulate the pressure, flow-rate, and temperature conditions proper to fuel channels in operating CANDU 600 reactors. The main components of this testing facility are: - fuel channels, cold also test sections, allowing the coupling of MID end upwardly and downwardly, corresponding to the direction of the water flow through the channel; - technological installation feeding with light water the testing sections of the facility in thermohydraulic conditions, similar to those in the reactor, allowing the cold and hot testings, respectively, of the MID end; - cold testing installation, water supply and oil control panel, feeding the hydraulic drives of the MID's end during the testings; - fixed bridge and mobile carrier for MID's end positioning against testing sections; - installation for functional testing of MID thrusters, before pre-admission and reception tests; - dedicated tools and devices; - raising and transport mechanical devices for handling and positioning the MID's end upon the carrier; - automation panel for controlling the stand equipment and MID's end; - process computer for conducting on-line tests. MID's end testing implies mainly the following operations: - regulation, calibration and functional testing of the MID thrusters carried out independently on a specialised stand; - regulation and calibration of MID's end sub-assemblages; - carrying out the cold and hot pre-admission tests consisting in automatic performing, without operator intervention, of 12 fuel changes, two of which being successive; - performing the cold and hot reception tests, consisting in automatic accomplishment of 4

  15. 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.

  16. Reactor potential for magnetized target fusion

    International Nuclear Information System (INIS)

    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. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    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)

  18. Nuclear data requirements for fusion reactor shielding

    International Nuclear Information System (INIS)

    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)

  19. Fusion reactor blanket: neutronic studies in France

    International Nuclear Information System (INIS)

    The problem of effective tritium regeneration is a crucial issue for the fusion reactor, especially 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 analysis. The results of these studies permit us to conclude that it is possible to expect an adequate tritium breeding ratio

  20. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    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)

  1. Environmental aspects of fusion reactors 1985

    International Nuclear Information System (INIS)

    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

  2. Shakedown analysis of fusion reactor first wall

    International Nuclear Information System (INIS)

    Shakedown analyses of a typical fusion reactor first wall including coolant channels subjected to cyclic thermal/steady primary and cyclic primary/steady thermal stresses are carried out. The stresses are assumed to be predominantly of the bending type. The first cycle of loading/unloading is analyzed using elastic-plastic beam bending theory. The general problem of shakedown is solved using the shakedown theorem of perfect plasticity

  3. Shielding design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report first describes the basic design philosophy of radiation shields for the fusion experimental reactor (FER) which has been proposed to be the next step machine to JT-60. Next, geometrical models and calculation parameters for shielding calculations were investigated to establish the standard design calculation methods, and accuracy of the calculation was evaluated. Further, irradiation properties of in-vessel components and bulk shielding properties were summarized in the useful form for the future design works. (author)

  4. Choice of coils for a fusion reactor

    OpenAIRE

    Alexander, Romeo; Garabedian, Paul R.

    2007-01-01

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

  5. The spheromak as a compact fusion reactor

    International Nuclear Information System (INIS)

    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

  6. Materials needs for compact fusion reactors

    International Nuclear Information System (INIS)

    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/m3 versus 0.3 to 0.5 MW/m3), 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.)

  7. Fission-reactor experiments for fusion-materials research

    International Nuclear Information System (INIS)

    The US Fusion Materials Program makes extensive use of fission reactors to study the effects of simulated fusion environments on materials and to develop improved alloys for fusion reactor service. The fast reactor, EBR-II, and the mixed spectrum reactors, HFIR and ORR, are all used in the fusion program. The HFIR and ORR produce helium from transmutations of nickel in a two-step thermal neutron absorption reaction beginning with 58Ni, and the fast neutrons in these reactors produce atomic displacements. The simultaneous effects of these phenomena produce damage similar to the very high energy neutrons of a fusion reactor. This paper describes irradiation capsules for mechanical property specimens used in the HFIR and the ORR. A neutron spectral tailoring experiment to achieve the fusion reactor He:dpa ratio will be discussed

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

    Science.gov (United States)

    Koch, Raymond

    2006-06-01

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

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

    International Nuclear Information System (INIS)

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

  10. Fokker-Planck Modelling of Delayed Loss of Charged Fusion Products in TFTR.

    Energy Technology Data Exchange (ETDEWEB)

    Edenstrasser, J.W.; Goloborod' ko, V.Ya.; Reznik, S.N.; Yavorskij, V.A.; Zweben, S.

    1998-08-01

    The results of a Fokker-Planck simulation of the ripple-induced loss of charged fusion products in the Tokamak Fusion Test Reactor (TFTR) are presented. It is shown that the main features of the measured "delayed loss" of partially thermalized fusion products, such as the differences between deuterium-deuterium and deuterium-tritium discharges, the plasma current and major radius dependencies, etc., are in satisfactory agreement with the classical collisional ripple transport mechanism. The inclusion of the inward shift of the vacuum flux surfaces turns out to be necessary for an adequate and consistent explanation of the origin of the partially thermalized fusion product loss to the bottom of TFTR.

  11. Fusion reactors-high temperature electrolysis (HTE)

    International Nuclear Information System (INIS)

    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 18000C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 14000C 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 H2 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 106 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

  12. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

    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

  13. Nuclear reactor fuel elements charging tool

    International Nuclear Information System (INIS)

    To assist the loading of nuclear reactor fuel elements in a reactor core, positioning blocks with a pyramidal upper face charged to guide the fuel element leg are placed on the lower core plate. A carrier equipped with means of controlled displacement permits movement of the blocks over the lower core plate

  14. Atomic collisions in fusion plasmas involving multiply charged ions

    International Nuclear Information System (INIS)

    A short survey is given on atomic collisions involving multiply charged ions. The basic features of charge transfer processes in ion-ion and ion-atom collisions relevant to fusion plasmas are discussed. (author)

  15. Report of Nuclear Fusion Reactor Engineering Research Meeting. 6. Advanced reactor engineering technology for nuclear fusion demonstration reactor

    International Nuclear Information System (INIS)

    This research meeting has been held every year, and the 6th meeting was held on January 17, 1995 at University of Tokyo. As the type of a demonstration reactor, tokamak type and helical type were set up, and the topics on the various subjects of their reactor engineering technology were presented, and active discussion was carried out. At the meeting, lectures were given on the reactor engineering technology required for a prototype reactor, the material technology supposed for a demonstration reactor, thermal-electric conversion and the direct electricity generation using Nernst effect, the advanced manufacturing technology of functional, structural materials, the application of high temperature superconductors to nuclear fusion reactors, the reactor engineering technology required for a helical type demonstration reactor, and tokamak demonstration reactor and the common technology of fission and fusion. This report is the summary of these lecture materials. The useful knowledges were obtained for considering the development of nuclear fusion reactor technology hereafter in this meeting. (K.I.)

  16. Survey of the laser-solenoid fusion reactor

    International Nuclear Information System (INIS)

    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

  17. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    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 1013Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 1010Bq (0.5 Ci) per day per ton of fuel

  18. Series lecture on advanced fusion reactors

    International Nuclear Information System (INIS)

    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.)

  19. Surveying of large fusion reactor components

    International Nuclear Information System (INIS)

    For CAD-supported remote maintenance of fusion machines a surveying system has to measure the geometrical changes of the ractor components. A study of principally applicable techniques indicated that triangulation with theodolites is well suited to update the CAD-models of fusion reactor components. The remote-controlled and CAD-supported surveying system GMS (Geometry Measurment System), developed by KfK, is equipped with two digital theodolites, a laser- and a camera-theodolite, completely controlled by a computer. The data transfer CAD-GMS will be realized with the standardized IGES-interface. Two show the feasibility of this draft a GMS-prototype, equipped with a single camera-theodolite, is built up presently. (author). 10 refs.; 7 figs

  20. Conceptual design of Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    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/m2, 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

  1. Investigation of shielding analysis method for fusion reactors

    International Nuclear Information System (INIS)

    An investigation has been made, at the shielding laboratory, into the status of shielding analysis method for fusion reactor based on conceptual designs of a variety of fusion power reactors and fusion experimental facilities, in cooperation with the Fusion Reactor Shielding Working Group in the Research Committee on Fast Neutron Shielding of the Atomic Energy Society of Japan. The reactors and facilities considered are CULHAM MKII(U.K), SPTR (Japan), TFTR(U.S.A.), STARFIRE(U.S.A.) and INTOR-USA(U.S.A.). (author)

  2. Organic materials for fusion-reactor applications

    International Nuclear Information System (INIS)

    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

  3. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  5. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-28

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

  6. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    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

  7. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    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)

  8. Evaluation of skyshine calculation method for fusion reactor and application to fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    In the design of the reactor room for a fusion reactor, the cost of the room strongly depends on the thickness of the roof because the area of the roof is generally large. The roof thickness is mostly determined by the requirement to reduce the skyshine dose rate level at the site boundary below the assigned value. Therefore the accurate evaluation of the skyshine dose becomes important for the design of the reactor room. Skyshine dose for a D-T fusion reactor has been evaluated by a number of researchers but the agreement is not so good. In this report, the first collision source is used with two-dimensional SN transport method to form DOT3.5-GRTUNCL-DOT3.5 coupled calculation flow. The validity of the methodology was first shown by calculating the skyshine dose from a 14 MeV neutron source and comparing the calculated results with the measured results. This methodology was then used to calculate the skyshine dose for the Fusion Experimental Reactor (FER). The calculated results were compared with those from several other methods to clarify the mutual difference. (author)

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

    International Nuclear Information System (INIS)

    It is possible to use a nuclear fusion reactor, of a somewhat less technologically challenging design than that contemplated purely for the generation of electricity, by employing fusion-derived neutrons to drive useful nuclear reactions. One device based on this concept is called the fusion hybrid reactor, or, perhaps more explicitly, the fusion-fission hybrid reactor. Neutrons from a fusion core would react with fertile and fissible material in a blanket surrounding the core, with the consequent creation of both fissile material for conventional nuclear reactor fuel and heat for generating electricity. Another such device, called the tritium-breeding fusion reactor, would breed tritium by reaction with lithium targets around the core. This report examines future circumstances in which these reactors might be needed and advantageous. Based on their technical, economic, and social aspects, it discusses the program content and pace at which these applications ought to be pursued. 46 refs., 35 figs., 31 tabs

  10. SOLASE conceptual laser fusion reactor study

    International Nuclear Information System (INIS)

    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/m2. 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 Li2O 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 Li2O 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

  11. Fusion Experimental Reactor (FER) design concept

    International Nuclear Information System (INIS)

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been conducted at JAERI within the frame of the longterm program of the Atomic Energy Commission issued in 1982. The major mission of the FER, which is the device planned to succeed the JT-60 tokamak device, is realizing a self-ignited, long burn D-T plasma and demonstrating engineering feasibility of fusion energy. The reference design concept is based on a quasi-steady state operation scenario. This scenario includes non-inductive current drive at low plasma density during startup and power dwell period and conventional inductive current drive for high density plasma burn periods (2000 s). Key design features and parameters are as follows: non-breeding blanket (shield blanket); single-null poloidal divertor; lifetime neutron fluence of 0.3 MW·y/m2; fusion power of 300 MW; average wall loading of 0.68 MW/m2; plasma major radius of 5.2 m; plasma minor radius of 1.12 m with the elongation of 1.5; and plasma current of 5.9 MA. (author). 4 refs, 10 figs, 9 tabs

  12. Hydrogen production from high temperature electrolysis and fusion reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  14. Multivariable optimization of fusion reactor blankets

    International Nuclear Information System (INIS)

    The optimization problem consists of four key elements: a figure of merit for the reactor, a technique for estimating the neutronic performance of the blanket as a function of the design variables, constraints on the design variables and neutronic performance, and a method for optimizing the figure of merit subject to the constraints. The first reactor concept investigated uses a liquid lithium blanket for breeding tritium and a steel blanket to increase the fusion energy multiplication factor. The capital cost per unit of net electric power produced is minimized subject to constraints on the tritium breeding ratio and radiation damage rate. The optimal design has a 91-cm-thick lithium blanket denatured to 0.1% 6Li. The second reactor concept investigated uses a BeO neutron multiplier and a LiAlO2 breeding blanket. The total blanket thickness is minimized subject to constraints on the tritium breeding ratio, the total neutron leakage, and the heat generation rate in aluminum support tendons. The optimal design consists of a 4.2-cm-thick BeO multiplier and 42-cm-thick LiAlO2 breeding blanket enriched to 34% 6Li

  15. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    OpenAIRE

    Lee, Seung Kyu; Kang, Byoung-Hwi; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spe...

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

    International Nuclear Information System (INIS)

    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. Design study of cooling system for tokamak fusion reactor

    International Nuclear Information System (INIS)

    Design study of the reactor cooling system for a tokamak fusion reactor has been carried out. In the cooling system of an experimental 150 MWt fusion reactor, to grasp the plant concept and clarify the R and D items the main cooling system and the tritium recovery system were designed and the auxiliary system was examined. In the cooling system of a commercial 2000 MWt fusion reactor, to study the plant and environment safety the main cooling system and the tritium recovery system were designed, including the evaluation of water leakage and tritium penetration in the steam generators. (auth.)

  18. EPRI Asilomar papers: on the possibility of advanced fuel fusion reactors, fusion-fission hybrid breeders, small fusion power reactors, Asilomar, California, December 15--17, 1976

    International Nuclear Information System (INIS)

    An EPRI Ad Hoc Panel met in Asilomar, California for a three day general discussion of topics of particular interest to utility representatives. The three main topics considered were: (1) the possibility of advanced fuel fusion reactors, (2) fusion-fission hybrid breeders, and (3) small fusion power reactors. The report describes the ideas that evolved on these three topics. An example of a ''neutron less'' fusion reactor using the p-11B fuel cycle is described along with the critical questions that need to be addressed. The importance to the utility industry of using fusion neutrons to breed fission fuel for LWRs is outlined and directions for future EPRI research on fusion-fission systems are recommended. The desirability of small fusion power reactors to enable the early commercialization of fusion and for satisfying users' needs is discussed. Areas for possible EPRI research to help achieve this goal are presented

  19. Natural Fueling of a Tokamak Fusion Reactor

    CERN Document Server

    Wan, Weigang; Chen, Yang; Perkins, Francis W

    2009-01-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

  20. Waste management for different fusion reactor designs

    International Nuclear Information System (INIS)

    Safety and Environmental Assessment of Fusion Power (SEAFP) waste management studies performed up to 1998 concerned three power tokamak designs. In-vessel structural materials consist of V-alloys or low activation martensitic (LAM) steel; tritium-producing materials are Li2O, Pb-17Li, Li4SiO4 with a Be-multiplier; coolants are helium or water. The strategy chosen reduces permanent radwaste by recycling the in-vessel materials and by clearance of the other structures. Limits of the contact dose rate and specific activity of the waste allowing such options are defined accordingly. SEAFP activities for 1999 enlarge the analysis to three additional reactors with in-vessel structures made with SiC/SiC composites. These materials cannot be recycled due to their form and, according to national regulations of E.C. countries, long-lived activation products hinder near-surface burial (NSB)

  1. Thermomagnetic burn control for magnetic fusion reactor

    International Nuclear Information System (INIS)

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

  2. Charge exchange recombination spectroscopy measurements in the extreme ultraviolet region of central carbon concentrations during high power neutral beam heating in TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The carbon concentration in the central region of TFTR discharges with high power neutral beam heating has been measured by charge-extracted recombination spectroscopy (CXRS) of the C+5 n = 3--4 transition in the extreme ultraviolet region. The carbon concentrations were deduced from absolute measurements of the line brightness using a calculation of the beam attenuation and the appropriate cascade-corrected line excitation rates. As a result of the high ion temperatures in most of the discharges, the contribution of beam halo neutrals to the line brightness was significant and therefore had to be included in the modeling of the data. Carbon concentrations have been measured in discharges with Ip = 1.0-1.6 MA and beam power in the range of 2.6-30 MW, including a number of supershots. The results are in good agreement with carbon concentrations deduced from the visible bremsstrahlung Zeff and metallic impurity concentrations measured by x-ray pulse-height analysis, demonstrating the reliability of the atomic rates used in the beam attenuation and line excitation calculations. Carbon is the dominant impurity species in these discharges; the oxygen concentration measured via CXRS in a high beam power case was 0.0006 of ne, compard to 0.04 for carbon. Trends with Ip and beam power in the carbon concentration and the inferred deuteron concentration are presented. The carbon concentration is independent of Ip and decreases from 0.13 at 2.6 MW beam power to 0.04 at 30 MW, while the deuteron concentration increases from 0.25 to 0.75 over the same range of beam power. These changes are primarily the result of beam particle fueling, as the carbon density did not vary significantly with beam power. The time evolutions of the carbon and deuteron concentrations during two high power beam pulses, one which exhibited a carbon bloom and one which did not, are compared. 30 refs., 12 figs., 2 tabs

  3. 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…

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

    International Nuclear Information System (INIS)

    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

  5. Application of Bondarenko formalism to fusion reactors

    International Nuclear Information System (INIS)

    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 93Nb, which exhibits a large number of capture resonances 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 93Nb on blanket parameters such as breeding ratio, total afterheat, radioactivity, magnet-coil heating and total energy depositions have been studied. Resonance self-shielding of 93Nb, 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

  6. Tritium issues for realization of a DT fusion reactor

    International Nuclear Information System (INIS)

    A trend of studies of production and consumption of tritium is described. Realization of DT fusion reactor is discussed by tritium balance obtained from the above studies. It consists of introduction, tritium introduced into plasma vessel, tritium inventory in plasma vessel, tritium loss at fueling cycle system, tritium breeding and loss in blanket system, tritium balance in DT fusion reactor and summary. Investigation of development of external tritium resources has to be started. Tritium flow in DT fusion reactor, comparison of tritium inventory in fusion reactor, schematic diagram of tritium behavior in plasma vessel, change of overall burning efficiency and overall plasma generation rate, tritium inventory in re-deposition layer, effects of recovery efficiency of tritium from re-deposition layer, various breeding efficiencies in solid blanket, tritium flow in inertial confinement reactor with first wall, a tabular comparison of tritium balance calculation values, and comparison between tritium production methods are illustrated. (S.Y.)

  7. Safety aspects of an inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study is intended to identify significant safety aspects of inertial confinement fusion power plant concepts and to relate them to the more familliar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Needs for safety related research and development specifically for inertial confinement fusion are pointed out. (orig./GG)

  8. Cost assessment of a generic magnetic fusion reactor

    International Nuclear Information System (INIS)

    A generic magnetic fusion reactor model is used to determine the conditions under which electricity generation from fusion would be economically viable. The use of a generic model helps to circumvent problems associated with present perceptions of magnetic configurations. It helps also to decouple those limitations set by generic constraints such as nuclear cross sections from those set by the state of development today. The model shows that only moderate advances are required in reactor characteristics over current designs to make an economically attractive magnetic fusion reactor

  9. Materials problems in magnetically confined pulsed fusion reactors

    International Nuclear Information System (INIS)

    It is noted that materials in fusion power reactors must function satisfactorily under conditions of intense high energy neutron flux heat loads, and temperatures ranging from cryogenic to about 13000K. The competition between fatigue and thermal creep will occur in all pulsed fusion reactors, but with the additional possibility of radiation creep being the dominant deformation mode. The response of structural support members to the effects of neutron irradiation, in pulses, of cyclic temperature transients, and therefore cyclic thermal stresses, and of elevated temperature must be evaluated for each different type of pulsed fusion reactor having different combinations of cyclic stress, steady stress, temperature, flux level, and metal choice. (U.S.)

  10. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    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 Li2O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  11. Engineering the fusion reactor first wall

    Energy Technology Data Exchange (ETDEWEB)

    Wurden, Glen [Los Alamos National Laboratory; Scott, Willms [Los Alamos National Laboratory

    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

  12. Engineering the fusion reactor first wall

    International Nuclear Information System (INIS)

    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

  13. Chromium-molybdenum steels for fusion-reactor applications

    International Nuclear Information System (INIS)

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

  14. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    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

  15. Health physics aspects of activation products from fusion reactors

    International Nuclear Information System (INIS)

    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

  16. Alternative fusion concepts and the prospects for improved reactors

    International Nuclear Information System (INIS)

    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

  17. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    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

  18. The role of direct energy conversion for the realization of fusion reactors

    International Nuclear Information System (INIS)

    The stable supply of electric power is an important subject for the development of human civilized society, and nuclear fusion energy is one of the energy sources when the use of fossil fuel becomes difficult in future, together with nuclear fission energy and solar energy. It is important to convert controlled nuclear fusion energy to electric power with efficiency as high as possible for the improvement of its economical efficiency and the reduction of the effect to living bodies, and the primary significance of high efficiency direct electric power generation is in this point. As the object of direct electric power generation in nuclear fusion, there are the electrostatic recovery of the kinetic energy of charged particles leaking from the core plasma of fusion reactors, the method of converting plasma energy and magnetic energy to electric power through the coupling of an electromagnetic field, the in situ MHD electric power generation by liquid metal Rankine cycle using the kinetic energy of neutrons and the energy of bremsstrahlung and synchrotron radiation coming from the core plasma of fusion reactors and so on. In this paper, the role of direct electric power generation in nuclear fusion and the way of contribution that direct electric power generation makes to the realization of fusion reactors are explained, showing the concrete examples. (Kako, I.)

  19. Neutronics design for a spheric tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Based on studies of spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. A set of plasma parameters suitable for the transmutation blanket is selected. Using the transport and burn-up calculation code BISON3.0 and its associated database, transmutation rate of MA nuclear waste, energy multiplication, and tritium breeder rate in the transmutation blanket are calculated

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

    Science.gov (United States)

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

    1988-01-01

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

  1. Blankets for tritium catalyzed deuterium (TCD) fusion reactors

    International Nuclear Information System (INIS)

    The TCD fusion fuel cycle - where the 3He from the D(D,n)3He reaction is transmuted, by neutron capture in the blanket, into tritium which is fed back to the plasma - was recently recognized as being potentially more promising than the Catalyzed Deuterium (Cat-D) fuel cycle for tokamak power reactors. It is the purpose of the present work to assess the feasibility of, and to identify promising directions for designing blankets for TCD fusion reactors

  2. Reversed-field pinch fusion reactor

    International Nuclear Information System (INIS)

    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 Li2O 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

  3. Colliding beam fusion reactor space propulsion system

    Science.gov (United States)

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

    2000-01-01

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

  4. Communication links for fusion reactor maintenance operations

    International Nuclear Information System (INIS)

    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

  5. Topics on cryogenic design of superconducting magnet for fusion reactor

    International Nuclear Information System (INIS)

    The design concepts and the research topics on cryogenic stability, quench and refrigeration of superconducting magnets for fusion experimental reactor are reviewed. Emphasis is given to introduce their fundamental ideas as well as the results obtained by various experiments at National Institute for Fusion Science, however, some of the other recent topics are also included. (author)

  6. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

    Results of a conceptual design study of a 233U-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

  7. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-12-01

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

  8. Heat transfer in inertial confinement fusion reactor systems

    International Nuclear Information System (INIS)

    The short time and deposition distance for the energy from inertial fusion products results in local peak power densities on the order of 1018 watts/m3. This paper presents an overview of the various inertial fusion reactor designs which attempt to reduce these peak power intensities and describes the heat transfer considerations for each design

  9. An analysis of the estimated capital cost of a fusion reactor

    International Nuclear Information System (INIS)

    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)

  10. Issues in the development of a commercial fusion reactor

    International Nuclear Information System (INIS)

    From the results reported in Outline Design, it is possible to define a route towards the construction of a fusion power reactor that would produce large amounts of power (nearly 1 to 2 GWe in a single unit)at a capital cost of around $5 per watt for the fusion plant. The main technological constraints of a fusion reactor results from economics, which favors a large neutron flux at the reactor first wall. This constraint has an impact on the viability, reliability, and life time of the blanket and divertor components which are subject to important mechanical and thermal stresses, and to a large neutron fluence. In this paper the parameters of the reactor will be defined by extrapolating from the ITER Outline Design, and the issues of the reactor physics and of the blanket, divertor and magnet systems will be reviewed, with a view towards balancing the constraints resulting from economics, safety and maintenance. 12 refs., 6 figs., 1 tab

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

    International Nuclear Information System (INIS)

    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.)

  12. Fusion reactor nucleonics: status and needs

    International Nuclear Information System (INIS)

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

  13. Brief review of the fusion--fission hybrid reactor

    International Nuclear Information System (INIS)

    Much of the conceptual framework of present day fusion-fission hybrid reactors is found in the original work of the early 1950's. Present day motivations for development are quite different. The role of the hybrid reactor is discussed as well as the current activities in the development program

  14. Defensins promote fusion and lysis of negatively charged membranes.

    OpenAIRE

    Fujii, G; Selsted, M E; Eisenberg, D.

    1993-01-01

    Defensins, a family of cationic peptides isolated from mammalian granulocytes and believed to permeabilize membranes, were tested for their ability to cause fusion and lysis of liposomes. Unlike alpha-helical peptides whose lytic effects have been extensively studied, the defensins consist primarily of beta-sheet. Defensins fuse and lyse negatively charged liposomes but display reduced activity with neutral liposomes. These and other experiments suggest that fusion and lysis is mediated prima...

  15. Thermal energy and bootstrap current in fusion reactor plasmas

    International Nuclear Information System (INIS)

    For DT fusion reactors with prescribed alpha particle heating power Pα, plasma volume V and burn temperature i> ∼ 10 keV specific relations for the thermal energy content, bootstrap current, central plasma pressure and other quantities are derived. It is shown that imposing Pα and V makes these relations independent of the magnitudes of the density and temperature, i.e. they only depend on Pα, V and shape factors or profile parameters. For model density and temperature profiles analytic expressions for these shape factors and for the factor Cbs in the bootstrap current formula Ibs ∼ Cbs(a/R)1/2βpIp are given. In the design of next-step devices and fusion reactors, the fusion power is a fixed quantity. Prescription of the alpha particle heating power and plasma volume results in specific relations which can be helpful for interpreting computer simulations and for the design of fusion reactors. (author) 5 refs

  16. Fusion reactor design towards radwaste minimum with advanced shield material

    International Nuclear Information System (INIS)

    A new concept of fusion reactor design is proposed to minimize the radioactive waste of the reactor. The main point of the concept is to clear massive structural components located outside the neutron shield from regulatory control. The concept requires some reinforcement of shielding with an advanced shield material such as a metal hydride, detriation, and tailoring of a detrimental element from the superconductor. Our assessment confirmed a large impact of the concept on radwaste reduction, in that it reduces the radwaste fraction of a fusion reactor A-SSTR2 from 92 wt.% to 17 wt.%. (author)

  17. Electromagnetic analysis for fusion reactors: status and needs

    International Nuclear Information System (INIS)

    Electromagnetic effects have far-reaching implications for the design, operation, and maintenance of future fusion reactors. Two-dimensional (2-D) eddy current computer codes are available, but are of limited value in analyzing reactors. Three-dimensional (3-D) codes are needed, but are only beginning to be developed. Both 2-D and 3-D codes need verification against experimental data, such as that provided by the upcoming FELIX experiments. Coupling between eddy currents and deflections has application in fusion reactor design and is being studied both by analysis and experiment

  18. 8th International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Leotta, G G; Muon-catalyzed fusion and fusion with polarized nuclei

    1988-01-01

    The International School of Fusion Reactor Technology started its courses 15 years ago and since then has mantained a biennial pace. Generally, each course has developed the subject which was announced in advance at the closing of the previous course. The subject to which the present proceedings refer was chosen in violation of that rule so as to satisfy the recent and diffuse interest in cold fusion among the main European laboratories involved in controlled thermonuclear research (CTR). In the second half of 1986 we started to prepare a workshop aimed at assessing the state of the art and possibly of the perspectives of muon- catalyzed fusion. Research in this field has recently produced exciting experimental results open to important practical applications. We thought it worthwhile to consider also the beneficial effects and problems of the polarization ofthe nuclei in both cold and thermonuclear fusion. In preparing the 8th Course on Fusion Reactor Technology, it was necessary to abandon the tradi...

  19. Second preliminary design of JAERI experimental fusion reactor (JXFR)

    International Nuclear Information System (INIS)

    Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

  20. On the requirements for a competitive cold fusion reactor

    International Nuclear Information System (INIS)

    This paper reports on the performed parametric study which imposes requirements upon the fundamental determinants of the muon catalyzed fusion: the expense of energy per generation of one muon and the number of fusions per muon. The obtained results laid at the formulation of more severe requirements to be fulfilled for the realization of a competitive cold fusion reactor. Nevertheless the observed continuous progress in this area and wide margins of uncertainty of pertinent physical data still leave room for promising perspectives of the muon catalyzed fusion

  1. Cold fusion reactors and new modern physics

    OpenAIRE

    Huang Zhenqiang Huang Yuxiang

    2013-01-01

    The author of the "modern physics classical particle quantization orbital motion model general solution", referred to as the “new modern physics” a book. “The nuclear force constraint inertial guidance cold nuclear fusion collides” patent of invention referred to as the “cold nuclear fusion reactor” detailed technical data. Now provide to you, hope you help spread and the mainstream of modern physics of academic and fusion engineering academic communication. We work together to promote the c...

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

    Directory of Open Access Journals (Sweden)

    Sahan Halide

    2015-01-01

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

  3. Cost assessment of a generic magnetic fusion reactor

    International Nuclear Information System (INIS)

    A generic reactor model is used to examine the economic viability of generating electricity by magnetic fusion. The simple model uses components that are representative of those used in previous reactor studies of deuterium-tritium-burning tokamaks, stellarators, bumpy tori, reversed-field pinches (RFPs), and tandem mirrors. Conservative costing assumptions are made. The generic reactor is not a tokamak; rather, it is intended to emphasize what is common to all magnetic fusion rectors. The reactor uses a superconducting toroidal coil set to produce the dominant magnetic field. To this extent, it is not as good an approximation to systems such as the RFP in which the main field is produced by a plasma current. The main output of the study is the cost of electricity as a function of the weight and size of the fusion core - blanket, shield, structure, and coils. The model shows that a 1200-MW(e) power plant with a fusion core weight of about 10,000 tonnes should be competitive in the future with fission and fossil plants. Studies of the sensitivity of the model to variations in the assumptions show that this result is not sensitively dependent on any given assumption. Of particular importance is the result that a fusion reactor of this scale may be realized with only moderate advances in physics and technology capabilities

  4. Power plant systems for fusion reactors

    International Nuclear Information System (INIS)

    To investigate and compare the applicability of thermal cycles for power generation to nuclear fusion, four plant systems, i.e. direct steam turbine, in-direct steam turbine, direct gas turbine and in-direct steam turbine with gas-cooled blanket, have been designed with the estimates of their thermal efficiencies. An plant designs here are based on the same power core: fusion power 2300 MW, external heating power 58 MW, thermal power of blanket 2420 MW and divertor 490 MW. In addition, it is assumed that the construction of the power plant is near future so that the structural material would be a ferritic/martensitic steel such as F82H to be used at temperatures lower than 500-550degC. Also the divertor is always cooled by water at pressure of 10 MPa, and inlet/outlet temperature of 150-200degC/200-250degC. The removal heat from the divertor is utilized to heat the coolant fed to the blanket inlet in all above plants. The direct steam turbine cycle employs supercritical pressure water at 25 MPa and blanket inlet/outlet temperatures of 280degC/500degC. The steam out from the blanket directly flows into a high pressure turbine. The steam intermediately extracted from the high pressure turbine and/or a part of main steam from the blanket outlet is utilized to reheat the steam coming out of the high pressure turbine. Also the regenerative cycle is applied by using steams extracted from high, medium and low pressure turbines. Eventually the obtained thermal efficiency is 41.4%. The in-direct steam turbine cycle consists of the primary loop which removes the heat from the blanket and the secondary (power generation) loop by using a steam generator at their interface. The primary coolant is supercritical pressure water similar to that of above direct steam cycle, i.e. 25 MPa, 290degC/510degC. The secondary coolant is also water but with the condition of a fast breeder fission reactor, i.e. 16.3 MPa, 210degC/480degC. With reheat and regenerative cycle, the thermal

  5. Estimation of decay heat in fusion DEMO reactor

    International Nuclear Information System (INIS)

    The decay heat of activated materials is important in safety assessment of fusion DEMO reactor against loss of coolant-flow accidents. Decay heat for reactor main components of the SlimCS DEMO reactor was studied with a one-dimensional code THIDA-2. The reactor main components consist of the inboard (IB) blanket module, outboard (OB) blanket module and divertor. For a reactor with a fusion output of 3.0 GW, the decay heat of IB blanket, OB blanket, divertor and radiation shield were estimated to be as high as 8.6 MW, 30.9 MW, 10.6 MW and 1.8 MW, respectively, immediately after the shutdown of operation. The total decay heat was as high as 52 MW immediately after the shutdown and 3.1 MW one month later. The blanket produces the largest portion of decay heat, about 76%. (author)

  6. Design studies of innovatively small fusion reactor based on biomass-fusion hybrid concept: GNOME

    International Nuclear Information System (INIS)

    Conceptual design of an innovatively small tokamak reactor 'GNOME' based on a non-fission biomass-fusion hybrid concept is proposed. This fusion plant concept intends to use high-temperature heat from the blanket to generate hydrogen or synthetic fuels out of waste biomass. Since energy multiplication is expected by utilizing chemical energy of biomass, the requirement for the fusion plasma for net plant energy output is reduced to Q ≥ 5. As a result, the GNOME reactor has been designed to produce 320 MW fusion power with a 5.2 m major radius, 3.1 normalized beta and 11 T maximum field. This relatively small maximum field can be achieved by using Nb3Sn superconducting magnets. Besides, this reactor allows 3.0 m diameter space for its center solenoid coil and requires 60 MW of the input power. These features require minimal technical extensions from ITER.

  7. Fusion reactivity graphs and tables for charged particle reactions

    International Nuclear Information System (INIS)

    Graphs and tables are presented on 31 light isotope fusion reaction parameters [, n, Q/sub +/, nQ/sub +/ (for n = 1020 fuel ion species/m3 and Q/sub +/ = energy release in charged particles)] in the kinetic temperature range 1 to 1000 keV

  8. Fusion reactor development scenarios for the laser solenoid concept

    International Nuclear Information System (INIS)

    A program is described which overcomes some size problems by utilizing the fusion-fission hybrid or symbiotic technology to produce fuel for the installed LWR capacity, eliminating reliance on early fusion reactors for base load power, and taking advantage of the reduced technological demands of the fusion-fission hybrid to allow earlier introduction of these systems. The use of the fusion-fission hybrid to breed fuel for the LWR economy not only takes advantage of a very effective breeder, but also combines the technological development of the breeder and fusion power into a single, more cost effective program. Once a fusion hybrid breeder economy is established, the advent of pure fusion power will involve a much smaller, relatively risk-free technological development. The proposed program is demonstrated by a series of conceptual designs using the laser solenoid fusion concept as an example. It will be shown that the fusion-fission hybrid power plant is a project whose engineering requirements appear quite reachable at the present time and that with better knowledge of the physics and technology, smaller fusion power plants which have very attractive characteristics for the utility industry should be possible at a later time

  9. Cold fusion reactors and new modern physics

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang Huang Yuxiang

    2013-10-01

    Full Text Available The author of the "modern physics classical particle quantization orbital motion model general solution", referred to as the “new modern physics” a book. “The nuclear force constraint inertial guidance cold nuclear fusion collides” patent of invention referred to as the “cold nuclear fusion reactor” detailed technical data. Now provide to you, hope you help spread and the mainstream of modern physics of academic and fusion engineering academic communication. We work together to promote the cause of science and technology progress of mankind to contribute

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

    Science.gov (United States)

    Miley, G. H.

    1978-01-01

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

  11. Effects of a liquid lithium curtain as the first wall in a fusion reactor plasma

    Institute of Scientific and Technical Information of China (English)

    Li Cheng-Yue; J.P. Allain; Deng Bai-Quan

    2007-01-01

    This paper explores the effect of a liquid lithium curtain on fusion reactor plasma, such curtain is utilized as the first wall for the engineering outline design of the Fusion Experimental Breeder (FEB-E). The relationships between the surface temperature of a liquid lithium curtain and the effective plasma charge, fuel dilution and fusion power production have been derived. Results indicate that under normal operation, the evaporation of liquid lithium does not seriously affect the effective plasma charge, but effects on fuel dilution and fusion power are more sensitive. As an example, it has investigated the relationships between the liquid lithium curtain flow velocity and the rise of surface temperature based on operation scenario Ⅱ of the FEB-E design with reversed shear configuration and high power density. Results show that even if the liquid lithium curtain flow velocity is as low as 0.5 m/s, the effects of evaporation from the liquid lithium curtain on plasma are negligible. In the present design, the sputtering of liquid lithium curtain and the particle removal effects of the divertor are not yet considered in detail. Further studies are in progress, and in this work implication of lithium erosion and divertor physics on fusion reactor operation are discussed.

  12. Structural integrity evaluation for HIP bonded fusion DEMO reactor component

    International Nuclear Information System (INIS)

    HIP (Hot Isostatic Pressing) is applicable to fabrication technology for in-vessel components of fusion DEMO reactor, which has been fabricated by ferritic steel F82H. Verification of structural integrity of HIP bonded joint is one of the most important subjects on the development of fusion DEMO reactor component. Even now structural integrity evaluation method for HIP bonded joint is not enough estimated. This paper describes the results of low cycle fatigue test and inelastic analysis using HIP-element model under the bending load of fusion DEMO reactor. Low cycle fatigue strength and fracture/deformation mode has been obtained. SEM-observation was performed at the fracture surface of HIP-element model and the bonded joints between cooling tubes and plate were inspected. (author)

  13. Conceptual design study of Fusion Experimental Reactor (FY87FER)

    International Nuclear Information System (INIS)

    The design study of Fusion Experimental Reactor(FER) which has been proposed to be the next step fusion device has been conducted by JAERI Reactor System Laboratory since 1982 and by FER design team since 1984. This is the final report of the FER design team program and describes the results obtained in FY1987 (partially in FY1986) activities. The contents of this report consist of the reference design which is based on the guideline in FY1986 by the Subcomitees set up in Nuclear Fusion Council of Atomic Energy Commission of Japan, the Low-Physics-Risk reactor design for achieving physics mission more reliably and the system study of FER design candidates including above two designs. (author)

  14. The TITAN Reversed-Field Pinch fusion reactor study

    International Nuclear Information System (INIS)

    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/m2 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/m2; 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

  15. The need for novel neutron energy conversion schemes for DT fusion reactors

    International Nuclear Information System (INIS)

    While it is universally accepted that the DT reaction will provide the easiest method of attaining the first generation of fusion power reactors, current conceptual designs indicate that such reactors may be 1.5-2 times as expensive as an equivalent fission plant. Since the fusion nuclear island is inherently more expensive than a fission nuclear island, coupling to a conventional steam cycle balance-of-plant can only lead to a more expensive total. Accordingly, in this paper, the authors explore methods of converting 14 MeV neutron energy to electricity without employing a complex and expensive steam cycle. They show that there is a natural division in these alternative methods into schemes which are Carnot-limited (i.e., produce electricity via an intermediate heat stage) and schemes which intercept the energy flow from neutron-induced charged particles before they slow down and cause bulk heating of the absorbing medium. Furthermore, several of the prospective methods are shown to be fusion-reactor-specific, i.e., they exploit the unique properties of DT fusion and are only applicable to such reactors. In particular, they examine two innovative fusion-specific neutron conversion schemes, namely, (a) radiation-catalyzed MHD conversion which employs the existing reactor magnetic field for in-situ MHD electricity generation, and (b) excimer-channeled UV conversion which converts neutrons to electricity through intermediate UV production via an energy-channeling excimer medium. With these novel methods, the higher cost of a fusion nuclear island over that for fission could be offset by a lower ''balance-of-plant'' cost, resulting in a power source with attractive economics and enhanced environmental and safety advantages

  16. Conceptual design of light ion beam inertia nuclear fusion reactors

    International Nuclear Information System (INIS)

    Light ion beam, inertia nuclear fusion system drew attention recently as one of the nuclear fusion systems for power reactors in the history of the research on nuclear fusion. Its beginning seemed to be the judgement that the implosion of fusion fuel pellets with light ions can be realized with the light ions which can be obtained in view of accelerator techniques. Of course, in order to generate practically usable nuclear fusion reaction by this system and maintain it, many technical difficulties must be overcome. This research was carried out for the purpose of discovering such technical problems and searching for their solution. At the time of doing the works, the following policy was adopted. Though their is the difference of fine and rough, the design of a whole reactor system is performed conformably. In order to make comparison with other reactor types and nuclear fusion systems, the design is carried out as the power plant of about one million kWe output. As the extent of the design, the works at conceptual design stage are performed to present the concept of design which satisfies the required function. Basically, the design is made from conservative standpoint. This research of design was started in 1981, and in fiscal 1982, the mutual adjustment among the design of respective parts was performed on the basis of the results in 1981, and the possible revision and new proposal were investigated. (Kako, I.)

  17. Power generation system using two models for an inertial confinement fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lien, L.C.; Harada, N. [Nagaoka Univ. of Technology, Niigata (Japan). Graduate School of Engineering

    2005-08-15

    In this study, the series cooling model and the parallel cooling model of inertial confinement fusion reactor were used as a heat source for driving the MHD/Gas Turbine combined power generation system. This reactor is designed with the first wall and the blanket, which are used to collect the products of fusion reactions (including X-ray, charged particles, and neutrons) and to convert the fusion energy into thermal energy. In the series cooling model, the coolant after being heated in the blanket is re-heated again in the first wall, therefore, >2000 K working gas can be obtained. In the parallel cooling model, 1300-1700 K working gas was extracted from the blanket for driving the Gas Turbine cycle and high temperature 2000-2400 K working gas can be extracted from the first wall for driving the MHD cycle. The system using the series cooling model reached a highest plant efficiency of 58.34 per cent whereas the system using the parallel cooling model reached a highest plant efficiency of 57.49 per cent. It was found that the enthalpy extraction and the fist wall output temperature both affected the fusion output power, therefore, the plant efficiency was greatly affected by these factors. With the increase of reactor output temperature, the plant efficiency increased, however, because of the temperature limitation of the Gas Turbine and blanket, an output temperature >2400 K from reactor cannot be used. (author)

  18. Utilization of fission reactors for fusion engineering testing

    Energy Technology Data Exchange (ETDEWEB)

    Deis, G.A.; Miller, L.G.

    1985-02-08

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful.

  19. Atomistic simulations of plasma-wall interactions in fusion reactors

    International Nuclear Information System (INIS)

    Atomistic computer simulations, especially molecular dynamics, but also kinetic Monte Carlo simulations and electronic structure calculations, have proven to be a valuable tool for studying radiation effects in fusion reactor materials. In this paper, I will first review a few cases where these methods have given additional insights into the interaction between a fusion plasma and the first wall of a reactor. Then I will, in the spirit of the workshop theme of 'new directions in plasma-wall interactions' discuss some possible future avenues of research

  20. Vanadium-base alloys for fusion reactor applications

    International Nuclear Information System (INIS)

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

  1. Revised graphs of activation data for fusion reactor applications

    International Nuclear Information System (INIS)

    Activation data are required for calculation of induced activity in a fusion reactor. This report gives in graphical form, the activation data which have been evaluated based on recent measurements and calculations, for use in the activation calculation code system THIDA-2. It shows transmutation and decay chain data, activation cross sections and decay gamma-ray emission data for 152 nuclides of interest in terms of fusion reactor design. This report is an updated and enlarged version of a similar report compiled in 1982 for the activation data of 116 nuclides, which had been shown to be extremely effective in referring the activation data and in locating and correcting inappropriate data. (author)

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

    Science.gov (United States)

    McGuire, Thomas

    2015-11-01

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

  3. ITER at the international conference on fusion reactor materials

    International Nuclear Information System (INIS)

    The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

  4. PARAMETRIC SENSITIVITY STUDY OF A FUSION REACTOR DIVERTOR COOLING FINGER

    OpenAIRE

    Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper the results of coupled thermal-mechanical Finite Element (FE) analysis on the design of a fusion reactor divertor cooling finger are presented. Beside its main purpose, to remove alpha particles, helium and other impurities from the plasma stream, a divertor has to remove approximately 15% of the total thermal power of the fusion reactor. The aim of the analysis is to assess the influence of a number of physical properties of the brazing layer (BL) of the cooling finger on the o...

  5. Fusion--fission hybrid reactors based on the laser solenoid

    International Nuclear Information System (INIS)

    Fusion-fission reactors, based on the laser solenoid concept, can be much smaller in scale than their pure fusion counterparts, with moderate first-wall loading and rapid breeding capabilities (1 to 3 tonnes/yr), and can be designed successfully on the basis of classical plasma transport properties and free-streaming end-loss. Preliminary design information is presented for such systems, including the first wall, pulse coil, blanket, superconductors, laser optics, and power supplies, accounting for the desired reactor performance and other physics and engineering constraints. Self-consistent point designs for first and second generation reactors are discussed which illustrate the reactor size, performance, component parameters, and the level of technological development required

  6. Review of direct energy conversion for fusion reactors

    International Nuclear Information System (INIS)

    The direct conversion to electrical energy of the energy carried by the leakage plasma from a fusion reactor and by the ions that are not converted to neutrals in a neutral-beam injector is discussed. The conversion process is electrostatic deceleration and direct particle collection as distinct from plasma expansion against a time-varying magnetic field or conversion in an EXB duct (both MHD). Relatively simple 1-stage plasma direct converters are discussed which can have efficiencies of about 50 percent. More complex and costly (measured in $/kW) 2-, 3-, 4-, and 22-stage concepts have been tested at efficiencies approaching 90 percent. Beam direct converters have been tested at 15 keV and 2 kW of power at 70 +- 2 percent efficiency, and a test of a 120-keV, 1-MW version is being prepared. Designs for a 120-keV, 4-MW unit are presented. The beam direct converter, besides saving on power supplies and on beam dumps, should raise the efficiency of creating a neutral beam from 40 percent without direct conversion to 70 percent with direct conversion for a 120-keV deuterium beam. The technological limits determining power handling and lifetime such as space-charge effects, heat removal, electrode material, sputtering, blistering, voltage holding, and insulation design, are discussed. The application of plasma direct converters to toroidal plasma confinement concepts is also discussed

  7. Fusion reactor technology impact of alternate fusion fuels

    International Nuclear Information System (INIS)

    The initial results of a study carried out to assess some of the technology implications of non-D-T fusion fuel cycles are presented. The primary emphasis in this paper is on D-D, catalyzed-D and D-3He fuel cycles. Tokamaks and field-reversed mirrors have been selected as sample confinement concepts. The technology areas considered include first wall design considerations, shielding requirements, fuel cycle requirements and some safety and environmental considerations. Conclusions resulting from the study are also presented

  8. Concept of a charged fusion product diagnostic for NSTX

    International Nuclear Information System (INIS)

    The concept of a new diagnostic for NSTX to determine the time dependent charged fusion product emission profile using an array of semiconductor detectors is presented. The expected time resolution of 1-2 ms should make it possible to study the effect of magnetohydrodynamics and other plasma activities (toroidal Alfven eigenmodes (TAE), neoclassical tearing modes (NTM), edge localized modes (ELM), etc.) on the radial transport of neutral beam ions. First simulation results of deuterium-deuterium (DD) fusion proton yields for different detector arrangements and methods for inverting the simulated data to obtain the emission profile are discussed.

  9. Fuel provision for nonbreeding deuterium-tritium fusion reactors

    International Nuclear Information System (INIS)

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

  10. Electricity production from a pulsed tokamak fusion reactor

    International Nuclear Information System (INIS)

    A study has been undertaken to investigate the use of a possible pulsed fusion reactor to supply the national grid. Detailed models of the individual components of a 1200 MWe reactor plant were developed, including the reactor blanket, boiler and turbine generator. Using a drum boiler as a thermal energy store, full output could be maintained for reactor off-periods up to only 40 seconds, compared with an expected off-period for a pulsed tokamak fusion reactor of up to 300 seconds. Two possible solutions to this mis-match problem are considered, the first involving an externally fired superheater and reheater, which would allow the off-period to be extended to 105 seconds, and a second involving an auxiliary boiler, which would allow an indefinite off-period. Under these conditions, the plant and operating costs are estimated to be higher than the estimated cost of incorporating non-inductive current drive into a tokamak, and therefore the study suggests that it would be advantageous to develop a continuously operating tokamak fusion reactor, although other possible solutions relevant to the pulse operation should be further investigated. (Author)

  11. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in a previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes were considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the ''in-situ'' replacement of first walls using atomic coating processes were considered. The vapor deposition of carbon was shown to be promising

  12. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in the previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes was considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the in-situ replacement of first walls using atomic coating processes was considered. The vapor deposition of carbon was shown to be promising

  13. Theoretical prerequisites for creating cold fusion reactor

    International Nuclear Information System (INIS)

    The paper proposes a theoretical model which explains the energy output in cold fusion reactions, while using nickel and copper. The technique, presented by the authors, allows to determine what other isotopes of chemical elements can be used to optimize the energy output. The results can be used to create a source of energy for industrial purposes

  14. Electric power from near-term fusion reactors

    International Nuclear Information System (INIS)

    Near-term fusion reactors such as FED or INTOR will probably have primary cooling systems which operate at temperatures lower than is optimal for power production using a conventional steam cycle. This limitation may be imposed by uncertainties in materials behavior or structural limitations. There are economic motivations to demonstrate electric power production from fusion at the earliest possible date. A greater motivation is the elicitation of public interest in and support of fusion as a viable power source. This paper examines requirements and possibilities of electric power production on near-term fusion reactors using low temperature cycle technology similar to that used in some geothermal power systems. Requirements include the need for a working fluid with suitable thermodynamic properties and which is free of oxygen and hydrogen to facilitate tritium management. Thermal storage will also be required due to the short system thermal time constants on near-term reactors. It is possible to use the FED shield in a binary power cycle, and results are presented of thermodynamic analyses of this system. Thermal storage is accomplished by using the latent heat of fusion of a PbBi eutectic. The secondary loop can use R-11, R-113, or hexafluorobenzene as a working fluid. Such a system would cost about $50 million and would generate about 10 MW of electric power

  15. Heat transfer in inertial confinement fusion reactor systems

    International Nuclear Information System (INIS)

    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 1018 watts/m3. 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

  16. Tokamak Fusion Test Reactor. Final conceptual design report

    International Nuclear Information System (INIS)

    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

  17. Development of controlled thermonuclear fusion reactors and their prospect

    International Nuclear Information System (INIS)

    Controlled thermonuclear fusion seen as the most desirable form of nuclear energy utilization has reached the stage of forecasting the bright future in these several years. One of its features is the superiority over fossil fuel and nuclear fission energy in the availability of resources. The others are no anxiety of re-criticality accident and less radioactivity accumulated in plants. The present status in the core type, blanket type, supply and heating of the fuel, magnetic field and shield, and zero power conditions are described. Then the reactors being presently investigated, namely laser type, mirror type, theta pinch type, and Tokamak, are explained regarding their plant concept and technical problems. Next stage plans are formed and being put into practice in succession to the medium type experimental facilities in the world presently in operation. These aim at the achievement of zero power conditions for plasma confinement. Related countries seem to consider that they can reach the scientific feasibility experiment in the first half of 1980s. The next stage is the operation of the experimental reactors for nuclear fusion, which are supposed to generate neutrons by confining and heating the plasma using deuterium and tritium, apart from practically obtaining electric power. If everything goes well, the start of operation of proven type fusion reactors can be seen by the first half of 1990s, thus the realization of practical nuclear fusion reactors is expected at the beginning of the twenty-first century. (Wakatsuki, Y.)

  18. A view of technology maturity assessment to realize fusion reactor by Japanese young researchers

    International Nuclear Information System (INIS)

    Japanese young researchers who have interest in realizing fusion reactor have analyzed Technology Readiness Levels (TRL) in Young Scientists Special Interest Group on Fusion Reactor Realization. In this report, brief introduction to TRL assessment and a view of TRL assessment against fusion reactor projects conducting in Japan. (J.P.N.)

  19. Parametric study of the Tormac fusion reactor concept

    International Nuclear Information System (INIS)

    A preliminary but comprehensive power balance for the D-T Tormac magnetic fusion reactor concept is examined parametrically in order to establish general scaling relationships, tradeoffs, and constraints. The results are based on the simplifying assumptions of steady-state operation, a homogeneous plasma, and ideal thin-sheath, mirror-like confinement. Crucial physics uncertainties requiring further theoretical and experimental research attention are identified. Representative reactor physics operating points are generated to illustrate anticipated Tormac reactor embodiments. This study should be considered preliminary to a more detailed physics and technology modeling effort and is intended only to scope and identify possible operating points, parametric sensitivities, and potential physics/technological problems

  20. Magnetic divertors for experimental Tokamaks and fusion reactors

    International Nuclear Information System (INIS)

    Brief reports of working group discussions. These covered the requirements for a divertor in a fusion reactor including reducing impurities, exhausting the plasma and controlling the plasma-wall interactions. Divertor configurations were also reviewed and their merits and disadvantages compared. Existing divertor experiments were summarised and recommendations for further work made. Then the problems anticipated in designing a divertor for a conceptual reactor were considered. The physics of divertors and the scrape-off layer was discussed with reference to present models of plasma in divertors. Finally, experiments needed to demonstrate the feasibility of divertors for reactors and the development of specialised diagnostics for such experiments were considered. (U.K.)

  1. Mechanical design of a magnetic fusion production reactor

    International Nuclear Information System (INIS)

    The mechanical aspects of a tandem mirror and tokamak concepts for the tritium production mission are compared, and a proposed breeding blanket configuration for each type of reactor is presented in detail, along with a design outline of the complete fusion reaction system. In both cases, the reactor design is developed sufficiently to permit preliminary cost estimates of all components. A qualitative comparison is drawn between both concepts from the view of mechanical design and serviceability, and suggestions are made for technology proof tests on unique mechanical features. Detailed cost breakdowns indicate less than 10% difference in the overall costs of the two reactors

  2. Fusion Reactor Safety Research Program annual report, FY-79

    International Nuclear Information System (INIS)

    The objective of the program is the development, coordination, and execution of activities related to magnetic fusion devices and reactors that will: (a) identify and evaluate potential hazards, (b) assess and disclose potential environmental impacts, and (c) develop design standards and criteria that eliminate, mitigate, or reduce those hazards and impacts. The program will provide a sound basis for licensing fusion reactors. Included in this report are portions of four reports from two outside contractors, discussions of the several areas in which EG and G Idaho is conducting research activities, a discussion of proposed program plan development, mention of special tasks, a review of fusion technology program coordination by EG and G with other laboratories, and a brief view of proposed FY-80 activities

  3. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report describes the results of the reactor configuration/structure design for the fusion experimental reactor (FER) performed in FY 1986. The design was intended to meet the physical and engineering mission of the next step device which was decided by the subcommittee on the next step device of the nuclear fusion council. The objectives of the design study in FY 1986 are to advance and optimize the design concept of the last year because the recommendation of the subcommittee was basically the same as the design philosophy of the last year. Six candidate reactor configurations which correspond to options C ∼ D presented by the subcommittee were extensively examined. Consequently, ACS reactor (Advanced Option-C with Single Null Divertor) was selected as the reference configuration from viewpoints of technical risks and cost performance. Regarding the reactor structure, the following items were investigated intensively: minimization of reactor size, protection of first wall against plasma disruption, simplification of shield structure, reactor configuration which enables optimum arrangement of poloidal field coils. (author)

  4. Ceramic materials for fission and fusion nuclear reactors

    International Nuclear Information System (INIS)

    A general survey on the ceramics for nuclear applications is presented. For the fission nuclear reactor, the ceramics materials are almost totally used as fuel e.g. (U,Pu)O2; other types of ceramics, e.g. Uranium-Plutonium carbide and nitride, have been investigated as potential nuclear fuels. The (U,Pu)N compound is to be the fuel for the space nuclear power reactor in the U.S.A. For the fusion nuclear reactor, the ceramics should be the fundamental materials for many components: first wall, breeder, RF heating systems, insulant and shielding parts, etc. In recent years many countries are involved on the research and development of ceramic compounds with the principal purpose of being used in the fusion powerplant (year 2010-2020 ?). An effort has been even made to verify if it is possible to use more ceramic components in the fission nuclear plant (probably differntly disigned) to improve the safety level

  5. Joint ICFRM-14 (14. international conference on fusion reactor materials) and IAEA satellite meeting on cross-cutting issues of structural materials for fusion and fission applications. PowerPoint presentations

    International Nuclear Information System (INIS)

    The Conference was devoted to the challenges in the development of new materials for advanced fission, fusion and hybrid reactors. The topics discussed include fuels and materials research under the high neutron fluence; post-irradiation examination; development of radiation resistant structural materials utilizing fission research reactors; core materials development for the advanced fuel cycle initiative; qualification of structural materials for fission and fusion reactor systems; application of charged particle accelerators for radiation resistance investigations of fission and fusion structural materials; microstructure evolution in structural materials under irradiation; ion beams and ion accelerators

  6. 1st International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Knoepfel, Heinz; Safety, Environmental Impact and Economic Prospects of Nuclear Fusion

    1990-01-01

    This book contains the lectures and the concluding discussion of the "Seminar on Safety, Environmental Impact, and Economic Prospects of Nuclear Fusion", which was held at Erice, August 6-12, 1989. In selecting the contributions to this 9th meeting held by the International School of Fusion Reactor Technology at the E. Majorana Center for Scientific Cul­ ture in Erice, we tried to provide a comprehensive coverage of the many interre­ lated and interdisciplinary aspects of what ultimately turns out to be the global acceptance criteria of our society with respect to controlled nuclear fusion. Consequently, this edited collection of the papers presented should provide an overview of these issues. We thus hope that this book, with its extensive subject index, will also be of interest and help to nonfusion specialists and, in general, to those who from curiosity or by assignment are required to be informed on these as­ pects of fusion energy.

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

    International Nuclear Information System (INIS)

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

  8. Economic, CO2 Emission and Energy Assessments of Fusion Reactors

    International Nuclear Information System (INIS)

    Full text: Global warming due to rapid greenhouse gas (GHG) emission is a serious environmental problem, and fusion reactors are expected as one of safe and abundant electric power generation systems to reduce GHG emission amounts. To search for economic, environment-friendly and energy-efficient fusion reactors, system studies have been done using PEC (Physics-Engineering- Cost) code taking care of life-cycle cost of electricity (COE), CO2 gas emission rate equivalently including other GHG emission, and energy payback ratio (EPR), for magnetic fusion reactors (tokamak (TR), spherical tokamak (ST) and helical (HR)) and inertial fusion reactors (IR). At first, reactor system modeling is described and typical design parameters are derived. The magnetic fusion reactor designs strongly depend on achievable plasma beta value and permissible magnetic field strength, and inertial fusion designs depend on the driver energy and driver repetition rate. Using the PEC code, COE, CO2 emission rate and EPR can be analyzed. The former two indecies were previously evaluated by the authors, and the latter parameter EPR is defined here as a ratio of electric output energy to input energy investments required for construction, operation, fuel, replacement and decommissioning. Especially, as for TR design to reduce COE and to raise EPR, high plasma-current-drive efficiency is required for low-beta (normalized βN e = 1 — 3 GW, plant availability favail = 0.65 — 0.85, normalized βN = 3 — 5 or averaged beta (3 — 5%), maximum magnetic field strength Bmax = 10 — 16 T, thermal efficiency fth = 0.37 — 0.59, operation year (20 — 40 Years) and isentrope parameter αF = 2 — 4. These formulas might be important for making a strategy of fusion research development. As future assessments, the accident risk probability and related accident settlement expenditures should be included in COE, in addition to CO2 environmental tax and nuclear fuel tax, for the comparisons with other

  9. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.

    1977-12-19

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements. (MOW)

  10. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

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

    Science.gov (United States)

    Lampe, Martin; Manheimer, Wallace M.

    1998-11-01

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

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

    International Nuclear Information System (INIS)

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

  13. Economic, safety and environmental prospects of fusion reactors

    International Nuclear Information System (INIS)

    Controlled fusion energy is one of the long term, non-fossil energy sources available to mankind. It has the potential of significant advantages over fission nuclear power in that the consequences of severe accidents are predicted to be less and the radioactive waste burden is calculated to be smaller. Fusion can be an important ingredient in the future world energy mix as a hedge against environmental, supply or political difficulties connected with the use of fossil fuel and present-day nuclear power. Progress in fusion reactor technology and design is described for both magnetic and inertial fusion energy systems. The projected economic prospects show that fusion will be capital intensive, and the historical trend is towards greater mass utilization efficiency and more competitive costs. Recent studies emphasizing safety and environmental advantages show that the competitive potential of fusion can be further enhanced by specific choices of materials and design. The safety and environmental prospects of fusion appear to exceed substantially those of advanced fission and coal. Clearly, a significant and directed technology effort is necessary to achieve these advantages. Typical parameters have been established for magnetic fusion energy reactors, and a tokamak at moderately high magnetic field (about 7 T on axis) in the first regime of MHD stability (β ≤ 3.5 I/aB) is closest to present experimental achievement. Further improvements of the economic and technological performance of the tokamak are possible. In addition, alternative, non-tokamak magnetic fusion approaches may offer substantive economic and operational benefits, although at present these concepts must be projected from a less developed physics base. For inertial fusion energy, the essential requirements are a high efficiency (≥ 10%) repetitively pulsed pellet driver capable of delivering up to 10 MJ of energy on target, targets capable of an energy gain of about 100, reactor chambers capable of

  14. Simulated ablation of carbon wall by alpha particles for a laser fusion reactor

    International Nuclear Information System (INIS)

    Thermal reactions of materials heated by charged particles may lead to serious damage in a laser fusion reactor. When charged particles irradiate and heat the wall material with high intensity like at above 109 W/cm2, the material can be ablated. Once the wall is ablated, expanding gas or plasma can disturb the propagation of laser light irradiating the fuel target if it stagnates long enough for next laser shot. In order to understand the ablation dynamics in detail, we have performed 1-D hydro simulation to evaluate this ablation. As a new feature, we introduce the calculation of energy deposition by charged particles focusing on the interaction between ablated material and charged particles

  15. Fusion reactor materials research in China

    International Nuclear Information System (INIS)

    The fusion materials research in China is introduced. Many kinds of structural materials (such as Ti-modified stainless steel, ferritic steel, HT-9, HT-7, oxide dispersion strengthening ferritic steel), tritium breeders (lithium, Li2O, γ-LiAlO2) and plasma facing materials (PFMs) (graphite with TiC and SiC coatings) have been developed or being developed. A systematic research activities on irradiation effects, compatibility, plasma materials interaction, thermal shock during disruption, tritium production, release and permeation, neutron multiplication in Be and Pb, etc. have been performed. The research activities are summarized and some experimental results are also given

  16. Development of aluminum nitride insulator coatings for fusion reactor applications

    International Nuclear Information System (INIS)

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of various concepts, including liquid metal, molten salt, water, and helium. This report discusses the requirements of the International Thermonuclear Experimental Reactor for a self-cooled blanket that uses liquid Li and for indirectly cooled blankets that use other alkali metals such as NaK. The report discusses the requirements for an electrically insulating coating on the first-wall structural material to minimize the MHD pressure drop during the flow of liquid metal in a magnetic field. The report addresses the thermodynamics of interactions between the liquid metals (e.g., Li and NaK) and structural materials (e.g., V-base alloys and Type 316 stainless steel) and the AlN candidate electrical insulator coating, together with associated corrosion/compatibility issues. Details are presented on the AlN coating fabrication methods, and experimental data are reported for microstructures, pretreatment of the substrate, and heat treatment of coatings, coating/substrate and coating/lithium interactions, and electrical resistance before and after exposure to lithium

  17. High speed plasma focus fusion reactor

    International Nuclear Information System (INIS)

    An electrical discharge thermonuclear reactor having a capacitor which is discharged into a reaction chamber through a low inductance distribution circuit funneling discharge current to a focus point in the reaction chamber so that the magnitude of the magnetic field intensity associated with the discharge current is generally inversely proportional to the square of the distance from the focus point. Then the circuit inductance is limited to an minimum value regardless of the absolute maximum distance from the capacitor to the focus point and thus the size of the capacitor. The distribution circuit has two outward-branching, interpenetrating three dimensional circuit networks of opposite polarity conveniently fabricated by stacking conductor plates having a generally cylindrical geometry. The distribution circuit spherically surrounds the reaction chamber so far as is practical so that the discharge rate, power and energy transfer to the reaction chamber are maximized and thus reducing the required size of the reactor

  18. Modular stellarator reactor: a fusion power plant

    International Nuclear Information System (INIS)

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

  19. Laser fusion hybrid reactor systems study

    International Nuclear Information System (INIS)

    The work was performed in three phases. The first phase included a review of the many possible laser-reactor-blanket combinations and resulted in the selection of a ''demonstration size'' 500 MWe plant for further study. A number of fast fission blankets using uranium metal, uranium-molybdenum alloy, and uranium carbide as fuel were investigated. The second phase included design of the reactor vessel and internals, heat transfer system, tritium processing system, and the balance of plant, excluding the laser building and equipment. A fuel management scheme was developed, safety considerations were reviewed, and capital and operating costs were estimated. Costs developed during the second phase were unexpectedly high, and a thorough review indicated considerable unit cost savings could be obtained by scaling the plant to a larger size. Accordingly, a third phase was added to the original scope, encompassing the redesign and scaling of the plant from 500 MWe to 1200 MWe

  20. Reactor design considerations for inertial confinement fusion

    International Nuclear Information System (INIS)

    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

  1. Fusion reactors: physics and technology. Annual progress report

    International Nuclear Information System (INIS)

    Fusion reactors are designed to operate at full power and generally at steady state. Yet experience shows the load variations, licensing constraints, and frequent sub-system failures often require a plant to operate at fractions of rated power. The aim of this study has been to assess the technology problems and design implications of startup and fractional power operation on fusion reactors. The focus of attention has been tandem mirror reactors (TMR) and we have concentrated on the plasma and blanket engineering for startup and fractional power operation. In this report, we first discuss overall problems of startup, shutdown and staged power operation and their influence on TMR design. We then present a detailed discussion of the plasma physics associated with TMR startup and various means of achieving staged power operation. We then turn to the issue of instrumentation and safety controls for fusion reactors. Finally we discuss the limits on transient power variations during startup and shutdown of Li17Pb83 cooled blankets

  2. High efficiency direct energy conversion in controlled thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Still many physical and engineering subjects must be overcome before the realization of commercial nuclear fusion reactors. But if its realization is aimed at in the middle period of 21st century, the competition with other energy sources, particularly advanced nuclear fission reactors, is the problem. In order that nuclear fusion reactors are accepted by the society, it is indispensable to verify their superiority in the fields of the safety, reliability and economical efficiency. As one of their many potential superiority, there is the high efficiency direct electric power generation related to nuclear fusion reactors. In principle, this is to convert the kinetic energy that plasma and ion beam possess directly to electric energy through electromagnetic fields, and high efficiency that largely surpasses that by conventional steam power generation is obtainable. In this report, as the methods of representative direct energy conversion, electrostatic energy conversion, energy conversion by electromagnetic waves and inductive energy conversion are explained by showing the concrete examples. (K.I.)

  3. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements

  4. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  5. Quantification of structural materials for reactor systems: synergy's in materials for fusion/fission reactors and advanced fission reactor

    International Nuclear Information System (INIS)

    In nuclear technology a lot of experience has been accumulated meanwhile from reactor programmes for ferritic alloys, austenitic steels and Ni-based alloys as main component materials during R and D, design, construction and operation. Generally materials are a key issue for a safe and reliable operation of -NPPs. Many grades investigated are of interest for the design of GenIVs and fusion reactors. Synergisms of materials, material technologies, mechanical data, corrosion and other topics -for the qualification of materials for nuclear systems are generally discussed and information on a qualification procedure is compiled. Also some lessons learned from fabrication, test programmes or operation of NPPs are provided. A special problem is the fusion system because a final validation for alloy performance in the long term will need irradiation under realistic -fusion condition anticipated in a high-energetic, fusion-specific intense neutron source such as (IFMIF), the International Fusion Materials Irradiation Facility. (author)

  6. Charged fusion product loss measurements using nuclear activation

    International Nuclear Information System (INIS)

    In ITER, α particle loss measurements will be required in order to understand the alpha particle physics. Techniques capable of operating in a fusion reactor environment need further development. Recent experimental studies on JET demonstrated the potential of nuclear activation to measure the flux of escaping MeV ions. New results from MeV ion induced activation of metallic, ceramic, and crystal samples placed near the plasma edge are reported. Activation products were measured as function of orientation with respect to the magnetic field as well as function of the distance to the plasma. Sample activity was measured using ultralow-level gamma-ray spectrometry. Distribution of 14.68 MeV fusion proton induced activation products is strongly anisotropic in agreement with simulations and falls off sharply with increasing distance to the plasma. Prospects for using the technique in ITER are discussed.

  7. Computer simulation of multi-elemental fusion reactor materials

    International Nuclear Information System (INIS)

    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

  8. Feasibility of Reduced Tritium Circulation in the Heliotron Reactor by Enhancing Fusion Reactivity Using ICRF

    OpenAIRE

    Yanagi, Nagato; SHYSHKIN, Oleg A.; Goto, Takuya; Kasahara, Hiroshi; MIYAZAWA, Junichi; SAGARA, Akio

    2011-01-01

    A scheme for reducing the tritium fraction in DT fusion reactors is investigated by means of enhancing the fusion reactivity using high-power ICRF heating in heliotron reactors. We assume a situation that the density fraction of tritons is less than 10%, and the minority tritons are accelerated by ICRF waves. We then analyze the increase of fusion reactivity by assuming an effective temperature of high-energy tritons and examine the possibility of realizing a fusion reactor with this concept....

  9. HIP technologies for fusion reactor blankets fabrication

    International Nuclear Information System (INIS)

    The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangement flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed

  10. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  11. Fusion reactor blanket/shield design study

    International Nuclear Information System (INIS)

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  12. Computer simulation of tritium releases in inertial fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perlado, J.M.; Velarde, M. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM (Spain)

    2000-07-01

    Accidental releases of tritium from Inertial Fusion reactors are presented. A well-established computer code, MACCS2, is used with realistic models. Release fractions of 1 - 10 - 50 - 100 % of inventories are considered, with height of emissions 10, 30, 60 m, and duration of 10 min. and 2 hours. Only early emergency phase is considered with mitigative actions and shielding factors. It is concluded that except in 100 % releases for some reactors and heights the effective doses to workers and general population does not exceed the regulatory limits. Differences with very conservative results can attain 2 orders of magnitude. (authors)

  13. A method of safety assurance for fusion experimental reactor

    International Nuclear Information System (INIS)

    The present report describes safety assurance method for fusion experimental reactor. The ALARA (As Low As Reasonably Achievable) principle for a normal condition and the defence in depth principle for states deviated from the normal condition can be used as basic principles of safety assurance of the reactor. The method includes safety design for systems, importance categorization method to impose suitable demands to their systems, safety evaluation method to validate the design and application of the method. It is considered that this method can be a strong candidate for safety assurance method. (author)

  14. Neutron irradiation of candidate ceramic breeder materials of fusion reactors

    International Nuclear Information System (INIS)

    In the context of the European programs for the future fusion reactors, the Process Chemistry Department of ENEA, Casaccia Center (Rome), has been involved in preparing ceramic blanket materials as tritium breeders; a special consideration has been addressed to the nuclear characterization of LiAlO2 and Li2ZrO3. In this paper are reported neutron irradiation of ceramic specimens in TRIGA reactor and γ-spectrometric measurements for INAA purposes; and isothermal annealing of the irradiated samples and tritium extraction, by using an 'out of pile' system. (author) 3 refs.; 4 figs.; 4 tabs

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

    Science.gov (United States)

    Colling, B R; Monk, S D

    2012-07-01

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

  16. Conceptual design study of Fusion Experimental Reactor (FY87FER)

    International Nuclear Information System (INIS)

    This report describes the results of the design coordination and the conceptual design study on plant systems which have been carried out as a part of the design work for the Fusion Experimental Reactor (FY87 FER). The former contains the selection of the reference concept for FY87 FER and giving it flexibility, directions for study and assessment of low physics risk reactors, and the revision of system integration, while the latter mainly describes the design philosophies, construction of systems, and the results of designs and analyses of processes and systems. (author)

  17. ELMO Bumpy Torus fusion-reactor design study

    International Nuclear Information System (INIS)

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is described. Those features that are unique to the EBT confinement concept are emphasized, with subsystems and balance-of-plant items that are generic to magnetic fusion being adopted from past, more extensive tokamak reactor designs. This overview paper stresses the design philosophy and asumptions that led to an economic, 35-m major-radius design that at 1.4 MW/m2 wall loading generates 4000 MWt with a 15% recirculating power fraction

  18. Maintenance of torus components for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Maintenance of torus components is one of key technologies for the Fusion Experimental Reactor (FER), which is the device planned to succeed the JT-60 tokamak device. The objective of the present study is to develop a reliable, feasible and simple maintenance systems for torus components such as divertor and movable shield modules. This paper describes the reactor structure and its maintenance scheme of FER and the maintenance systems for torus components. A 1/4-scale mock-up of FER is also introduced, which was made to demonstrate the feasibility of the maintenance system for torus components of FER

  19. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 2000C. The design description and results of the prototype capsule performance are presented

  20. Simulations of carbon sputtering in fusion reactor divertor plates

    International Nuclear Information System (INIS)

    The interaction of edge plasma with material surfaces raises key issues for the viability of the International Thermonuclear Reactor (ITER) and future fusion reactors, including heat-flux limits, net material erosion, and impurity production. After exposure of the graphite divertor plate to the plasma in a fusion device, an amorphous C/H layer forms. This layer contains 20-30 atomic percent D/T bonded to C. Subsequent D/T impingement on this layer produces a variety of hydrocarbons that are sputtered back into the sheath region. We present molecular dynamics (MD) simulations of D/T impacts on amorphous carbon layer as a function of ion energy and orientation, using the AIREBO potential. In particular, energies are varied between 10 and 150 eV to transition from chemical to physical sputtering. These results are used to quantify yield, hydrocarbon composition and eventual plasma contamination

  1. Feasibility of HTS magnet option for fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies are being carried out on the application of high-temperature superconducting (HTS) conductors and coils to the magnet systems of fusion reactors. A 100-kA-class HTS conductor is required to be applied at high magnetic fields of > 12 T. A simple stack of YBCO tapes embedded in copper and stainless-steel jackets is found to be a practical approach to producing large-scale conductors that exhibit high cryogenic stability and mechanical rigidity. The feasibility of the segmented fabrication method for large complex HTS coils, such as the helical coils in the LHD-type helical fusion reactor FFHR-d1, is being investigated by developing mechanical bridge-type lap joint technology of HTS conductors. (author)

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

    Science.gov (United States)

    Tabarés, Francisco L.

    2016-01-01

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

  3. Comparative breeding characteristics of fusion and fast reactors.

    Science.gov (United States)

    Fortescue, P

    1977-06-17

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

  4. Hydrogen production from fusion reactors coupled with high temperature electrolysis

    International Nuclear Information System (INIS)

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and complement 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. Processes which may be considered for this purpose include electrolysis, thermochemical decomposition or thermochemical-electrochemical hybrid cycles. Preliminary studies at Brookhaven indicate that high temperature electrolysis has the highest potential efficiency for production of hydrogen from fusion. Depending on design electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets

  5. Fusion Reactor Materials Program Plan. Section IV. Special purpose materials

    International Nuclear Information System (INIS)

    Components that were considered include breeding materials, coolants, materials for tritium service, graphite (boronated) and silicon carbide, ceramics, heat-sink materials, and magnet materials. The Task Group on Special Purpose Materials has limited its purview to crucial and generic materials problems that must be resolved if a given class of devices such as mirrors or tokamaks is to succeed. For the moment, the materials problems associated with the fusion-reactor balance of plant have been ignored; but it must be recognized that, at a later date, this area could become a major source of problems. Assumptions made in this analysis were that the goal of the program is to demonstrate commercial fusion power by the end of this century; that only pure fusion systems were considered; that only normal operating conditions were considered for long-life applications; and that radioactive waste disposal is a manageable problem

  6. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  7. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

  8. Study on spent fuel rejuvenation in PROMETHEUS fusion reactor

    International Nuclear Information System (INIS)

    This study presents the spent fuel rejuvenation potential of the PROMETHEUS-H fusion reactor. For this purpose, three different spent fuels were selected, i.e. (1) CANDU (2) PWR-UO2 and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the modified PROMETHEUS-H fusion reactor. The FZ was cooled with high pressure helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE 4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years with a 75% plant factor (η) under a first wall neutron load (P) of 4.7 MW/m2. The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δt) of 1/12 year (one month) by using the interface program (code). In all investigated spent fuel cases, the tritium self sufficiency is maintained for the DT fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU reactor after a regeneration time of ∼5.5 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO2 and PWR-MOX spent fuels after 41 and 35 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during the spent fuel rejuvenation. Consequently, the modified PROMETHEUS-H fusion reactor has high neutronic performance for the rejuvenation of the spent fuels

  9. Hydrogen isotopes transport parameters in fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Serra, E. [Politecnico di Torino (Italy). Dipartimento di Energetica; Benamati, G. [ENEA Fusion Division, CR Brasimone, 40032 Camungnano, Bologna (Italy); Ogorodnikova, O.V. [Moscow State Engineering Physics Institute, Moscow 115409 (Russian Federation)

    1998-06-01

    This work presents a review of hydrogen isotopes-materials interactions in various materials of interest for fusion reactors. The relevant parameters cover mainly diffusivity, solubility, trap concentration and energy difference between trap and solution sites. The list of materials includes the martensitic steels (MANET, Batman and F82H-mod.), beryllium, aluminium, beryllium oxide, aluminium oxide, copper, tungsten and molybdenum. Some experimental work on the parameters that describe the surface effects is also mentioned. (orig.) 62 refs.

  10. Hydrogen isotopes transport parameters in fusion reactor materials

    Science.gov (United States)

    Serra, E.; Benamati, G.; Ogorodnikova, O. V.

    1998-06-01

    This work presents a review of hydrogen isotopes-materials interactions in various materials of interest for fusion reactors. The relevant parameters cover mainly diffusivity, solubility, trap concentration and energy difference between trap and solution sites. The list of materials includes the martensitic steels (MANET, Batman and F82H-mod.), beryllium, aluminium, beryllium oxide, aluminium oxide, copper, tungsten and molybdenum. Some experimental work on the parameters that describe the surface effects is also mentioned.

  11. Contributions to the sixth international conference on fusion reactor materials

    International Nuclear Information System (INIS)

    The ICFRM series has documented progress in the field of fusion reactor materials since the first conference held in Tokyo in 1984. The conference series has continually increased its coverage to the point where it now includes the comprehensive range of materials science and technology areas that enable systems designers to meet the needs of current experiments and to present innovative solutions for future energy systems. This publication contains five contributions to the sixth international conference which have each been indexed separately

  12. Evaluation of structural integrity for maintenance of nuclear fusion reactors

    International Nuclear Information System (INIS)

    Aiming at optimization of maintenance of nuclear fusion reactors, we propose a new concept of maintenance that components including defects under control can be kept in use without replacement of them. For the purpose, a life-cycle simulation of fatigue crack behavior was carried out, taking an example of the ITER vacuum vessel. The simulation shows that there is no possibility that virtual defects penetrate the vessel, which implies no inspection is required based on the present concept. (author)

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

    International Nuclear Information System (INIS)

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

  14. Mechanical properties along interfaces of bonded structures in fusion reactors

    International Nuclear Information System (INIS)

    Proper assessment of the mechanical properties along interfaces of bonded structures currently used in many fusion reactor designs is essential to compare the different fabrication techniques. A Mechanical Properties Microprobe (MPM) was used to measure hardness and Young's modules along the interfaces of Be/Cu bonded structure. The MPM was able to distinguish different fabrication techniques by a direct measurement of the hardness, Young's modules, and H/E2 which reflects the ability of deformation of the interfacial region

  15. Helium generation in fusion reactor materials

    International Nuclear Information System (INIS)

    The work performed under this giant included an analysis of the multiple-step helium production mechanism discovered in iron following long-term mixed-spectrum reactor exposure, the measurement of a large number of samples irradiated in fast-neutron environments for cross section determinations, the initial mapping of the neutron fluence distribution for a high-fluence T(d,n) irradiation experiment, the initial measurements of helium production in materials irradiated by 10-MeV neutrons, and the initiation of a joint experiment with ANL to measure the spectrum-integrated Be(n,2n) cross section at lower neutron energies. This work is summarized in the present report. The work is ongoing, and this document thus provides a status report rather than final numerical data

  16. Electrical insulator requirements for mirror fusion reactors

    International Nuclear Information System (INIS)

    The requirements for mirror fusion electrical insulators are discussed. Insulators will be required at the neutral beam injectors, injector power supplies, direct converters, and superconducting magnets. Insulators placed at the neutral beam injectors will receive the greatest radiation exposure, 1014 to 1016 neutrons/m2.s and 0.3 to 3 Gy/s (105 to 106 R/h) of gamma rays, with shielding. Direct converter insulators may receive the highest temperature (up to 13000K), but low voltage holding requirements. Insulators made from organic materials (e.g., plastics) for the magnet coils may be satisfactory. Immediate conductivity increases of all insulators result from gamma irradiation. With an upper limit to gamma flux exposures of 300 Gy/s in a minimally shielded region, the conductivity could reach 10-6 S/m. Damage from neutron irradiation may not be serious during several years' exposure. Surface changes in ceramics at the neutral beam injector may be serious. The interior of the injector will contain atomic hydrogen, and sputtering may transfer material away from or onto the ceramic insulators. Unknown and potentially damaging interactions between irradiation, electric fields, temperature gradients, cycling of temperature, surface and joint reactions, sputtering, polarization, and electrotransport in the dielectrics are of concern. Materials research to deal with these problems is needed

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

    Science.gov (United States)

    Costley, A. E.

    2016-06-01

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

  18. Neutron dosimetry for radiation damage in fission and fusion reactors

    International Nuclear Information System (INIS)

    The properties of materials subjected to the intense neutron radiation fields characteristic of fission power reactors or proposed fusion energy devices is a field of extensive current research. These investigations seek important information relevant to the safety and economics of nuclear energy. In high-level radiation environments, neutron metrology is accomplished predominantly with passive techniques which require detailed knowledge about many nuclear reactions. The quality of neutron dosimetry has increased noticeably during the past decade owing to the availability of new data and evaluations for both integral and differential cross sections, better quantitative understanding of radioactive decay processes, improvements in radiation detection technology, and the development of reliable spectrum unfolding procedures. However, there are problems caused by the persistence of serious integral-differential discrepancies for several important reactions. There is a need to further develop the data base for exothermic and low-threshold reactions needed in thermal and fast-fission dosimetry, and for high-threshold reactions needed in fusion-energy dosimetry. The unsatisfied data requirements for fission reactor dosimetry appear to be relatively modest and well defined, while the needs for fusion are extensive and less well defined because of the immature state of fusion technology. These various data requirements are examined with the goal of providing suggestions for continued dosimetry-related nuclear data research

  19. High density, high magnetic field concepts for compact fusion reactors

    International Nuclear Information System (INIS)

    One rather discouraging feature of our conventional approaches to fusion energy is that they do not appear to lend themselves to a small reactor for developmental purposes. This is in contrast with the normal evolution of a new technology which typically proceeds to a full scale commercial plant via a set of graduated steps. Accordingly' several concepts concerned with dense plasma fusion systems are being studied theoretically and experimentally. A common aspect is that they employ: (a) high to very high plasma densities (∼1016cm-3 to ∼1026cm-3) and (b) magnetic fields. If they could be shown to be viable at high fusion Q, they could conceivably lead to compact and inexpensive commercial reactors. At least, their compactness suggests that both proof of principle experiments and development costs will be relatively inexpensive compared with the present conventional approaches. In this paper, the following concepts are considered: (1) The staged Z-pinch, (2) Liner implosion of closed-field-line configurations, (3) Magnetic ''fast'' ignition of inertial fusion targets, (4) The continuous flow Z-pinch

  20. Study meeting on 'criteria for materials of nuclear fusion reactors'

    International Nuclear Information System (INIS)

    This study meeting was held on March 1 and 2, 1984, at the Institute of Plasma Physics, Nagoya University. Recently, the problems required for the materials of nuclear fusion reactors have become considerably clear. The problem of the high concentration damage due to 14 MeV neutrons and the problem of surface materials have been well known from the beginning, but moreover, the radioactivation of materials, the problem of safety, and the feasibility of remote operation related to it have become urgent problems. Besides, the plan of large scale facilities as the means of promoting research is one of the important themes. The research on materials must take part in the whole technological problems which enable the construction of actual nuclear fusion devices. This study meeting was held as a part of the R project of the Institute of Plasma Physics, Nagoya University, but it aimed at grasping the present status and discussing the future perspective of the materials of nuclear fusion reactors, and examining the criteria for nuclear fusion materials. The gists of 23 lectures presented at the meeting are collected in this report. (Kako, I.)

  1. Inertial fusion reactors using Compact Fusion Advanced Rankine (CFARII) MHD conversion

    International Nuclear Information System (INIS)

    This study evaluates the potential performance (efficiency and cost) of inertial fusion reactors assumed capable of vaporizing blankets of various working materials to a temperature (10,000-20,000 K) suitable for economical MHD conversion in a Compact Fusion Advanced Rankine II (CFARII) power cycle. Using a conservative model, 1-D neutronics calculations of the fraction of fusion yield captured as a function of the blanket thickness of Flibe, lithium and lead-lithium blankets are used to determine the optimum blanket thickness for each material to minimize CoE for various assumed fusion yields, 'generic' driver costs, and target gains. Lithium-hydride blankets are also evaluated using an extended neutronics model. Generally optimistic ('advanced') combinations of lower driver cost/joule and higher target gain are assumed to allow high enough fusion yields to vaporize and ionize target blankets thick enough to stop most 14 MeV neutrons, and to breed tritium. A novel magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the CFARII Balance-of-Plant (BoP), consisting of a supersonic plasma jet, MHD generator, and 'raindrop' condensor. High fusion yields (20 to 80 GJ) are found necessary to heat and ionize the Flibe, lithium, and lead-lithium blankets for MHD conversion, with initial solid thicknesses sufficient to capture most of the fusion yield. Much smaller fusion yields (1 to 20 GJ) are required for lithium-hybride blankets. For Flibe, lithium, and lead-lithium blankets, improvements in target gain and/or driver cost/joule, characterized by a 'Bang per Buck' figure-of-merit of > or ∼20 joules yield per driver Dollar, would be required for competitive CoE, while a figure-of-merit of > or ∼1 joule yield per driver Dollar would suffice for lithium-hybride blankets. Advances in targets/driver costs would benefit any IFE reactor, but the very low CFARII BoP costs (contributing only 3 mills/kWh for CoE) allows this

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-03-01

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

  4. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  5. Fusion reactor development using high power particle beams

    International Nuclear Information System (INIS)

    The present paper outlines major applications of the ion source/accelerator to fusion research and also addresses the present status and future plans for accelerator development. Applications of ion sources/accelerators for fusion research are discussed first, focusing on plasma heating, plasma current drive, plasma current profile control, and plasma diagnostics. The present status and future plan of ion sources/accelerators development are then described focusing on the features of existing and future tokamak equipment. Positive-ion-based NBI systems of 100 keV class have contributed to obtaining high temperature plasmas whose parameters are close to the fusion break-even condition. For the next tokamak fusion devices, a MeV class high power neutral beam injector, which will be used to obtain a steady state burning plasma, is considered to become the primary heating and current drive system. Development of such a system is a key to realize nuclear fusion reactor. It will be entirely indebted to the development of a MeV class high current negative deuterium ion source/accelerator. (N.K.)

  6. A comparison of the radiologic impact of electricity production by means of pressurized water reactors or Tokamak fusion reactors

    International Nuclear Information System (INIS)

    The impacts of respectively light water reactors and a planned fusion reactor, for which tritium-deuterium fusion reactions will act as energy source, have been compared. The comparison is based on a generated capacity of 1 GWe.y, using the following criteria: fuel inventories, radioactive releases, collective effective dose equivalent commitments to the public and the volume of waste. The accidental risk is not introduced. Fusion reactor parameters are still subject to uncertainties, which prevent accurate quantification of radionuclide releases (tritium apart) from the nuclear plant. Only orders of magnitude extrapolated from values for the NET Tokamak are given. Despite these uncertainties, it would seem more interesting, from the dosimetric point of view, to use fusion reactors to produce electricity, although problems of radioactive releases, handling and long-term storage of radioactive waste would remain. Fusion reactors also generate high-level wastes with long-term exposure rates that are lower than those of light water reactors

  7. Neutron streaming evaluation for the DREAM fusion power reactor

    International Nuclear Information System (INIS)

    Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction. (author)

  8. Conceptual design study of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. During two years from 1984 to 1985 FER concept was reviewed and redesigned. This report is the summary of the results obtained in the review and redesign activities in 1984 and 85. In the first year FER concept was discussed again and its frame work was reestablished. According to the new frame work the major reactor components of FER were designed. In the second year the whole plant system design including plant layout plan was conducted as well as the more detailed design analysis of the reactor conponents. The newly established frame for FER design is as follows: 1) Plasma : Self-ignition. 2) Operation scenario : Quasi-steady state operation with long burn pulse. 3) Neutron fluence on the first wall : 0.3 MWY/M2. 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

  9. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report describes the results of conceptual design study on plant systems for the Fusion Experimental Reactor (FY86 FER). Design studies for FER plant systems have been continued from FY85, especially for design modifications made in accordance with revisions of plasma scaling parameters and system improvements. This report describes 1) system construction, 2) site and reactor building plan, 3) repaire and maintenance system, 4) tritium circulation system, 5) heating, ventilation and air conditioning system, 6) tritium clean-up system, 7) cooling and baking system, 8) waste treatment and storage system, 9) control system, 10) electric power system, 11) site factory plan, all of which are a part of FY86 design work. The plant systems described in this report generally have been based on the FY86 FER (ACS Reactor) which is an one of the six candidates for FER. (author)

  10. Technology and economics of hydrogen production from fusion reactors

    International Nuclear Information System (INIS)

    The technology, economics, and environmental effects of producing synthetic fuels (H2 gas, H2 liquid, and methanol) based on fusion (CTR) reactors are assessed. Four United States energy systems (2020 A.D.) with different degrees of CTR implementation are compared: in System A, no CTR input is assumed; in System B, CTRs replace 50 percent of nuclear fission electric; in System C, CTRs supply all electrical demand, produce synthetic fuels to replace all oil and gas imports, and eliminate strip mining; and in System D, CTRs supply all electrical demand and virtually all fuel demand. CTR reactor costs are analyzed in detail for a range of containment parameters, reactor outputs, and first well loadings for DT and catalyzed DD fuel cycles

  11. Metrology/viewing system for next generation fusion reactors

    International Nuclear Information System (INIS)

    Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system

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

    International Nuclear Information System (INIS)

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

  13. Status of fusion power reactor design activity in JAERI

    International Nuclear Information System (INIS)

    A number of fusion power reactor design studies conducted in JAERI over the past 22 years are reviewed and the present status of reactor studies is introduced. A helium gas cooled power reactor which has a blanket with solid lithium ceramics such as Li2O for the breeding material and incolloy-800 for the structural material was proposed in 1973. This is one of the first reactor design using a lithium ceramics blanket concept. Another power reactor design called the Swimming Pool-type Tokamak Reactor (SPTR-P) was completed in 1983. In SPTR-P, the tokamak reactor is submerged in a water pool to utilize water as the shield to reduce long term radioactive waste and to achieve easy repair and maintenance. In 1990, the Steady State Tokamak Reactor (SSTR) concept was proposed as a realistic fusion reactor to be built in the near future. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required to maintain steady state operation. It is cooled by pressurized water and uses low activation ferritic steel as the structural material. A study of SSTR-2 to improve the safety and economic aspects of SSTR by replacing the water coolant by a mixture of helium gas and fine solid particles was carried out in 1992. Light yet highly heat resistant material, titanium aluminide intermetallic compound (TiAl), is used as the structural material. The net thermal efficiency larger than 40 % can be achieved with the gas-particulate mixture at 5 MPa and the outlet coolant temperature of 700degC. A concept of a Drastically Easy Maintenance (DREAM) tokamak reactor has recently been proposed. For easy replacement of blanket and divertor, a plasma configuration with high aspect ratio around 6 and a small number of torus sectoring of 12 are selected. A 1/12 torus sector is horizontally pulled out between the TF coils with a straight radial motion. It is estimated that the availability of 85 % can be achieved. (author)

  14. Enhanced Charged Higgs Production through W^\\pm-Higgs Fusion

    CERN Document Server

    Arhrib, Abdesslam; Lee, Jae Sik; Lu, Chih-Ting

    2015-01-01

    We study the associated production of a charged Higgs boson with a bottom quark and a light quark at the LHC via p p \\to H^\\pm\\,b\\,j in the Two Higgs Doublet Models (2HDMs). Using the effective W approximation, we show that there is exact cancellation among various Feynman diagrams in high energy limit. This may imply that the production of charged Higgs can be significantly enhanced in the presence of large mass differences among the neutral Higgs bosons via W^\\pm-Higgs fusion in the p p \\to H^\\pm\\,b\\,j process. Particularly, we emphasize the potential enhancement due to a light pseudoscalar boson $A$, which is still allowed by the current data by which we explicitly calculate the allowed regions in (M_A,\\,\\tan\\beta) plane, and show that the production cross section can be as large as 0.1 pb for large $\\tan\\beta$. We also show that the transverse momentum distribution of the b quark can potentially distinguish the W^\\pm-A fusion diagram from the top diagram. Finally, we point out further enhancement when we ...

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

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

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

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

    OpenAIRE

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

    2015-01-01

    Fusion has a practically limitless fuel supply and is attractive as an energy source. The main goal of fusion research is to construct and operate an energy generating system. Fusion researches also contains fusion structural materials used fusion reactors. Material issues are very important for development of fusion reactors. Therefore, a wide range of fusion structural materials have been considered for fusion energy applications. Zirconium (Zr), Niobium (Nb) and Tantalum (Ta) containing al...

  17. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 1014 cm-3.s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

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

    International Nuclear Information System (INIS)

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

  19. Simulation of Plasma Performance of the Tokamak DEMO Nuclear Fusion Reactors Design

    International Nuclear Information System (INIS)

    The DEMOnstration Power Plant is intended to be the first fusion reactor to generate electrical power. It is designed to produce up to 500 megawatts of electricity which will require a thermal output of around 1500 megawatts. If DEMO is built at roughly the same size as ITER, it will require much higher heat flux through the reactor walls and improved plasma performance. As a reference design, DEMO reactor was designed with reference to the SSTR (steady-state tokamak reactor) using the BALDUR code. Plasma evaluation analysis of the fusion reactor DEMO using BALDUR code is performed to predict the performance of fusion power for the DEMO reactor.

  20. Mobile reactor concepts as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ignition-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a four-rail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were unplugged and returned to a large, centrally located hot shop. A similar concept is envisioned for compact fusion reactor testing

  1. The mobile reactor concept as applied to testing of compact fusion reactors

    International Nuclear Information System (INIS)

    Compact fusion reactor concepts have recently received increased emphasis because of advantages principally related to their low cost and short development time. Physics experiments are underway and test results are sufficiently encouraging to merit consideration of ''ignition''-type experiments. Since experiments of this nature involve radioactivity, the requirement for test facilities which incorporate remote handling capabilities becomes apparent. One approach to a test facility concept which has particularly attractive features is based on the mobile test reactor concept employing facilities such as are found at the Idaho National Engineering Laboratory (INEL). The mobile reactor test concept was developed in the 1950s and was used extensively in the testing of aircraft nuclear propulsion reactors at the INEL. In this instance, test reactors were assembled on a dolly and were transported to and from test facilities on a fourrail track system. Nuclear operations were conducted from heavily shielded underground control rooms and, for major maintenance operations, the reactors were ''unplugged'' and returned to a large, centrally located ''hot'' shop. A similar concept is envisioned for compact fusion reactor testing

  2. Modified wetted-wall inertial fusion reactor concept

    International Nuclear Information System (INIS)

    Limitations on reactor pulse repetition rate and uncertainties with respect to assurance of first wall protection in LASL wetted-wall inertial fusion reactor concepts, in which restoration of cavity conditions to those required for acceptable driver energy pulse transmission following pellet microexplosion is accomplished by exhaust of ablated liquid metal through nozzles and protective films are formed by forcing liquid metal through porous first walls, can be circumvented through alternative methods of cavity clearing and protective film formation. Exploratory analyses indicate that our modified wetted-wall concept, in which protective liquid metal films are injected directly onto cavity walls through slit nozzles to ensure first wall protection and are held there by centrifugal forces and cavity clearing occurs by condensation of vapor on film liquid not ablated as a result of pellet x ray and debris ion energy deposition, can be operated at substantially higher repetition rates. The new mode of operation appears to be attractive for heavy ion fusion, for which constraints on cavity design options may be more severe, as well as laser fusion. Numerical results of the exploratory analyses, plus discussion of aspects of the new concept requiring further work, are presented

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-09-01

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

  5. Workshop a data and planning center of NIFS in 2003. Assessment of fusion R and D. The critical issues of demonstration fusion reactor

    International Nuclear Information System (INIS)

    In order to asses the fusion research and development, we discussed on proposals which might have possibility to overcome the critical issues of a demonstration fusion reactor. Advanced Tokamak reactor concept and a fast ignition method of laser fusion are proposed to reduce the high capital cost which is common for both of magnetic fusion and inertial fusion. Hopeful divertor concepts are also proposed for both of Tokamak and Helical reactor. These summarized presentations are appended. (author)

  6. A Fusion-Fission Reactor Concept based on Viable Technologies

    International Nuclear Information System (INIS)

    Full text: The world needs a great deal of carbon free energy for civilization to continue. Nuclear power is attractive for helping cut carbon emissions and reducing imports of fossil fuel. It is commonly realized that it needs hard work before pure fusion energy could be commercially and economically utilized. Some countries are speeding up the development of their fission industry. In China, the government has decided to develop nuclear power with a mid-term target of ∼40 GWe in 2020. If only PWR is used to meet the huge nuclear capacity requirement, there may be a shortage of fissile uranium and an increase of long-lived nuclear wastes. Therefore, any activity to solve the problems has been welcome. A lot of research activities had been done to evaluate the possibility of the hybrid systems in the world, however, most of them were based on advanced fusion and fission technologies. In this contribution, three types of fusion-fission hybrid reactor concepts, i.e. the energy multiplier named FDS-EM, the fuel breeder named FDS-FB, waste transmuter named FDS-WT, have been proposed for the re-examination of feasibility, capability and safety and environmental potential of fission-fusion hybrid systems. Then based on the re-evaluation activity, a multi-functional fusion-fission reactor concept named FDS-MF simultaneously for nuclear waste transmutation, fissile fuel breeding and thermal energy production based on viable technologies i.e. available or limitedly extrapolated nuclear, processing and fusion technologies is proposed. The tokamak can be designed based on relatively easy-achieved plasma parameters extrapolated from the successful operation of the Experimental Advanced Superconducting Tokamak (EAST) in China and other tokamaks in the world, and the subcritical blanket can be designed based on the well-developed technology of PWR. The design and optimization of fusion plasma core parameters, fission blanket and fuel cycle have been presented. And the

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

    International Nuclear Information System (INIS)

    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

  8. Neutronic predesign tool for fusion power reactors system assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, J.-C., E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Li Puma, A. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martínez Arroyo, J. [ETSEIB, Internship in CEA (Spain)

    2013-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility. The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.

  9. Fracture toughness test methods and examples for fusion reactor materials

    International Nuclear Information System (INIS)

    This paper introduces the importance of the evaluation of fracture toughness in nuclear fusion reactor structural materials, and the fracture toughness evaluation methods that are used as the standards and their actual examples. It also discusses the problems involved in the standardized approach and the efforts for the technology improvement. To evaluate the material life under nuclear fusion reactor environment, fracture toughness measurement after neutron irradiation is indispensable. Due to a limitation in the irradiation area size of an irradiation reactor, and to avoid the temperature difference in a specimen, the size of the specimen is required to be minimized, which is different from the common standards. As for the size effect of the test specimen, toughness value tends to decrease when ligament length is 7 mm or below. The main problems and challenges are as follows. (1) As for the tendency that fracture toughness value decreases along with the miniaturization of the ligament length, it is necessary to elucidate the mechanism of size effects, and to develop the correction method for size effects. (2) As for the issues of the curve shape and application to irradiation time in the master curve method, it is necessary to review the data checking method and plastic constraint conditions for crack tip M = 30 that is stipulated in ASTM E1921, and to elucidate the material dependence of master curve shape. (A.O.)

  10. Remote maintenance design for Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Design of Fusion Experimental Reactor, FER, has been conducted by Japan Atomic Energy Research Institute (JAERI) since 1981. Two typical reactors can be classified in general from the viewpoints of remote maintenance among four design concepts of FER. In the case of the type 1 FER, the torus module consists of shield structure and blanket, and the connective joints between toruses provided at the outer region of the reactor. As for the type 2 FER, the shield structure is joined with the vacuum cryostat, and only the blanket module is allowed to move, but connection between toruses are located in the inner region of the reactor. Comparing type 1 with type 2 FER, this paper describes on the remote maintenance of FER including reactor configurations, work procedures, remote systems/equipments, repairing facility and future R and D problems. Reviewing design studies and investigation for the existing robotics technologies, R and D for FER remote maintenance technology should be performed under the reasonable long-term program. The main items of remote technology required to start urgently are multi-purpose manipulator system with performance of dextrousity, tele-viewing system which reduces operator fatigue and remote tests for commercially available components

  11. Application of induction MHD generator to inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    The purpose of the present paper is to examine applicability of induction-type MHD generators to inertial confinement fusion reactors. Alternative fusion explosions by laser or REB heat and evaporate the blanket of liquid metal. The vapor accelerates a piston made of liquid metal, which interacts with the magnetic field and induces electricity in solenoids. The temperature of working vapor is very high at the inlet of cylinder so that a large amount of the vapor is ionized at the inlet and gradually recombined into the neutral gas along the cylinder. The ionization and recombination of vapor act as a cushion of energy, resulting in the degradation of generator performance. The induction MHD generators attain the thermal efficiency of about 60%. The output voltage has shapes of very sharp pulses, leading to a difficulty of the electricity supply to commercial networks. This suggests that the travelling magnetic field may be suitable for this kind of generators. The period of micro fusion explosions is a few Hz. This is fitting to usual designs of laser fusion concepts. The thermal efficiency increases with the expansion ratio. The division of solenoid is required for effective interactions. The inner radius of cylinder has a large influence not on the thermal efficiency but on the period of explosions. (author)

  12. Different types of cryogenics Pellet injection systems (PIS for fusion reactor

    Directory of Open Access Journals (Sweden)

    Devarshi Patel

    2014-05-01

    Full Text Available Fusion reactor is the one of the most capable option for generating the large amount of energy in future. Fusion means joining smaller nuclei (the plural of nucleus to make a larger nucleus and release energy in the form of neutrons.The sun uses nuclear fusion of hydrogen atoms into helium atoms. This gives off heat and light and other radiation. Hydrogen is used as the fuel in the fusion reactor. We have to inject the solid hydrogen pellet into the tokamak as per the requirement. For injecting the pellet we use the pellet injection system. Pellet injection system (PIS is the fuel injection system of the fusion reactor.

  13. Economic analysis of a magnetic fusion production reactor

    International Nuclear Information System (INIS)

    The magnetic fusion reactor for the production of nuclear weapon materials, based on a tandem mirror design, is estimated to have a capital cost of $1.5 billion and to produce 10 kg of tritium/year for $22,000/g or 940 kg/year of plutonium in the plutonium mode for $250/g plus heavy metal processing. A tokamak-based design is estimated to cost $1.5 billion and to produce 10 kg of tritium/year for $29 thousand/g. For comparison, a commercially sized tandem mirror fusion breeder selling excess electricity and fissile material to commercial markets is estimated to cost $3.6 billion and to produce tritium for $2.6 thousand/g and plutonium for $34/g plus heavy metal processing

  14. Tokamak fusion reactor plant assembly, logistics and critical issues

    International Nuclear Information System (INIS)

    A tokamak fusion reactor will be an installation with linear sizes of a single assembly element of ∼10-20 m and a weight >1000 t. Similar weights can be handled at a construction site by commercially available equipment. But a problem is created by the combination of high weights and large dimensions of the components with tight installation tolerances driven by plasma physics requirements. To overcome this problem, the best available metrology techniques with 3-D real time computer modeling and specially selected assembly procedures are needed at all stages of production and assembly. Some of the biggest components i.e. poloidal coils must be produced at the construction site. Some smaller components must be preassembled at the site to accelerate construction. Adequate space with proper environmental control must be provided for these operations. These and other considerations determine the logistics of a fusion plant assembly

  15. Evaluation of the impact of a committed site on fusion reactor development

    International Nuclear Information System (INIS)

    The technical and economic merits of a committed fusion site for development of tokamak, mirror, and EBT reactor from ignition through demo phases were evaluated. Schedule compression resulting from evolving several reactor concepts and/or phases on a committed site as opposed to sequential use of independent sites was estimated. Land, water, and electrical power requirements for a committed fusion site were determined. A conceptual plot plan for siting three fusion reactors on a committed site was configured. Reactor support equipment common to the various concepts was identified as candidates for sharing. Licensing issues for fusion plants were briefly addressed

  16. Fissile fuel breeding in DT fusion reactor blankets

    International Nuclear Information System (INIS)

    Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For 233U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of 233U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO2 fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For 239Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for 233U breeding, yield 0.76 atoms of 233U and 0.38 atoms of 239Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated

  17. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Low cost tokamak fusion power reactor based on a reversed shear plasma: CREST

    International Nuclear Information System (INIS)

    The Cost down of tokamak fusion reactors is one of the key issues that must be dealt with in order to deploy fusion energy for commercial use. Our previous study of the costs of tokamak power plants has shown that a very high Troyon coefficient βN (up to 5 or 6) is required in order to realize a cost-competitive tokamak reactor. A reversed shear configuration, which has been observed recently in several tokamak experiments, might results in such a low cost tokamak. In this study, we propose a compact commercial reactor based on the shear reversed high beta equilibrium, which is named the Compact REversed Shear Tokamak (CREST). The optimized parameters of CREST are; major radius R = 5.4 m, maximum toroidal field Bmax = 12.5 T (5.6 T at plasma center), and fusion power Pf = 3 GW with 4.5 MW/m2 in the mean neutron wall load. The ideal MHD instabilities are all stable in this equilibrium while the βN value reaches 5.5. The plasma configuration of CREST is close to that of ITER advanced mode plasma. This will encourage an investigation of the feasibility of the CREST concept by ITER. Although many further studies and developments are necessary, such compact tokamak can be cost-competitive as an electric power source in the 21st century and could be a promising tool in the development of a commercial tokamak reactor succeeding the ITER project. (author)

  18. The cost of tritium production in a fusion reactor

    International Nuclear Information System (INIS)

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

  19. Application of uncertainty analysis in conceptual fusion reactor design

    International Nuclear Information System (INIS)

    The theories of sensitivity and uncertainty analysis are described and applied to a new conceptual tokamak fusion reactor design--NUWMAK. The responses investigated in this study include the tritium breeding ratio, first wall Ti dpa and gas productions, nuclear heating in the blanket, energy leakage to the magnet, and the dpa rate in the superconducting magnet aluminum stabilizer. The sensitivities and uncertainties of these responses are calculated. The cost/benefit feature of proposed integral measurements is also studied through the uncertainty reductions of these responses

  20. Structural design of fusion reactor: past, present and future

    International Nuclear Information System (INIS)

    The potentialities of fusion as a source of clean, intrinsically safe energy are discussed. The structural problems of tokamaks are presented and it is shown how the patterns of electromagnetic and thermal loads acting on them define the required structural systems and the overall reactor topology. These general concepts are applied to mechanical and thermal problems faced by the author in several research machines, namely, TBR-E and TBR-II (Brazil), REX (Italy) and TST (USA). 171 refs., 184 figs., 17 tabs

  1. Assessment of nucleonic methods and data for fusion reactors

    International Nuclear Information System (INIS)

    An assessment is provided of nucleonic methods, codes, and data necessary for a sound experimental fusion power reactor (EPR) technology base. Gaps in the base are identified and specific development recommendations are made in three areas: computational tools, nuclear data, and integral experiments. The current status of the first two areas is found to be sufficiently inadequate that viable engineering design of an EPR is precluded at this time. However, a program to provide the necessary data and computational capability is judged to be a low-risk effort

  2. High conductivity Be-Cu alloys for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lilley, E.A. [NGK Metals Corp., Reading, PA (United States); Adachi, Takao; Ishibashi, Yoshiki [NGK Insulators, Ltd., Aichi-ken (Japan)

    1995-09-01

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

  3. Feasibility of HTS Magnet Option for Fusion Reactors

    OpenAIRE

    Yanagi, Nagato; Ito, Satoshi; TERAZAKI, Yoshiro; NATSUME, Kyohei; TAMURA, Hitoshi; Hamaguchi, Shinji; MITO, Toshiyuki; HASHIZUME1, Hidetoshi; Morikawa, Junji; OGAWA, Yuichi; IWAKUMA, Masataka; SAGARA, Akio

    2014-01-01

    Conceptual design studies are being carried out on the application of high-temperature superconducting (HTS) conductors and coils to the magnet systems of fusion reactors. A 100-kA-class HTS conductor is required to be applied at high magnetic fields of > 12 T. A simple stack of YBCO tapes embedded in copper and stainless-steel jackets is found to be a practical approach to producing large-scale conductors that exhibit high cryogenic stability and mechanical rigidity. The feasibility of the s...

  4. Conceptual design of laser fusion reactor KOYO-fast

    Energy Technology Data Exchange (ETDEWEB)

    Tomabechi, K. [Former Advisor of Central Research Institute of Electric Power Industry (Japan); Kozaki, Y.; Norimatsu, T. [Osaka Univ., Institute of Laser Engineering, (Japan)

    2006-06-15

    A conceptual design of the laser fusion reactor KOYO-F based on the fast ignition scheme is reported including the target design, the laser system and the design for chamber. A Yb-YAG ceramic laser operated at 200 K is the primary candidate for the compression laser and an OPCPA (optical parametric chirped pulse amplification) system is the one for the ignition laser. The chamber is basically a wet wall type but the fire position is vertically off-set to simplify the protection scheme of the ceiling. The target consists of foam insulated, cryogenic DT shells with a LiPb, reentrant guide-cone. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-21

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

  7. Advanced fusion reactor design using remountable HTc SC magnet

    International Nuclear Information System (INIS)

    A new concept of fusion reactor design is proposed using remountable high critical temperature (HTc) superconducting (SC) magnet. There are two advantages using this system. First one is that the magnet system can be composed by parts, which means it easy to replace the damaged magnet module. The second one is that it becomes possible to access the reactor first wall easily. In order to realize this system, we have performed experiments using HTc SC tape. The experimental results indicate that the resistance of the jointed region becomes about 60 μΩ, which shows the feasibility of this concept. Using this system the remountable first wall system also has the feasibility based on thermomechanical analysis. (author)

  8. Systems study of tokamak fusion--fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations.

  9. Thermal response of fusion reactor containment to lithium fire

    International Nuclear Information System (INIS)

    The lithium pool combustion model LITFIRE was used to study the consequences of lithium fire within fusion reactor containments. Calculations based on the UWMAK-III design show that without any special fire protection measures, the containment may reach over-pressures of up to 2.2 atm when one coolant loop is spilled inside the reactor building. Temperatures as high as 11000C would also be experienced by some of the containment structures. These consequences were found to diminish greatly by the incorporation of a number of design strategies including initially subatmospheric containment pressures, initially low oxygen concentrations, and active post-accident cooling of the containment gas. Compartmentalization of the containment, as in the EBTR design, was found to limit the consequences of lithium fire and hence offers a potential safety advantage

  10. A Z-Pinch Driven Fusion Reactor Concept

    Science.gov (United States)

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

    1999-11-01

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

  11. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  12. Maintenance features of the Compact Ignition Tokamak fusion reactor

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs

  13. Nuclear group constant set FUSION-J3 for fusion reactor nuclear calculations based on JENDL-3

    International Nuclear Information System (INIS)

    Based on evaluated nuclear data file JENDL-3, published in April 1990, we produced a nuclear group constant set 'FUSION-J3' for fusion reactor nuclear calculation by ANISN code instead of GICX40 produced in 1977. The set FUSION-J3 is the coupled group constant set with neutron 125 and gamma-ray 40 group structure, and has the maximum order of 5 as Legendre expansion in scattering cross section. Forty nuclides included in FUSION-J3 can be used in fusion reactor nuclear calculations. Considering mobility in two-dimensional calculations and fixed group structure in induced activity calculation code system as the GICX40 structure, we composed also FUSION-40 group constant set with neutron 42 group and gamma-ray 21 group structure. The set FUSION-40 includes the same maximum order of the Legendre expansion and the same nuclides as FUSION-J3. From the results in experimental analysis and benchmark calculations, it became proved that JENDL-3 is at higher level of accuracy than ENDF/B-IV and -V. The set FUSION-J3 can be clear applicable to fusion reactor nuclear calculations. (author)

  14. Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

    International Nuclear Information System (INIS)

    The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

  15. A preliminary conceptual design study for Korean fusion DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb3Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters

  16. Low power density ion cyclotron arrays for fusion reactors

    International Nuclear Information System (INIS)

    Highlights: • A low power density, high directivity, FW heating and current drive launching structure is proposed for use in a commercial fusion reactor. • The structure, integrated in the reactor blanket first wall, is modular, unobtrusive and imposes no specific constraints to the blanket functions. • It may significantly reduce the undesirable effects of FW evanescence in the plasma scrape off layer such as increased thermal wall loading, localized hot spots, impurity production, and enhanced E × B0 particles convection. - Abstract: Ion Cyclotron Radio Frequency (ICRF) Heating and Current Drive (H&CD) is a well established technique of auxiliary heating in present tokamaks, as it features high on-axis heating and current drive efficiencies associated with proven and low cost technology. An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is linked to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing high RF standing and DC rectified voltages at the plasma periphery, possible voltage breakdowns and enhanced local wall loading. Further to previous work (Bosia et al., Ion Cyclotron and Lower Hybrid Arrays applicable to Current Drive in Fusion Reactors, in: AIP Proc. of 20th Topical Conf on RF Power in plasmas No. 1580, 2013, 215) developing new concepts for Ion Cyclotron and Lower Hybrid Heating & Current Drive arrays, based on the use of periodic structures, a practical example for an in-blanket IC array for DEMO1 is presented in this study

  17. Analysis of a laser-initiated, inertially-confined reactor for a fusion engineering research facility (LA FERF)

    International Nuclear Information System (INIS)

    With the increasing optimism for the viability of laser-induced, inertially-confined fusion-reactor power plants, comes an increased interest in a high flux 14-MeV neutron generator. This generator can be used to investigate 14-MeV neutron damage in first-wall material, and various neutron energy direct conversion schemes. The associated charged particles can be used to investigate various first-wall designs proposed for laser-fusion reactor systems as well as the direct conversion of charged particle energy by expansion of a fusion fireball against a magnetic field imposed from outside the neutron generator first wall. A systematic parameter study is made on a laser-initiated, inertially-confined reactor for a fusion engineering research facility (LA FERF). The parameters investigated are the variations of plant cost, plant power, and plant performance as a function of the fusion gain and power input to the laser system. Design point envelopes are presented for the laser input power and fusion gain (defined as the ratio of the thermonuclear power to power into the laser system) based on reference plant costs, powers and performance. A design point was selected (based on technology available in the not-too-distant future) which produces a flux of 1.8 x 1018 n/m2 . s over an experimental area of approximately 1 m2 for a cost of 160 megadollars using a power of 140 MVA. A comparison is made with a current mirror FERF study proposed for use in CTR engineering studies. The LA FERF could be built in the mid 1980's using near-term short pulse CO2 laser technology, producing thermonuclear power of 6 to 20 MW with fusion gains of approximately 0.1

  18. Application of Kelvin Probe to Studies of Fusion Reactor Materials under Irradiation

    Institute of Scientific and Technical Information of China (English)

    Luo Guangnan; K. Yamaguchi; T. Terai; M. Yamawaki

    2005-01-01

    Recently, the work function (WF) changes in metallic and ceramic materials to be potentially used in future fusion reactors have been examined by means of Kelvin probe (KP),under He ion irradiation in high energy (MeV) and / or low energy (500 eV) ranges. The results of polycrystalline Ni samples indicate that the 1 MeV beam only induces decrease in the WF within the experimental fluence range; whereas the irradiation of 500 eV beam results in decrease in the WF firstly, then increase till saturation. A dual layer surface model is employed to explain the observed phenomena, together with computer simulation results by SRIM code. Charges buildup on the surface of lithium ceramics has been found to greatly influence the probe output, which can be explained qualitatively using a model concerning an induction electric field due to external field and free charges on the ceramic surface.

  19. Final optics for laser-driven inertial fusion reactors

    International Nuclear Information System (INIS)

    If Inertial Confinement Fusion (ICF) power plants utilizing laser drivers are to be considered for electrical power generation, a method for delivering the driver energy into the reactor must be developed. This driver-reactor interface will necessarily employ final optics, which must survive in the face of fast neutrons, x-rays, hot vapors and condensates, and high-speed droplets. The most difficult to protect against is fast neutron damage since no optically transmissive shielding material for 14-MeV neutrons is available. Multilayer dielectric mirrors are judged to be unsuitable because radiation-induced chemical change, diffusion, and thickness changes will destroy their reflectivity within a few months of plant operation. Recently, grazing incidence metal mirrors were proposed, but optical damage issues are unresolved for this approach. In this paper, the authors consider the use of refractive optics. a major question to be answered is: what duration of reactor operation can this optic withstand?To answer this question the authors have reviewed the literature bearing on radiation-induced optical damage in fused silica and assessed its implications for reactor operation with the baseline final optics scheme. It appears possible to continuously anneal the neutron damage in the silica by keeping the wedge at a modestly elevated temperature

  20. Tritium handling in the Mirror Fusion Hybrid Reactor

    International Nuclear Information System (INIS)

    A reference design study for a Mirror Fusion Hybrid Reactor has been completed which examines the tritium handling problems. Breeding pins composed of aluminum alloys contain lithium hydride with a four-year residence time for tritium production. The slip-stream helium tritium capture system is designed to handle a 0.1 percent pin failure and will reduce environmental losses to below 3 Ci/day. The neutral beam injectors and direct converters utilize small, thin electrode tubes at 7000C for accelerating the deuterium or tritium, and they will by triton implantation permeate about 3 x 105 Ci/day into the internal helium coolant flow. A capture system will reduce these losses to 6 Ci/day, combined. The reactor hall is designed with a low humidity, air atmosphere which is continuously processed in order to handle leakage and permeability losses from the nuclear island at 180 Ci/day while still maintaining levels of tritium below MPC. The precessor is also able to handle severe accidental releases of tritium at the 26 kilogram level and permit worker re-entry (with ventilated suits) in a matter of about one week. These approaches to fusion power plant are found to be technically feasible today and economically attractive

  1. Summary of conceptual design study of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 FER design is now being reviewed and redesigned. This report is the summary of the report which describes the results obtained in the review and redesign activities in 1984. The following three steps are followed in those activities ; critical issues study step in which FER critical issues were reviewed and the frame of FER design was revised, torus structure selection step in which a few options within the frame for FER were examined and design step in which major components of the torus structure were designed. The newly established frame for FER design is as follows : 1) Plasma : Self-ignition, 2) Operation scenario : Quasi-steady state operation with long burn pulse, 3) Neutron fluence on the first wall : 0.3 MWY/m2, 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development, 5) Magnets : Superconducting Magnets. (author)

  2. An Ion Switch Regulates Fusion of Charged Membranes

    Science.gov (United States)

    Siepi, Evgenios; Lutz, Silke; Meyer, Sylke; Panzner, Steffen

    2011-01-01

    Here we identify the recruitment of solvent ions to lipid membranes as the dominant regulator of lipid phase behavior. Our data demonstrate that binding of counterions to charged lipids promotes the formation of lamellar membranes, whereas their absence can induce fusion. The mechanism applies to anionic and cationic liposomes, as well as the recently introduced amphoteric liposomes. In the latter, an additional pH-dependent lipid salt formation between anionic and cationic lipids must occur, as indicated by the depletion of membrane-bound ions in a zone around pH 5. Amphoteric liposomes fuse under these conditions but form lamellar structures at both lower and higher pH values. The integration of these observations into the classic lipid shape theory yielded a quantitative link between lipid and solvent composition and the physical state of the lipid assembly. The key parameter of the new model, κ(pH), describes the membrane phase behavior of charged membranes in response to their ion loading in a quantitative way. PMID:21575575

  3. Conceptual design of laser fusion reactor KOYO-fast - Concepts of reactor system and laser driver

    Energy Technology Data Exchange (ETDEWEB)

    Kozaki, Y.; Miyanaga, N.; Norimatsu, T.; Nakatsuka, M.; Jitsuno, T.; Fujita, H.; Kawanaka, J.; Tsubakimoto, K.; Fujimoto, Y.; Lu, J. [Osaka Univ., Institute of Laser Engineering (Japan); Soman, Y. [Mitsubishi Heavy Industry (Japan); Hayashi, T. [Japan Atomic Energy Research Institute (Japan); Furukawa, H. [Institute for Laser Technology (Japan); Yoshida, K.; Nakano, H.; Kubomura, H.; Kawashima, T.; Nishimae, J.; Suzuki, Y.; Tsuchiya, N.; Kanabe, T.; Matsuoka, S.; Ikegawa, T.; Owadano, Y.; Ueda, K.; Tomabechi, K. [IFE Forum (Japan)

    2006-06-15

    We have carried out the design studies of KOYO-Fast laser fusion power plant, using fast ignition cone targets, diode-pumped solid state lasers, and LiPb liquid wall chambers. Using fast ignition targets, we could design a middle sized 300 MWe reactor module, with 200 MJ fusion pulse energy and 4 Hz rep-rates, and 1200 MWe modular power plants with 4 reactor modules and a 16 Hz laser driver. The liquid wall chambers with free surface cascade flows are proposed for cooling surface quickly enough to a 4 Hz pulse operation. We examined the potential of Yb-YAG ceramic lasers operated at 150 - 225 K for both implosion and heating laser systems required for a 16-Hz repetition and 8 % total efficiency. (authors)

  4. Silicon carbide composites as fusion power reactor structural materials

    International Nuclear Information System (INIS)

    Silicon carbide was first proposed as a low activation fusion reactor material in the mid 1970s. However, serious development of this material did not begin until the early 1990s, driven by the emergence of composite materials that provided enhanced toughness and an implied ability to use these typically brittle materials in engineering application. In the decades that followed, SiC composite system was successfully transformed from a poorly performing curiosity into a radiation stable material of sufficient maturity to be considered for near term nuclear and non-nuclear systems. In this paper the recent progress in the understanding and of basic phenomenon related to the use of SiC and SiC composite in fusion applications will be presented. This work includes both fundamental radiation effects in SiC and engineering issues such as joining and general materials properties. Additionally, this paper will briefly discuss the technological gaps remaining for the practical application of this material system in fusion power devices such as DEMO and beyond.

  5. Considerations for tritium protection at a fusion reactor

    International Nuclear Information System (INIS)

    The view on the radiological hazard associated with future fusion power stations as presented in this discussion is rarely supported by reasonably certain or reliably accurate prediction. This fact should not be taken as indicating a major programmatic deficiency. In fact, it is expected that large uncertainty would be present in health effect at the current level of technological development. The details of tritium exposure will be clarified, waiting for the operation of the Tritium System Test Assembly. Once the data base for the TSTA is established, future fusion design can be made based on economic cost/radiation exposure risk benefit. The actual execution of this cost/benefit analysis is complex because three populations are of interest: occupational work force, local population and global population. The knowledge of tritium management must be increased if D-T fusion reactors are to become compatible with the needs of utility companies. In order to exploit the differing hazard between HT and HTO, it is necessary to know much more about the mechanism of uncatalyzed conversion over a wide range of concentration and about the change caused by the variety of potential catalytic sequence in potential tritium leak. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

  7. A feasibility study of a linear laser heated solenoid fusion reactor. Final report

    International Nuclear Information System (INIS)

    This report examines the feasibility of a laser heated solenoid as a fusion or fusion-fission reactor system. The objective of this study, was an assessment of the laser heated solenoid reactor concept in terms of its plasma physics, engineering design, and commercial feasibility. Within the study many pertinent reactor aspects were treated including: physics of the laser-plasma interaction; thermonuclear behavior of a slender plasma column; end-losses under reactor conditions; design of a modular first wall, a hybrid (both superconducting and normal) magnet, a large CO2 laser system; reactor blanket; electrical storage elements; neutronics; radiation damage, and tritium processing. Self-consistent reactor configurations were developed for both pure fusion and fusion-fission designs, with the latter designed both to produce power and/or fissile fuels for conventional fission reactors. Appendix A is a bibliography with commentary of theoretical and experimental studies that have been directed at the laser heated solenoid

  8. Nuclear energy and fusion-fission hybrid reactor for pure energy production

    International Nuclear Information System (INIS)

    The next two decades are very critical for nuclear energy development. The commercial fast reactor may be in use around 2035; it is also possible that magnetically confined fusion, laser fusion and z-pinch fusion will be demonstrated at that time. A fusion demonstration reactor can be a pure fusion or a fusion-fission hybrid. The latter can lower the fusion power and mitigate the radiation damage of high energy neutrons to materials. On the other hand, the supply of deuterium and tritium as fuel for fusion can only last a few hundred years. We describe here a hybrid for pure energy use which can make full use of uranium and is proliferation resistant, as no separation of uranium and plutonium is needed in post-processing. The union of fission, fusion, and a pure energy hybrid can contribute to the large scale use of nuclear energy in the near future, and supply mankind for more than a thousand years. (authors)

  9. Radiation damage studies in fusion reactor materials. Part of a coordinated programme on energetic particle interactions with materials of importance for fusion reactors

    International Nuclear Information System (INIS)

    This paper constitutes the final report (IAEA Research Contract No. 1882/RB) on Radiation Damage Studies in Fusion Reactor Materials (Sputtering and Blistering) performed at the Research Centre of Bhabha, India

  10. Analysis of tritium production in TRIGA Mark II reactor at JSI for the needs of fusion research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jazbec, Anze; Zerovnik, Gasper; Snoj, Luka; Trkov, Andrej [Jozef Stefan Institute, Ljubljana (Slovenia)

    2013-12-15

    In future, electricity could be produced in fusion power plants. One of the steps towards development of fusion power plants is the construction of an experimental fusion reactor ITER where deuterium (D) and tritium (T) will be fused and energy will be released. As natural concentrations of T are extremely low, the T as fusion fuel will have to be produced artificially. A series of calculations were made to investigate the possibility of producing small quantities of T for experimental fusion reactors such as JET and ITER in a small research reactor like the TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). The T production is the largest if all irradiation channels in reactor's reflector are filled with LiAlO{sub 2} samples. When samples are inserted, the excess reactivity decreases by around 200 pcm. In the second part of the work an estimate was made of how long the reactor can operate with current fuel supplies. Calculations were made with the TRIGLAV computer code. TRIGA can operate at full power for at least 2,860 days, during which 152 mg of T could be produced. We conclude that small TRIGA reactors can not produce any significant quantities of T for the needs of the future experimental fusion reactors. (orig.)

  11. Analysis of tritium production in TRIGA Mark II reactor at JSI for the needs of fusion research reactors

    International Nuclear Information System (INIS)

    In future, electricity could be produced in fusion power plants. One of the steps towards development of fusion power plants is the construction of an experimental fusion reactor ITER where deuterium (D) and tritium (T) will be fused and energy will be released. As natural concentrations of T are extremely low, the T as fusion fuel will have to be produced artificially. A series of calculations were made to investigate the possibility of producing small quantities of T for experimental fusion reactors such as JET and ITER in a small research reactor like the TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). The T production is the largest if all irradiation channels in reactor's reflector are filled with LiAlO2 samples. When samples are inserted, the excess reactivity decreases by around 200 pcm. In the second part of the work an estimate was made of how long the reactor can operate with current fuel supplies. Calculations were made with the TRIGLAV computer code. TRIGA can operate at full power for at least 2,860 days, during which 152 mg of T could be produced. We conclude that small TRIGA reactors can not produce any significant quantities of T for the needs of the future experimental fusion reactors. (orig.)

  12. Nuclear-pumped laser concepts for laser fusion or laser-heated solenoid reactors

    International Nuclear Information System (INIS)

    The combination of a nuclear-pumped laser with a fusion reactor, using some of the neutrons emitted from the fusion reactor to pump the laser, is described. This paper describes several concepts which might be used, points out potential advantages of using a nuclear-pumped laser, and describes several areas which must be investigated to prove feasibility

  13. Helium generation in fusion-reactor materials. Progress report, January-March 1981

    International Nuclear Information System (INIS)

    The objectives of this program are to measure helium generation rates of materials for Magnetic Fusion Reactor applications in the various neutron environments used for fusion reactor materials testing, to characterize these neutron test environments, and to develop helium accumulation neutron dosimeters for neutron fluence and energy spectrum dosimetry in these test environments

  14. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  15. Neutron spectrum effects of the defect production in fusion reactor candidate alloys

    International Nuclear Information System (INIS)

    In the present work, irradiation effects of fission and fusion neutrons on fusion reactor candidate alloys, V-4Cr-4Ti and F82H were studied using the FNS facility and the Kyoto University Reactor (KUR). The comparison of defect structures in two fusion reactor candidate alloys between fusion neutron irradiation and fission neutron irradiation was performed. Even though the irradiation doses were low, the defect formation was detected by positron annihilation spectroscopy. Higher irradiation doses and different irradiation temperatures are required to detect the effects of neutron spectra more precisely. (author)

  16. Economic analysis of advanced fuel fusion reactors and derivation of scaling law for COE

    International Nuclear Information System (INIS)

    Social acceptance of fusion reactors depends largely on their economic viability. To investigate this issue, we estimate and compare the cost of electricity (COE) among D-T, D-3He, and D-D fusion reactors. Three types of confinement systems are evaluated: the tokamak reactor (TR), the spherical tokamak reactor (STR), and helical reactor (HR). For each reactor type, COE parameter surveys are performed and new scaling laws for COE are derived. The COE for D-3He and D-D is high and depends more strongly on plasma beta value and maximum magnetic field strength than that of D-T. (author)

  17. Design of a fusion reactor for eutectic alloys Pb- Li; Diseno de un reactor de fusion para aleaciones autecticas Pb-Li

    Energy Technology Data Exchange (ETDEWEB)

    Quinones, J.; Barrena Perez, M. I.; Gomez de Salazar, J. M.; Serrano, L.; Duran, S.; Conde, E.; Barrado, A. I.; Fernandez, M.; Sedano, L.; Soria Munoz, A.

    2010-07-01

    Given the interest that have the Pb-Li eutectic alloys in the field of production of tritium, designed to optimize energy through nuclear fusion processes, in this paper, we present the design, construction and commissioning of a fusion reactor of Pb-Li alloys eutectic and the optimal process conditions, for these alloys.

  18. On the feasibility of a fusion-fission hybrid reactor driven by dense magnetized plasmas

    International Nuclear Information System (INIS)

    The feasibility of a fusion-fission hybrid reactor driven by dense magnetized plasmas was analyzed from the point of view of the technical requirements for the fusion and fission components of the reactor. In the conceptual design, a 200 MW hybrid fusion-fission reactor is considered to be used as a heat source for district heating. The fission heat-generating blanket is based on the CANDU reactor technology, while the fusion fast neutrons are provided by a high-density pinch plasma. As far as the fission components of the reactor are concerned, the hybrid reactor turns out to be entirely feasible based on existing technologies. On the other hand extensive development will be needed to meet the requirements for the fusion component of the reactor. The basic conditions for a dense magnetized plasma fusion device to be used for the proposed hybrid concept are not concerned only with the attainment of high neutron yield per pulse (at least 5 x 10 18), but also with a relatively high repetition rate (in the range 1-10 Hz). An important feature of the proposed design is its inherent safety feature: no active component are necessary within the reactor containment area, all the hybrid system control being ensured by the fusion component of the reactor. (authors)

  19. Procurement of tritium for fusion reactor. A design study of facility for production of fusion fuel tritium

    International Nuclear Information System (INIS)

    Tritium, a developmental fuel for use in fusion reactors, has been produced in fission research reactors in Japan by extraction from neutron-irradiated 6Li-targets. This paper describes the preliminary design of a large-scale production facility capable of producing 500 g of tritium annually. The present status of tritium production technology in Japan is also discussed. (author)

  20. Conceptual design study of fusion experimental reactor (FY86FER)

    International Nuclear Information System (INIS)

    This report describes the results of applicability studies for the negative ion-based neutral beam injector to the Fusion Experimental Reactor (FER). The operation scenario of FER has been proposed to adopt the neutral injection method as one of candidates, which has three functions of heating, current drive and profile control. One of the fundamental requirements is the tangential injection of the neutral beam. For neutral beam injectors, three port sections are available. Supposing to adopt the beam line with the straight long neutralizer which has been designed in JAERI, the geometrical arrangement was determined so as to avoid any trouble to the reactor structure. The conceptual study for major components which compose the beam line system was carried out including the estimation of the neutron streaming. The power supply system was studied also and the work was concentrated on the acceleration power supply which requires the output voltage of 500 kV and fast cut-off time. A basic concept, in which a inverter with a AC switch is used and the frequency of the supplied AC line is increased was proposed. In these works, the configuration of the neutral beam injection system was detailed and it was shown that the beam line seems to be well implemented with the geometrical constraints related to the reactor configuration. (author)

  1. Economic Feasibility of Stellarator and Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Studies of model designs of stellarator-type fusion reactors are presented. These serve to high-light key technological and plasma-physics problems which require solution. The basic conflict is between estimates of the critical β for equilibrium and stability and the cost of high magnetic fields from superconducting multipolar coils. The effect of diffusion-driven currents on the critical β and the effect of electric fields on estimates of the cross-field diffusion rate are considered. The inclusion of a 1.5-m blanket in a system with a short-range multipolar field makes it difficult to achieve economic operation in small sizes. However, the unit cost for this type of reactor falls rapidly with increasing size. Thus if present estimates for the critical β can be realized and if low-cost superconductors can be developed, we expect a 10 000-MW(e) stellarator reactor to meet target economic costs which include a penalty for the large size. (author)

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

    Science.gov (United States)

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

    1989-01-01

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

  3. Development and prospect of nano-structured ODS steels for fusion reactor first wall application

    International Nuclear Information System (INIS)

    The first wall structural material of fusion reactors must meet extremely rigorous operation environment requirements and it is one of the key factors restricting the development of fusion reactors. Nano-structured oxide dispersion strengthened (ODS) steels are the leading candidates for fusion first wall due to their excellent irradiation resistance coming from their characteristic microstructures. This paper described briefly the current state of development and understanding of ODS steels. (author)

  4. Modelling of PFC Life-Time in Tokamak Fusion Reactor (KIT Scientific Reports ; 7612)

    OpenAIRE

    Igitkhanov, Yuri; Bazylev, B.; Landman, I.

    2012-01-01

    The performance of materials in fusion reactor DEMO has long been recognized as fundamental issue affecting the ultimate technological and economic feasibility of fusion power. Many factors influence the choice of a functional and structural material in a fusion reactor. Three effects limit component lifetime in the steady-state operation: radiation damage, disruptions, and sputtering erosion. Our design strategy is to determine the structure and coating thickness, which maximize component lifet

  5. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of helical-type fusion reactors

    International Nuclear Information System (INIS)

    Currently, there is active interest in the research and development of helical systems. New large devices using superconducting magnets (LHD in Japan and W7-X in Germany) are expected to produce highly improved plasmas comparable to those recently obtained in the large tokamaks. Because current-less steady operation is advantageous, these aggressive programs have accelerated several design studies of helical-type reactors, which are promising alternatives to demonstration reactors. A reference design for the Force Free Helical Reactor (FFHR) is presented, the main feature of which is the force-free-like configuration of the helical coils. Another feature is the selection of molten-salt Flibe as a self-cooling tritium breeder, which enhances safety. Demo-relevant engineering issues in the concept definition phase are discussed. (author)

  6. Radiation Damage Studies on a Laser Fusion Reactor

    International Nuclear Information System (INIS)

    Full text: A laser fusion-fission (hybrid) has been investigated with a multi-layered spherical blanket, composed of a first wall made of oxide dispersed steel (ODS, 2 cm); neutron multiplier and coolant zone made of LiPb; ODS-separator (2 cm); a molten salt Flibe coolant and fission zone; ODS-separator (2 cm); graphite reflector. In the second phase, LiPb coolant zone behind the first wall has been removed. But instead, a flowing liquid protective first wall is included of in front of the solid first wall in order to reduce material damage and residual radioactivity after final disposal of the latter. Tritium breeding ratio (TBR) has been calculated as TBR > 1.05 over 8 to 9 years of continuous plant operation period, which would supply fusion fuel for self-sustaining fusion reactor operation. Blanket energy multiplication factor (M) remains nearly stable between 2.4 to 6.8 over 10 years for TRISO fractions 2 to 8 vol-% in the coolant. Without an internal liquid wall protection, major damage mechanisms have been calculated as DPA = 50 and He = 170 appm/year at the ODS first wall. It faces directly to the fusion chamber, and will be the primary component subject to material damage. This will oblige to change the ODS first wall every ∼ 3 years. Hydrogen production is calculated as 650 appm/year. Hydrogen will not reside permanently in the metallic lattice as helium atoms, but diffuse out of the structure by high operation temperatures. The alternative version to include a Flibe zone of ∼ 50 cm thickness as flowing wall liquid protection in front of the solid ODS first wall reduces material damage below permissible limits. It allows shallow burial of structure after final reactor decommissioning. Calculations have yielded very high burn up grades (> 400 000 MW.D/MT) for the fissionable RG-PuC/ThC fuel without fuel change, making the LIFE engine a sustainable power source. Peach Bottom experiments have shown that TRISO fuel particles can withstand burn ups of 740

  7. Application of controlled thermonuclear reactor fusion energy for food production

    International Nuclear Information System (INIS)

    Food and energy shortages in many parts of the world in the past two years raise an immediate need for the evaluation of energy input in food production. The present paper investigates systematically (1) the energy requirement for food production, and (2) the provision of controlled thermonuclear fusion energy for major energy intensive sectors of food manufacturing. Among all the items of energy input to the ''food industry,'' fertilizers, water for irrigation, food processing industries, such as beet sugar refinery and dough making and single cell protein manufacturing, have been chosen for study in detail. A controlled thermonuclear power reactor was used to provide electrical and thermal energy for all these processes. Conceptual design of the application of controlled thermonuclear power, water and air for methanol and ammonia synthesis and single cell protein production is presented. Economic analysis shows that these processes can be competitive. (auth)

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

    International Nuclear Information System (INIS)

    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

  9. Study on plasma ignition of JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Heating the plasma in JAERI Experimental Fusion Reactor up to the equilibrium operating state has been studied with a time dependent zero-dimensional model. The neoclassical or pseudoclassical scaling-law plays a leading part of the plasma diffusion in the low temperature region below several keV and the trapped-ion scaling-law does so in the higher region. The plasma temperature is raised to 1 keV by 10 sec Joule-heating. The plasma is heated up to the equilibrium operating state of plasma temperature 7 keV and electron density 1.1 x 1020 m-3 by 10 sec neutral beam injection heating with injection power 28 MW and fueling rate 3 x 1019 m-3s-1. (auth.)

  10. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

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

  11. High temperature indentation tests on fusion reactor candidate materials

    International Nuclear Information System (INIS)

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 oC. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 oC. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods

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

    International Nuclear Information System (INIS)

    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 m3/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

  13. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  14. First Results from a Charged Fusion Products Diagnostic at MAST

    Science.gov (United States)

    Perez, Ramona V.; Allan, Scott Y.; Boeglin, Werner U.; Cecconello, Marco; McClements, Ken G.; Darrow, Douglass S.; MAST Team

    2013-10-01

    We designed, built and installed in MAST a 4-channel solid-state detector array for the detection of the charged deuterium-deuterium fusion products protons and tritons. The array has been mounted at the end of the reciprocating probe arm in MAST allowing it to sample a range of radial positions. First data have been taken in August 2013. The detector signals have been digitized with a 60 MHz sampling rate and have been continuously recorded during plasma discharges. Protons and tritons were readily identified and counted. The observed count rates showed clear dependence on the neutral beam power and were modulated synchronous with saw-teeth. Comparison with data obtained from the MAST neutron camera and the fission chamber neutron detector is planned. We found that time resolutions as low as at least 1 ms were achievable. The detector performance and first analysis results for various plasma scenarios will be presented. Supported in part by DOE grant DE-SC0001157.

  15. Production of nuclear fusion reactor fuel by ceramic tritium breeder material

    International Nuclear Information System (INIS)

    Fuel tritium is generated from the nuclear reaction between the fusion neutron and the lithium of the breeder material arranged in the blanket that encloses the fusion plasma in the fusion reactor. However, the release process of the generated tritium has not been completely clarified. Recently, Japan Atomic Energy Agency started the tritium generation and recovery experiment in using nuclear fusion neutron source (FNS). In this report, the recent results of study on breeder material and its manufacturing technology is presented. (author)

  16. Flibe use in fusion reactors: An initial safety assessment

    International Nuclear Information System (INIS)

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

  17. LIBRA - a light ion beam fusion conceptual reactor design

    International Nuclear Information System (INIS)

    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 carbind tubes with Li17Pb83 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 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with Li17Pb83 that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequently without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of a 331 MWe LIBRA design is estimated to be 2843 Dollar/kWe while a 1200 MWe LIBRA design will cost approximately 1300 Dollar/kWe. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  20. Development of high purity vanadium alloys for fusion reactors

    International Nuclear Information System (INIS)

    Vanadium alloys are most attractive candidate materials for liquid Li self-cooled blanket system of fusion reactors. This paper summarizes the program and its activities of the NIFS (National Institute for Fusion Science), Japan for developments of high purity V-4Cr-4Ti alloys. The results from NIFS-Heats show various benefits by reducing the level of oxygen. Significant improvement of the impact properties of the welded joint by reducing oxygen level is one of examples in recent studies. Collaboration is in progress, in which those heats are being characterized by a number of research groups including Japanese universities, and international collaboration partners in the US, Russia and China. The impact tests of irradiated specimens are in progress for further investigation. Significant progress has been made recently on the insulator ceramic coating in static conditions in the Japan-USA Cooperation Program. The understanding on the condition of in-situ CaO coating in liquid Li was enhanced. Based on these achievements, a flowing loop test is being planned to investigate the effects of temperature gradient and Li chemistry. (Y. Tanaka)

  1. Saddle point condition for D - 3He tokamak fusion reactor

    International Nuclear Information System (INIS)

    In this paper the concept of a generalized ignition contour map, showing bar PhtT2E, NTE, and T, is used to study the ignition criterion for a D-3He fusion reactor with plasma temperature and density profiles. Direct heating scenarios to the D - 3He ignition regime without the help of deuterium-tritium burning are considered. The machine size and enhancement factor for the confinement time required to reach D - 3He ignition can be simple determined by comparing the height of the operation path with Goldston L-mode scaling and the height of the generalized saddle point. A confinement enhancement factor of 2 to 3 is required in the case of a large plasma current (30 to 80 MA) in a small-aspect-ratio tokamak. On the other hand, for a small plasma current (approx-lt 10 MA), large-aspect-ratio tokamak, an enhancement factor of 5 to 6 is necessary to reach ignition. Fuel dilution effects by fusion products and impurities, the confinement degradation effect due to 14-MeV protons, and the operation paths are also considered. To lower the height of the saddle point, and hence the auxiliary heating power, we optimize the fuel composition and examine operation in the hot ion mode

  2. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    It is conjectured that the number of fusion events catalyzed by a single muon is orders of magnitude greater for deuterium and tritium concentrated in a transition metal than in gaseous form and that the recent observation of 2.5-MeV neutrons from a D2O electrolytic cell with palladium and titanium cathodes can thereby be interpreted in terms of cosmic muon-catalyzed deuterium-deuterium fusion. This suggests a new fusion reactor reactor consisting of deuterium and tritium concentrated in transition metal fuel elements in a fusion core that surrounds an accelerator-produced muon source. The feasibility of net energy production in such a reactor is established in terms of requirements on the number of fusion events catalyzed per muon. The technological implications for a power reactor based on this concept are examined. The potential of such a concept as a neutron source for materials testing and tritium and plutonium production is briefly discussed

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

    International Nuclear Information System (INIS)

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

  6. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report describes the results of a conceptual study on the RF system in the typical candidates for the Fusion Experimental Reactor (FER), which were picked out through the '86FER scoping studies. According to the FER operation scenario, three RF systems, that is, ICRF (heating), LHRF (current drive and heating), ECRF (auxiliary heating) were studied. Main concern in these RF systems is the launcher, which may be so designed that required power match the geometrical constraints of the reactor. Then studies were concentrated on the launcher configuration. A prug-in concept of the launcher was adopted in each system and vacancies except transmission space were filled with water. The ICRF launcher had the 2 x 2 loop arrays antenna and the faraday shield area of 1.5 m x 1 m to provide a power of 20 MW. The LHRF launcher had the grillantenna with 28 x 8 open waveguides, and included multi junction-type power splitters which were connected to 56 transmission wave guides. The grild was designed to have two functions of current drive and heating, and provide a power of 20 MW each. The ECRF launcher had a boundle of open wave guides which a reflection mirror each, and three plain mirrors. Assuming a oscillator unit size of 200 kW, it had 40 oversized wave guides to provide a power of 3 MW. (author)

  7. Intense neutron source requirements for fusion reactor materials development

    International Nuclear Information System (INIS)

    Materials research should precede machine construction by at least ten years because considerable time is required for the materials development. When the next generation machine is under discussion, materials scientists and engineers should consider next-next generation device as DEMO for establishing the materials database in time. In this sense, development of an intense high energy neutron source is an urgent problem. Characteristic features of radiation effects with 14 MeV neutrons will be briefly reviewed. Then, the reasons why we need intense source will be discussed. These discussions will lead to identify requirements for the intense neutron sources. There are both near term and long term materials issues which can be studied with such intense neutron sources depending on their capacity. One should also recognize that development of such an intense source will require considerable time and maximum use of existing intense fission reactor neutrons will be one of the practical options for the moment. In other words, the intense neutron sources under discussion should be superior for the study of fusion radiation effects than the existing fission reactors. Items are listed for the evaluation of the sources and some critical comments will be made on several kinds of sources currently being proposed. (author)

  8. Metabolic and environmental aspects of fusion reactor activation products: niobium

    Energy Technology Data Exchange (ETDEWEB)

    Easterly, C.E.; Shank, K.E.

    1977-11-01

    A summary of the metabolic and environmental aspects of niobium is presented. The toxicological symptoms from exposure to niobium are given, along with lethal concentration values for acute and chronic exposures. Existing human data are presented; animal uptake and retention data are analyzed for various routes of administration. Recommended metabolic values are also presented along with comments concerning their use and appropriateness. The natural distribution of niobium is given for freshwater, seawater, and the biosphere. Concentration factors and retention of /sup 95/Nb in the environment are discussed with reference to: plant retention via leaf absorption; plant retention via root uptake; uptake in terrestrial animals from plants; uptake in freshwater organisms; uptake in marine organisms; and movement in soil. Conclusions are drawn regarding needs for future work in these areas. This review was undertaken because niobium is expected to be a key metal in the development of commercial fusion reactors. It is recognized that niobium will likely not be used in the first generation reactors as a structural material but will appear as an alloy in such materials as superconducting wire.

  9. Methane impurity production in the fusion reactor environment

    International Nuclear Information System (INIS)

    Fusion requires temperatures of the order of 108 degrees C. In order to attain the required temperature it will be essential to minimise the energy losses from the plasma. Impurities are a major cause of plasma cooling. Ionization of impurity species in the plasma leads to a subsequent decay and emission of radiation. The most common low Z contaminants to be consideed are water and methane produced by reaction of hydrogen isotopes with oxygen and carbon. This review focuses on the methane production problem. We will be concerned with the sources of carbon in the reactor and also with the reactivity of carbon with hydrogen molecules, atoms and ions and the synergistic effects which can arise from coincident fluxes of electrons and photons and the effects of radiation-induced damage of the materials involved. While the reactor first wall will provide the most hostile environment for methane producton, most of the reactions discussed can occur in breeder blankets and also in other tritium facilities such as fuel handling, purification and storage facilities

  10. Conceptual design study of the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A conceptual design study of the Fusion Experimental Reactor (FER) is presented. FER is planned, on the basis of a domestic programme, as a device to explore reasonable minimum physics and technological issues necessary to proceed to DEMO. Among various concepts, including the improvement of ITER-like design, the reference design of FER is chosen as size-minimum and a detailed design study is performed. LHCD assist and a single null divertor configuration are employed to reduce the device size. Simplification and the resultant reliability are attained by the appropriate choice of the fluence, 0.3 MW·a·m-2. A variety of new ideas are explored to develop the FER concept. A layered structure of the divertor for a reliable maintenance scheme and a uniform vacuum vessel with a thin double wall structure for providing structural simplicity and tritium double containment are typical examples. R and D programmes for these key reactor components are now actively being promoted in conjunction with this design activity. (author). 4 refs, 2 figs, 1 tab

  11. Preliminary Estimation of Activated Corrosion Products in the Coolant System of Fusion Demo Reactor

    International Nuclear Information System (INIS)

    The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of 57Co and 57Ni were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

  12. Material Challenges For Plasma Facing Components in Future Fusion Reactors

    International Nuclear Information System (INIS)

    Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

  13. Neutronics and pumping power analyses on the Tokamak reactor for the fusion-biomass hybrid concept

    International Nuclear Information System (INIS)

    Highlights: • MCNP analyses on a Tokamak with LiPb-cooled components shows concentrations of nuclear heating at the in-board region in addition to the out-board region. • Required pumping power of LiPb coolants for the nuclear heating exponentially increases as fusion power increases. • Pumping power analysis for the divertor also indicates the increasing pumping power as the fusion power increases. -- Abstract: The authors aim to develop a fusion-biomass combined plant concept with a small power fusion reactor. A concern for the small power reactor is the coolant pumping power which may significantly decreases the apparent energy outcome. Thus pressure loss and corresponding pumping power were studied for a designed Tokamak reactor: GNOME. First, 3-D Monte-Carlo Neutron transport analysis for the reactor model with dual-coolant blankets was taken in order to simulate the tritium breeding ability and the distribution of nuclear heat. Considering calculated concentration of nuclear heat on the in-board blankets, pressure loss of the liquid LiPb at coolant pipes due to MHD and friction forces was analyzed as a function of fusion power. It was found that as the fusion power increases, the pressure loss and corresponding pumping power exponentially increase. Consequently, the proportion of the pumping power to the fusion power increases as the fusion power increases. In case of ∼360 MW fusion power operation, pumping power required for in-board cooling pipes was estimated as ∼1% of the fusion power

  14. Safety and environment aspects of Tokamak-type fusion power reactor - an overview

    International Nuclear Information System (INIS)

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R and D on safety and environmental aspects of Tokamak type fusion reactor. (author)

  15. Towards a reduced activation structural materials database for fusion reactors

    International Nuclear Information System (INIS)

    Full text: The development of First Wall, Blanket and Divertor materials which are capable of withstanding many years the high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide a related materials database for design, construction, licensing and safe operation of the ITER Test Blanket Modules and of a DEMO reactor, a wealth of R and D results on the European reduced activation ferritic-martensitic steel EUROFER, and on oxide dispersion strengthened (ODS) variants have become available, mainly in the temperature window 250-700 deg. C. Industrial EUROFER-batches of 3.5 and 8.0 tons have been produced with a variety of semi-finished, quality-assured product forms. Extensive chipless shaping and joining experience taking into account different welding procedures and powder technology product forms have demonstrated that EUROFER type steel complies with a wide range of established manufacturing processes. EUROFER is also resistant to high temperature aging, and the existing creep-rupture properties (∼30000 h) indicated long term stability and predictability. To increase the thermal efficiency of blankets beyond 45%, high temperature resistant SiCf/SiC channel inserts for liquid metal coolant tubes are developed. Mechanical and thermal properties of various SiCf/SiC composits have been measured after neutron radiation. Regarding radiation damage resistance of blanket structural materials, a broad based reactor irradiation programme counts several steps from 2 needs to be removed, the design is presently based on tiles made of W (∼2000 deg. C), as well as on structural materials like W-alloy (∼700-1300 deg. C) and RAF(M)-ODS steel (∼650 deg. C). Severe plastic deformation of pure W and W alloys improves ductility, but does not prevent from re-crystallisation between 850 and 1200 deg. C. For the

  16. System analysis study for Korean fusion DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: ► A conceptual design study for a steady-state K-DEMO has been initiated. ► The major radius is designed to be below 6.5 m, considering engineering feasibilities. ► Magnetic field at the plasma center around 8 T is achieved by using Nb3Sn technology. ► Feasibility of near-future DEMO reactor is studied with a system analysis code. ► A net electric generation on the order of 300 MWe can be achieved below the βN of 5. -- Abstract: A conceptual design study for a steady-state Korean fusion DEMO reactor (K-DEMO) has been initiated. Two peculiar features need to be noted. First, the major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8 T is expected to be achieved by using current state-of-the-art high performance Nb3Sn strand technology. Second, a two-stage development plan is being considered. In the first stage, K-DEMO will demonstrate a net electricity generation but will also act as a component test facility. Then, after a major upgrade, K-DEMO is expected to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). Feasibility of such a practical, near-future demonstration reactor is studied in this paper, based on a zero dimensional system analysis code study. It was shown that a net electric generation on the order of 300 MWe can be achieved below the optimistic βN limit of 5. The elongation of K-DEMO is around 1.8 with single null configuration. Detailed optimization process and the resultant various plasma parameters are described

  17. Plasma transport control and self-sustaining fusion reactor

    International Nuclear Information System (INIS)

    The possibility of a high performance/low cost fusion reactor concept which can simultaneously satisfy (1) high beta, (2) high bootstrap fraction (self-sustaining), and (3) high confinement is discussed. In CDX-U, a tokamak configuration was created and sustained solely by internally generated bootstrap currents, in which a seed current is created through a non-classical current diffusion process. Recent theoretical studies of MHD stability limits in spherical torus .g., the National Spherical Torus Experiment (NSTX) reduced a promising regime with stable beta of 45% and bootstrap current fraction of ≥99%. Since the bootstrap current is generated by the pressure gradient, to satisfy the needed current profile for MHD stable high beta regimes, it is essential to develop a means to control the pressure profile. It is suggested that the most efficient approach for pressure profile control is through a creation of transport barriers (localized regions of low plasma transport) in the plasma. As a tool for creating the core transport barrier, poloidal-sheared-flow generation by ion Bernstein waves (IBW) near the wave absorption region appears to be promising. In PBX-M, application of IBW power produced a high-quality internal transport barrier where the ion energy and particle transport became neoclassical in the barrier region. The observation is consistent with the IBW-induced-poloidal-sheared-flow model. An experiment is planned on TFTR to demonstrate this concept with D-T reactor-grade plasmas. For edge transport control, a method based on electron ripple injection (ERI), driven by electron cyclotron heating (ECH), is being developed on CDX-U. It is estimated that both the IBW and ERI methods can create a transport barrier in reactor-grade plasmas (e.g., ITER) with a relatively small amount of power (∼10 MW much-lt Pfusion)

  18. Different types of cryogenics Pellet injection systems (PIS) for fusion reactor

    OpenAIRE

    Devarshi Patel; Alkesh Mavani

    2014-01-01

    Fusion reactor is the one of the most capable option for generating the large amount of energy in future. Fusion means joining smaller nuclei (the plural of nucleus) to make a larger nucleus and release energy in the form of neutrons.The sun uses nuclear fusion of hydrogen atoms into helium atoms. This gives off heat and light and other radiation. Hydrogen is used as the fuel in the fusion reactor. We have to inject the solid hydrogen pellet into the tokamak as per the require...

  19. Measurements of charged fusion product diffusion in TFTR

    International Nuclear Information System (INIS)

    The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator [ZnS(Ag)] and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current (approx-gt 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model

  20. Measurements of charged fusion product diffusion in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R.L.

    1991-12-01

    The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator [ZnS(Ag)] and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current ({approx_gt} 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model.

  1. Measurements of charged fusion product diffusion in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R.L.

    1991-12-01

    The single particle confinement of charged fusion products, namely the 1 MeV triton and the 3 MeV proton, has been studied using a detector located near the outer midplane of TFTR. The detector, which measure the flux of escaping particles, is composed of a scintillator (ZnS(Ag)) and a system of collimating apertures, which permit pitch angle, energy and time resolution. It is mounted on a movable probe which can be inserted 25 cm into the vacuum vessel. Measurements indicate a level of losses higher than expected from a first-orbit loss mechanism alone. The primary candidate for explaining the observed anomalous losses is the toroidal field (TF) stochastic ripple diffusion, theoretically discovered by Goldston, White and Boozer. This loss mechanism is expected to be localized near the outer midplane where, at least at high current ({approx gt} 1.0 MA) it would locally dominate over first-orbit losses. Calculations made with a mapping particle orbit code (MAPLOS) show a semi-quantitative agreement with the measurements. The predominant uncertainties in the numerical simulations were found to originate from the modeling of the first wall geometry and also from the assumed plasma current and source profiles. Direct measurements of the diffusion rate were performed by shadowing the detector with a second movable probe used as an obstacle. The diffusion rate was also measured by moving the detector behind the radius of the RF limiters, located on the outer wall. Comparisons of these experimental results with numerical simulations, which include diffusive mechanisms, indicate a quantitative agreement with the TF stochastic ripple diffusion model.

  2. Some safety studies for conceptual fusion--fission hybrid reactors. Final report

    International Nuclear Information System (INIS)

    The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

  3. Coil Design and Related Studies for the Fusion-Fission Reactor Concept SFLM Hybrid

    OpenAIRE

    Hagnestål, Anders

    2012-01-01

    A fusion-fission (hybrid) reactor is a combination of a fusion device and a subcritical fission reactor, where the fusion device acts as a neutron source and the power is mainly produced in the fission core. Hybrid reactors may be suitable for transmutation of transuranic isotopes in the spent nuclear fuel, due to the safety margin on criticality imposed by the subcritical fission core. The SFLM Hybrid project is a theoretical project that aims to point out the possibilities with steady-state...

  4. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  5. System model for analysis of the mirror fusion-fission reactor

    International Nuclear Information System (INIS)

    This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

  6. System model for analysis of the mirror fusion-fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Carlson, G.A.

    1977-10-12

    This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters.

  7. Evaluation of CO2 emission in the life cycle of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Global warming problem is one of the most serious problems which human beings are currently face. Carbon Dioxide (CO2) from power plants is considered one of the major causes of the global warming this study, CO2 emission from Tokamak fusion power plants are compared with those from conventional present power generating technologies. Plasma parameters are calculated by a systems code couples the ITER physics, TF coil shape, and cost calculation. CO2 emission from construction and operation is evaluated from summing up component volume times CO2 emission intensities of the composing materials. The uncountable components on such as reactor building, balance of plants, etc., are scaled from the ITER referenced power reactor (ITER-like) by use of Generomak model. Two important findings are revealed. Most important finding- is that CO2 emissions from fusion reactors are less than that from PV, and less than double of that from fission reactor. The other findings are that (i) most CO2 emissions from fusion reactors are from materials, (ii) CO2 emissions from reactor construction becomes almost 60% to 70%, rest from reactor operation, and (m) the RS reactor can reduce CO2 emission half compared with the ITER-like reactor. In conclusion, tokamak fusion reactors are excellent because of their small CO2 emission intensity, and they can be one of effective energy supply technologies to solve global warming. (author)

  8. RACC-PULSE, Neutron Activation in Fusion Reactor System

    International Nuclear Information System (INIS)

    1 - Description of program or function: CCC-0388/RACC was specifically developed to compute the radioactivity and radioactivity-related parameters (e.g. afterheat, biological hazard potential, etc.) due to neutron activation within Inertial Fusion Energy and Magnetic Fusion energy reactor systems. It can also be utilized to compute the radioactivity in fission, accelerator or any other neutron generating and neutron source system. This new designated RACC-PULSE is based on CCC-0388 and has the capability to model irradiation histories of varying flux levels having varying pulse widths (on times) and dwell periods (off times) and varying maintenance periods. This provides the user with the flexibility of modeling most any complexity of irradiation history beginning with simple steady state operating systems to complex multi-flux level pulse/intermittent operating systems. 2 - Method of solution: The solution method implemented within the RACC-PULSE code is a matrix based method which relies on the evaluation of the Matrix Exponential for the pulse period (on period), dwell period (off time) and post shutdown periods. For the pulsed and dwell periods, the Matrix Exponential was evaluated using the squaring and scaling technique outlined in a review article by Molar and Van Loan entitled Nineteen Dubious Ways to Compute the Exponential of a Matrix. A balanced binary tree method utilized for parameter storage in information systems was employed to evaluate the linear chains constructed for the post shutdown period. The RACC-Pulse code retains the capability of modeling the standard slab, cylinder, sphere and torus geometries in multi-dimensions as well as the point or zero-dimension geometry for Monte Carlo code interfacing. It provides easy interfacing with many of the standard multigroup, multidimensional neutron/photon transport code systems currently employed by the fusion community and implemented on the UNICOS Cray 2 System at NERSC. An auxiliary code is provided

  9. Comparison between a steady-state fusion reactor and an inductively driven pulse reactor

    International Nuclear Information System (INIS)

    In the present report, a comparison is made between tokamak reactors of steady state operation -SSTR- and pulse operation. The former design uses neutral beams as a current driver to realize steady state operation. The latter is inductively operated basic tokamak with burn time of one hour to a half day. This time is determined by dimensions of the central solenoid coil and these dimensions also determine the basic design concept of the pulse tokamak. The dimension includes effect of fatigue due to pulse operation. Performance as a power plant is evaluated with a schematic design of heat transport and power generation system. Heat accumulation in the primary coolant loop is studied in order to make up for a dwell time of a pulse reactor. It is shown that large heat accumulator is necessary to suppress a drop in output during the dwell time. The dwell time has an optimum length with respect to the dwell time. Comparison of fusion plant with other energy source reveals that reduction of the size is essential in order that the fusion is competitive with other sources. (author)

  10. Conceptual design study of quasi-steady state fusion experimental reactor (FEQ-Q), part 1

    International Nuclear Information System (INIS)

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 JER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included; core plasma, reactor structure, reactor core components, magnets. (author)

  11. EURAC: accelerator-based material testing device for a fusion reactor

    International Nuclear Information System (INIS)

    The European Communities' Joint Research Center (JCR) has studied the feasibility of spallation neutrons to simulate the fusion reactor first wall conditions. It can be shown that spallation neutrons, produced by 600 MeV protons impinging on a thin lead target are simulating the fusion reactor first wall conditions as well as, or even better than, neutron sources based on the D-Li stripping or D-T fusion reaction. A D-T fusion cycle produces five times more neutrons per unit of energy released than a fission cycle, with about twice the damage energy and the capability to produce ten times more hydrogen, helium and transmutation products than fission neutrons. They determine, together with other parameters, the lifetime of the construction materials for the low plasma-density fusion reactors (Tokamak, Tandem-Mirror, etc.), which require a first wall. 15 refs., 1 fig

  12. Expected effect of fusion reactor on global environment. Nuclear fusion as a global warming mitigation technology

    International Nuclear Information System (INIS)

    This paper outlines the use of nuclear fusion as a global warming mitigation technology. Life cycle CO2 emission from a nuclear fusion plant is quite low; it is comparable to that of nuclear fission. Nuclear fusion has the potential to contribute future energy systems and environment. The technological feasibility of nuclear fusion should be demonstrated in order to begin clarifying the potential contribution of nuclear fusion as well as to educate those outside of the fusion community about its potential. (author)

  13. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

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

    International Nuclear Information System (INIS)

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

  15. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    Science.gov (United States)

    Hançerliogulları, Aybaba; Cini, Mesut

    2013-10-01

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

  16. Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

  17. General Atomic Company fusion experimental power reactor conceptual design

    International Nuclear Information System (INIS)

    The results of a two-year, conceptual design study of a fusion experimental power reactor (EPR) are presented. For this study, the primary objectives of the EPR are to obtain plasma ignition conditions and produce net electrical power. The design features a Doublet plasma configuration with a major radius of 4.5 meters. The average plasma beta is 10 percent which yields a thermonuclear power level of 410 MW during a 105 second burn period. With a duty factor of 0.84, the gross electrical output is 124 MW(e) while the net output is 37 MW(e). The design features a 25 cm thick, helium cooled, modular, stainless steel blanket with a 1 cm thick, thermal radiation-cooled silicon carbide first wall. Sufficient shielding is provided to permit contact maintenance outside the shield envelop within 24 hours after shutdown. An overall facility concept was developed, including a superheated steam cycle power conversion system. Preliminary cost estimates and construction schedules were also developed

  18. Design of separated first wall for fusion experimental reactor

    International Nuclear Information System (INIS)

    Design studies and R and D activities on the separate first wall for tokamak fusion experimental reactors have been progressed at JAERI. The first wall has a high probability of unexpected damage because of the uncertainties in local heat and particle loads and it requires easy replacement in case of failure. In order to satisfy the requirement of assembly and maintenance, the first wall mechanically separated and separately cooled from a massive blanket module has been proposed as a promising concept with a number of advantageous features, such as easy handling during assembly/disassembly due to light weight (∝350 kg), short down-time for maintenance operation, minimized amount of radwaste and so on. A fail-safe structure, which is consistent with in-service-inspection requirements, has been realized by employing a reliable double-walled thin shell structure sandwiching metal mesh. A quilting structure of austenitic stainless steel (SS316) cooled by low pressure (2 MPa), low temperature (100-150 C) water is employed to accommodate high surface heat flux of more than 0.3 MW m-2 and nuclear heating together with large electromagnetic loads up to 2 MPa. This paper describes the outlines of the structural design of the separated first wall, cooling and manifolds, mechanical connection to blanket structure, fabrication procedure, results of thermo-mechanical analyses and related R and D activities performed at JAERI. (orig.)

  19. Conceptual design study of fusion experimental reactor (FY86FER)

    International Nuclear Information System (INIS)

    This report describes the results of the capacity estimation for the electrical power system on the typical two candidates for the FER (Fusion Experimental Reactor) which were picked out through the process of '86 FER scoping studies. Main concern in the electrical systems is coil power supplies which have a capacity of about 1 GW, and this is dominated by poloidal coil power supplies. Then, studies to reduce the converter capacity are concentrated on the poloidal coil power system in relation to the sypplying poloidal flux at the initial phase of plasma ramp-up. A quench protection circuit was proposed on the toroidal coil power supply. On the position control power supply, a circuit with reasonable functions was proposed. Under these system studies, general specifications were determined and the capacity of each power supply unit was estimated. On the poloidal coil power supply system, the accumulated capacity of converters amounted to 885 MW for the one candidate and 782 MW for another. (author)

  20. SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization

    International Nuclear Information System (INIS)

    1 - Description of problem or function: Given the material density profiles which describe a one-dimensional reference system with a neutron source, SWAN will calculate, and optionally implement, density changes so as to optimize a single functional parameter of the system. 2 - Method of solution: The one-dimensional discrete-ordinate transport code ANISN is used to calculate flux and adjoint distributions for specified sources. The code SWIF calculates first-order estimates of the effect of material density changes on a goal functional, and from these evaluates effectiveness functions for the substitution of one material for another. Density distribution changes are then calculated which would optimize the goal functional, optionally subject to a constraint of holding another functional constant (to first order). 3 - Restrictions on the complexity of the problem: SWAN is not designed to analyze critical systems; it assumes that there is a fixed source, as in shielding or fusion reactor applications. Otherwise it is compatible with ANISN. All arrays are variably-dimensioned, so that there are no restrictions on individual dimensions

  1. Studies on plasma shutdown of JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Shutdown of the plasma with a time-dependent one-point model is described. The pseudoclassical scaling law plays a role in the plasma diffusion in the low energy region below several keV and the trapped ion scaling law in the higher energy region. In this shutdown model, only deuterium is inserted during 20-second shutdown process. In the first 10 sec, while the plasma temperature, electron density and plasma current decrease from 7 keV to 1 keV, 1.1 x 1020m-3 to 1019m-3 and 4 MA respectively the fusion power falls down with gradual decrease of heating power. During the second 10 sec, while the plasma temperature, electron density and plasma current decrease from 1 keV to 100 keV, 1019m-3 to 1018m-3 and 1 MA to 100 kA respectively, the plasma thermal energy is removed. Plasma one-turn voltages are -4.0 volt and -0.5 -- -1.0 volt which fall the plasma current down to 1 MA and 100 kA during the first 10 sec and the second 10 sec, respectively. Decrease of plasma current largely lowers plasma density and energy since particle and energy confinement times decrease as plasma current decreases. Deuterium insertion rate below that in the equilibrium operation little lowers plasma density and energy. This plasma shutdown scheme is effective in driven-type reactors. (auth.)

  2. Nonlinear analysis of a superconducting magnet of fusion reactor

    International Nuclear Information System (INIS)

    Superconducting magnets being considered for the design of experimental fusion reactors consist of multiphase materials, involving superconductors, stabilizers, structural reinforcement and insulators. From the structural standpoint, the magnets will behave differently for the corresponding loads. Under operating conditions, the magnets will probably respond in a linear elastic manner. However, under abnormal conditions elastic-plastic deformations may become important. Furthermore, since the magnets are laminated structures, delamination of the layers and interlaminar slip may occur due to the loss of bonding. Such action would in turn reduce the overall stiffness of the structure and cause stress redistribution and strain accumulation in localized region. In this paper, two constitutive models are proposed to represent the structural response of the superconducting magnet. The constitutive models for the two-dimensional and three-dimensional states of deformation are implemented into a nonlinear finite element program. Numerical results are obtained for a typical superconducting magnet design to demonstrate the effects of elastic-plastic deformations and interlaminar slip. (Auth.)

  3. Design of the TFTR [Tokamak Fusion Test Reactor] maintenance manipulator

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) plans to generate a total of 3 x 1021 neutrons during its deuterium-tritium run period in 1900. This will result in high levels of radiation, especially within the TFTR vacuum vessel. The maintenance manipulator's mission is to assist TFTR in meeting Princeton Plasma Physics Laboratory's personnel radiation exposure criteria and in maintaining as-low-as-reasonably-achievable principals by limiting the radiation exposure received by operating and maintenance personnel. The manipulator, which is currently being fabricated and tested by Kernforschungszentrum Karlsruhe, is designed to perform limited, but routine and necessary, functions within the TFTR vacuum torus after activation levels within the torus preclude such functions being performed by personnel. These functions include visual inspection, tile replacement, housekeeping tasks, diagnostic calibrations, and leak detection. To meet its functional objectives, the TFTR maintenance manipulator is required to be operable in TFTR's very high vacuum environment (typically 2 x 10-8 Torr). It must also be bakeable at 150 degree C and able to withstand the radiation environment

  4. Thermodynamics of ceramic breeder materials for fusion reactors

    International Nuclear Information System (INIS)

    Based on known or deduced phase relationships in ternary lithium oxygen systems such as Li-Al-O, Li-Si-O and Li-Zr-O, the unknown free enthalpy of formation values of ternary compounds are calculated starting from the known data of the compounds of the binary border systems. Criterion for the data assessment is interconsistency of the data of all the compounds within a given multi-component system. With the help of these data the development of partial pressures during the breeding process can be calculated for all the compounds of interest. In order to facilitate a compatibility assessment the quaternary systems Cr-Li-Si-O, Fe-Li-Si-O and Be-Li-Si-O were also investigated and thermodynamic data of pertinent ternary and quaternary compounds determined. Both the partial pressure development and the compatibility behaviour of a lithium containing compound are criteria for its qualification as a breeder material for a fusion reactor. (orig.)

  5. Energy analysis and carbon dioxide emission of Tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Energy gain and carbon dioxide (CO2) emission of tokamak fusion power reactors are evaluated in this study compared with other reactor types, structural materials, and other Japanese energy sources currently in use. The reactors treated in this study are (1) a conventional physics performance international thermonuclear experimental reactor (ITER), like a reactor based upon the ITER engineering design activity (ITER-EDA), (2) a RS (reversed shear) reactor using the reversed shear safety-factor/plasma current profile, and (3) a ST (spherical torus) reactor based upon the final version of the advanced reactor innovative engineering study ST (ARIES-ST). The input energy and CO2 emission from these reactors are calculated by multiplying the weight or cost of the fusion reactor components by the energy intensity and/or with the CO2 intensity data, which are updated as often as possible. The ITER cost estimation is estimated based on the component unit costs. The following results were obtained: (1) The RS and the ST reactor can double the energy gain and reduce CO2 emission by one-half compared with the ITER-like reactor. (2) Silicon carbide (SiC) used as the structural material of inner vessel components is best for energy gain and CO2 emission reduction. (3) The ITER-like reactor is slightly superior to a photovoltaic (PV) with regard to CO2 emission. (4) The energy gain and CO2 emission intensity of the RS reactor and the ST reactor are as excellent as those of a fission reactor and a hydro-powered generator. These results indicate that a tokamak fusion power reactor can be one of the most effective power-generating technologies both in high-energy payback gains and reduction of CO2

  6. Perspective on fusion research in China (2) fusion activities in China with special intonation on hybrid reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Lijian, Qiu

    2001-09-01

    Chinese fusion research was started from 1958, but with more clear problem definition it has been set up as the national program for development of the hybrid reactor in 1986. In this paper, it will be described how the organized program is going on.

  7. Environmental and Economic Assessments of Magnetic and Inertial Confinement Fusion Reactors

    International Nuclear Information System (INIS)

    Full text: Global warming due to rapid greenhouse gas (GHG) emission is one of the present-day crucial problems, and fusion reactors are expected as abundant electric power generation systems to reduce GHG emission amounts. To search for environment-friendly and economic fusion reactor system, comparative system studies have been done for several magnetic confinement fusion (MCF) reactors, and recently extended to include inertial confinement fusion (ICF) reactors. At first, economic assessment models for MCF reactors including tokamak (TR), helical (HR) and spherical tokamak (ST) systems are described. These MCF reactor designs strongly depend on achievable plasma beta value and magnetic field strength. In the system code we confirmed the advantage of high-beta TR designs in cost of electricity (COE). After wide parameter scans, we obtained the new COE scaling formulas for MCF reactors as functions of electric power (1-3 GW), plant availability (0.65-0.85), normalized beta (3-5) or averaged beta (3-5%), maximum magnetic field strength (10-16 T), thermal efficiency (0.37-0.59) and operation year (20-40 Year). In the case of ICF reactors, fast ignition concept is adopted here. The target gain and driver repetition rate are evaluated assuming driver energy, driver efficiency (∼ 0.075), compression efficiency (∼ 0.05) and heating efficiency (∼ 0.10), which critically determine the fusion core system. The life-cycle CO2 emissions equivalently including methane gas are also evaluated using the input-output table method. High power plant assessment and carbon tax effect on COE are also evaluated, and fusion power plants are compared with other electric power plants. The fusion reactors emit less GHG than fossil fuel thermal power plant. In comparison with fission reactors, the fusion has a disadvantage in COE, but has an advantage in GHG emission. When the carbon tax of around 3,000 yen/t-CO2equi is introduced, the COE of fusion reactor might be at the same level

  8. Economic evaluation of D-T, D-3He, and catalyzed D-D fusion reactors

    International Nuclear Information System (INIS)

    Because the D-3He reaction generates no neutrons and the D-D reaction can use abundant fuel resources, these reactions are expected to be used in advanced fuel fusion reactors. Economic considerations and engineering problems are important for realizing such reactors as commercial plants. Therefore, we estimate and compare the cost of electricity (COE) from D-T, D-3He, and catalyzed D-D (cat D-D) fusion reactors. D-3He and cat D-D reactors have a low neutron wall load. Therefore, the D-3He reactor has no wall replacement cost. In addition, no tritium breeding system is needed for the D-3He reactor, but 3He gas is rare. Because the reaction rates of the D-3He and D-D reactions are less, D-3He and D-D reactors require highly efficient confinement properties and operation at high ion temperatures. Furthermore, the power densities of D-3He and D-D reactors are smaller than that of the D-T reactor; thus, D-3He and D-D reactors require a large plasma volume. Assuming a high ion temperature (= 60 keV) and high normalized beta (= 7-8), the COE of a D-3He reactor is expected to be similar to that of a D-T reactor. In terms of cost, cat D-D is disadvantageous in comparison with D-3He and D-T reactors. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    McAdams, R

    2014-02-01

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

  12. Hydrogen Spectral Line Shape Formation in the SOL of Fusion Reactor Plasmas

    Directory of Open Access Journals (Sweden)

    Valery S. Lisitsa

    2014-05-01

    Full Text Available The problems related to the spectral line-shape formation in the scrape of layer (SOL in fusion reactor plasma for typical observation chords are considered. The SOL plasma is characterized by the relatively low electron density (1012–1013 cm−3 and high temperature (from 10 eV up to 1 keV. The main effects responsible for the line-shape formation in the SOL are Doppler and Zeeman effects. The main problem is a correct modeling of the neutral atom velocity distribution function (VDF. The VDF is determined by a number of atomic processes, namely: molecular dissociation, ionization and charge exchange of neutral atoms on plasma ions, electron excitation accompanied by the charge exchange from atomic excited states, and atom reflection from the wall. All the processes take place step by step during atom motion from the wall to the plasma core. In practice, the largest contribution to the neutral atom radiation emission comes from a thin layer near the wall with typical size 10–20 cm, which is small as compared with the minor radius of modern devices including international test experimental reactor ITER (radius 2 m. The important problem is a strongly non-uniform distribution of plasma parameters (electron and ion densities and temperatures. The distributions vary for different observation chords and ITER operation regimes. In the present report, most attention is paid to the problem of the VDF calculations. The most correct method for solving the problem is an application of the Monte Carlo method for atom motion near the wall. However, the method is sometimes too complicated to be combined with other numerical codes for plasma modeling for various regimes of fusion reactor operation. Thus, it is important to develop simpler methods for neutral atom VDF in space coordinates and velocities. The efficiency of such methods has to be tested via a comparison with the Monte Carlo codes for particular plasma conditions. Here a new simplified method

  13. Quantitative analysis of economy and environmental compatibility of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    The current worth of the economy, energy gain, carbon dioxide (CO2) emission, and waste disposal of tokamak fusion power reactors are quantitatively evaluated compared with other current Japanese energy sources. The following results were obtained : (1) CO2 emission intensity (i.e., CO2 emission per unit kWh) from the International Thermonuclear Experimental Reactor-Engineering Design Activity (ITER-EDA) scale power reactor (referred to here as the ITER-like reactor), whose physics performance is conventional, can be 25% lower than that of a common household photovoltaic. The energy gain of the ITER-like reactor is comparable to that of a coalfired power plant. The cost is four times higher than that of a fission reactor; however, note that this cost evaluation is based upon FOAK (first-of-a-kind) cost evaluation. (2) The CO2 emission intensities and energy gains of RS and ST reactors are comparable to those of fission reactors. (3) Radioactive waste disposal volume for the ITER-like reactor is similar to that for a fission reactor. We believe that continuing tokamak fusion research and development is worthy, since tokamak fusion is an environmentally compatible future technology. (author)

  14. Swelling, mechanical properties, and microstructure of Type 316 stainless steel at fusion reactor damage levels

    International Nuclear Information System (INIS)

    Alloys such as AISI 316 stainless steel exhibit more swelling and larger decreases in ductility when irradiated to produce fusion reactor He and dpa levels than at fast reactor He and dpa levels. For T approx. 0C to ensure adequate ductility for long-term service

  15. Workshop summaries for the third US/USSR symposium on fusion-fission reactors

    International Nuclear Information System (INIS)

    Workshop summaries on topics related to the near-term development requirements for fusion-fission (hybrid) reactors are presented. The summary topics are as follows: (1) external factors, (2) plasma engineering, (3) ICF hybrid reactors, (4) blanket design, (5) materials and tritium, and (6) blanket engineering development requirements

  16. Importance of effects due to fusion α-particles for tokamak reactor design

    International Nuclear Information System (INIS)

    Issues related to the presence of fusion α-particles which are of importance for the design of a tokamak reactor are listed and shortly discussed. It is concluded that these issues, although to a large extent directly connected with the general problems of tokamak physics, require more attention to provide the information needed for designing a tokamak reactor. (orig.)

  17. Workshop summaries for the third US/USSR symposium on fusion-fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L. (ed.)

    1979-07-01

    Workshop summaries on topics related to the near-term development requirements for fusion-fission (hybrid) reactors are presented. The summary topics are as follows: (1) external factors, (2) plasma engineering, (3) ICF hybrid reactors, (4) blanket design, (5) materials and tritium, and (6) blanket engineering development requirements. (MOW)

  18. Design of tritium breeding experiments for the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Among intense fusion-neutron generators of the 1980's, the unique features of the TFTR are a geometrically extended D-T fusion-neutron source and a neutron spectrum, including backscattered neutrons, characteristic of a practical toroidal fusion reactor. It is planned to install a tritium-breeding module on the TFTR in order to take advantage of this opportunity to obtain reactor-relevant integral neutronics data and breeding rate profiles. These data will be combined with the measured neutron source parameters and the spatially dependent fusion-neutron fluence for comparison with the predictions of neutronics design codes. The results of this program will help determine the blanket coverage factors needed to achieve tritium self-sufficiency in future toroidal reactors. A preliminary conceptual design of a TFTR blanket module has been completed, utilizing lithium oxide as the tritium breeding material

  19. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (Pfus ∼ 10-100 MW, βN ∼ 2-3, Qp ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  20. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    Science.gov (United States)

    Stacey, W. M.

    2001-02-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (Pfus approx 10-100 MW, βN approx 2-3, Qp approx 2-5, R approx 3-5 m, I approx 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-à-vis a neutron source is reviewed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  2. Irradiation effects on the ductility of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Austenitic and ferritic-martensitic stainless steels have been proposed as first wall structural materials for the next generation of fusion devices. In order to study the effect of high temperature irradiation on their tensile properties, specimens of the steel AISI 316 L (CEC reference), of the martensitic steel W. Nr 1.4914 and of the duplex ferritic-martensitic steel EM12 have been irradiated in the BR2 reactor in Mol. The austenitic steel was irradiated at 4700C to about 1.1 1022n/cm2 ( E>0.1 MeV) while the ferritic-martensitic steels were irradiated at 5900C to about 7.7 1022n/cm2 (E>0.1 MeV). The tensile tests of the 316 L steel have been performed between 250 and 7500C. Below around 5500C, the yield stress after irradiation increased from about 160 to 270 MPa and the total elongation decreased from 42 to about 26%. At 7500C, the yield stress increase was small but the total elongation decreased from 60 to only 10%. At this temperature, the rupture of the irradiated specimen was intergranular while all the other specimens presented a transgranular rupture. At 6500C the variations were intermediate. The change of the ultimate tensile strength was small at all test temperatures. The EM12 and W. Nr 1.4914 steels tested only at 5500C, showed a decrease of the yield and tensile strength as well as an increase of the total elongation. The same tests performed on specimens which have been heat treated in parallel showed that the observed changes were due, in a large part, if not completely, to the maintenance of steels at high temperature

  3. HYLIFE-II inertial confinement fusion reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1990-12-14

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

  4. The current status of fusion reactor blanket thermodynamics

    International Nuclear Information System (INIS)

    The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li3N, Li2O, and Li2C2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

  5. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Wang, Shijia; Wang, Shaojie

    2015-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  9. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    1993-01-01

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

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

    Science.gov (United States)

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

    2015-10-01

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

  13. Economical and life-cycle energy assessment of magnetic fusion power reactors

    International Nuclear Information System (INIS)

    We analyzed several types of fusion reactors, tokamak (TR), spherical tokamak (ST), helical (HR), and inertial fusion reactor (IR) using physics, engineering and cost (PEC) code, which evaluates economic and lifecycle energy amount quantitatively. We compared the cost of electricity (COE) and the energy payback ratio (EPR) of each fusion reactors with those of fission power plant. Especially, we focus on the EPR of TR with several blanket and shield designs having scarce materials such as silicon carbide (SiC), vanadium alloy (V), and ferritic steel (FS). As the result, we found that the EPR of TR with SiC/LiPb blanket/shield model is the lowest. The COEs and the input energy of TR (βN = 4.0) and IR are lower than those of ST and HR. The COE of fusion reactor is two times higher than that of fission power plants. However the EPR of fusion reactor is as high as that of fission reactor. (author)

  14. Feasibility of a Fusion Hybrid Reactor Based on the Gasdynamic Mirror

    International Nuclear Information System (INIS)

    Full text: A Comprehensive analysis of the feasibility of a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM) is presented. Since the primary role of the fusion component is to supply neutrons to a blanket laden with fertile material, it can operate at or near “breakeven” condition which is a much less stringent condition than that required for a pure fusion reactor. As a high beta device, with demonstrated MHD and kinetic stability, the GDM is chosen for utilization in such a reactor because it can also operate in steady state. Using extensive multigroup neutronic analysis, we show that such a reactor is capable of breeding fissile material and burning it to produce tens to hundreds of megawatts of thermal power per centimeter of length “safely” since it will be “subcritical”, and “securely” because of the use of a thorium fuel cycle which is known to be resistant to “proliferation” and clandestine operations. Moreover, we demonstrate that D-D fusion reactions are more suitable for use in a hybrid reactor since the energy of the neutrons produced by these reactions is closer to “thermalization” than those produced by D-T leading to a much more manageable waste disposal problem. Finally, since the reactor in question is “self-fueling” it can be designed to operate for an extensive period of time without refueling. (author)

  15. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  16. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    International Nuclear Information System (INIS)

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si3N4. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

  17. Conceptual Design of a Fast-Ignition Laser Fusion Reactor FALCON-D

    OpenAIRE

    Goto, T.; Ogawa, Y; Hiwatari, R.; Asaoka, Y; Okano, K.; Someya, Y; Sunahara, A.; T. Johzaki

    2008-01-01

    A new conceptual design of the laser fusion power plant FALCON-D (Fast ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast ignition method can achieve the sufficient fusion gain for a commercial operation (~100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5~6m radius). 1-D/2-D hydrodynamic simulations showed...

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

    Energy Technology Data Exchange (ETDEWEB)

    Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id [Department of Nuclear Physics, Faculty of Mathematic and Natural Sciences, Institut Teknologi Bandung (Indonesia); Yazid, Putranto Ilham [Research and Development of Nuclear Association (Indonesia)

    2015-09-30

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-09-01

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

  1. Effects of the difference between the charge and matter deformations on fusion reactions of unstable nuclei

    CERN Document Server

    Rumin, T; Rumin, Tamanna; Takigawa, Noboru

    2002-01-01

    Relativistic mean field calculations suggest that the charge and matter deformations significantly differ in some of the unstable neutron and proton rich nuclei. We discuss the effects of the difference on the fusion reactions induced by them at energies near and below the Coulomb barrier by taking the $^{19,25,37}$Na + $^{208}$Pb reactions as examples. We also discuss whether one can probe the difference by the so called fusion barrier distribution analysis.

  2. Direct conversion of plasma energy to electricity for mirror fusion reactors

    International Nuclear Information System (INIS)

    Direct conversion of both plasma and ion beam energy to electricity for mirror fusion reactors is described. Selective leakage, magnetic expansion, electron separation, deceleration, and collection of ions are discussed. An experiment testing all processes except selective leakage gave an overall efficiency of 86.5+-1.5% for a 22-stage collector. Computer calculations of these same processes gave an efficiency of 88.6+-1.5%, considered excellent agreement. Experiments on a two-stage collector gave 65% efficiency compared to 69% calculated efficiency. One-, two-, and 22-stage converters were studied under reactor-like conditions with a large computer simulation code that accounts for space charge, secondary electrons, and finite, realistic electrode shapes. We estimate the converters' efficiencies to be 48%, 58% and 63%, respectively. The efficiencies under the assumed reactor conditions are lower than under controlled laboratory conditions (86.5%). The one-stage converter is more cost-effective than the two-stage converter, which, in turn, is more cost-effective than the 22-stage converter; although more collector stages give higher efficiency, the simpler converters can handle more power. Conversely, if higher efficiency were desired, our studies show that it can always be achieved - but at a price. Laboratory tests of mono-energetic beam direct conversion gave 96% efficiency for a 2-keV low power d.c. beam and 70% for a d.c. 20 keV, 1-kW beam. The same computer code mentioned above is being used to design practical direct converters that reduce power requirements of neutral-beam injectors in the 100 to 200-keV range. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

  5. ITER: design, construction and operation of te first experimental nuclear fusion reactor

    International Nuclear Information System (INIS)

    The ITER project's mission is to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes and its now under construction at Saint Paul-Lez-Durance (France). The ITER reactor is based on the 'tokamak' concept of plasma magnetic confinement, in which the fusion (deuterium-tritium) fuel is contained in a doughnut-shaped vessel. The plasma is kept away from the vessel walls by strong magnetic fields produced by superconducting coils surrounding the vessel and by an electrical current driven in the plasma. The ITER reactor is designed to generate 500 MW of fusion power for periods of 300 to 500 seconds with a fusion power multiplication factor, Q, of at least 10 (Q ≥ 10). ITER will also aim at demonstrating long fusion power production pulses, of at least 1000 seconds, with a fusion power multiplication factor of 5 and, ultimately, of 1 hour duration (only limited by hardware design limits) when full non-inductive operation is demonstrated. The presentation will cover the following aspects of the ITER reactor design, construction and foreseen operation: a) The basis for power production by magnetic confinement tokamak fusion reactors. b) The main features of the ITER tokamak reactor design. c) The key design principles of the ITER tokamak and of the key ancillary systems required for the operational scenarios considered to achieve the project's mission. d) The design and progress in qualification and manufacturing of the key ITER tokamak infrastructure, tokamak components and ancillary systems. e) The operational plan from the initial commissioning phase, through operation with non-nuclear hydrogen-helium plasmas to nuclear operation with deuterium-tritium plasmas and the demonstration of high Q fusion power. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

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

  7. Beryllium and lithium resource requirements for solid blanket designs for fusion reactors

    International Nuclear Information System (INIS)

    The lithium and beryllium requirements are analyzed for an economy of 106 MW(e) CTR3 capacity using solid blanket fusion reactors. The total lithium inventory in fusion reactors is only approximately 0.2 percent of projected U. S. resources. The lithium inventory in the fusion reactors is almost entirely 6Li, which must be extracted from natural lithium. Approximately 5 percent of natural lithium can be extracted as 6Li. Thus the total feed of natural lithium required is approximately 20 times that actually used in fusion reactors, or approximately 4 percent of U. S. resources. Almost all of this feed is returned to the U. S. resource base after 6Li is extracted, however. The beryllium requirements are on the order of 10 percent of projected U. S. resources. Further, the present cost of lithium and the cost of beryllium extraction could both be increased tenfold with only minor effects on CTR capital cost. Such an increase should substantially multiply the economically recoverable resources of lithium and beryllium. It is concluded that there are no lithium or beryllium resource limitations preventing large-scale implementation of solid blanket fusion reactors. (U.S.)

  8. Economic and environmental assessment modeling of magnetic and inertial fusion reactors

    International Nuclear Information System (INIS)

    In order to search for economically and environmentally optimized fusion reactors, physics properties, engineering designs and the cost of electricity (COE) are evaluated by the PEC (Physics-Engineering-Cost) system code for several magnetic confinement fusion reactors including tokamak (TR), helical (HR) and spherical tokamak (ST) reactors. The life-cycle CO2 emission amounts are also evaluated for various blanket designs using input-output table. This code has recently been upgraded to apply to inertial fusion reactor (IR) designs. The advantage of high-beta TR designs in COE and the advantage of compact ST designs in life-cycle CO2 emission reduction are clarified in the present economical and environmental assessments. The probable merits of IR design in both values are also clarified in the present model. The increase in net electric fusion power from 1GW to 3GW leads to 38% reduction in COE and 23% reduction in CO2 emission amounts. The scaling formulas of COE and CO2 emissions are derived as a function of plasma beta and net electric power. When the carbon tax of around 3,000 yen/t-CO2 is introduced, the COE of fusion reactor might be same level on that of coal-fired electric power plant and 1.5 times lower than that of oil-fired electric power plant. (author)

  9. Romanian research in the field of Tokamak fusion reactors

    International Nuclear Information System (INIS)

    To re-create the conditions of the sun and stars for the production of fusion energy on earth, scientists most accomplish three major tasks. They have already passed the first task by achieving the necessary temperatures. In same cases, they have attained temperatures as high as 510 million degrees, 20 times more then the temperature at the center of the sun. Secondly, they need to demonstrate sustained reactions where substantial amounts of energy are produced. The third major milestone for fusion would be operation of a demonstration fusion power plant. Many different magnetic confined schemes have been studied. The one which is receiving the greatest attention in the international magnetic fusion energy programme is the tokamak concept, and represents actually the most advanced fusion devices. The advantage of fusion are: - abundant fuel supply; - no risk of a nuclear accident; - no air pollution; - no high-level nuclear waste; - no generation of weapons material. The present objectives and research priorities of the fusion community are: - continuation of ongoing research; - concept improvements; - long term technology. Our research programme in the field of tokamak fusion reactions is performed mainly in the frame of international cooperation with 'I.V. Kurchatov' Nuclear Fusion Institute from Moscow, Institute of Applied Mathematics from Grenoble, Research Center from Cadarache, 'Max-Planck' Institute for Plasma Physics from Garching at Munich and Columbia University from New York. The activities carried out under our programme are closely coordinated with those of the European Atomic Energy Community and are related to current problems concerning equilibrium, stability, transport and diagnostics of tokamak plasmas. Our results are mentioned in the International Atomic Energy Agency's World Survey of Activities in Controlled Fusion Research in 1997 and the European Community's Reports EUR FUR BRU from 1993 and 1996. (author)

  10. Collection of Summaries of reports on result of research at basic experiment device for nuclear fusion reactor blanket design, 1994

    International Nuclear Information System (INIS)

    The development of nuclear fusion reactors reached such stage that the generation of fusion power output comparable with the input power into core plasma is possible. At present, the engineering design of the international thermonuclear fusion experimental reactor, ITER, is advanced by the cooperation of Japan, USA, Europe and Russia, aiming at the start of operation at the beginning of 21st century. This meeting for reporting the results has been held every year, and this time, it was held on May 19, 1995 at University of Tokyo with the theme ''The interface properties of fusion reactor materials and the control of particle transport''. About 50 participants from academic, governmental and industrial circles discussed actively on the theme. Three lectures on the topics of fusion reactor engineering and materials and seven lectures on the basic experiment of fusion reactor blanket design related to the next period project were given at the meeting. (K.I.)

  11. Charged fusion product loss measurements using nuclear activation

    Czech Academy of Sciences Publication Activity Database

    Bonheure, G.; Hult, M.; González de Orduña, R.; Arnold, D.; Dombrowski, H.; Laubenstein, M.; Wieslander, E.; Vermaercke, P.; Murari, A.; Popovichev, S.; Mlynář, Jan

    2010-01-01

    Roč. 81, č. 10 (2010), 10D331. ISSN 0034-6748. [TOPICAL CONFERENCE ON HIGH-TEMPERATURE PLASMA DIAGNOSTICS/18th./. Wildwood, New Jersey, 16.05.2010-20.05.2010] R&D Projects: GA ČR GAP205/10/2055; GA MŠk LA08048 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * fast particle loss * diagnostics * activation Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.598, year: 2010 http://rsi.aip.org/resource/1/rsinak/v81/i10/p10D331_s1

  12. A preliminary concept of stochastic model of the tritium cycle in a fusion reactor

    International Nuclear Information System (INIS)

    A preliminary concept of stochastic model of the tritium circulation in a fusion reactor was elaborated in purpose of determining the necessary minimum and current tritium inventory in real circumstances. A random character of reactor operation was assumed what is especially valid in the starting phase being of particularly low reliability of the assembly. A system of differential equations with random initial conditions describing the tritium cycle was solved for both operation and break states of the reactor. The distribution of the moments and of the number of breaks in the reactor operation was discussed and the possibilities of further development of the present model are indicated. 5 refs., 2 figs. (author)

  13. A concept of an advanced inertia fusion reactor; TAKANAWA-I

    International Nuclear Information System (INIS)

    A concept of an advanced inertia fusion reactor: TAKANAWA-I is proposed. A pellet with DT ignitor and DD major fuel, Pb wet walls, C or SiC blocks for shielding, and SiC vessels in the water pool are employed. This reactor does not need blanckets for T breeding, since T is supplied through DD reaction, and has low induced radioactivities. These and a simple structure might give a hopeful prediction of economical and safe advantages and mitigate difficulties of reactor technologies, especially remote maintenance of the reactor. (author)

  14. Critical issues of burning plasma, engineering, economic and environmental assessments on steady-state fusion reactors

    International Nuclear Information System (INIS)

    For burning plasma simulation and reactor system analysis on steady-state high beta fusion reactors, TOTAL physics code and PEC engineering code have been developed. From TOTAL analysis, it is clarified that by choosing appropriate external current drive profile, high bootstrap-current fraction is achieved in steady-state. From PEC analysis, it is found that the current drive efficiency should be raised for cost of electricity (COE) and CO2 reductions in rather low-beta reactors. Newly derived scaling formulas on COE and life-cycle CO2 emission rate might contribute to the future reactor design projection. (author)

  15. Parasitic components from charge transfer in neutral beams for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, O.A.

    1978-02-01

    Charge exchange within accelerating grids in neutral beam systems produces parasitic beam components which degrade the performance of the systems. These components also change the plasma confinement properties at the target. This note discusses parasitic beams produced in three types of grid systems: (1) TFTR/MFTF sources, (2) accel-decel grids for low energy beams, and (3) the JSC negative ion system.

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

    Science.gov (United States)

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

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A., E-mail: Azizov-EA@nrcki.ru; Ignatiev, V. V.; Subbotin, S. A., E-mail: subbotinSA@dhtp.nrcki.ru; Tsibulskiy, V. F., E-mail: sibulskiy-VF@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

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

  18. The European fusion program and the role of the research reactors

    International Nuclear Information System (INIS)

    The main objectives of the European long-term Fusion Technology Program are i) investigation of DEMO breeding blankets options, ii) development of low activation materials resistant to high neutron fluence, iii) construction of IFMIF for validation of DEMO materials, and iv) promotion of modelling efforts for the understanding of radiation damage. A large effort is required for the development and performance verification of the materials subjected to the intense neutron irradiation encountered in fusion reactors. In the absence of a strong 14.1 MeV neutron source fission materials research reactors are used. Elaborate in-pile and post-irradiation examinations are performed. In addition, the modelling effort is increased to predict the damage by a 'true' fusion spectrum in the future. Even assuming that a positive decision for IFMIF construction can be reached, the operation of a limited number of materials test reactors is needed to perform irradiation studies on large samples and for screening. (author)

  19. Concept definition of an FRC/DD-3He advanced fusion reactor

    International Nuclear Information System (INIS)

    Posibilities of advanced fusion fuel cycle reactors are investigated. Characteristics of various D - D fusion fuel cycles are clarified and which magnetic confinement method can fit the most efficient advanced fuel cycle reactor is examined. A concept definition is considered for an advanced fusion reactor with DD - 3He fuel cycle in which the plasma is confined in a field-reversed configuration or field-reversed mirror. The concept definition is developed with emphasis on the feasibility of a steady-state self-ignited DD - 3He plasma with temperatures of 100 keV, the production method, the formation of ambipolar potential in the ambient plasma and the design of plasma energy direct convertor. (author)

  20. Risk assessment of computer-controlled safety systems for fusion reactors

    International Nuclear Information System (INIS)

    The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is in its infancy. Combined plant hardware and computer software systems are often treated by making simple assumptions about software performance. This method is not acceptable for assessing risks in the complex fusion systems, and a new technique for risk assessment of combined plant hardware and computer software systems has been developed. This technique is an extension of the traditional fault tree analysis and uses structured flow charts of the software in a manner analogous to wiring or piping diagrams of hardware. The software logic determines the form of much of the fault trees