WorldWideScience

Sample records for the next step thermonuclear reactor

  1. The international thermonuclear reactor project

    International Nuclear Information System (INIS)

    James, T.R.

    1993-01-01

    The International Thermonuclear Experimental Reactor Project is a 6-year collaborative effort involving the U.S., Europe, Japan, and the Russian Federation to produce a detailed engineering design for the next-step fusion device

  2. Controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1992-01-01

    The author gives a chronological account of the research about thermonuclear fusion and presents the principle of JET thermonuclear reactor based upon the magnetic confinement. The problems of heating and confining a thermonuclear plasma may be regarded as solved. They make possible the definition of the size and geometry needed to realize a next-step tokamak (ITER, NET projects)

  3. Important problems of future thermonuclear reactors*

    Directory of Open Access Journals (Sweden)

    Sadowski Marek J.

    2015-06-01

    Full Text Available This paper concerns important and difficult problems connected with a design and construction of thermonuclear reactors, which have to use nuclear fusion reactions of heavy isotopes of hydrogen, i.e., deuterium (D and tritium (T. There are described conditions in which such reactions can occur, and different methods of a high-temperature plasma generation, i.e., high-current electrical discharges, intense microwave pulses, and injection of energetic neutral atoms (NBI. There are also presented experimental facilities which can contain hot plasma for an appropriate period, and particularly so-called tokamaks. The second part presents the technical problems which must be solved in order to build a thermonuclear reactor, that might be used for energetic purposes. There are considered problems connected with a choice of constructional materials for a vacuum chamber, its internal parts, external windings generating a magnetic field, and necessary shields. The next part considers the handling of radioactive tritium; the using of alpha particles (4He for additional heating of plasma; recuperation of hydrogen isotopes absorbed in the tokamak internal parts, and a removal of a helium excess. There is presented a scheme of a future thermonuclear power plant and critical comments on a road map which should enable the construction of an industrial thermonuclear reactor (DEMO.

  4. Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

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

  5. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-10

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

  8. Structure of thermonuclear reactor wall

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro.

    1991-01-01

    In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

  9. Research into thermonuclear fusion

    International Nuclear Information System (INIS)

    Schumacher, U.

    1989-01-01

    The experimental and theoretical studies carried out in close international cooperation in the field of thermonuclear fusion by magnetic plasma confinement have achieved such progress towards higher plasma temperatures and densities, longer confinement times and, thus, increased fusion product, that emphasis now begins to be shifted from problems of physics to those of technology as a next major step is being prepared towards a large international project (ITER) to achieve thermonuclear burning. The generation and maintenance of a burning fusion plasma in an experimental physics phase will be followed by a phase of technical materials studies at high fluxes of fusion neutrons. These goals have been pursued since 1983 by an international study group at Garching working on the design of a Next European Torus (NET). Since May 1988, an international study group comprising ten experts each from the USSR, USA, Japan, and the European Community has begun to work on a design draft of ITER (International Thermonuclear Experimental Reactor) in Garching under the auspices of IAEA. (orig.) [de

  10. [International Thermonuclear Experimental Reactor support

    International Nuclear Information System (INIS)

    Dean, S.O.

    1990-01-01

    This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to ''support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: ''to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to ''assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

  11. Inertia-confining thermonuclear molten salt reactors

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  12. Reactor wall in thermonuclear device

    International Nuclear Information System (INIS)

    Shibui, Masanao.

    1988-01-01

    Purpose: To always monitor the life of armours in reactor walls and automatically shutdown the reactor if it should be operated in excess of the limit of use. Constitution: Monitoring material of lower melting point than armours (for example beryllium pellets) as one of the reactor wall constituents of a thermonuclear device are embedded in a region leaving the thickness corresponding to the allowable abrasion of the armour. In this structure, if the armours are abrased due to particle loads of a plasma and the abrasion exceeds a predetermined allowable level, the monitoring material is exposed to the plasma and melted and evaporated. Since this can be detected by impurity monitors disposed in the reactor, it is possible to recognize the limit for the working life of the armours. If the thermonuclear reactor should be operated accidentally exceeding the life of the armours, since a great amount of the monitoring materials have been evaporated, they flow into the plasma to increase the plasma radiation loss thereby automatically eliminate the plasma. (K.M.)

  13. The Canadian initiative to bring the international thermonuclear experimental reactor to Canada

    International Nuclear Information System (INIS)

    James, R.A.

    1996-01-01

    The International Thermonuclear Experimental Reactor (ITER) is the next step in fusion research. It is expected to be the last major experimental facility, before the construction of a prototype commercial reactor. The Engineering Design Activities (EDA) of ITER are being funded by the USA, Japan, the Russian Federation, and the European Union, with each of the major parties contributing about 25% of the cost. Canada participates as part of the European coalition. The EDA is due to be completed in 1998, and the major funding partners are preparing for the decision on the siting and construction of ITER. The Canadian Fusion Fuels Technology Project (CFFTP) formed a Canadian ITER Siting Task Group to study siting ITER in Canada. The study indicated that hosting ITER would provide significant benefits, both technological and economic, to Canada. We have also confirmed that there would be substantial benefits to the ITER Project. CFFTP then formed a Canadian ITER Siting Board, with representation from a broad range of stakeholders, to champion, 'Canada as Host'. This paper briefly outlines the ITER Project, and the benefits to both Canada and the Project of a Canadian site. With this as background, the paper discusses the international scene and assesses Canada's prospects of being chosen to host ITER. (author)

  14. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    Johner, J.

    1989-04-01

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  15. Implications of fusion results for a reactor: a proposed next step device-JIT

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1989-01-01

    Simulations with a critical-temperature model have been made of proposed future devices (NET, ITER, JIT, etc.). These show that only machines with a current capability of ∼ 30MA have a sufficient ignition domain to cope with more realistic operating conditions (i.e. taking into account sawteeth effects, impurity dilution and semi-continuous operation). The importance of dilution and Bremsstrahlung radiation are clearly demonstrated; a mean temperature > 7keV is required for ignition. This prevents higher field, lower current devices from reaching ignition. Transient operations with monster sawteeth or H-mode allow such devices (>30MA) to reach ignition at lower density without additional heating. To investigate the problems of a controlled burning plasma for days in semi-continuous operation, the plasma of the next-step tokamak should be similar in size and performance to an energy producing reactor. The scientific and technical aims of such a machine should be to study burning plasma, test wall technology, provide a test-bed for breeding blankets and most importantly to demonstrate the potential and viability of fusion as an energy source. The main design characteristics of a Thermonuclear Furnace-JIT-dedicated to these objectives are presented. Watercooled copper magnets are used to benefit from proven technology. A single-null divertor configuration ensures helium exhaust and possibly benefits from an H-mode to reach the ignition domain. The X-point position relative to the dump plates would be swept to limit wall loading

  16. Some safety considerations in laser-controlled thermonuclear reactors. Final report

    International Nuclear Information System (INIS)

    Botts, T.E.; Breton, D.; Chan, C.K.; Levy, S.I.; Sehnert, M.; Ullman, A.Z.

    1978-07-01

    A major objective of this study was to identify potential safety questions for laser controlled thermonuclear reactors. From the safety viewpoint, it does not appear that the actual laser controlled thermonuclear reactor conceptual designs present hazards very different than those of magnetically confined fusion reactors. Some aspects seem beneficial, such as small lithium inventories, and the absence of cryogenic devices, while other aspects are new, for example the explosion of pressure vessels and laser hazards themselves. Major aspects considered in this report include: (a) general safety considerations, (b) tritium inventories, (c) system behavior during loss of flow accidents, and (d) safety considerations of laser related penetrations

  17. Baking method for thermonuclear reactor

    International Nuclear Information System (INIS)

    Kobayashi, Shigetada.

    1986-01-01

    Purpose: To improve the heat transmission property to the reactor core structures thereby shortening the baking time for the reactor core in thermonuclear reactors. Constitution: High temperature airs are supplied from a baking system to cooling pipeways disposed within reactor core structures and helium gas is supplied from a helium gas supply system through the reactor core structures to the inside of the reactor core for scavenging. The scavenging operation may be combined with vacuum suction. Further, the inside of the reactor is scavenged while maintaining at such a negative pressure as within a range not degrading the heat conduction property. Since the helium gas is chemically inert and poor in the depositing property, it shows no adsorbability even for the material heated to high temperature. Further, since the diffusion and heat conduction properties are high, the heat conduction property to the materials upon baking can be improved to shorten the baking time. No disadvantages are caused by the introduction of the helium gas upon baking. (Kawakami, Y.)

  18. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Blevins, J.D.; Stasko, R.R.

    1989-09-01

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

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

    International Nuclear Information System (INIS)

    Okano, Kunihiko; Kurihara, Kenichi; Tobita, Kenji

    2006-01-01

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

  20. Capacitor requirements for controlled thermonuclear experiments and reactors

    International Nuclear Information System (INIS)

    Boicourt, G.P.; Hoffman, P.S.

    1975-01-01

    Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested. (U.S.)

  1. Capacitor requirements for controlled thermonuclear experiments and reactors

    International Nuclear Information System (INIS)

    Boicourt, G.P.; Hoffman, P.S.

    1975-01-01

    Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested

  2. Use of code DTF-4 for determining the coefficient of back-reflection of the neutron within the thermonuclear plasma of a thermonuclear reactor controlled by the rate of the fission reactions. Pt. 1

    International Nuclear Information System (INIS)

    Cristea, G.

    1975-01-01

    The neutron problems are discussed of the thermonuclear reactor controlled by the rate of the fission reactions. The results obtained by rolling the DTF-4 program in a spherical geometry in the case of an ''external source'' problem permit to draw conclusions concerning the problems of the neutronics system of this thermonuclear reactor type. A relation is deduced for estimating the coefficient of back-reflection of the neutrons within the thermonuclear plasma and the focussion system is discussed of the neutronics of this reactor type

  3. The international thermonuclear reactor (ITER)

    International Nuclear Information System (INIS)

    Fowler, T.K.; Henning, C.D.

    1987-01-01

    Four governmental groups, representing Europe, Japan, USSR and U.S. met in March 1987 to consider a new international design of a magnetic fusion device for the 1990's. An interim group was appointed. The author gives a brief synopsis of what might be thought of as a draft charter. The starting point is the objective of the ITER device, which is summarized as demonstrating both scientific and technical feasibility of fusion. The paper presents an update on the current thinking and technical aspects for the International Thermonuclear Experimental Reactor (ITER). This covers not only what is happening in the U.S. but also some reports of preliminary thinking of the last technical work that occurred in Vienna

  4. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  5. Effect of plasma physics on choices of first wall materials and structures for a thermonuclear reactor

    International Nuclear Information System (INIS)

    Meade, D.M.

    1975-01-01

    Impurity ions adversely affect the behavior of present-day tokamaks, and control of impurities is expected to be a key element in determining the feasibility of thermonuclear fusion reactors. The plasma-surface interactions for tokamaks and several techniques for controlling impurities are described. The plasma-surface problem of next generation devices PLT, PDX, DIII and TFTR is expected to be similar to those encountered in a reactor. For these devices calculations indicate that most of the particle energy efflux will be in the 1 keV region. Ironically this energy region has not yet been investigated thoroughly by the surface physicists

  6. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

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

  7. The International Thermonuclear Experimental Reactor configuration evolution

    International Nuclear Information System (INIS)

    Lousteau, D.C.; Nelson, B.E.; Lee, V.D.; Thomson, S.L.; Miller, J.M.; Lindquist, W.B.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) conceptual design activities consist of two phases: a definition phase, completed in September 1988, and a design phase, now in progress. The definition phase was successful in identifying a consistent set of technical characteristics and the broad definition of the required reactor configuration and hardware. Scheduled for completion in November 1990, the design phase is producing a more detailed definition of the required components, a first cost estimate, and a description of site requirements. A major activity in the ITER design phase is the period of joint work conducted at the Max Planck Institute for Plasma Physics, Garching, Federal Republic of Germany, from June through October 1989. An official report of the findings and conclusions of this activity will be submitted to and published by the International Atomic Energy Agency (IAEA). This paper highlights the evolution of the reactor mechanical configuration since the conclusion of the definition phase. 8 figs., 2 tabs

  8. Major achievements of the European shield blanket R and D during the ITER EDA, and their relevance for future next step machines

    Energy Technology Data Exchange (ETDEWEB)

    Daenner, W. E-mail: daenner@ipp.mpg.de; Cardella, A.; Jones, L.; Lorenzetto, P.; Maisonnier, D.; Malavasi, G.; Peacock, A.; Rodgers, E.; Tavassoli, F

    2000-11-01

    In the frame of the international thermonuclear experimental reactors (ITER) collaboration, the European home team (EU HT) has committed significant efforts on the R and D for the Shield Blanket. This paper summarises the main achievements of this programme, which have been obtained over the last 7 years. The depth of R and D extends from generic activities up to the manufacture of prototypes, but has, in accordance with the design progress, reached different stages of maturity for the various components. New ITER options being considered since early 1998 have not made these activities irrelevant. With few exceptions, the results are still applicable for less ambitious next step machines, or transferable to components with similar functions or requirements.

  9. ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

    International Nuclear Information System (INIS)

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

  10. Vacuum problems of thermonuclear reactor design

    International Nuclear Information System (INIS)

    Paty, L.

    1981-01-01

    A thermonuclear reactor can be considered to be a vacuum system in which constant concentration should be maintained of reacting particles while permanently discharging the undesirable particles using a system of pumps. The discharging proceeds in two stages: in the former, the reactor is degassed using external pumps connected to the reactor chamber through a pumping pipe. The latter in which hydrogen is admitted, uses high pump-rate machines based on the principle of the binding of the gas to the pump surface and must not introduce molecules of higher atomic mass in the system. Turbomolecular pumps of diffusion oil pumps are most suitable for the former stage while condensation, cryosorption, titanium pumping machines and special pumping methods are most suitable for the latter stage. Examples are shown of the pump system design for Tokamak 10 and for facilities at the Euratom laboratory in Fontenay-aux-Roses. (M.D.)

  11. Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Majumdar, S.

    1991-01-01

    Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

  12. Metrology/viewing system for next generation fusion reactors

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M.; Dagher, M.A.

    1997-01-01

    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

  13. Thermonuclear power plants and the environment

    International Nuclear Information System (INIS)

    Becka, J.

    1978-01-01

    Environmental safety and protection from the effects of the thermonuclear power plants are discussed. Factors are assessed which should be considered in the choice of fuel and breeding material of a thermonuclear reactor, the problems of structural material activation and the overall reactor concepts. Main specifications are given of the US thermonuclear power plant projects with D-T reaction based reactors. The overall amounts of tritium in the reactor cycles are shown. The potential biological risk is evaluated for the different materials considered for the UWMAK-1 project. Discussed are possible pathways of activity release in normal plant operation, non-radioactive aspects, such as waste heat, the magnetic field effect on personnel and population, etc., as well as possible environmental impacts in case of accidents. (B.S.)

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

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.

    1975-06-01

    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)

  15. Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

    Science.gov (United States)

    Shimomura, Y.

    1994-05-01

    The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

  16. The development of beryllium plasma spray technology for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Castro, R.G.; Elliott, K.E.; Hollis, K.J.; Watson, R.D.

    1999-01-01

    Over the past five years, four international parties, which include the European Communities, Japan, the Russian Federation and the United States, have been collaborating on the design and development of the International Thermonuclear Experimental Reactor (ITER), the next generation magnetic fusion energy device. During the ITER Engineering Design Activity (EDA), beryllium plasma spray technology was investigated by Los Alamos National Laboratory as a method for fabricating and repairing and the beryllium first wall surface of the ITER tokamak. Significant progress has been made in developing beryllium plasma spraying technology for this application. Information will be presented on the research performed to improve the thermal properties of plasma sprayed beryllium coatings and a method that was developed for cleaning and preparing the surface of beryllium prior to depositing plasma sprayed beryllium coatings. Results of high heat flux testing of the beryllium coatings using electron beam simulated ITER conditions will also be presented

  17. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

  18. Tandem mirror next step conceptual design

    International Nuclear Information System (INIS)

    Doggett, J.N.; Damm, C.C.; Bulmer, R.H.

    1980-01-01

    A study was made to define the features of the experimental mirror fusion device - The Tandem Mirror Next Step, or TMNS - that will bridge the gap between present mirror confinement experiments and a power-producing reactor. We outline the project goals, describe some initial device parameters, and relate the technological requirements to ongoing development programs

  19. Magnet systems for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

    The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs

  20. Status of the mirror-next-step (MNS) study

    International Nuclear Information System (INIS)

    Damm, C.C.; Doggett, J.N.; Bulmer, R.H.

    1979-09-01

    A study was made to define the features of the experimental mirror fusion device - the Mirror Next Step, or MNS - that will bridge the gap between present mirror confinement experiments and a power-producing reactor. The project goals and organization of the study are outlined, some initial device parameters are described, and the technological requirements are related to ongoing development programs

  1. Continuously renewed wall for a thermonuclear reactor

    International Nuclear Information System (INIS)

    Livshits, A.I.; Pustovojt, YU.M.; Samartsev, A.A.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii)

    1982-01-01

    The possibility of creating a continuously renewed first wall of a thermonuclear reactor is experimentally investigated. The following variants of the wall are considered: the wall is double, its part turned to plasma is made of comparatively thin material. The external part separated from it by a small gap appears to be protected from interaction with plasma and performs structural functions. The gap contains the mixture of light helium and hydrogen and carbon-containing gas. The light gas transfers heat from internal part of the wall to the external part. Carbon-containing gas provides continuous renewal of carbon coating of the operating surface. The experiment is performed with palladium membrane 20 μm thick. Carbon is introduced into the membrane by benzol pyrolysis on one of the surfaces at the membrane temperature of 900 K. Carbon removal from the operating side of the wall due to its spraying by fast particles is modelled by chemical itching with oxygen given to the operating membrane wall. Observation of the carbon release on the operating surface is performed mass-spectrometrically according to the observation over O 2 transformation into CO and CO 2 . It is shown that in cases of benzol pressure of 5x10 -7 torr, carbon current on the opposite surface is not less than 3x10 12 atoms/sm 2 s and corresponds to the expected wall spraying rate in CF thermonuclear reactors. It is also shown that under definite conditions the formation and maintaining of a through protective carbon coating in the form of a monolayer or volumetric phase is possible

  2. Neutronic calculation of the next fuel elements for the Argonaut reactor

    International Nuclear Information System (INIS)

    Oliveira, C.R.E.; Brito Aghina, L.O. de

    1981-01-01

    The best parameters of the next fuel elements of the Argonaut reactor, at IEN (Instituto de Engenharia Nuclear - Brazil), were determined. The next fuel elements will be rods of an uranium mixture (19.98% enriched), graphite and bakelite. The parameters to be determined are: mixture density, percentage of uranium in the mixture, pellet radius, rod material and elements arrangement (step). The calculations routines consisted in the analysis of several steps, using the LEOPARD computer code for cell calculations and RMAT1D for one dimensional spatial calculations (criticality) with four energy groups. Finally a neutronic study of the Argounat reactors present configuration was done, using the HAMMER computer code (cell), the EXTERMINATOR computer code (two-dimensional calculations) and RAMAT1D. (Author) [pt

  3. Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

    International Nuclear Information System (INIS)

    Paris, P.J.; Yassen, F.; Assis, A.S. de; Raposo, C.

    1988-07-01

    The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.) [pt

  4. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  5. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C.

    2010-01-01

    More and more 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 CO 2 -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 meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) 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. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm -2 , meanwhile 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

  6. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

    2010-07-01

    More and more 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 CO{sub 2}-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 meterials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PEMs) 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. Materials for plasma facing components have to fulfill a number of requirements. First of all the materials have to be plasma compatible, i.e. they should exhibit a low atomic number to avoid radiative losses whenever atoms from the wall material will be ionized in the plasma. In addition, the materials must have a high melting point, a high thermal conductivity, and adequate mechanical properties. To select the most suitable material candidates, a comprehensive data base is required which includes all thermo-physical and mechanical properties. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm{sup -2}, meanwhile 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{sup -2} for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs

  7. Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

    International Nuclear Information System (INIS)

    Simos, N.

    2011-01-01

    operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.

  8. Thermonuclear controlled fusion: international cooperation

    International Nuclear Information System (INIS)

    Conscience, J.-F.

    2001-01-01

    This report summarizes the current worldwide status of research in the field of thermonuclear controlled fusion as well as the international research programme planed for the next decades. The two main projects will be the ITER facility (International Thermonuclear Experimental Reactor) that should produce 10 times more energy than the energy injected, and the IFMIF (International Fusion Materials Irradiation Facility) designed to study the reactions of materials under intense neutron fluxes. The future of the pioneering JET facility (Joint European Torus) is also discussed. The engagement of the various countries (USA, Japan, Germany, Russian Federation and Canada) and international organisations (EURATOM and IEA) in terms of investment and research is described. Switzerland is involved in this program through an agreement with EURATOM and is mainly dedicated to experimental studies with the TCV machine in Lausanne and numerical studies of plasma configurations. It will participate to the development of the microwave plasma heating system for the ITER machine

  9. Ecological problems of thermonuclear energetics. Review

    Energy Technology Data Exchange (ETDEWEB)

    Sivintsev, Yu V

    1980-01-01

    A review of preliminary quantitative estimates of radiation hazard of thermonuclear reactors is presented. Main attention is given to three aspects: nonradiation effect on environment, radionuclide blow-ups at normal operation and emergency situations with their consequences. The given data testify to great radiological advantages of thermonuclear energetics as compared with the modern nuclear energetics with thermal and prospective fast reactors.

  10. The ITER fusion reactor and its role in the development of a fusion power plant

    International Nuclear Information System (INIS)

    McLean, A.

    2002-01-01

    Energy from nuclear fusion is the future source of sustained, full life-cycle environmentally benign, intrinsically safe, base-load power production. The nuclear fusion process powers our sun, innumerable other stars in the sky, and some day, it will power the Earth, its cities and our homes. The International Thermonuclear Experimental Reactor, ITER, represents the next step toward fulfilling that promise. ITER will be a test bed for key steppingstones toward engineering feasibility of a demonstration fusion power plant (DEMO) in a single experimental step. It will establish the physics basis for steady state Tokamak magnetic containment fusion reactors to follow it, exploring ion temperature, plasma density and containment time regimes beyond the breakeven power condition, and culminating in experimental fusion self-ignition. (author)

  11. Vacuum pumping for controlled thermonuclear reactors

    International Nuclear Information System (INIS)

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

    1976-01-01

    Thermonuclear reactors impose unique vacuum pumping problems involving very high pumping speeds, handling of hazardous materials (tritium), extreme cleanliness requirements, and quantitative recovery of pumped materials. Two principal pumping systems are required for a fusion reactor, a main vacuum system for evacuating the torus and a vacuum system for removing unaccelerated deuterium from neutral beam injectors. The first system must pump hydrogen isotopes and helium while the neutral beam system can operate by pumping only hydrogen isotopes (perhaps only deuterium). The most promising pumping techniques for both systems appear to be cryopumps, but different cryopumping techniques can be considered for each system. The main vacuum system will have to include cryosorption pumps cooled to 4.2 0 K to pump helium, but the unburned deuterium-tritium and other impurities could be pumped with cryocondensation panels (4.2 0 K) or cryosorption panels at higher temperatures. Since pumping speeds will be limited by conductance through the ducts and thermal shields, the pumping performance for both systems will be similar, and other factors such as refrigeration costs are likely to determine the choice. The vacuum pumping system for neutral beam injectors probably will not need to pump helium, and either condensation or higher temperature sorption pumps can be used

  12. Analysis and evaluation of the hydrogen risk in a thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Chaudron, V.; Arnould, F.; Latge, C.; Laurent, A.

    2001-01-01

    After a recall of the principle of controlled thermonuclear fusion, the ITER reactor project is briefly described. The integrity of the reactor must be preserved in the case of a potential explosion of the hydrogen generated inside the reactor, in order to avoid any dispersion radioactive, chemical or toxic materials in the environment. The fundamental principles of safety developed to fulfill these objectives, in particular the defense-in-depth concept, are presented. The main potential source of hydrogen production is the oxidation of beryllium, which is used as protection material in the first wall of the torus, and the accidental presence of water, as reported in several scenarios. The confinement strategy is then described with the qualification of the role of the different barriers. Finally, the hydrogen explosion risk is analyzed and evaluated with respect to the sources, to the reference envelope scenarios and to the location of hydrogen inside the ITER reactor. It appears, at the engineering stage, that the vacuum toric vessel, the discharge reservoir and the exchanger compartments are the most worrying parts. (J.S.)

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

    International Nuclear Information System (INIS)

    Pan Chuanhong

    2010-01-01

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

  14. Thermonuclear reactor materials composed of glassy carbons

    International Nuclear Information System (INIS)

    Kazumata, Yukio.

    1979-01-01

    Purpose: To improve the durability to plasma radiation by the use of glassy carbon as the structural materials for the first wall and the blanket in thermonuclear devices. Constitution: The glassy carbon (glass-like carbon) is obtained by forming specific organic substances into a predetermined configuration and carbonizing them by heat decomposition under special conditions. They are impermeable carbon material of 1.40 - 1.70 specific gravity, less graphitizable and being almost in isotropic crystal forms in which isotropic structure such as in graphite is scarcely observed. They have an extremely high hardness, are less likely to be damaged when exposed to radiation and have great strength and corrosion resistance. Accordingly, the service life of the reactor walls and the likes can remarkably be increased by using the materials. (Horiuchi, T.)

  15. Next generation light water reactors

    International Nuclear Information System (INIS)

    Omoto, Akira

    1992-01-01

    In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

  16. TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

    The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

  17. Developing Boundary/PMI Solutions for Next-Step Fusion Devices

    Science.gov (United States)

    Guo, H. Y.; Leonard, A. W.; Thomas, D. M.; Allen, S. L.; Hill, D. N.; Unterberg, Z.

    2014-10-01

    The path towards next-step fusion development requires increased emphasis on the boundary/plasma-material interface. The new DIII-D Boundary/Plasma-Material Interactions (PMI) Center has been established to address these critical issues on a timescale relevant to the design of FNSF, adopting the following transformational approaches: (1) Develop and test advanced divertor configurations on DIII-D compatible with core plasma high performance operational scenarios in FNSF; (2) Validate candidate reactor PFC materials at reactor-relevant temperatures in DIII-D high-performance plasmas, in collaboration with the broad material research/development community; (3) Integrate validated boundary-materials interface with high performance plasmas to provide viable boundary/PMI solutions for next-step fusion devices. This program leverages unique DIII-D capabilities, promotes synergistic programs within the broad PMI community, including linear material research facilities. It will also enable us to build a compelling bridge for the US research on long-pulse facilities. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344, DE-AC05-00OR2725.

  18. Divertor plate for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Sato, Keisuke; Nishio, Satoshi.

    1993-01-01

    In a divertor plate for a thermonuclear reactor, adjacent cooling pipes are electrically insulated from each other and pipes made of a gradient functional material prepared by compositing ceramics having an insulation property and metals are metallurgically joined to at least one portion of each of the cooling pipes. Electric current caused upon occurrence of plasma disruption is interrupted by the insulation portion, so that a large circuit is not formed and electromagnetic force is decreased to such a extent that the divertor plate is not ruptured. Since a header of the cooling pipes can be installed at any optional position, the installation space can be reduced. Further, since inlet and exit collection headers can be disposed on both ends of the cooling pipes, it is possible to shorten the length of the cooling pipe of the divertor plate corresponded to high heat fluxes and reduce the pressure loss on the side of coolants to about 1/2. Further, turn back portions of small radius of curvature of the cooling pipes are eliminated to reduce the cost and extend the lifetime and, in addition, protection tiles can be attached easily. (N.H.)

  19. Plasma and controlled thermonuclear reaction

    Energy Technology Data Exchange (ETDEWEB)

    Kapitsa, P L [AN SSSR, Moscow. Inst. Fizicheskikh Problem

    1980-06-01

    Two contemporary trends of research are characterized aiming at the thermonuclear reactor, viz., tokamak type equipment and pulsed heating of a deuterium-tritium mixture using focused laser light. There is a third trend based on the use of high-power continuous wave (CW) microwave generators which allow producing a rope discharge. The design is described of an anticipated CW thermonuclear reactor. Using current experimental facilities, a continuous high-frequency discharge can be obtained at a pressure of 25 atm and electron temperature of 50 million K. The major problem involved in the design of a CW reactor is the heating of ions to the same temperature as the electron temperature and the reduction in ion gas thermal conductivity.

  20. International Thermonuclear Experimental Reactor (ITER). Engineering Design Activities (EDA). Agreement and protocol 1

    International Nuclear Information System (INIS)

    1992-01-01

    This document contains protocol 1 to the agreement among the European Atomic Energy Community, the government of Japan, the Government of the Russian Federation, and the Government of the United States of America on cooperation in the engineering design activities for the International Thermonuclear Experimental Reactor, which activities shall be conducted under the auspices of the International Atomic Energy Agency

  1. Stabilization of burn conditions in a thermonuclear reactor using artificial neural networks

    Science.gov (United States)

    Vitela, Javier E.; Martinell, Julio J.

    1998-02-01

    In this work we develop an artificial neural network (ANN) for the feedback stabilization of a thermonuclear reactor at nearly ignited burn conditions. A volume-averaged zero-dimensional nonlinear model is used to represent the time evolution of the electron density, the relative density of alpha particles and the temperature of the plasma, where a particular scaling law for the energy confinement time previously used by other authors, was adopted. The control actions include the concurrent modulation of the D-T refuelling rate, the injection of a neutral He-4 beam and an auxiliary heating power modulation, which are constrained to take values within a maximum and minimum levels. For this purpose a feedforward multilayer artificial neural network with sigmoidal activation function is trained using a back-propagation through-time technique. Numerical examples are used to illustrate the behaviour of the resulting ANN-dynamical system configuration. It is concluded that the resulting ANN can successfully stabilize the nonlinear model of the thermonuclear reactor at nearly ignited conditions for temperature and density departures significantly far from their nominal operating values. The NN-dynamical system configuration is shown to be robust with respect to the thermalization time of the alpha particles for perturbations within the region used to train the NN.

  2. Stabilization of burn conditions in a thermonuclear reactor using artificial neural networks

    International Nuclear Information System (INIS)

    Vitela, J.E.; Martinell, J.J.

    1998-01-01

    In this work we develop an artificial neural network (ANN) for the feedback stabilization of a thermonuclear reactor at nearly ignited burn conditions. A volume-averaged zero-dimensional nonlinear model is used to represent the time evolution of the electron density, the relative density of alpha particles and the temperature of the plasma, where a particular scaling law for the energy confinement time previously used by other authors, was adopted. The control actions include the concurrent modulation of the D-T refuelling rate, the injection of a neutral He-4 beam and an auxiliary heating power modulation, which are constrained to take values within a maximum and minimum levels. For this purpose a feedforward multilayer artificial neural network with sigmoidal activation function is trained using a back-propagation through-time technique. Numerical examples are used to illustrate the behaviour of the resulting ANN-dynamical system configuration. It is concluded that the resulting ANN can successfully stabilize the nonlinear model of the thermonuclear reactor at nearly ignited conditions for temperature and density departures significantly far from their nominal operating values. The NN-dynamical system configuration is shown to be robust with respect to the thermalization time of the alpha particles for perturbations within the region used to train the NN. (author)

  3. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    International Nuclear Information System (INIS)

    1997-01-01

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest

  4. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-18

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

  5. Breeder control fusion reactor. Topical interview

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-09-01

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

  6. Transport simulation of ITER [International Thermonuclear Engineering Reactor] startup

    International Nuclear Information System (INIS)

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

    1989-01-01

    The present International Thermonuclear Engineering Reactor (ITER) reference configurations are the ''Technology Phase,'' in which the plasma current is maintained noninductively at a subignition density, and the ''Physics Phase,'' which is ignited but requires inductive maintenance of the current. The WHIST 1.5-D transport code is used to evaluate the volt-second requirements of both configurations. A slow current ramp (60-80's) is required for fixed-radius startup in ITER to avoid hollow current density profiles. To reach the operating point requires about 203 V·s for the Technology Phase (18 MA) and about 270 V·s for the Physics Phase (22 MA). The resistive losses can be reduced with expanding-radius startup. 5 refs., 4 figs

  7. Estabilishing requirements for the next generation of pressurized water reactors--reducing the uncertainty

    International Nuclear Information System (INIS)

    Chernock, W.P.; Corcoran, W.R.; Rasin, W.H.; Stahlkopf, K.E.

    1987-01-01

    The Electric Power Research Institute is managing a major effort to establish requirements for the next generation of U.S. light water reactors. This effort is the vital first step in preserving the viability of the nuclear option to contribute to meet U.S. national electric power capacity needs in the next century. Combustion Engineering, Inc. and Duke Power Company formed a team to participate in the EPRI program which is guided by a Utility Steering committee consisting of experienced utility technical executives. A major thrust of the program is to reduce the uncertainties which would be faced by the utility executives in choosing the nuclear option. The uncertainties to be reduced include those related to safety, economic, operational, and regulatory aspects of advanced light water reactors. This paper overviews the Requirements Document program as it relates to the U.S. Advanced Light Water Reactor (ALWR) effort in reducing these uncertainties and reports the status of efforts to establish requirements for the next generation of pressurized water reactors. It concentrates on progress made in reducing the uncertainties which would deter selection of the nuclear option for contributing to U.S. national electric power capacity needs in the next century and updates previous reports in the same area. (author)

  8. An electromagnetic spherical phased array thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Okress, E.C.

    1983-01-01

    Discussed are salient physics aspects of a microwave singly reentrant spherical periodic phased array of uniformally distributed identical coaxial radiation elements in an essentially simulated infinite array environment. The array is capable of maintaining coherence or phase control (to the limit of the order of 300 GHz) of its spherically converging electromagnetic transverse magnetic mode radiation field, for confinement (and heating) of thermonuclear plasma in steady-state or inertial thermonuclear fusion. The array also incorporates capability for coaxial directional coupler extraction of fusionpower. The radiation elements of the array are shielded against DT Thermonuclear plasma emissions (i.e., neutrons and bremsstrahlung) by either sufficiently (available) low less tangent and cooled, spherically concentric shield (e.g., Titanium oxide); or alternately by identical material dome windows mounted on each radiation element's aperture of the array. The pump microwave power required for thermonuclear fusion feasibility comprises an array of phase-locked available klystron amplifiers (comparable gyratron amplifiers remain to be developed)

  9. Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Petti, D.A.; Haire, J.C.

    1993-12-01

    The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project

  10. Structure of pipeline or duct for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayashi, Takeshi; Fujioka, Junzo; Nishio, Satoshi; Okawa, Yoshinao; Sato, Keisuke.

    1992-01-01

    An electrically insulating material comprising a gradient function material is bonded metallurgically to a pipeline or a duct to be disposed to a magnetic field-confining type thermonuclear reactor. The gradient material has an ingredient approximate to ceramics on the side of an electrically insulative ceramic portion and an ingredient approximate to a metal on the other side. The intermediate portion between them, has a continuous gradient ingredient. Further, in the gradient portion of the electrically insulative portion, a heat expansion coefficient is also varied continuously or stepwise in addition to the electrical insulative property. Accordingly, even when a temperature distribution is caused during operation and welding upon production, thermal stresses applied to the pipelines is moderated. Further, since the electrically insulative ceramics are interposed with no support by an electric conductor, sufficient electrical insulation can be ensured. (T.M.)

  11. Cooling device for thermonuclear reactor and modular packing block for the wall realization of a such device

    International Nuclear Information System (INIS)

    Archer, J.; Stalport, G.; Besson, D.; Faron, R.; Coulon, M.

    1988-01-01

    The cooling device for a thermonuclear reactor wall is made by modular thermally conductive heat-resistant blocks (graphite by example), a prismatic head on one face of each block, the opposite face bearing against cooling tubes, a base to each block with an aperture and rods passing through the apertures reversibly fixing each row of blocks to a support [fr

  12. Stochastic models of edge turbulent transport in the thermonuclear reactors

    International Nuclear Information System (INIS)

    Volchenkov, Dima

    2005-01-01

    Two-dimensional stochastic model of turbulent transport in the scrape-off layer (SOL) of thermonuclear reactors is considered. Convective instability arisen in the system with respect to perturbations reveals itself in the strong outward bursts of particle density propagating ballistically across the SOL. The criterion of stability for the fluctuations of particle density is formulated. A possibility to stabilize the system depends upon the certain type of plasma waves interactions and the certain scenario of turbulence. A bias of limiter surface would provide a fairly good insulation of chamber walls excepting for the resonant cases. Pdf of the particle flux for the large magnitudes of flux events is modeled with a simple discrete time toy model of I-dimensional random walks concluding at the boundary. The spectra of wandering times feature the pdf of particle flux in the model and qualitatively reproduce the experimental statistics of transport events

  13. Development of technology for next generation reactor - Development of next generation reactor in Korea -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

    1993-09-01

    The project, development of next generation reactor, aims overall related technology development and obtainment of related license in 2001. The development direction is to determine the reactor type and to build up the design concept in 1994. For development trend analysis of foreign next generation reactor, level-1 PSA, fuel cycle analysis and computer code development are performed on System 80+ and AP 600. Especially for design characteristics analysis and volume upgrade of AP 600, nuclear fuel and reactor core design analysis, coolant circuit design analysis, mechanical structure design analysis and safety analysis etc. are performed. (Author).

  14. Organization of the ITER [International Thermonuclear Experimental Reactor] Project - Sharing of information and procurements

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1990-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is expected to fully confirm the scientific feasibility and to address the technological feasibility of fusion power. Consequently, the machine must be designed for controlled ignition and extended burn of deuterium-tritium plasma. It must also demonstrate and perform integrated testing of components required to utilize fusion power for practical purposes. Cooperation among four countries/organizations (United States, Soviet Union, Japan, and EURATOM) to build a single experimental reactor will reduce the cost for each country and provide an international pool of scientific and engineering resources. This paper describes ITER organization for conceptual design activity, schedule for conceptual design activities, ITER operating parameters, conceptual project schedule and cost, future plans, basic principles and problems related to task sharing, and basic principles in handling of intellectual property

  15. A high-recycle divertor for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Werley, K.A.; Bathke, C.G.

    1988-01-01

    A coupled one-dimensional (axial/radial) edge-plasma model (SOLAR) has been used to investigate tradeoffs between collector-plate and edge-plasma conditions in a doublenull, open, high-recycle divertor (HRD) for a preliminary International Thermonuclear Experimental Reactor (ITER) design. A steady-state HRD produces in attractive high-density edge plasma (5 /times/ 10 19 m/sup /minus/3/) with sufficiently low plasma temperature (10-20eV) at a tungsten plat that the sheath-accelerated ions are below sputtering threshold energies. Manageable plate heat fluxes (3-6 MW/m 2 ) are achieved by positioning the plate poloidal cross section at a minimum angle of 15-30/degree/ with respect to flux surfaces. 6 refs., 9 figs

  16. Industrial opportunities on the International Thermonuclear Experimental Reactor (ITER) project

    International Nuclear Information System (INIS)

    Ellis, W.R.

    1996-01-01

    Industry has been a long-term contributor to the magnetic fusion program, playing a variety of important roles over the years. Manufacturing firms, engineering-construction companies, and the electric utility industry should all be regarded as legitimate stakeholders in the fusion energy program. In a program focused primarily on energy production, industry's future roles should follow in a natural way, leading to the commercialization of the technology. In a program focused primarily on science and technology, industry's roles, in the near term, should be, in addition to operating existing research facilities, largely devoted to providing industrial support to the International Thermonuclear Experimental Reactor (ITER) Project. Industrial opportunities on the ITER Project will be guided by the amount of funding available to magnetic fusion generally, since ITER is funded as a component of that program. The ITER Project can conveniently be discussed in terms of its phases, namely, the present Engineering Design Activities (EDA) phase, and the future (as yet not approved) construction phase. 2 refs., 3 tabs

  17. Controlled thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Walstrom, P.L.

    1976-01-01

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

  18. Controlled thermonuclear fusion

    CERN Document Server

    Bobin, Jean Louis

    2014-01-01

    The book is a presentation of the basic principles and main achievements in the field of nuclear fusion. It encompasses both magnetic and inertial confinements plus a few exotic mechanisms for nuclear fusion. The state-of-the-art regarding thermonuclear reactions, hot plasmas, tokamaks, laser-driven compression and future reactors is given.

  19. Controlled thermonuclear fusion and the latest progress on China's HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Li Jiangang; Yang Yu

    2003-01-01

    After 50 years of research on controlled thermonuclear fusion, a new stage will be reached in 2003, when a site for the International Thermonuclear Experimental Reactor project will be chosen to start the construction. Scientists hope that this project could herald a new era in which the energy problem will be solved completely. The great progress made on the HT-7 superconducting tokamak in China has provided positive and powerful support for fusion research. The HT-7 is one of the only two superconducting tokamaks in the world that can carry out minute-scale high temperature plasma research, and has achieved a duration of 63.95s for the hot plasma discharge. This is a major step towards real steady-state operation of the tokamak configuration. We present an overview of the latest progress on the tokamak experiments in the Institute of Plasma Physics, Chinese Academy of Sciences

  20. Radiological dose rate calculations for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Khater, H.Y.; Santoro, R.T.

    1996-01-01

    Two-dimensional biological dose rates were calculated at different locations outside the International Thermonuclear Experimental Reactor (ITER) design. An 18 degree sector of the reactor was modeled in r-θ geometry. The calculations were performed for three different pulsing scenarios. This included a single pulse of 1000 s duration, 10 pulses of 1000 s duration with a 50% duty factor, and 9470 pulses of 1000 s duration with a 50% duty factor for a total fluence of 0.3 MW.a/m 2 . The dose rates were calculated as a function of toroidal angle at locations in the space between the toroidal field (TF) coils and cryostat, and in the space between the cryostat and the biological shield. The two-dimensional results clearly showed the toroidal effect, which is dominated by contribution from the activation of the cryostat and the biological shield. After one pulse, full access to the machine is possible within a few hours following shutdown. After 10 pulses, full access is also possible within the first day following shutdown. At the end of the Basic Performance Phase (BPP), full access is possible at any of the locations considered after one week following shutdown. 5 refs., 5 figs., 2 tabs

  1. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    Alesso, H.P.; Majumdar, K.C.

    1991-01-01

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

  2. The impact of confinement scaling on ITER [International Thermonuclear Experimental Reactor] parameters

    International Nuclear Information System (INIS)

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

    1988-09-01

    Energy confinement scaling is a major concern in the design of the International Thermonuclear Experimental Reactor (ITER). The existing database for tokamaks can be fitted with a number of different confinement scaling expressions that have similar degrees of approximation. These scaling laws predict confinement times for ITER that vary by over an order of magnitude. The uncertainties in the form and magnitude of these scaling laws must be substantially reduced before the plasma performance of ITER can be predicted with adequate reliability. The TETRA systems code is used to calculate the dependence of major ITER parameters on the scaling laws currently in use. Design constraints of interest in the present phase of ITER consideration are used, and the minimum-cost devices arising from these constraints are reviewed. 9 refs., 13 figs., 4 tabs

  3. Modeling of secondary emission processes in the negative ion based electrostatic accelerator of the International Thermonuclear Experimental Reactor

    OpenAIRE

    G. Fubiani; H. P. L. de Esch; A. Simonin; R. S. Hemsworth

    2008-01-01

    The negative ion electrostatic accelerator for the neutral beam injector of the International Thermonuclear Experimental Reactor (ITER) is designed to deliver a negative deuterium current of 40 A at 1 MeV. Inside the accelerator there are several types of interactions that may create secondary particles. The dominating process originates from the single and double stripping of the accelerated negative ion by collision with the residual molecular deuterium gas (≃29% losses). The resulting seco...

  4. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  5. Conception of thermonuclear reactor with a shielding layer of the first wall

    International Nuclear Information System (INIS)

    Marin, S.V.

    1979-01-01

    Considered is the way of the shielding of the first wall of a thermonuclear reactor by the layer of ISSEC (Internal spectral shifter and Energy Converter). It is a constructive non-power element placed between a plasma and the first wall, and intended for the softening of the spectrum and intensity reduction of particle fluxes falling on the first wall. Results of neutron-physical calculations of the UWMAK-type reactor blanket (in the S 4 -P 3 approximation) are presented. While comparing five materials (C, Mo, Nb, V,W) by the rate of radiation damage formation, gas production, radioactivity level and energy output in the blanket with the 316 stainless steel first wall, it is obvious that the conception of ISSEC permits to prolong the service period of the first wall. Construction elements should be then in the same irradiation conditions as those in fast reactors. Molybdenum has been taken as the best ISSEC material. It reduces the number of displaced atoms of the first wall by 20% and decreases helium production by about 100%, increases energy output in the blanket by 15-18%. However, graphite is advantageous, while comparing it to molybdenum in values of residual energy output, radioactivity level, costs and manufacture simplicity. One problem stays unsolved, which is connected with chemical sputtering of graphite at the formation of C 2 H 2 in the high temperature range. So it is hard to prefer any material now

  6. Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Honda, T.; Maki, K.; Okazaki, T.

    1994-01-01

    Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

  7. Thermonuclear device

    International Nuclear Information System (INIS)

    Yagi, Yasuomi; Takahashi, Ken; Hashimoto, Hiroshi.

    1984-01-01

    Purpose: To improve the plasma confining performances by bringing the irregular magnetic fields nearly to zero and decreasing the absolute value of the irregular magnetic fields at every positions. Constitution: The winding direction of a plurality of coil elements, for instance, double pan cake coils of toroidal coils in a torus type or mirror type thermonuclear device are reversed to each other in their laminating direction, whereby the irregular magnetic fields due to the coil-stepped portions in each toroidal coils are brought nearly to zero. This enables to bring the average irregular magnetic fields as a whole in the thermonuclear device nearly to zero, as well as, decrease the absolute value of the irregular magnetic fields in each positions. Thus, the plasma confining performances can be improved. (Moriyama, K.)

  8. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

    1999-01-01

    The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λthe transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

  9. Next Step Spherical Torus Design Studies

    International Nuclear Information System (INIS)

    Neumeyer, C.; Heitzenroeder, P.; Kessel, C.; Ono, M.; Peng, M.; Schmidt, J.; Woolley, R.; Zatz, I.

    2002-01-01

    Studies are underway to identify and characterize a design point for a Next Step Spherical Torus (NSST) experiment. This would be a ''Proof of Performance'' device which would follow and build upon the successes of the National Spherical Torus Experiment (NSTX) a ''Proof of Principle'' device which has operated at PPPL since 1999. With the Decontamination and Decommissioning (DandD) of the Tokamak Fusion Test Reactor (TFTR) nearly completed, the TFTR test cell and facility will soon be available for a device such as NSST. By utilizing the TFTR test cell, NSST can be constructed for a relatively low cost on a short time scale. In addition, while furthering spherical torus (ST) research, this device could achieve modest fusion power gain for short-pulse lengths, a significant step toward future large burning plasma devices now under discussion in the fusion community. The selected design point is Q=2 at HH=1.4, P subscript ''fusion''=60 MW, 5 second pulse, with R subscript ''0''=1.5 m, A=1.6, I subscript ''p''=10vMA, B subscript ''t''=2.6 T, CS flux=16 weber. Most of the research would be conducted in D-D, with a limited D-T campaign during the last years of the program

  10. A conceptual design of the International Thermonuclear Experimental Reactor for the Central Solenoid

    International Nuclear Information System (INIS)

    Heim, J.R.; Parker, J.M.

    1990-01-01

    Conceptual design of the International Thermonuclear Experimental Reactor (ITER) superconducting magnet system is nearing completion by the ITER Design Team, and one of the Central Solenoid (CS) designs is presented. The CS part of this magnet system will be a vertical stack of eight modules, approximately 16 m high, each having a approximate dimensions of: 4.1-m o.d., 2.8-m i.d., 1.9-m h. The peak field at the bore is approximately 13.5 T. Cable-in-conduit conductor with Nb 3 Sn composite wire will be used to wind the coils. The overall coil fabrication will use the insulate-wind-react-impregnate method. Coil modules will be fabricated using double-pancake coils with all splice joints located in the low-field region on the outside of the coils. All coils will be structurally graded with high-strength steel reinforcement which is co-wound with the conductor. We describe details of the CS coil design and analysis

  11. Thermonuclear land of plenty

    Science.gov (United States)

    Gasior, P.

    2014-11-01

    Since the process of energy production in the stars has been identified as the thermonuclear fusion, this mechanism has been proclaimed as a future, extremely modern, reliable and safe for sustaining energetic needs of the humankind. However, the idea itself was rather straightforward and the first attempts to harness thermonuclear reactions have been taken yet in 40s of the twentieth century, it quickly appeared that physical and technical problems of domesticating exotic high temperature medium known as plasma are far from being trivial. Though technical developments as lasers, superconductors or advanced semiconductor electronics and computers gave significant contribution for the development of the thermonuclear fusion reactors, for a very long time their efficient performance was out of reach of technology. Years of the scientific progress brought the conclusions that for the development of the thermonuclear power plants an enormous interdisciplinary effort is needed in many fields of science covering not only plasma physics but also material research, superconductors, lasers, advanced diagnostic systems (e.g. spectroscopy, interferometry, scattering techniques, etc.) with huge amounts of data to be processed, cryogenics, measurement-control systems, automatics, robotics, nanotechnology, etc. Due to the sophistication of the problems with plasma control and plasma material interactions only such a combination of the research effort can give a positive output which can assure the energy needs of our civilization. In this paper the problems of thermonuclear technology are briefly outlined and it is shown why this domain can be a broad field for the experts dealing with electronics, optoelectronics, programming and numerical simulations, who at first glance can have nothing common with the plasma or nuclear physics.

  12. Thermonuclear fusion

    International Nuclear Information System (INIS)

    Weisse, J.

    2000-01-01

    This document takes stock of the two ways of thermonuclear fusion research explored today: magnetic confinement fusion and inertial confinement fusion. The basic physical principles are recalled first: fundamental nuclear reactions, high temperatures, elementary properties of plasmas, ignition criterion, magnetic confinement (charged particle in a uniform magnetic field, confinement and Tokamak principle, heating of magnetized plasmas (ohmic, neutral particles, high frequency waves, other heating means), results obtained so far (scale laws and extrapolation of performances, tritium experiments, ITER project), inertial fusion (hot spot ignition, instabilities, results (Centurion-Halite program, laser experiments). The second part presents the fusion reactor and its associated technologies: principle (tritium production, heat source, neutron protection, tritium generation, materials), magnetic fusion (superconducting magnets, divertor (role, principle, realization), inertial fusion (energy vector, laser adaptation, particle beams, reaction chamber, stresses, chamber concepts (dry and wet walls, liquid walls), targets (fabrication, injection and pointing)). The third chapter concerns the socio-economic aspects of thermonuclear fusion: safety (normal operation and accidents, wastes), costs (costs structure and elementary comparison, ecological impact and external costs). (J.S.)

  13. International Thermonuclear Experimental Reactor: Physics issues, capabilities and physics program plans

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1997-01-01

    Present status and understanding of the principal plasma-performance determining physics issues that affect the physics design and operational capabilities of the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2 (International Atomic Energy Agency, Vienna, 1994)] are presented. Emphasis is placed on the five major physics-basis issues emdash energy confinement, beta limit, density limit, impurity dilution and radiation loss, and the feasibility of obtaining partial-detached divertor operation emdash that directly affect projections of ITER fusion power and burn duration performance. A summary of these projections is presented and the effect of uncertainties in the physics-basis issues is examined. ITER capabilities for experimental flexibility and plasma-performance optimization are also described, and how these capabilities may enter into the ITER physics program plan is discussed. copyright 1997 American Institute of Physics

  14. Safety considerations in next step fusion design and beyond

    International Nuclear Information System (INIS)

    Holland, D.F.

    1990-01-01

    Recent U.S. and international design studies provide insights into the potential safety and environmental advantages of fusion as well as the development needed to realize this potential. We in the Fusion Safety Program at EG ampersand G Idaho have analyzed the Compact Ignition Tokamak (CIT), the International Thermonuclear Engineering Reactor (ITER), and the Advanced Reactor Innovative Engineering Study (ARIES). I have reviewed these three designs to determine issues related to meeting the safety and the environmental goals that guide fusion development in the U.S. The paper lists safety and environmental issues that are generic to fusion and approaches to favorably resolve each issue. The technical developments that have the highest potential of contributing to improving the safety and environmental attractiveness of fusion are identified and discussed. These developments are in the areas of low-activation materials, plasma- facing components, and plasma physics relating to off-normal plasma events and tritium burn-up. 8 refs., 7 tabs

  15. FENIX [Fusion ENgineering International eXperimental]: A test facility for ITER [International Thermonuclear Experimental Reactor] and other new superconducting magnets

    International Nuclear Information System (INIS)

    Slack, D.S.; Patrick, R.E.; Miller, J.R.

    1990-01-01

    The Fusion ENgineering International eXperimental (FENIX) Test Facility which is nearing completion at Lawrence Livermore National Laboratory, is a 76-t set of superconducting magnets housed in a 4-m-diameter cryostat. It represents a significant step toward meeting the testing needs for the development of superconductors appropriate for large-scale magnet applications such as the International Thermonuclear Experimental Reactor (ITER). The magnet set is configured to allow radial access to the 0.4-m-diameter high-field region where maximum fields up to 14 T will be provided. The facility is fitted with a thermally isolated test well with a port to the high-field region that allows insertion and removal of test conductors without disturbing the cryogenic environment of the magnets. It is expected that the facility will be made available to magnet developers internationally, and this paper discusses its general design features, its construction, and its capabilities

  16. Thermonuclear fusion: Current status and future prospects

    International Nuclear Information System (INIS)

    Bruhns, H.; Maisonnier, Ch.

    1992-01-01

    Thermonuclear Fusion holds great promises for becoming an important energy source for the future. Fusion research and development is undertaken in al major countries of the world. The European Community pursues fusion in a large programme which embraces all R and D in the field of magnetic confinement fusion in the Member States, and to which Sweden and Switzerland are fully associated. The long-term objective of the programme is the joint creation of safe, environmentally sound prototype reactors. The main R and D line of the Community Fusion Programme is fusion by toroidal magnetic confinement on the basis of the Tokamak concept. Some related concepts are also studied which possibly could offer advantages for a reactor, and keep-in-touch activities exist for other approaches. Several small and medium sized specialised devices in Associated Laboratories have been built by the Community Fusion Programme as well as the Joint European Torus (JET Joint Undertaking) which is the largest and the most successful fusion device in the world. Recently, fusion power in the megawatt range has been achieved in JET. The long timescale and the large effort needed for the development of fusion as an energy source have been important elements to foster international collaboration. Engineering Design Activities for an International Thermonuclear Experimental Reactor (ITER) are undertaken, under the auspices of the IAEA, by the European Community, Japan, the Russian Federation and the United States of America. The objective of ITER is to achieve self-sustained thermonuclear burn and its control under long-pulse operation and to provide basic data for the engineering of a demonstration fusion reactor. (author)

  17. Controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    An outline is given of the present position of research into controlled fusion. After a brief reminder of the nuclear reactions of fusion and the principle of their use as a source of energy, the results obtained by the method of magnetic confinement are summarized. Among the many solutions that have been imagined and tried out to achieve a magnetic containing vessel capable of holding the thermonuclear plasma, the devices of the Tokamak type have a good lead and that is why they are described in greater detail. An idea is then given of the problems that arise when one intends conceiving the thermonuclear reactor based on the principle of the Tokamaks. The last section deals with fusion by lasers which is a new and most attractive alternative, at least from the viewpoint of basis physics. The report concludes with an indication of the stages to be passed through to reach production of energy on an industrial scale [fr

  18. Modelling of thermal and thermalhydraulic in a heat exchanger of a fusion thermonuclear reactor using 'GENEPI' computer code

    International Nuclear Information System (INIS)

    Langlais, Gilles

    1999-01-01

    The work presented in this report has been performed in the frame of fusion safety studies for thermonuclear reactors of ITER type (International Thermonuclear Experimental Reactor). It is particularly related to the thermal and two-phases thermalhydraulic studies of heat exchangers facing plasma. These components are submitted to unidirectional high heat flux between 1 to 10 MW/m 2 . The cooling fluid is then heat by an anisotropic heat flux. This non-uniform distribution induces the presence of different heat transfer on the cooling channel (single phase forced convection, subcooled nucleate boiling). The thermal and the thermalhydraulic three-dimensional study has been performed using experimental data and coupled computer calculations developed in the frame of this thesis work. The heat transfer between solid and fluid are modelled using correlations selected after the bibliography study. These heat exchange correlations as well as the CHF ones have been assessed by comparison to the available experimental data. This allowed to modify the single phase heat transfer correlation and to select two CHF correlations. (author) [fr

  19. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

  20. Inertial thermonuclear fusion by laser

    International Nuclear Information System (INIS)

    Watteau, J.P.

    1993-12-01

    The principles of deuterium tritium (DT) magnetic or inertial thermonuclear fusion are given. Even if results would be better with heavy ions beams, most of the results on fusion are obtained with laser beams. Technical and theoretical aspects of the laser fusion are presented with an extrapolation to the future fusion reactor. (A.B.). 34 refs., 17 figs

  1. Advance in physics of laser thermonuclear fusion

    International Nuclear Information System (INIS)

    Afanasev, J.; Basov, N.; Gamalij, J.; Krokhin, O.; Rozanov, V.

    1977-01-01

    A survey is given of current advance in the physics of laser thermonuclear fusion (LTF). The LTF physical model is discussed with regard to the optimal laser-target systems not only for attaining the physical limit but also for future thermonuclear reactors. The basic physical principles of LTF are formulated which make use of the fact that in focusing laser radiation on the surface of a substance a high density may be attained of the energy flux (10 5 to 10 6 J) and thereby also a high velocity of energy release in the substance. A detailed description is given of the processes which take place in laser irradiation of a spherical target. The problem is discussed of hydrodynamic stability in the compression of matter in laser thermonuclear targets, the concept is explained of the physical threshold of a thermonuclear reaction in laser excitation as are the conditions for attaining this threshold. The quantitative criterion is examined of the attainment of the physical threshold of LTF for pulsed systems. (B.S.)

  2. Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

    CERN Document Server

    2002-01-01

    Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

  3. Audit of United States portion of the International Thermonuclear Experimental Reactor project

    International Nuclear Information System (INIS)

    1993-01-01

    Worldwide efforts in fusion energy research are designed to develop fusion power as a safe, environmentally sound, and economically competitive source of energy. The International Thermonuclear Experimental Reactor (ITER) project is a worldwide effort to demonstrate the scientific and technological feasibility of fusion power. The European Community, Japan, the Russian Federation, and the United States are collaborating on ITER, with each of the four parties expected to equally share costs and benefits. Shared costs for the current engineering design phase of the project are estimated at $1 billion in 1989 dollars, excluding certain management and support costs to be absorbed by each partner, with an early estimate of $6 billion, also in 1989 dollars, for construction of the reactor. Engineering design formally began in July 1992, and this phase is in its formative stages. The US had already spent about $100 million since 1987 on ITER conceptual design activities and other preparatory activities in advance of the engineering design phase. Because of its cost significance, the importance of ITER to the US fusion energy program, and the project's unique aspects which may provide a framework for future international endeavors, we initiated an audit of the ITER project. The purpose of the audit was to evaluate management controls over the US portion of the ITER project. Our objectives was to determine whether key front-end controls were in place to ensure that the project could be managed in an efficient and effective manner

  4. Thermonuclear device

    International Nuclear Information System (INIS)

    Honda, Takuro; Maki, Koichi.

    1997-01-01

    The present invention provides a thermonuclear device, in which integrity of a measuring device is kept, the reactor wall temperature and wear of armour materials are monitored accurately even under intense radiation rays, so that the flow rate of coolants and plasma power can be controlled by using the signals. Infrared rays generated from the surface of the armour materials disposed on a first wall are detected to measure the reactor wall temperature. Coolant flow rate and plasma power are controlled based on the obtained reactor wall temperature. In addition, infrared rays generated from the back of the armour materials are detected to obtain the surface temperature in order to avoid intense radiation rays from plasmas. The coolant flow rate and the plasma power are controlled based on the obtained temperature on the surface of the reactor thereby controlling the temperature of the first wall and the armour material to 300degC or lower in a case of the first wall made of stainless steel and 1000degC or lower in a case of the armour material made of graphite. (I.S.)

  5. Controlled thermonuclear fusion: research on magnetic fusion

    International Nuclear Information System (INIS)

    Paris, P.J.

    1988-12-01

    Recent progress in thermonuclear fusion research indicates that the scientists' schedule for the demonstration of the scientific feasibility will be kept and that break-even will be attained in the course of the next decade. To see the implementation of ignition, however, the generation of future experiments must be awaited. These projects are currently under study. With technological research going on in parallel, they should at the same time contribute to the design of a reactor. Fusion reactors will be quite different from the fission nuclear reactors we know, and the waste of the plants will also be of a different nature. It is still too early to define the precise design of a fusion reactor. On the basis of a toric machine concept like that of the tokamak, we can, however, envisage that the problems with which we are confronted will be solved one after the other. As we have just seen, these will be the objectives of the future experimental installations where ignition will be possible and where the flux of fast neutrons will be so strong that they will allow the study of low-activation materials which will be used in the structure of the reactor. But this is also a task in which from now onwards numerous laboratories in Europe and in the world participate. The works are in fact punctiform, and often the mutual incidences can only be determined by an approach simulated by numerical codes. (author) 19 figs., 6 tabs., 8 refs

  6. New stainless steels of ferrite-martensite grade and perspectives of their application in thermonuclear facilities and fast reactors

    International Nuclear Information System (INIS)

    Ajtkhozhin, Eh.S.; Maksimkin, O.P.

    2007-01-01

    Review of scientific literature for last 5 years in which results on study of radiation effect on ferrite-martensite steels - construction materials of fast reactors and most probable candidates for first wall and blanket of the thermonuclear facilities ITER and Demo - are presented. Alongside with this a prior experimental data on study of microstructure changing and physical- mechanical properties of ferrite-martensite steel EhP-450 - the material of hexahedral case of spent assembly of BN-350 fast reactor- are cited. Principal attention was paid to considering of radiation effects of structural components content changing and ferrite-martensite steel swelling irradiated at comparatively low values of radiation damage climb rate

  7. NextSTEP Hybrid Life Support

    Data.gov (United States)

    National Aeronautics and Space Administration — NextSTEP Phase I Hybrid Life Support Systems (HLSS) effort assessed options, performance, and reliability for various mission scenarios using contractor-developed...

  8. Blue energy - The story of thermonuclear fusion energy

    International Nuclear Information System (INIS)

    Laval, G.

    2007-01-01

    The author has written a story of thermonuclear fusion as a future source of energy. This story began about 50 years ago and its last milestone has been the decision of building the ITER machine. This decision has been taken by an international collaboration including a large part of the humanity which shows how great are the expectations put on fusion and that fusion deserves confidence now. For long years fusion energy has been the subject of large controversy due to the questioning about the overcoming of huge theoretical and technological difficulties. Different machines have been built to assess new theoretical developments and to prepare the next step. The physics of hot plasmas has been understood little by little at the pace of the discovery of new instabilities taking place in fusion plasmas. The 2 unique today options: the tokamak-type machine and the laser-driven inertial confinement machine took the lead relatively quickly. (A.C.)

  9. Advanced Electric Propulsion NextSTEP BAA Activity

    Data.gov (United States)

    National Aeronautics and Space Administration — The goal of the AES Advanced Electric Propulsion Next Space Technologies for Exploration Partnerships (NextSTEP) Broad Agency Announcement (BAA) activity is to...

  10. Study on structural materials used in thermonuclear fusion technology

    International Nuclear Information System (INIS)

    Billa, R.; Amaral, D.

    1995-01-01

    The main problem related to the construction of a thermonuclear fusion reactor is the absence of suitable materials for the process, concerning to temperature limits, heat flux and life time. The first wall is the most critical part of the structure, being submitted to radiation effects, ionic corrosion and coolant, besides thermal fatigue and tension produced by cyclical burning. The AISI 316(17-12SPH) stainless steel is used as structural material, which has a wide known database. This work proposes an alternative material study to be used in the future thermonuclear fusion reactors. As a option a study on the utilization of Cr-Mn(Fe-17 Mn-10 Cr-0,1 C) steels and their alloy variations is presented

  11. The FREYA project at VENUS-F - the next step towards MYRRHA

    Energy Technology Data Exchange (ETDEWEB)

    Krasa, A.; Baeten, P.; Kochetkov, A.; Uyttenhove, W.; Vittiglio, G.; Wagemans, J. [SCK.CEN, Boeretang 200, 2400 Mol (Belgium); Mercatali, L. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany); Bianchini, G.; Carta, M.; Fabrizio, V.; Peluso, V. [ENEA, C.R. Casaccia, via Anguillarese, 301-00060 S. Maria di Galeria (Italy)

    2015-07-01

    The VENUS-F reactor is a fast, zero-power research reactor. Since 2009 it has used 30% wt. metallic uranium as fuel and solid lead as reflector and coolant simulator. It can also be coupled to a particle accelerator with a neutron producing target and make up a subcritical Accelerator Driven System (ADS). Within the currently ongoing FREYA project, the main tasks being investigated are the validation of online reactivity monitoring of an ADS, the validation of nuclear data and neutronic codes, and the experimental characterization of fast critical and subcritical cores representative for MYRRHA. MYRRHA will be an ADS demonstrator with the option to be operated in critical mode too. So far the only critical core at VENUS-F was assembled using 97 fuel assemblies, each of them consisting of 9 U and 16 Pb rodlets. The core was thoroughly characterized: flux traverses, spectral indices, control rod worth, delayed neutron parameters were measured and compared with Monte Carlo and deterministic calculations. A sensitivity study was also performed focusing on the influence of material impurities, positioning uncertainties, structures and detectors around the core. It was a simple core without any perturbations in the active zone. In the next critical cores that are planned to be investigated in 2015 and 2016, additional elements will be introduced in order to become more representative for MYRRHA. As MYRRHA will use MOX fuel, oxygen will be added into the following core. This will be done by introducing Al{sub 2}O{sub 3} into the fuel assemblies. New fuel assemblies have been already made and are ready to be loaded. Each consists of 13 U, 8 Pb, and 4 Al{sub 2}O{sub 3} rodlets. The new critical core will consist of approximately 40 such fuel assemblies and will be loaded in January 2015. The first results of the characterization measurements will be presented. Additionally, it is planned to investigate coolant void effects. As MYRRHA will be a rather complex system using a

  12. Tritium Issues in Next Step Devices

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; G. Federici

    2001-09-05

    Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

  13. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    C.H. Skinner; G. Federici

    2001-01-01

    Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

  14. Plasma-materials interaction issues for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Cohen, S.A.; Werley, K.A.

    1992-02-01

    Analysis of proposed operating scenarios for the International Thermonuclear Experimental Reactor has yielded predictions for the power and particle fluxes onto the material surfaces facing the plasma. The particles, mostly deuterium, tritium, and helium ions, would have energies in the range of 50--2000 eV and fluxes up to 5 x 10 23 /m 2 s. Lower fluxes of multi-MeV electrons and alpha particles may also strike the plasma-facing surfaces, primarily during transient events. The peak power fluxes onto the plasma-facing surfaces during normal operation are expected to be 5--100 MW/m 2 , but much higher during transient events. At the extreme conditions expected for steady-state operation, commonly used heat-removal structures are unable to withstand either the high sputter erosion rates or power loads. To reduce the time-averaged power flux, active control of the plasma position is specified to sweep the plasma heat load across larger areas of plasma-facing components. However, the cyclic heat load creates fatigue lifetime problems. Solutions to these lifetime and reliability problems by (1) changes in machine design and operation, (2) redeposition mechanisms, and (3) changes in materials, will be discussed. A proposed accelerated-life test facility for prototype divertor plate development is described

  15. Research report on the users' needs for next research reactor

    International Nuclear Information System (INIS)

    Takahashi, Hiroyuki; Tamura, Itaru; Hosoya, Toshiaki; Horiguchi, Hironori

    2015-03-01

    JRR-3 has been operated for more than 25 years for that it is time to investigate the role of a next research reactor. A task force under the Committee for Promotion of JRR-3 Neutron Beam Application has been organized by Department of Research Reactor and Tandem Accelerator to survey neutron beam application trends in the future. This is a report on the survey results and users' requirements for the next research reactor have been summarized in this report carried by the task force. (author)

  16. Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems

    International Nuclear Information System (INIS)

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

    1988-01-01

    The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The TF winding pack provides support against in-plane separating loads but offers little support against out-of-plane loads, unless shear-bonding of the conductors can be maintained. The removal of heat due to nuclear and ac loads has not been a fundamental limit to design, but certainly has non-negligible economic consequences. We present here preliminary ITER magnetic systems design parameters taken from trade studies, designs, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit. The work presented here reflects the efforts of many, but the responsibility for the opinions expressed is the authors'. 4 refs., 3 figs., 4 tabs

  17. Tandem mirror next step: remote maintenance

    International Nuclear Information System (INIS)

    Doggett, J.N.; Damm, C.C.; Hanson, C.L.

    1980-01-01

    This study of the next proposed experiment in the Mirror Fusion Program, the Tandem Mirror Next Step (TMNS), has included serious consideration of the maintenance requirements of such a large source of high energy neutrons with its attendant throughput of tritium. Although maintenance will be costly in time and money, our conclusion is that with careful attention to a design for maintenance plan such a device can be reliably operated

  18. Reactor structure and superconducting magnet system of ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

    1993-01-01

    Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

  19. Design considerations for ITER [International Thermonuclear Experimental Reactor] toroidal field coils

    International Nuclear Information System (INIS)

    Kalsi, S.S.; Lousteau, D.C.; Miller, J.R.

    1987-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Europe, Japan, the Union of Soviet Socialist Republics (USSR), and the United States. This paper describes a magnetic and mechanical design methodology for toroidal field (TF) coils that employs Nb/sub 3/Sn superconductor technology. Coil winding is sized by using conductor concepts developed for the US TIBER concept. The nuclear heating generated during operation is removed from the windings by helium flowing through the conductor. The heat in the coil case is removed through a separate cooling circuit operating at approximately 20 K. Manifold concepts are presented for the complete coil cooling system. Also included are concepts for the coil structural arrangement. The effects of in-plane and out-of-plane loads are included in the design considerations for the windings and case. Concepts are presented for reacting these loads with a minimum amount of additional structural material. Concepts discussed in this paper could be considered for the ITER TF coils. 6 refs., 5 figs., 1 tab

  20. Next-Step Spherical Torus Experiment and Spherical Torus Strategy in the Fusion Energy Development Path

    International Nuclear Information System (INIS)

    Ono, M.; Peng, M.; Kessel, C.; Neumeyer, C.; Schmidt, J.; Chrzanowski, J.; Darrow, D.; Grisham, L.; Heitzenroeder, P.; Jarboe, T.; Jun, C.; Kaye, S.; Menard, J.; Raman, R.; Stevenson, T.; Viola, M.; Wilson, J.; Woolley, R.; Zatz, I.

    2003-01-01

    A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction

  1. MHD equilibrium methods for ITER [International Thermonuclear Experimental Reactor] PF [poloidal field] coil design and systems analysis

    International Nuclear Information System (INIS)

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

    1989-03-01

    Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs

  2. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  3. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  4. Structural characteristics of proposed ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coil conductor

    International Nuclear Information System (INIS)

    Gibson, C.R.; Miller, J.R.

    1988-01-01

    This paper analyzes the effect of transverse loading on a cable-in-conduit conductor which has been proposed for the toroidal field coils of the International Thermonuclear Experimental Reactor. The primary components of this conductor are a loose cable of superconducting wires, a thin-wall tube for helium containment, and a U-shaped structural channel. A method is given where the geometry of this conductor can be optimized for a given set of operating conditions. It is shown, using finite-element modeling, that the structural channel is effective in supporting loads due to transverse forces and internal pressure. In addition, it is shown that the superconducting cable is effectively shielded from external transverse loads that might otherwise degrade its current carrying capacity. 10 refs., 10 figs., 3 tabs

  5. Device for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Yutaro; Kawarazaki, Yuki; Sugiyama, Yu.

    1996-01-01

    A member comprising hydrogen occluding materials is introduced to a reactor incorporated with U-235 as fuels in order to moderate and breed fast neutrons and to control the reactor. Since the amount of light hydrogen or heavy hydrogen is substantially the same as that of metal, etc. of hydrogen occluding material, a moderating efficiency substantially equal with that of a moderator comprising H 2 O can be obtained. In addition, since the member acting as a moderator has an effect of multiplying neutrons, use of only natural uranium 0.72% as nuclear fuels causes chain reaction to provide a function as a nuclear reactor. Further, the hydrogen occluding material can be used also as a control rod for controlling the reactor. The hydrogen occluding material may be Ti, Zr, Pd, proton conductor, Ag, Pt, Rh or oxides thereof or alloys thereof. The member comprising hydrogen occluding materials is preferably coated with a material not permeating hydrogen. (N.H.)

  6. Thermonuclear detonation

    International Nuclear Information System (INIS)

    Feoktistov, L.P.

    1998-01-01

    The characteristics of, and energy transfer mechanisms involved in, thermonuclear detonation are discussed. What makes the fundamental difference between thermonuclear and chemical detonation is that the former has a high specific energy release and can therefore be employed for preliminary compressing the thermonuclear mixture ahead of the burning wave. Consequently, with moderate (mega joule) initiation energies, a steady-state detonation laboratory experiment with unlimited energy multiplication becomes a possibility

  7. Proposal for a decision of the Council concerning the planning of a research- and education-program (1982-1986) on the field of thermonuclear fusion

    International Nuclear Information System (INIS)

    The thermonuclear fusion is in an early development state and has however in principle possible advantages which could be especially valuable for Europe. The primary fusion fuels (D, Li) are plentiful existent, wide spread and cheap (1 g natural Lithium could generate 15 MHW); both fuels and the end product of the reactions - Helium - are stable. From the nuclear-technological point of view a thermonuclear reactor could be built with high safety; the doubling time for breeding of new fuels in principle could be very short. These potential advantages however are balanced by certain disadvantages, e.g. high costs for the construction of a thermonuclear reactor etc. The research program, other possibilities and the costs are outlined. (orig./HT) [de

  8. Thermonuclear detonation

    International Nuclear Information System (INIS)

    Feoktistov, L P

    1998-01-01

    The characteristics of, and energy transfer mechanisms involved in, thermonuclear detonation are discussed. What makes the fundamental difference between thermonuclear and chemical detonation is that the former has a high specific energy release and can therefore be employed for preliminarily compressing the thermonuclear mixture ahead of the burning wave. Consequently, with moderate (megajoule) initiation energies, a steady-state detonation laboratory experiment with unlimited energy multiplication becomes a possibility. (from the history of physics)

  9. Nuclear reactor technology: the next 50 years

    International Nuclear Information System (INIS)

    Sollychin, R.; Subki, H.; Adelfang, P.; Koshy, T.

    2013-01-01

    hybrid energy systems therefore have the potential to be a key solution to meet the energy needs of the emerging economies in the next few decades. In order to be operated as part of distributed hybrid energy systems, SMRs should be available in various power rating and be able to be engineered for multiple-applications. They should also be user-friendly, and able to be integrated easily with renewable energy systems as well as, if necessary, with adjacent energy application/storage systems. With these considerations, a number of special design requirements for SMR have been identified. The multiple-applications of the SMR may include provision of neutron utilization for industrial applications and research in the future, depending upon the needs in the region where they are built. The difference between a research reactor and such SMR may therefore become indistinct. In any case, operating and utilizing research reactor and small and/or prototype SMR can be taken as a first step in the preparation for a larger nuclear power program, as both the small nuclear systems and the large nuclear power systems require personnel with similar skills sets and the support of the same/similar infrastructure including regulatory body. A strive-for-excellence culture normally required for the operation of a large nuclear system should therefore be cultivated starting with the deployment of the small nuclear systems including the research reactors. Organization well-embedded with such culture put a strong emphasis on safety, health and environmental protection. They are disciplined and well-motivated to perform work systematically, according to plans and established procedures. They communicate effectively within the organization and with stakeholders, and have a continuous drive to improve quality. These considerations should be included in the design of SMRs and research reactors to be deployed in the next 50 years. (author)

  10. Thermonuclear device

    International Nuclear Information System (INIS)

    Tezuka, Masaru.

    1993-01-01

    Protrusions and recesses are formed to a vacuum vessel and toroidal magnetic coils, and they are engaged. Since the vacuum vessel is generally supported firmly by a rack or the like by support legs, the toroidal magnetic field coils can be certainly supported against tumbling force. Then, there can be attained strong supports for the toroidal magnetic field coils, in addition to support by wedges on the side of inboard and support by share panels on the side of outboard, capable of withstanding great electromagnetic forces which may occur in large-scaled next-generation devices. That is, toroidal magnetic field coils excellent from a view point of deformation and stress can be obtained, to provide a thermonuclear device of higher reliability. (N.H.)

  11. Dust processing device for inside of vacuum vessel of thermonuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Atsushi; Tsujimura, Seiichi; Takahashi, Kenji; Ueda, Yasutoshi; Kuwata, Masayasu; Onozuka, Masaki

    1995-05-02

    The device of the present invention can occasionally recover dusts in a vacuum vessel of a thermonuclear reactor. In addition, fine powdery dusts are never scattered to the vacuum vessel. Namely, a processing device main body comprises a locally sealed space in the vacuum vessel. A blow-up device blows up and floats dusts accumulated in the vacuum vessel to the processing device main body. A discharge plate electrically charges the floating dusts by discharge. An electrode collects the charged dusts. Collected dusts are recovered together with a pressurized gas through a dust recovering port to the outside of the processing device. With such a constitution, it is not necessary to release the vacuum vessel to the atmosphere and evacuate after the completion of the collection of the dusts on every time when the dusts are generated as in the prior art. It is no more necessary for an operator to enter into the vacuum vessel and recover the dusts. Since fine powdery dusts are never scattered in the vacuum vessel, no undesired effects are given to exhaustion facilities and instruments of the vacuum vessel. (I.S.).

  12. Dust processing device for inside of vacuum vessel of thermonuclear reactor

    International Nuclear Information System (INIS)

    Okumura, Atsushi; Tsujimura, Seiichi; Takahashi, Kenji; Ueda, Yasutoshi; Kuwata, Masayasu; Onozuka, Masaki.

    1995-01-01

    The device of the present invention can occasionally recover dusts in a vacuum vessel of a thermonuclear reactor. In addition, fine powdery dusts are never scattered to the vacuum vessel. Namely, a processing device main body comprises a locally sealed space in the vacuum vessel. A blow-up device blows up and floats dusts accumulated in the vacuum vessel to the processing device main body. A discharge plate electrically charges the floating dusts by discharge. An electrode collects the charged dusts. Collected dusts are recovered together with a pressurized gas through a dust recovering port to the outside of the processing device. With such a constitution, it is not necessary to release the vacuum vessel to the atmosphere and evacuate after the completion of the collection of the dusts on every time when the dusts are generated as in the prior art. It is no more necessary for an operator to enter into the vacuum vessel and recover the dusts. Since fine powdery dusts are never scattered in the vacuum vessel, no undesired effects are given to exhaustion facilities and instruments of the vacuum vessel. (I.S.)

  13. Cleaning and air conditioning device for atmosphere in thermonuclear reactor chamber

    International Nuclear Information System (INIS)

    Ishida, Seiji.

    1993-01-01

    The device of the present invention removes tritium efficiently and attains ventilation and conditioning of a great amount of air flow. That is, there are disposed a humidity separator, a filter, a heater, a catalyst filled layer, a water jetting type humidifying heat insulation cooler and a cooler in this order from an inlet side (upstream) of contaminated room atmospheric gases. The catalyst filled layer, etc. are incorporated integrally into the ventilation air conditioning facility for ventilating air in the chamber of the thermonuclear reactor, to clean a tritium atmosphere at the same time. Accordingly, the device is made compact as a whole. A limit for the air flow rate owing to the use of the conventional catalyst tower and adsorbing tower is eliminated. Then a ventilating air conditioning for a great flow rate can be attained. Tritium is removed by cooling and dehumidification without using any adsorbent. Accordingly, an adsorbing tower is no more necessary and conventional regeneration operation is not required. As a result, space for installation is reduced, the system is simplified and the cost for construction and facility can be reduced. (I.S.)

  14. The CTB Treaty and next steps towards nuclear disarmament

    International Nuclear Information System (INIS)

    Epstein, W.

    1998-01-01

    The achievement of a CTBT is of course the indispensable first step in any program of nuclear disarmament. It will not be easy for Conference on Disarmament to agree by consensus to establish the committee on nuclear disarmament and will be even more difficult for it to agree on its mandate or terms of reference. In the meantime the nuclear powers should seek to implement a number of steps towards nuclear disarmament which thy can argue and implement by their own accord. The next steps are described in this report

  15. Processing of W-Cu functionally graded materials (FGM) through the powder metallurgy route: application as plasma facing components for ITER-like thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Raharijaona, J.J.

    2009-11-01

    The aim of this study was to study and optimize the sintering of W-Cu graded composition materials, for first wall of ITER-like thermonuclear reactor application. The graded composition in the material generates graded functional properties (Functionally Graded Materials - FGM). Rough thermomechanical calculations have shown the interest of W-Cu FGM to improve the lifetime of Plasma Facing Components (PFC). To process W-Cu FGM, powder metallurgy route was analyzed and optimized from W-CuO powder mixtures. The influence of oxide reduction on the sintering of powder mixtures was highlighted. An optimal heating treatment under He/H 2 atmosphere was determined. The sintering mechanisms were deduced from the analysis of the effect of the Cu-content. Sintering of W-Cu materials with a graded composition and grain size has revealed two liquid migration steps: i) capillary migration, after the Cu-melting and, ii) expulsion of liquid, at the end of sintering, from the dense part to the porous part, due to the continuation of W-skeleton sintering. These two steps were confirmed by a model based on capillary pressure calculation. In addition, thermal conductivity measurements were conducted on sintered parts and showed values which gradually increase with the Cu-content. Hardness tests on a polished cross-section in the bulk are consistent with the composition profiles obtained and the differential grain size. (author)

  16. Manufacturing device for vacuum vessel of thermonuclear reactor and manufacturing method therefor

    International Nuclear Information System (INIS)

    Yanagi, Hiroshi; Shibui, Masanao; Uchida, Takaho

    1998-01-01

    The present invention provides a method of manufacturing a vacuum vessel of a thermonuclear reactor with no welding deformation. Namely, there are disposed a manufacturing device comprises a welding machine equipped with a plurality of welding torches which can conduct synchronizing welding and a torch positioning mechanism for positioning the plurality of welding torches each at an optional distance. Then, both ends of a splice plate can be welded by the plurality of welding torches under synchronization. Accordingly, joining portions of sectors of a vacuum vessel can be welded in the site with no deviation of beveling at joining portions between an outer wall and an inner wall with the splice plate due to welding deformation. In addition, the welding machine is mounted on a travelling type clamping mechanism stand or a travelling type clamping mechanism. With such a constitution, since the peripheries of the joining portions on the inner wall are clamped with each other by the travelling type clamping mechanism, no angular distortion is caused in any welded portion of the outer wall. (I.S.)

  17. Evaluation of innovative means of hydrogen risk mitigation in thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Maruejouls, C.

    2003-01-01

    One of the main accidents in ITER-type thermonuclear fusion reactors is the loss of coolant leading to hydrogen production. Within the framework of the studies on the ITER fusion reactor, a mitigation strategy for this risk must be devised by focusing on a system, which can be placed near the hydrogen source. The uncertainty as to the air content during such a scenario forbids the use of classic methods based on the hydrogen/oxygen reaction such as passive catalytic recombiners. Former studies have proposed a process based on the reduction of metallic oxides and more particularly that of the manganese dioxide enhanced by silver oxide mixture. The reaction studied is H 2 + MnO 2 → MnO + H 2 O (reaction enhanced by Ag 2 O). The purpose is to study the kinetic. The method used consists in comparing the experimental results obtained on the pilot facility CIGNE with those provided by a model. The experimental results were obtained from tests made on a pilot facility with a solid/gas reaction in a fixed bed. These underlined the importance of favoring the solid/gas contact surface. The modeling used in the MITRHY simulation program, coupled to an optimizer helped determine the kinetic parameters and the data on the material and temperature transfers. The kinetic is of first order rate for hydrogen with an activation energy of 29428 J/mol and a kinetic coefficient of 142 m.s -1 . Integrated in the MITRHY program, the kinetic parameters were used to simulate the hydrogen elimination in the accident conditions on the ITER experimental reactor. This study achieved a pre-design basis of the device (bed of about 30 cm with grains of a diameter of less than 5 mm) to be implemented. It also underlined the need to favor the specific surface to improved process efficiency. (author)

  18. Proposal for a decision of the EC Council concerning the planning of a research- and education-program (1982-1986) on the field of controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The thermonuclear fusion is in a early development state and has, however, in principle possible advantages which could be especially valuable for Europe: the primary fusion fuels (D, Li) are plentiful existent, wide spread and cheap (1 g natural Lithium could generate 15 MWh); both fuels and the end product of the reactions - Helium - are stable. From the nuclear-technological point of view a thermonuclear reactor could be built with high safety; the doubling time for breeding of new fuels in principle could be very short. These potential advantages, however, are balanced by certain disadvantages, e.g. high costs for the construction of a thermonuclear reactor etc. The research program, other possibilities and the costs are outlined. (orig./HT) [de

  19. Tritium inventory and recovery in next-step fusion devices

    International Nuclear Information System (INIS)

    Causey, R.A.; Brooks, J.N.; Federici, G.

    2002-01-01

    Future fusion devices will use tritium and deuterium fuel. Because tritium is both radioactive and expensive, it is absolutely necessary that there be an understanding of the tritium retention characteristics of the materials used in these devices as well as how to recover the tritium. There are three materials that are strong candidates for plasma-facing-material use in next-step fusion devices. These are beryllium, tungsten, and carbon. While beryllium has the disadvantage of high sputtering and low melting point (which limits its power handling capabilities in divertor areas), it has the advantages of being a low-Z material with a good thermal conductivity and the ability to get oxygen from the plasma. Due to beryllium's very low solubility for hydrogen, implantation of beryllium with deuterium and tritium results in a saturated layer in the very near-surface with limited inventory (J. Nucl. Mater. 273 (1999) 1). Unfortunately, there are nuclear reactions generated by neutrons that will breed tritium and helium in the material bulk (J. Nucl. Mater. 179 (1991) 329). This process will lead to a substantial tritium inventory in the bulk of the beryllium after long-term neutron exposure (i.e. well beyond the operation life time of a next-step reactor like ITER). Tungsten is a high-Z material that will be used in the divertor region of next-step devices (e.g. ITER) and possibly as a first wall material in later devices. The divertor is the preferred location for tungsten use because net erosion is very low there due to low sputtering and high redeposition. While experiments are still continuing on tritium retention in tungsten, present data suggest that relatively low tritium inventories will result with this material (J. Nucl. Mater. 290-293 (2001) 505). For tritium inventories, carbon is the problem material. Neutron damage to the graphite can result in substantial bulk tritium retention (J. Nucl. Mater. 191-194 (1992) 368), and codeposition of the sputtered carbon

  20. Next Steps in Signaling (NSIS): Framework

    NARCIS (Netherlands)

    Hancock, R.; Loughney, J.; van den Bosch, S.; Hancock, R.; Karagiannis, Georgios; Loughney, J.; van den Bosch, S.

    The Next Steps in Signaling (NSIS) working group is considering protocols for signaling information about a data flow along its path in the network. The NSIS suite of protocols is envisioned to support various signaling applications that need to install and/or manipulate such state in the network.

  1. Overview of the European Fusion Programme

    International Nuclear Information System (INIS)

    Maisonnier, C.; Toschi, R.

    1989-01-01

    An overview of the European Fusion Programme is given and its near-term and long-term strategies are outlined. With the long-term energy problem worldwide as background, the role of thermonuclear fusion research is discussed in the context of energy sources having the potential to supply a substantial fraction of the electrical energy needs in the future. The European Fusion Programme, which is designed to lead in due course to the joint construction of prototypes with a view to their industrial production and marketing, is implemented by a sliding programme concept, i.e. through five-year programmes which overlap for about two years. The main objectives of the proposed 1987-1991 programme are outlined, with emphasis on the role of the Next Step (a Next European Torus or an International Thermonuclear Experimental Reactor), of the JET Joint Undertaking, of the Associated Laboratories, and of the European industry; and on the importance of international cooperation which has been established by bilateral framework agreements on fusion, by several multilateral implementing agreements in the frame of the IEA (OECD), and by the quadripartite cooperation of EURATOM, Japan, USA and USSR in the conceptual design of an International Thermonuclear Experimental Reactor under the auspices of the IAEA. (orig.)

  2. Economic impacts on the United States of siting decisions for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Peerenboom, J.P.; Wolsko, T.D.; Hanson, M.E.

    1997-01-01

    This paper presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively

  3. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R.

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively

  4. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R. [and others

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  5. Towards diagnostics for a fusion reactor

    International Nuclear Information System (INIS)

    Costley, A. E.

    2009-01-01

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

  6. Next-Step scientific objectives, targets, and parameters for reversed-field-pinch (RFP) magnetic fusion energy (MFE) systems: Preliminary thoughts

    International Nuclear Information System (INIS)

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

    1993-01-01

    The purpose of this document is the quantitative definition of objectives, targets, and parameters of the Next-Step device to follow the present RFX experiment; this device is given the name RFXNS. Although developed over five years ago, much of the material distilled into the 1988 RFP tactical plan is useful in establishing the goals and parameters of RFXNS. This earlier plan established tentative parameters of an RFP next step based on: predictions of RFP ignition and commercial-reactor devices; and the assumed successful operation of highly complementary RFP experiments RFX and ZTH/CPRF. Programmatic changes and evolution that have occurred since 1988 strongly impact the role and characteristics of an RFXNS: the Los Alamos ZTH/CPRF project and fusion program was terminated in mid-construction for reasons of MFE cost savings and concept focusing; great progress has been made in launching ITER; and reactor projections for the tokamak have increased in detail and variety, but not in commercial promise and competitiveness. A brief status of and perspective from each of the above three points is necessary before the key issues and their implementation to form the basis of the RFXNS definition are given

  7. Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems: Revision 1

    International Nuclear Information System (INIS)

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

    1988-01-01

    The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The TF winding pack provides support against in-plane separating loads but offers little support against out-of-plane loads, unless shear-bonding of the conductors can be maintained. The removal of heat due to nuclear and ac loads has not been a fundamental limit to design, but certainly has non-negligible economic consequences. We present here preliminary ITER magnet systems design parameters taken from trade studies, designs, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit. The work presented here reflects the efforts of many, but the responsibility for the opinions expressed is the authors'. 4 refs., 3 figs., 4 tabs

  8. Recommendations for a cryogenic system for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Slack, D.S.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Soviet Union, and the United States. ITER will be a large machine requiring up to 100 kW of refrigeration at 4.5 K to cool its superconducting magnets. Unlike earlier fusion experiments, the ITER cryogenic system must handle pulse loads constituting a large percentage of the total load. These come from neutron heating during a fusion burn and from ac losses during ramping of current in the PF (poloidal field) coils. This paper presents a conceptual design for a cryogenic system that meets ITER requirements. It describes a system with the following features: Only time-proven components are used. The system obtains a high efficiency without use of cold pumps or other developmental components. High reliability is achieved by paralleling compressors and expanders and by using adequate isolation valving. The problem of load fluctuations is solved by a simple load-leveling device. The cryogenic system can be housed in a separate building located at a considerable distance from the ITER core, if desired. The paper also summarizes physical plant size, cost estimates, and means of handling vented helium during magnet quench. 4 refs., 4 figs., 3 tabs

  9. Multi-User Interactive TV: the Next Step in Personalization

    NARCIS (Netherlands)

    van Brandenburg, Ray; van Deventer, M. Oskar; Karagiannis, Georgios; Schenk, Mike

    2010-01-01

    In the past few years there has been an increasing trend towards personalization in the TV world. IMS-based IPTV is a good example of a highly personalized IPTV architecture, featuring an advanced identity management subsystem. This article studies a next step in the personalization of the

  10. Safety criteria for the next generation of European reactors

    International Nuclear Information System (INIS)

    Dominguez Bautista, M.T.

    1995-01-01

    For the next generation of reactors, European companies operating in the electricity sector have drawn up a document called European Utilities Requirement (EUR), which sets out the requirements to be met by the designers of future reactors. The main objective of these new requirements is to increase the safety in existing reactors, making good use of operating experience available and the technological developments of the last decade. This paper offers an in-depth analysis of the most significant characteristics, describing how the EUR requirements have been prepared and how they are being implemented by the designers. Areas covered are: - Combining deterministic and probabilistic criteria - Automation of control systems - Design extension for severe accidents - Containment design - Emergency plans - Autonomy versus manual operation

  11. Radiation damages of material surfaces by plasma emission in thermonuclear devices. Methods of study of surface phenomena and simulation effect of thermonuclear plasma

    International Nuclear Information System (INIS)

    Rybalko, V.F.

    1978-01-01

    Phenomena that can introduce a controlling contribution into the erosion of the first wall surface in thermonuclear reactor are reviewed. Considered are the main characteristics of the physical disintegration: dependence of the disintegration coefficient upon the energy and the incidence angle of the bombarding particles, upon the atomic number of the material of the target and the type of bombarding particles. Stressed is the lack of reliable data on the disintegration of materials by light ions, which are of a maximum interest in relation to the controlled thermonuclear synthesis. The chemical disintegration and some regularities of it for the carbon-hydrogen and carbon-oxygen systems are discussed briefly. Listed are the main properties of blistering and its contribution to the erosion of crystalline surfaces

  12. The role of materials in controlled thermonuclear research

    Energy Technology Data Exchange (ETDEWEB)

    Craston, J L; Hancox, R; Robson, A E [U.K. Atomic Energy Authority, AERE, Harwell (United Kingdom); Kaufman, S; Miles, H T; Ware, A A; Wesson, J A [AEI Research Laboratory, Aldermaston (United Kingdom)

    1958-07-01

    It is the purpose of this paper to examine the processes occurring at the wall and to discuss their importance in the choice of materials both for present equipment and for future designs. The emphasis is laid primarily on plasma contamination but other effects are considered, such as thermal stress fatigue and radiation damage of the wall. The principal problems associated with the choice of wall material for a high current discharge tube have been discussed, both under the conditions which exist in present systems and under the conditions which are anticipated in a thermonuclear reactor.

  13. Nuclear reactor technology: the next 50 years

    Energy Technology Data Exchange (ETDEWEB)

    Sollychin, R.; Subki, H.; Adelfang, P.; Koshy, T. [International Atomic Energy Agency, Vienna (Austria)

    2013-07-01

    hybrid energy systems therefore have the potential to be a key solution to meet the energy needs of the emerging economies in the next few decades. In order to be operated as part of distributed hybrid energy systems, SMRs should be available in various power rating and be able to be engineered for multiple-applications. They should also be user-friendly, and able to be integrated easily with renewable energy systems as well as, if necessary, with adjacent energy application/storage systems. With these considerations, a number of special design requirements for SMR have been identified. The multiple-applications of the SMR may include provision of neutron utilization for industrial applications and research in the future, depending upon the needs in the region where they are built. The difference between a research reactor and such SMR may therefore become indistinct. In any case, operating and utilizing research reactor and small and/or prototype SMR can be taken as a first step in the preparation for a larger nuclear power program, as both the small nuclear systems and the large nuclear power systems require personnel with similar skills sets and the support of the same/similar infrastructure including regulatory body. A strive-for-excellence culture normally required for the operation of a large nuclear system should therefore be cultivated starting with the deployment of the small nuclear systems including the research reactors. Organization well-embedded with such culture put a strong emphasis on safety, health and environmental protection. They are disciplined and well-motivated to perform work systematically, according to plans and established procedures. They communicate effectively within the organization and with stakeholders, and have a continuous drive to improve quality. These considerations should be included in the design of SMRs and research reactors to be deployed in the next 50 years. (author)

  14. Transient temperature variations during the self-heating of a plasma by thermonuclear reactions

    Energy Technology Data Exchange (ETDEWEB)

    Greyber, Howard D [University of California Radiation Laboratory, Livermore, CA (United States)

    1958-07-01

    The motivation for this work arose from an observation by Rosenbluth that in a different but related physical situation, the electron temperature) could exceed ion temperature, during transient heating. We have undertaken to trace the transient temperatures to be expected in an idealized physical situation that still bears some resemblance to what one envisions for the Controlled Thermonuclear Reactor.

  15. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  16. Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

    International Nuclear Information System (INIS)

    Elsheikh, B.M.

    2017-01-01

    Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

  17. Ignition in the next step tokamak

    International Nuclear Information System (INIS)

    Johner, J.

    1990-07-01

    A 1/2-D model for thermal equilibrium of a thermonuclear plasma with transport described by an empirical global energy confinement time is described. Ignition in NET and ITER is studied for a number of energy confinement time scaling expressions. Ignited operation of these machines at the design value of the neutron wall load is shown to satisfy both beta and density constraints. The value of the confinement time enhancement factor required for such operation is found to be lower for the more recently proposed scaling expressions than it is for the oldest ones. With such new scalings, ignition could be obtained in H-mode in NET and ITER even with relatively flat density profiles

  18. Breeding blankets for thermonuclear reactors

    International Nuclear Information System (INIS)

    Rocaboy, Alain.

    1982-06-01

    Materials with structures suitable for this purpose are studied. A bibliographic review of the main solid and liquid lithiated compounds is then presented. Erosion, dimensioning and maintenance problems associated with the limiter and the first wall of the reactor are studied from the point of view of the constraints they impose on the design of the blankets. Detailed studies of the main solid and liquid blanket concepts enable the best technological compromises to be determined for the indispensable functions of the blanket to be assured under acceptable conditions. Our analysis leads to four classes of solution, which cannot at this stage be considered as final recommendations, but which indicate what sort of solutions it is worthwhile exploring and comparing in order to be in a position to suggest a realistic blanket at the time when plasma control is sufficiently good for power reactors to be envisaged. Some considerations on the general architecture of the reactor are indicated. Energy storage with pulsed reactors is discussed in the appendix, and a first approach made to minimizing the total tritium recovery [fr

  19. Deuterides of light elements: low-temperature thermonuclear burn-up and applications to thermonuclear fusion problems

    International Nuclear Information System (INIS)

    Frolov, A.M.; Smith, V.H.; Smith, G.T.

    2002-01-01

    Thermonuclear burn-up and thermonuclear applications are discussed for a number of deuterides and DT hydrides of light elements. These deuterides and corresponding DT hydrides are often used as thermonuclear fuels or components of such fuels. In fact, only for these substances thermonuclear energy gain exceeds (at some densities and temperatures) the bremsstrahlung loss and other high-temperature losses, i.e., thermonuclear burn-up is possible. Herein, thermonuclear burn-up in these deuterides and DT hydrides is considered in detail. In particular, a simple method is proposed to determine the critical values of the burn-up parameter x c for these substances and their mixtures at different temperatures and densities. The results for equimolar DT mixtures coincide quite well with the results of previous calculations. Also, the natural or Z limit is determined for low-temperature thermonuclear burn-up in the deuterides of light elements. (author)

  20. Thermonuclear pulsors engineering

    International Nuclear Information System (INIS)

    Ramos, Ruben F.

    2001-01-01

    The neutronic radiation has several applications, such as activation analysis of different substances, neutron radiography, molecular structures study, cancer therapy, humidity detection and materials surface treatment, among others. The main obstacle for these applications is the generation of neutronic beams. Nuclear reactors, isotopic sources and particle accelerators are neutron generators commonly used. They share the disadvantages of being non-portable, and quite expensive. This work is mainly focused on the development of neutron generators suitable to the applications mentioned before, in which traditional generators are non-applicable. The main characteristics should be transportability and to be non-contaminating, which would allow in-situ tests. Plasma focus generators, which produce neutron pulses by thermonuclear fusion reactions, satisfy these requirements and are economically convenient. This last feature would assure competitively in the neutron sources market. (author)

  1. Surface effects in controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Kaminsky, M.

    1975-08-01

    During the operation of large size plasma facilities and future controlled thermonuclear fusion reactors the surfaces of such major components as container walls, beam limiters, diverter walls and beam-dump walls of the injector region will be exposed to particle and photon bombardment from primary plasma radiations and from secondary radiations. Such radiations can cause, for example, physical and chemical sputtering, blistering, particle- and photon-impact induced desorption, secondary electron and x-ray emission, backscattering, nuclear reactions, photo-decomposition of surface compounds, photocatalysis, and vaporization. Such effects in turn can (a) seriously damage and erode the bombarded surface and (b) release major quantities of impurities which will contaminate the plasma. The effects of some of the major surface phenomena on the operation of plasma facilities and future fusion reactors are discussed

  2. A Game Theoretic Model of Thermonuclear Cyberwar

    Energy Technology Data Exchange (ETDEWEB)

    Soper, Braden C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-08-23

    In this paper we propose a formal game theoretic model of thermonuclear cyberwar based on ideas found in [1] and [2]. Our intention is that such a game will act as a first step toward building more complete formal models of Cross-Domain Deterrence (CDD). We believe the proposed thermonuclear cyberwar game is an ideal place to start on such an endeavor because the game can be fashioned in a way that is closely related to the classical models of nuclear deterrence [4–6], but with obvious modifications that will help to elucidate the complexities introduced by a second domain. We start with the classical bimatrix nuclear deterrence game based on the game of chicken, but introduce uncertainty via a left-of-launch cyber capability that one or both players may possess.

  3. An analysis of the impact of the thermonuclear pilot project ITER on industry and research in Austria

    International Nuclear Information System (INIS)

    Hangel, G.

    2007-03-01

    An analysis of the influence of the thermonuclear pilot project ITER on Austrian research and industrial activities is presented in terms of the following subjects: fusion research history, ITER technique, security, nuclear fusion, ITER (reactor, project specifications for quotations), possibilities for Austrian companies and fusion research in Austria. (nevyjel)

  4. Nuclear reactors and technology in the next stage

    International Nuclear Information System (INIS)

    Orlov, V.

    2000-01-01

    Author deals with the perspectives of development of nuclear power. It is possible to create in a fairly short time reactors and fuel technology that would meet the main requirements for large-scale power production, i.e.: (a) to afford a 100-fold reduction in the specific consumption of uranium, by utilizing thousands of tonnes of Pu accumulated in the spent fuel from the reactors of the fl t stage; .to rule out nuclear disasters, by taking advantage of the intrinsic properties and behavior of reactor, coolant, fuel, etc., with the plants made simpler and cheaper; (b) to hit a balance between the radiotoxicity of waste and that of feed uranium, by providing neutron transmutation; (c) to create power reactors and fuel cycle technology that would not afford extraction of weapon-grade materials. To fulfil all these requirements, it is necessary to provide substantial neutron excess in a chain reaction for Pu breeding, to use fuel with an equilibrium composition, to bum actinides and LLFPs. All this can be done only in fast reactors. Fast reactors can also provide fuel for thermal reactors that might still be used for some applications, operating in a Th/U cycle, which is the best option for such facilities. Novel engineering solutions will be necessary: high-density heat-conductive fuel (UPuN), chemically inert high-boiling coolant (Pb), dry reprocessing. These issues have been studied well enough to allow embarking on the development of advanced fast reactors. Minatom institutions are finalizing a detailed design of a demonstration BREST-300 plant, complete with an on-site fuel cycle that will meet the requirements of large-scale nuclear power. Hopefully, construction of this plant at Beloyarsk site with its subsequent trial operation would open a door to the next stage in nuclear power development. (author)

  5. Choice of economical optimum blanket of hybrid reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blinkin, V L; Novikov, V M

    1981-01-01

    The economical effectiveness of symbiotic power systems depends on the choice of the correlation between energy production and fissile fuel production in blankets of controlled thermonuclear fusion reactor (CTR), what is investigated here. It is shown that the optimum value of this correlation essentially depends on the ratio between the specific costs for energy production in hybrid thermonuclear reactors and that in fission reactors as part of the symbiotic system.

  6. Status report on controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    1990-06-01

    The International Fusion Research Council (IFRC), an advisory body to the International Atomic Energy Agency, reports on the current status of fusion; this report updates its 1978 status report. This report contains a General Overview and Executive Summary, and reports on all current approaches to fusion throughout the world; a series of technical reports is to be published elsewhere. This report is timely in that it not only shows progress which has occurred over the past, but interfaces with possible future devices, in particular the International Thermonuclear Experimental Reactor (ITER), whose conceptual design phase is nearing completion. 5 refs, 6 figs

  7. Synfuels production from fusion reactors

    International Nuclear Information System (INIS)

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

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

  8. The Tokamak Fusion Test Reactor D-T modifications and operations

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  9. Implementation of defence in depth for next generation light water reactors

    International Nuclear Information System (INIS)

    1997-12-01

    The publication of this IAEA technical document represents the conclusion of a task, initiated in 1995, devoted to defence in depth in future reactors. It focuses mainly on the next generation of LWRs, although many general considerations may also apply to other types of reactors

  10. Safety reviews of next-generation light-water reactors

    International Nuclear Information System (INIS)

    Kudrick, J.A.; Wilson, J.N.

    1997-01-01

    The Nuclear Regulatory Commission (NRC) is reviewing three applications for design certification under its new licensing process. The U.S. Advanced Boiling Water Reactor (ABWR) and System 80+ designs have received final design approvals. The AP600 design review is continuing. The goals of design certification are to achieve early resolution of safety issues and to provide a more stable and predictable licensing process. NRC also reviewed the Utility Requirements Document (URD) of the Electric Power Research Institute (EPRI) and determined that its guidance does not conflict with NRC requirements. This review led to the identification and resolution of many generic safety issues. The NRC determined that next-generation reactor designs should achieve a higher level of safety for selected technical and severe accident issues. Accordingly, NRC developed new review standards for these designs based on (1) operating experience, including the accident at Three Mile Island, Unit 2; (2) the results of probabilistic risk assessments of current and next-generation reactor designs; (3) early efforts on severe accident rulemaking; and (4) research conducted to address previously identified generic safety issues. The additional standards were used during the individual design reviews and the resolutions are documented in the design certification rules. 12 refs

  11. An approach to next step device optimisation

    International Nuclear Information System (INIS)

    Salpietro, E.

    2000-01-01

    The requirements for ITER EDA were to achieve ignition with a good safety margin, and controlled long inductive burn. These requirements lead to a big device, which requested a too ambitious step to be undertaken by the world fusion community. More realistic objectives for a next step device shall be to demonstrate the net production of energy with a high energy gain factor (Q) and a high boot strap current fraction (>60%) which is required for a Fusion Power Plant (FPP). The Next Step Device (NSD) shall also allow operation flexibility in order to explore a large range of plasma parameters to find out the optimum concept for the fusion power plant prototype. These requirements could be too demanding for one single device and could probably be better explored in a strongly integrated world programme. The cost of one or more devices is the decisive factor for the choice of the fusion power development programme strategy. The plasma elongation and triangularity have a strong impact in the cost of the device and are limited by the plasma vertical position control issue. The distance between plasma separatrix and the toroidal field conductor does not vary a lot between devices. It is determined by the sum of the distance between first wall-plasma sepratrix and the thickness of the nuclear shield required to protect the toroidal field coil insultation. The thickness of the TF coil is determined by the allowable stresses and superconducting characteristics. The outer radius of the central solenoid is the result of an optimisation to provide the magnetic flux to inductively drive the plasma. Therefore, in order to achieve the objectives for Q and boot-strap current fractions at the minimum cost, the plasma aspect ratio and magnetic field value shall be determined. The paper will present the critical issues for the next device and will make considerations on the optimal way to proceed towards the realisation of the fusion power plant

  12. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    International Nuclear Information System (INIS)

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-01-01

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided

  13. Thermonuclear fusion in the UK: towards a new abundant and durable energy source; La fusion nucleaire au Royaume-Uni: vers une nouvelle source d'energie abondante et durable

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    The ITER treaty (International thermonuclear experimental reactor) was signed in Paris on November 21, 2006, by the European Union, China, the USA, Japan and Russia. This treaty is devoted to the construction and exploitation of the biggest thermonuclear facility ever, capable to generate 500 MW during a reaction of 10 minutes. ITER is a priori the last experimental step before the construction of a fusion power plant for power generation at the industrial scale. The goal of ITER is to obtain a quasi-unexhaustible and less polluting energy source by the mid-21. century. The British research work has largely contributed to the development of this technology through a large number of projects that have preceded ITER but also through its present day involvement in the creation of the future reactor of Cadarache. This document presents: the UK fusion program, the projects carried out at the Culham science centre (Compass-D, Joint European Torus (JET), Small Tight Aspect Ratio Tokamak (START), Mega-Ampere Spherical Tokamak (MAST), EASY-2005 (European activation system)), the British involvement in ITER project and the transfer of technologies, and the nuclear fusion research in British universities (PPRG Imperial College London, CFSA Warwick university, Dalton nuclear institute (DNI), department of physics York university). (J.S.)

  14. ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

    International Nuclear Information System (INIS)

    1982-03-01

    ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

  15. Working together in future: next steps

    International Nuclear Information System (INIS)

    Copeland, S.

    2001-01-01

    Some questions about public trust have to find answers: the language of safety, how much is enough, how much is too much for stake holder interaction, how do you deal with the issue of responding, i.e. demonstrating action on issues when the licensing decision is not what public groups want, how do you measure the success of consultations, what are the indicators for trust, benchmarking and measuring public perceptions, need for consistency of approach (can determine public confidence by differences between countries) are so many questions that could help for next steps. (N.C.)

  16. Development of Stepping Endurance Test Plan on CRDM of a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, DongHyun; Kim, Hyeonil; Park, Suki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Various types of the irradiation targets can be loaded and unloaded during power operation, according to the purpose of research reactor utilization. And their reactivity worth varies as well. The insertion rate of reactivity is dependent to reactivity worth of targets, travel length during loading or unloading and transfer device speed. Due to the reactivity transition during loading and unloading, neutron power is changed and reaches an action point of the reactor regulating system. Based on the measured neutron rate of change, reactor power control system controls the power with its own algorithm. It generates the signals and transmits these to the CRDM for motor driving. Stepping motors on the CRDM move the control rods with step signals. The process repeats until power is stabilized. Accordingly, the stepping behaviours of CRDM should be modelled upon an understanding of the control process and reactor responses. Methodology for a stepping endurance test plan on the CRDM of a research reactor is developed since CRDM endurance is very important for reactor controller and should be ensured for a certain period of time throughout the life of a research reactor. Therefore, it is expected to provide a reasonable stepping test plan. In the future, the simulation will be performed with specific design values.

  17. Preliminary design of a Tandem-Mirror-Next-Step facility

    International Nuclear Information System (INIS)

    Damm, C.C.; Doggett, J.N.; Bulmer, R.H.

    1980-01-01

    The Tandem-Mirror-Next-Step (TMNS) facility is designed to demonstrate the engineering feasibility of a tandem-mirror reactor. The facility is based on a deuterium-tritium (D-T) burning, tandem-mirror device with a fusion power output of 245 MW. The fusion power density in the central cell is 2.1 MW/m 3 , with a resultant neutron wall loading of 0.5 MW/m 2 . Overall machine length is 116 m, and the effective central-cell length is 50.9 m. The magnet system includes end cells with yin-yang magnets to provide magnetohydrodynamic (MHD) stability and thermal-barrier cells to help achieve a plasma Q of 4.7 (where Q = fusion power/injected power). Neutral beams at energies up to 200 keV are used for plasma heating, fueling, and barrier pumping. Electron cyclotron resonant heating at 50 and 100 GHz is used to control the electron temperature in the barriers. Based on the resulting engineering design, the overall cost of the facility is estimated to be just under $1 billion. Unresolved physics issues include central-cell β-limits against MHD ballooning modes (the assumed reference value of β exceeds the current theory-derived limit), and the removal of thermalized α-particles from the plasma

  18. Modeling of secondary emission processes in the negative ion based electrostatic accelerator of the International Thermonuclear Experimental Reactor

    Directory of Open Access Journals (Sweden)

    G. Fubiani

    2008-01-01

    Full Text Available The negative ion electrostatic accelerator for the neutral beam injector of the International Thermonuclear Experimental Reactor (ITER is designed to deliver a negative deuterium current of 40 A at 1 MeV. Inside the accelerator there are several types of interactions that may create secondary particles. The dominating process originates from the single and double stripping of the accelerated negative ion by collision with the residual molecular deuterium gas (≃29% losses. The resulting secondary particles (positive ions, neutrals, and electrons are accelerated and deflected by the electric and magnetic fields inside the accelerator and may induce more secondaries after a likely impact with the accelerator grids. This chain of reactions is responsible for a non-negligible heat load on the grids and must be understood in detail. In this paper, we will provide a comprehensive summary of the physics involved in the process of secondary emission in a typical ITER-like negative ion electrostatic accelerator together with a precise description of the numerical method and approximations involved. As an example, the multiaperture-multigrid accelerator concept will be discussed.

  19. Development of next-generation light water reactor

    International Nuclear Information System (INIS)

    Ishibashi, Fumihiko; Yasuoka, Makoto

    2010-01-01

    The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

  20. Next generation advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Turgut, M. H.

    2009-01-01

    Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

  1. Plasma and controlled thermonuclear reaction

    International Nuclear Information System (INIS)

    Kapitsa, P.

    1980-01-01

    The principle and prospects are given of three methods of achieving controlled thermonuclear reaction. The original and so far most promising TOKAMAK method is presented invented in the USSR. Another method is the heating of a sphere about 1 mm in diameter from a mixture of deuterium and tritium by focused laser light from all sides. The third method consists in continuous plasma heating. A rope-like plasma discharge at a temperature of more than a million K results in the gas from microwave oscillations. The discharge is placed in a magnetic field and the ion temperature is increased by magneto-acoustic waves. A reactor is proposed operating on this principle and problems are pointed out which will have to be resolved. (M.S.)

  2. Plasma and controlled thermonuclear reaction

    Energy Technology Data Exchange (ETDEWEB)

    Kapitsa, P

    1980-06-01

    The principle and prospects are given of three methods of achieving controlled thermonuclear reaction. The original and so far most promising TOKAMAK method is presented invented in the USSR. Another method is the heating of a sphere about 1 mm in diameter from a mixture of deuterium and tritium by focused laser light from all sides. The third method consists in continuous plasma heating. A rope-like plasma discharge at a temperature of more than a million K results in the gas from microwave oscillations. The discharge is placed in a magnetic field and the ion temperature is increased by magneto-acoustic waves. A reactor is proposed operating on this principle and problems are pointed out which will have to be resolved.

  3. 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+

  4. Approaches to safety, environment and regulatory approval for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Saji, G.; Bartels, H.W.; Chuyanov, V.; Holland, D.; Kashirski, A.V.; Morozov, S.I.; Piet, S.J.; Poucet, A.; Raeder, J.; Rebut, P.H.; Topilski, L.N.

    1995-01-01

    International Thermonuclear Experimental Reactor (ITER) Engineering Design Activities (EDA) in safety and environment are approaching the point where conceptual safety design, topic studies and research will give way to project oriented engineering design activities. The Joint Central Team (JCT) is promoting safety design and analysis necessary for siting and regulatory approval. Scoping studies are underway at the general level, in terms of laying out the safety and environmental design framework for ITER. ITER must follow the nuclear regulations of the host country as the future construction site of ITER. That is, regulatory approval is required before construction of ITER. Thus, during the EDA, some preparations are necessary for the future application for regulatory approval. Notwithstanding the future host country's jurisdictional framework of nuclear regulations, the primary responsibility for safety and reliability of ITER rests with the legally responsible body which will operate ITER. Since scientific utilization of ITER and protection of the large investment depends on safe and reliable operation of ITER, we are highly motivated to achieve maximum levels of operability, maintainability, and safety. ITER will be the first fusion facility in which overall 'nuclear safety' provisions need to be integrated into the facility. For example, it will be the first fusion facility with significant decay heat and structural radiational damage. Since ITER is an experimental facility, it is also important that necessary experiments can be performed within some safety design limits without requiring extensive regulatory procedures. ITER will be designed with such a robust safety envelope compatible with the fusion power and the energy inventories. The basic approach to safety will be realized by 'defense-in-depth'. (orig.)

  5. Divertor, thermonuclear device and method of neutralizing high temperature plasma

    International Nuclear Information System (INIS)

    Ikegami, Hideo.

    1995-01-01

    The thermonuclear device comprises a thermonuclear reactor for taking place fusion reactions to emit fusion plasmas, and a divertor made of a hydrogen occluding material, and the divertor is disposed at a position being in contact with the fusion plasmas after nuclear fusion reaction. The divertor is heated by fusion plasmas after nuclear fusion reaction, and hydrogen is released from the hydrogen occluding material as a constituent material. A gas blanket is formed by the released hydrogen to cool and neutralize the supplied high temperature nuclear fusion plasmas. This prevents the high temperature plasmas from hitting against the divertor, elimination of the divertor by melting and evaporation, and solve a problem of processing a divertor activated by neutrons. In addition, it is possible to utilize hydrogen isotopes of fuels effectively and remove unnecessary helium. Inflow of impurities from out of the system can also be prevented. (N.H.)

  6. Controlled thermonuclear reactions and Tora Supra program

    International Nuclear Information System (INIS)

    1988-01-01

    The research programs for the nuclear energy production by means of thermonuclear fusion are shown. TORA SUPRA, Joint European Torus, Next European Torus and those developed at the Atomic Energy Center are described. The controlled fusion necessary conditions, the energy and confinement balance, and the research of a better tokamak configuration are discussed. A description of TORA SUPRA, the ways of achieving the project and the expected delays are shown. The Controlled Fusion Research Department functions, concerning these programs, are described. The importance of international cooperation and the perspectives about the use of controlled fusion are underlined [fr

  7. Analysis of the two accelerator concepts foreseen for the neutral beam injector of the International Thermonuclear Experimental Reactor

    Directory of Open Access Journals (Sweden)

    G. Fubiani

    2009-05-01

    Full Text Available Typical high-energy negative ion electrostatic accelerators such as the ones designed for fusion applications produce a significant amount of secondary particles. These particles may originate from coextracted electrons, which flow from the ion source, impacting the accelerator grids or as by-products of collisions between accelerated negative ions and the residual background gas, in the accelerator. Secondary emission particles may carry a non-negligible power and consequently must be precisely studied. The electrostatic-accelerator-Monte-Carlo-simulation code (EAMCC [G. Fubiani et al., Phys. Rev. ST Accel. Beams 11, 014202 (2008PRABFM1098-440210.1103/PhysRevSTAB.11.014202] was developed in order to provide a three-dimensional characterization of power and current deposition on all parts of the accelerator. The code includes all the relevant physics associated with secondary emission processes and consequently may be used as a tool for design improvement. In this paper, the two accelerator designs considered for the International Thermonuclear Experimental Reactor, that is, the multiaperture-multigrid and the single gap single aperture (SINGAP designs, are discussed and their predicted performances compared. Simulations have been compared with measurements on prototype accelerators of the SINGAP type. Reasonable agreement between EAMCC calculations and measurements of backstreaming ions and transmitted electrons was found.

  8. Fusion reactors for hydrogen production via electrolysis

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  9. Development of source term evaluation method for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Keon Jae; Cheong, Jae Hak; Park, Jin Baek; Kim, Guk Gee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-10-15

    This project had investigate several design features of radioactive waste processing system and method to predict nuclide concentration at primary coolant basic concept of next generation reactor and safety goals at the former phase. In this project several prediction methods of source term are evaluated conglomerately and detailed contents of this project are : model evaluation of nuclide concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant(NPP), investigation of prediction parameter of source term evaluation, basic parameter of PWR, operational parameter, respectively, radionuclide removal system and adjustment values of reference NPP, suggestion of source term prediction method of next generation NPP.

  10. Feasibility and application on steam injector for next-generation reactor

    International Nuclear Information System (INIS)

    Narabayashi, Tadashi; Ishiyama, Takenori; Miyano, Hiroshi; Nei, Hiromichi; Shioiri, Akio

    1991-01-01

    A feasibility study has been conducted on steam injector for a next generation reactor. The steam injector is a simple, compact passive device for water injection, such as Passive Core Injection System (PCIS) of Passive Containment Cooling System (PCCS), because of easy start-up without an AC power. An analysis model for a steam injector characteristics has been developed, and investigated with a visualized fundamental test for a two-stage Steam Injector System (SIS) for PCIS and a one-stage low pressure SIS for PCCS. The test results showed good agreement with the analysis results. The analysis and the test results showed the SIS could work over a very wide range of the steam pressure, and is applicable for PCIS or PCCS in the next generation reactors. (author)

  11. Thermonuclear energy and the power industry in the future

    International Nuclear Information System (INIS)

    Velikhov, E.P.

    1986-01-01

    The leader of the USSR thermonuclear program, the vicepresident of the Academy of Science, comrade Velikhov tells about the modern state and perspective of thermonuclear investigations, as well as about the problems on the international cooperation in this field

  12. Space nuclear reactors: energy gateway into the next millennium

    International Nuclear Information System (INIS)

    Angelo, J.A. Jr.; Buden, D.

    1981-01-01

    Power - reliable, abundant and economic - is the key to man's conquest of the Solar System. Space activities of the next few decades will be highlighted by the creation of the extraterrestrial phase of human civilization. Nuclear power is needed both to propel massive quantities of materials through cislunar and eventually translunar space, and to power the sophisticated satellites, space platforms, and space stations of tomorrow. To meet these anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100-kW(e) heat pipe nuclear reactor. The objectives of this program are to develop components for a space nuclear power plant capable of unattended operation for 7 to 10 years; having a reliability of greater than 0.95; and weighing less than 1910 kg. In addition, this heat pipe reactor is also compatible for launch by the US Space Transportation System

  13. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  14. The next step in biology: a periodic table?

    Science.gov (United States)

    Dhar, Pawan K

    2007-08-01

    Systems biology is an approach to explain the behaviour of a system in relation to its individual components. Synthetic biology uses key hierarchical and modular concepts of systems biology to engineer novel biological systems. In my opinion the next step in biology is to use molecule-to-phenotype data using these approaches and integrate them in the form a periodic table. A periodic table in biology would provide chassis to classify, systematize and compare diversity of component properties vis-a-vis system behaviour. Using periodic table it could be possible to compute higher- level interactions from component properties. This paper examines the concept of building a bio-periodic table using protein fold as the fundamental unit.

  15. Complex workplace radiation fields at European high-energy accelerators and thermonuclear fusion facilities

    CERN Document Server

    Bilski, P; D'Errico, F; Esposito, A; Fehrenbacher, G; Fernàndez, F; Fuchs, A; Golnik, N; Lacoste, V; Leuschner, A; Sandri, S; Silari, M; Spurny, F; Wiegel, B; Wright, P

    2006-01-01

    This report outlines the research needs and research activities within Europe to develop new and improved methods and techniques for the characterization of complex radiation fields at workplaces around high-energy accelerators and the next generation of thermonuclear fusion facilities under the auspices of the COordinated Network for RAdiation Dosimetry (CONRAD) project funded by the European Commission.

  16. A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

    The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

  17. International research co-operation in the field of controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Conscience, J.-F.

    2003-01-01

    This final report for the Swiss Federal Office of Education and Science presents a review of activities carried out in 2002 within the framework of the International Experimental Thermonuclear Reactor (ITER) project that involves contributions from Canada, Japan, the Russian Federation and the European Union. Further agreements on the development of a fusion reactor with other countries, including Switzerland, the USA and China, are mentioned. The first chapter describes the current state of research on electricity production using nuclear fusion and discusses feasibility, safety, environmental, fuel supply and economic aspects. A second chapter reviews global efforts in the fusion area, including ITER and EURATOM projects and the activities running under the European Fusion Development Agreement EFDA and the JET Implementing Agreement. Finally, a third chapter deals with fusion research activities in Switzerland and the contributions made to international research by Swiss universities and institutes

  18. Thermonuclear research development

    International Nuclear Information System (INIS)

    Velikhov, E.

    1977-01-01

    Tokamak 10, the world's largest thermonuclear facility was commissioned in 1975. Soviet scientists thus achieved enormous success in producing high-temperature plasma and constructing a thermonuclear fusion source. The problems which remain to be solved include finding a method of regenerating the deuterium-tritium fuel mixture and a method of purifying the reacting high-temperature plasma of heavy elements. The project is designed for a more powerful facility, namely the Tokamak 20 whose toroidal chamber will accommodate a current of 5 to 6 MA and whose plasma volume will be 400 m 3 . (Oy)

  19. Thermonuclear research development

    Energy Technology Data Exchange (ETDEWEB)

    Velikhov, E

    1977-04-01

    Tokamak 10, the world's largest thermonuclear facility was commissioned in 1975. Soviet scientists thus achieved enormous success in producing high-temperature plasma and constructing a thermonuclear fusion source. The problems which remain to be solved include finding a method of regenerating the deuterium-tritium fuel mixture and a method of purifying the reacting high-temperature plasma of heavy elements. The project is designed for a more powerful facility, namely the Tokamak 20 whose toroidal chamber will accommodate a current of 5 to 6 MA and whose plasma volume will be 400 m/sup 3/.

  20. Benchmark problem suite for reactor physics study of LWR next generation fuels

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Ikehara, Tadashi; Ito, Takuya; Saji, Etsuro

    2002-01-01

    This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70 GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO 2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management. (author)

  1. Thermo-mechanical analysis of an acceleration grid for the international thermonuclear experimental reactor-neutral beam injection system

    International Nuclear Information System (INIS)

    Fujiwara, Yukio; Hanada, Masaya; Okumura, Yoshikazu; Suzuki, Satoshi; Watanabe, Kazuhiro

    2001-01-01

    In the engineering design of a negative-ion beam source for a high-power neutral beam injection (NBI) system, one of the most important issues is thermo-mechanical design of acceleration grids for producing several tens of MW ion beams. An acceleration grid for the international thermonuclear experimental reactor-neutral beam injection (ITER-NBI) system will be subjected to the heat loading as high as 1.5 MW. In the present paper, thermo-mechanical characteristics of the acceleration grid for the ITER-NBI system were analyzed. Numerical simulation indicated that maximum aperture-axis displacement of the acceleration grid due to thermal expansion would be about 0.7 mm for the heat loading of 1.5 MW. From the thin lens theory of beam optics, beamlet deflection angle by the aperture-axis displacement was estimated to be about 2 mrad, which is within the requirement of the engineering design of the ITER-NBI system. Numerical simulation also indicated that no melting on the acceleration grid would occur for a heat loading of 1.5 MW, while local plastic deformation would happen. To avoid the plastic deformation, it is necessary to reduce the heat loading onto the acceleration grid to less than 1 MW

  2. Major NSSS design features of the Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, Insk; Kim, Dong-Su

    1999-01-01

    In order to meet national needs for increasing electric power generation in the Republic of Korea in the 2000s, the Korean nuclear development group (KNDG) is developing a standardized evolutionary advanced light water reactor (ALWR), the Korean Next Generation Reactor (KNGR). It is an advanced version of the successful Korean Standard Nuclear Power Plant (KSNP) design, which meets utility needs for safety enhancement, performance improvement and ease of operation and maintenance. The KNGR design starts fro the proven design concept of the currently operating KSNPs with uprated power and advanced design features required by the utility. The KNGR design is currently in the final stage of the basic design, and the paper describes the major nuclear steam supply system (NSSS) design features of the KNGR together with introduction of the KNGR development program. (author)

  3. Challenges for Plasma Diagnostic in a Next Step Device (FIRE)

    International Nuclear Information System (INIS)

    Young, Kenneth M.

    2002-01-01

    The physics program of any next-step tokamak such as FIRE [Fusion Ignition Research Experiment] sets demands for plasma measurement which are at least as comprehensive as on present tokamaks, with the additional capabilities needed for control of the plasma and for understanding the effects of the alpha-particles. The diagnostic instrumentation must be able to provide the fine spatial and temporal resolution required for the advanced tokamak plasma scenarios. It must also be able to overcome the effects of neutron- and gamma-induced electrical noise in ceramic components or detectors, and fluorescence and absorption in optical components. There are practical engineering issues of minimizing radiation streaming while providing essential diagnostic access to the plasma. Many diagnostics will require components at or close to the first wall, e.g., ceramics and MI cable for magnetic diagnostics and mirrors for optical diagnostics; these components must be mounted to operate, and survive, i n fluxes which require special material selection. A better set of diagnostics of alpha-particles than that available for the TFTR [Tokamak Fusion Test Reactor] is essential; it must be qualified well before moving into D-T [deuterim-tritium] experiments. A start has been made to assessing the potential implementation of key diagnostics for the FIRE device. The present status is described

  4. Methodology on the sparger development for Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K.

    1999-06-01

    In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloud acceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs

  5. Methodology on the sparger development for Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K

    1999-06-01

    In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloudacceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs.

  6. The world must build two atomic reactors each day the next hundred years

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    In summarising and commenting on the ideas presented in Mesarovic and Pestel's book 'Mankind at the turning point' it is pointed out that the global energy crisis makes comprehensive long term planning a necessity. Assuming, optimistically, that nuclear power alone is able to supply the total projected energy demand in 100 years, it is stated that this will require 3000 nuclear power stations, each with 8 fast breeder reactors, totally 100GW(t). This means a net rate of construction of four reactors per week, which again means, allowing for a 30 year life, two reactors per day, every day, for the next hundred years. Fuelling these reactors will require the production and transport of 15 x 10 6 kg of Pu239 per year. It is therefore obvious that the energy crisis is not only a technological, but also a political, social and even psychological problem. (JIW)

  7. Powerful lasers for thermonuclear fusion

    International Nuclear Information System (INIS)

    Basov, N.; Krokhin, O.; Sklizkov, G.; Fedotov, S.

    1977-01-01

    The parameters are discussed of the radiation of powerful lasers (internal energy of the plasma determined by the volume, density and temperature of the plasma, duration of the heating pulse, focusing of the laser pulse energy in a small volume of matter, radiation contrast) for attaining an effective thermonuclear fusion at minimum microexplosion energy. A survey is given of the methods of shaping laser pulses with limit parameters, and the principle of the construction of powerful laser systems is described. The general diagram and parameters are given of the Delfin thermonuclear apparatus and a diagram is presented of the focusing system of high luminosity for spherical plasma heating using spherical mirrors. A diagram is presented of the vacuum chamber and of the complex diagnostic apparatus for determining the basic parameters of thermonuclear plasma in the Delfin apparatus. The prospects are indicated of the further development of thermonuclear laser apparatus with neodymium and CO 2 lasers. (B.S.)

  8. Next steps in the Energy Frontier - Hadron colliders workshop at LPC@FNAL

    CERN Document Server

    2014-01-01

    With the observation of the Standard Model Higgs boson, the high energy physics community is investigating possible next steps for entering into a new era in particle physics. The aim of this workshop is to bring together physics, instrumentation/detector and accelerator experts to present, outline and discuss all aspects needed for the next steps in the energy frontier. The workshop will focus on the lessons learned with 7 and 8 TeV LHC, physics requirements and subsequent detector technologies for HL-LHC, as well as development needs for future 100 TeV proton collider. The goal is to identify synergies and common approaches where further collaboration between various initiatives could be fruitful. The discovery potential for a future 100 TeV proton collider will depend on the detector / instrumentation capabilities in order to explore the highest energy and phenomena. Many of these detection capabilities will need further studies such as muon detection at several 10s of TeV range, calorimeters capable of me...

  9. 3D Simulation of a Loss of Vacuum Accident (LOVA in ITER (International Thermonuclear Experimental Reactor: Evaluation of Static Pressure, Mach Number, and Friction Velocity

    Directory of Open Access Journals (Sweden)

    Jean-François Ciparisse

    2018-04-01

    Full Text Available ITER (International Thermonuclear Experimental Reactor is a magnetically confined plasma nuclear reactor. Inside it, due to plasma disruptions, the formation of neutron-activated powders, which are essentially made out of tungsten and beryllium, occurs. As many windows for diagnostics are present on the reactor, which operates at very low pressure, a LOVA (Loss of Vacuum Accident could be possible and may lead to dust mobilisation and a toxic and radioactive fallout inside the plant. This study is aimed at reproducing numerically the first seconds of a LOVA in ITER, in order to get information about the dust resuspension risk. This work has been carried out by means of a CFD (Computational Fluid Dynamics simulation of the beginning of the pressurisation transient inside the whole Tokamak. It has been found that the pressurization transient is extremely slow, and that the friction speed on the walls is very high, and therefore a high mobilization risk of the dust is expected on the entire internal surface of the reactor. It has been observed that a LOVA in a real-scale reactor is more severe than the one reproduced in reduced-scale facilities, as STARDUST-U, because the speeds are higher, and the dust resuspension capacity of the flow is greater.

  10. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  11. Enhanced probabilistic decision analysis for radiological confinement barriers of the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Golinescu, R.P.; Kazimi, M.S.

    1998-01-01

    To ensure a defence-in-depth approach, several radiological confinement barriers surrounding a tokamak plant can be employed. A methodology using probabilistic risk assessment (PRA) techniques is a useful tool for evaluating the performance of each confinement barrier within the context of a limited allowable risk of accidental radioactivity releases. Such a methodology was developed and applied to the confinement strategy for the International Thermonuclear Experimental Reactor (ITER). Accident sequence models were constructed for each of the confinement barriers to evaluate the probabilities of events leading to radioactive releases from the corresponding confinement barrier. The current ITER design requirements set radioactive release and dose limits for individual event sequences grouped in categories by frequency. To limit the plant's overall risk and account for event uncertainties in both frequency and consequence, an analytical form for a limit line is derived here as a complementary cumulative frequency (CCF) of radioactive releases to the environment. By comparing the releases from each confinement barrier against the limit line, a decision can be made about the number of barriers required to comply with the design requirements. The first barrier is the vacuum vessel (VV) and the primary heat transfer systems. The second confinement barrier consists of the cryostat vessel (CV) and the heat transfer system vaults. In case the outer building is needed to act as a third barrier for ITER, a decision model using the multi-attribute utility theory was constructed to help the designer choose the best type of tokamak building. The decision model allows for performing sensitivity analysis on relevant parameters and for design features of new options for the ITER tokamak building. (orig.)

  12. Conceptual designs of power tokamak-type thermonuclear reactors

    International Nuclear Information System (INIS)

    Shejndlin, A.E.; Nedospasov, A.V.

    1978-01-01

    Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

  13. Zone-plate coded imaging of thermonuclear burn

    International Nuclear Information System (INIS)

    Ceglio, N.M.

    1978-01-01

    The first high-resolution, direct images of the region of thermonuclear burn in laser fusion experiments have been produced using a novel, two-step imaging technique called zone-plate coded imaging. This technique is extremely versatile and well suited for the microscopy of laser fusion targets. It has a tomographic capability, which provides three-dimensional images of the source distribution. It is equally useful for imaging x-ray and particle emissions. Since this technique is much more sensitive than competing imaging techniques, it permits us to investigate low-intensity sources

  14. Transition to thermonuclear burn in fusion plasmas

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1991-01-01

    An analytical investigation is made of the time evolution of the 1-D temperature profile in a fusion reactor plasma where the nonlinear energy balance equation is dominated by alpha-particle heating and thermal conduction losses. Special emphasis is given to the problem of establishing sufficient conditions for the transition to thermonuclear burn for given initial temperature profiles. In particular, it is demonstrated that for strongly nonlinear alpha-particle heating, temperature profiles initially peaked on-axis are more easily ignited than profiles similar in form to the equilibrium profile of the energy balance equation. Simple analytical criteria for ignition are established and are shown to compare favourably with results of numerical calculations. (author)

  15. Thermonuclear fusion: from fundamental research to energy production? Science and technology report No. 26

    International Nuclear Information System (INIS)

    Laval, Guy; Blanzat, Bernard; Aspect, Alain; Aymar, Robert; Bielak, Bogdan; Decroisette, Michel; Martin, Georges; Andre, Michel; Schirmann, Daniel; Garbet, Xavier; Jacquinot, Jean; Laviron, Clement; Migus, Arnold; Moreau, Rene; Pironneau, Olivier; Quere, Yves; Vallee, Alain; Dercourt, Jean; Bayer, Charles; Juraszek, Denis; Deutsch, Claude; Le Garrec, Bruno; Hennequin, Pascale; Peysson, Yves; Rax, Jean-Marcel; Pesme, Denis; Bauche, Jacques; Monier-Garbet, Pascale; Stamm, Roland; Zerah, Gilles; Ghendrih, Philippe; Layet, Roland; Grosman, Andre; Alamo, Ana; Giancarli, Luciano; Poitevin, Yves; Rigal, Emmanuel; Chieze, Jean-Pierre

    2007-01-01

    This work has been commissioned by the French ministry of Education, Sciences and Research, its aim is to provide a reliable account of the state of development of thermonuclear fusion. This report makes a point on the scientific knowledge accumulated on the topic and highlights the research programs that are necessary to overcome the technological difficulties and draws the necessary steps before an industrial application to electricity production. This report is divided into 10 chapters: 1) tokamak technology and ITER, 2) inertial fusion, 3) magnetized hot plasmas, 4) laser-plasma interaction and peta-watt lasers, 5) atomic physics and fusion, 6) computer simulation, 7) plasma-wall interaction, 8) materials for fusion reactors, 9) safety analysis, and 10) inertial fusion and astrophysics. This report has been written by a large panel of experts gathered by the French Academy of Sciences. The comments on the issue by the 3 French organizations: Cea, Cnrs and SFP (French Society of Physics) follow the last chapter

  16. Development of Next-Generation LWR (Light Water Reactor) in Japan

    International Nuclear Information System (INIS)

    Yamamoto, T.; Kasai, S.

    2011-01-01

    The Next-Generation Light Water Reactor development program was launched in Japan in April 2008. The primary objective of the program is to cope with the need to replace existing nuclear power plants in Japan after 2030. The reactors to be developed are also expected to be a global standard design. Several innovative features are envisioned, including a reactor core system with uranium enrichment above 5%, a seismic isolation system, the use of long-life materials and innovative water chemistry, innovative construction techniques, safety systems with the best mix of passive and active concepts, and innovative digital technologies to further enhance reactor safety, reliability, economics, etc. In the first 3 years, a plant design concept with these innovative features is established and the effectiveness of the program is reevaluated. The major part of the program will be completed in 2015. (author)

  17. Composite electric generator equipped with steam generator for heating reactor coolant

    International Nuclear Information System (INIS)

    Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

    1997-01-01

    The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

  18. Control method for thermonuclear plasma

    International Nuclear Information System (INIS)

    Azuma, Kingo; Oda, Yasushi.

    1997-01-01

    CT (Compact Troid) is a doughnut-like shaped plasmas having a toroidal current and a poloidal current at the inside and forming a poloidal magnetic fluxes and toroidal magnetic flux. The structure of the CT is collapsed at a time of stationary state, accordingly, when it is injected to thermonuclear plasmas, particles can be supplied locally, and the state of the plasmas to be supplied can be changed by changing the direction of the injection. If a CT which is reverse to the poloidal magnetic fields is injected, plasmas with excessive ions can be supplied locally thereby enabling to form magnetic field in the thermonuclear plasmas. If the magnetic fields are formed in the vicinity of the surface of the thermonuclear plasmas, fast ions which have come over the magnetic field structure can be returned to the central portion of the plasmas. Then, confining performance of thermonuclear plasmas can be greatly improved, the efficiency for fuel supply can be increased, and energy required for ignition can be suppressed. (N.H.)

  19. A survey on the human reliability analysis methods for the design of Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W.

    2000-03-01

    Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design

  20. A survey on the human reliability analysis methods for the design of Korean next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W

    2000-03-01

    Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design.

  1. Design of the ITER (International Thermonuclear Experimental Reactor) neutral beam system beamline, United States concept

    International Nuclear Information System (INIS)

    Purgalis, P.; Anderson, O.A.; Cooper, W.S.; DeVries, G.E.; Lietzke, A.F.; Kunkel, W.B.; Kwan, J.W.; Matuk, C.A.; Nakai, T.; Stearns, J.W.; Soroka, L.; Wells, R.P.; Lindquist, W.B.; Neef, W.S.; Reginato, L.L.; Sedgley, D.W.; Brook, J.W.; Luzzi, T.E.; Myers, T.J.

    1989-01-01

    Design of a neutral beamline for ITER (International Thermonuclear Experimental Reactor) is described. The design incorporates a barium surface conversion D - source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to watercooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules that can be removed for remote maintenance. The neutral beam system delivers 75 MW of D degree into three ports with a total of nine modules arranged in stacks of three modules per port. To increase reliability each module is designed to deliver up to 10 MW at 1.3 MeV; this allows eight modules operating at partial capacity to deliver the required power in the event one module is removed from service. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 35 m from the port into the torus. Neutron shielding in the drift duct provides the added feature of limiting conductance and thus reducing gas flow to and from the torus. Alternative component choices are also discussed for the evolving design. 8 refs., 4 figs., 1 tab

  2. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  3. Development step toward fusion power plant and role of experimental reactor ITER

    International Nuclear Information System (INIS)

    Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

    2005-01-01

    The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

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

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

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

  6. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  7. Thermonuclear reaction generation method and device

    International Nuclear Information System (INIS)

    Imazaki, Kazuo

    1998-01-01

    The present invention provides a method of and a device for causing thermonuclear reaction capable of obtaining extremely high profits (about 1000 times), capable of forming a target which is strong against instability upon implosion as a problem of an inertia process and capable of realizing utilization of nuclear fusion. Namely, elementary particles such as pion, muon and K particles are deposited a portion or some portion of thermonuclear fuel materials by using high energy ions and highly brilliant γ rays generated from a high energy accelerator. The thermonuclear fuel materials are compressed to high density. The nuclear fusion reaction is promoted to ignite and burn thermonuclear fuels. A portion of nuclear fuels is ignited selectively by the means. High profits can be obtained. Since there is no need to attain implosion rate required for self ignition of nuclear fuels, a target of low aspect ratio can be used. (I.S.)

  8. Recycling, inventory and permeation of hydrogen isotopes and helium in the first wall of a thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Gervasini, G.; Reiter, F.

    1989-01-01

    The work was divided into three parts. The first part, which is theoretical, examines the behaviour of hydrogen in metals. After an introduction on the presence of hydrogen isotopes in fusion reactors, the main phenomena connected with hydrogen-metal interaction are summarised: solubility, diffusivity and trapping in material defects. The metal temperature is highlighted as the main parameter in the description of the phenomena. The second part of the work, also theoretical, concerns the interaction between helium and metals. We have tried as much as possible to show analogies and differences in the comparisons of the behaviour of hydrogen. The main types of damage caused by helium in metallic structures, which are the most important consequence of helium-metal interaction, were summarised. The characteristics of helium were treated in greater depth than those of hydrogen, because the latter are very well known. Also, there is a vast literature on the hydrogen-metal interaction. In the third and last part of the work a model was identified which allows the simulation of the evolution of a system formed from a metal in which hydrogen and helium isotopes have been introduced. A system of algebraic-differential equations was used to study the temporal evolution of the concentrations, the recycling, the inventory and the permeation of tritium and helium considering that these atoms diffuse in the metallic lattice and remain trapped in the vacancies created inside the metal by the bombardment of the neutrons from the fusion reactions. For the numerical simulation a series of data intended to represent the situation inside a thermonuclear reactor as precisely as possible were used for the numerical simulation. Analysis of the system was preceded by the analytical resolution of the steady state equations so that they could be compared with the simulation results

  9. Integration of element technology and system supporting thermonuclear fusion

    International Nuclear Information System (INIS)

    2003-01-01

    A special committee for integrated system technology survey on thermonuclear fusion (TNF) was begun on June, 1999, under an aim to generally summarize whole of shapes on technology to realize TNF reactor to summarize present state of every technologies and their positioning in whole of their TNF technology. On a base of survey of these recent informations, this report is comprehensively summarized for an integrated system technology on TNF. It has outlines on magnetic field enclosing method, outlines on inertia enclosing method, element technology supporting TNF, new power generation techniques, and ripple effects on TNF technology. (G.K.)

  10. Tying the knot with next-generation reactors: Can the industry afford a second marriage?

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This article examines the future of nuclear power beyond the year 2000. The nuclear industry just celebrated 50 years of nuclear technology, but no new plants have been ordered in the US since 1978 and some European countries are giving up on the nuclear option. This article discusses the four US advanced light-water reactor design and safety features, specific design features and parameters for the advanced designs, advanced designs from Europe, features utilities look for in a reactor, evolutionary versus passive designs, gaining public acceptance for new designs, and what alternatives are there to installing next-generation nuclear systems?

  11. Use of virtual environments to reduce the construction costs of the next generation nuclear power reactors

    International Nuclear Information System (INIS)

    Whisker, V.E.; Baratta, A.J.

    2007-01-01

    The near term deployment of the next generation of reactors will only be successful if they are built on time and without the costly overruns experienced in the previous generation. One critical factor in achieving these goals is to ensure the design is optimized for constructability. In this work the authors explored the effectiveness of full-scale virtual reality simulation in the optimization of the design and construction of the next generation of nuclear reactors. The research tested the suitability of immersive virtual reality display technology in aiding engineers in evaluating potential cost reductions that can be realized by the optimization of design and installation and construction sequences. The intent of this research is to see if this type of technology can be used in capacities similar to those currently filled by full-scale physical mockups and desktop simulations. Using a fully-immersive five sided virtual reality system, known as a CAVE, the authors constructed a series of virtual mockups that represented two next generation nuclear power plants, the Westinghouse AP-1000 and the Pebble Bed Modular Reactor (PBMR). These virtual mockups were then tested as a design tool to help locate and correct problem areas, to optimize the construction sequence, and to assist with familiarizing trades people with the performance of maintenance activities. A series of experiments were performed to assess the usefulness of these virtual mockups in accomplishing these tasks. (authors)

  12. Controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Sakanaka, P.H.

    1984-01-01

    A simplified review on the status of the controlled thermonuclear fusion research aiming to present the motivation, objective, necessary conditions and adopted methods to reach the objective. (M.C.K.) [pt

  13. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1976-09-01

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  14. Alternative Aviation Fuels: Overview of Challenges, Opportunities, and Next Steps

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-03-28

    The Alternative Aviation Fuels: Overview of Challenges, Opportunities, and Next Steps report, published by the U.S. Department of Energy’s Bioenergy Technologies Office (BETO) provides an overview of the current state of alternative aviation fuels, based upon findings from recent peer-reviewed studies, scientific working groups, and BETO stakeholder input provided during the Alternative Aviation Fuel Workshop.

  15. The importance of collaboration in the advancement of current and next generation reactors

    International Nuclear Information System (INIS)

    Jackson, Kate; Goossen, John; Anness, Mike; Meston, Tom

    2010-01-01

    The sections of the contribution are as follows: Tradition of innovation. Growing demand for nuclear power; Collaboration drivers; Responses. Knowledge transfer and management is critical. What kind of focus? Equipment reliability. Advanced repair, replacement and construction approaches. Materials. Plant safety margins. Spent fuel management. Examples of European collaboration. Zorita materials examination. Collaboration in the development of next generation reactors; Westinghouse R and D priorities; A look to the future. (P.A.)

  16. The Scientific Prototype - a proposed next step for the American MFE effort

    Science.gov (United States)

    Manheimer, Wallace

    2013-10-01

    The Scientific prototype is the only logical next step for the American magnetic fusion effort. This poster is divided into two parts. The first is a description of the scientific prototype, a tokamak about the size of TFTR, JET and JT-60, but which runs steady state in DT and breeds its own tritium. The second is an examination of other proposed approaches for American MFE and why none constitute a viable alternative. W. Manheimer, J. Fusion Energy, 32, 419-421, 2013.

  17. FIRE, A Next Step Option for Magnetic Fusion

    International Nuclear Information System (INIS)

    Meade, D.M.

    2002-01-01

    The next major frontier in magnetic fusion physics is to explore and understand the strong nonlinear coupling among confinement, MHD stability, self-heating, edge physics, and wave-particle interactions that is fundamental to fusion plasma behavior. The Fusion Ignition Research Experiment (FIRE) Design Study has been undertaken to define the lowest cost facility to attain, explore, understand, and optimize magnetically confined fusion-dominated plasmas. The FIRE is envisioned as an extension of the existing Advanced Tokamak Program that could lead to an attractive magnetic fusion reactor. The FIRE activities have focused on the physics and engineering assessment of a compact, high-field tokamak with the capability of achieving Q approximately equal to 10 in the ELMy H-mode for a duration of about 1.5 plasma current redistribution times (skin times) during an initial burning-plasma science phase, and the flexibility to add Advanced Tokamak hardware (e.g., lower-hybrid current drive) later. The configuration chosen for FIRE is similar to that of ARIES-RS, the U.S. Fusion Power Plant study utilizing an Advanced Tokamak reactor. The key ''Advanced Tokamak'' features are: strong plasma shaping, double-null pumping divertors, low toroidal field ripple ( 5) for a duration of 1 to 3 current redistribution times

  18. Annual report 1990. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1990-01-01

    In 1990 the operation of the High Flux Reactor was carried out as planned. The availability was 96% of scheduled operating time. The average utilization of the reactor was 71% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  19. Annual report 1989 operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1989-01-01

    In 1989 the operation of the High Flux Reactor Petten was carried out as planned. The availability was more than 100% of scheduled operating time. The average occupation of the reactor by experimental devices was 72% of the practical occupation limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons and for radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  20. Isochronous cyclotron for thermonuclear reactors driving

    International Nuclear Information System (INIS)

    Alenitskij, Yu.G.

    1998-01-01

    The main requirements to an accelerator as a part of an electronuclear power plant are considered. The range of the parameters of the accelerated proton and deuteron beams, for which the isochronous cyclotron is the most profitable, is proposed. An opportunity of using the cyclotron to drive the research reactors of various types is considered

  1. The six P’s of the next step in electronic patient records in the Netherlands

    NARCIS (Netherlands)

    Michel-Verkerke, Margreet B.; Stegwee, Robert A.; Spil, Antonius A.M.

    2015-01-01

    The objective of this study was to evaluate a decade of Electronic Patient Record development. During the study a second question was added: How to take the next step in the Netherlands? This paper describes the developments but the main results create a framework for the future situation. The USE

  2. Magnetohydrodynamics and the thermonuclear problem

    Energy Technology Data Exchange (ETDEWEB)

    Alfven, H [Department of Electronics, Royal Institute of Technology, Stockholm (Sweden)

    1958-07-01

    The importance of magnetohydrodynamics and plasma physics for the solution of thermonuclear problem is presented in the paper. Methods for capture of a plasma by a magnetic field are discussed. From the study it is concluded that in principle it is possible to shoot heated plasma into a magnetic field and capture it there. A possible method of capturing plasma which is shot into a magnetic field is illustrated. Magnetohydrodynamic research performed during the last decade in Stockholm is presented. Following a long series of investigations of relatively cool plasmas, it has been started a series of experimental investigations on hot plasmas, concentrating on the fundamental properties of the plasma. New ways of the approach to the thermonuclear problem are analysed. Experiments have been with discharges of a few hundred kiloamps to produce fast-moving magnetized plasmas, in order to investigate whether they could be captured by magnetic fields in the discussed way.

  3. Safety and environmental advantages of breeding blanketless fusion reactors

    International Nuclear Information System (INIS)

    Zucchetti, M.; Merola, M.; Matera, R.

    1994-01-01

    Next-step reactors will use DT cycle. However, environmental advantage will be the main chance for fusion to compete with other energy sources. The environmental problems of DT cycle due to tritium and neutron activation, are examined. Fusion commercial reactors could be based on alternative fuel cycles like D-He3. Advantages and disadvantages of this fuel cycle are outlined. All the technologies related with the self-breeding of tritium and the concept of breeding blanket itself may be not reactor relevant. In the frame of the Next-step studies, the potential advantages of intermediate DT devices without breeding blanket are discussed. Simplified design, lower cost, higher safety are the main ones. The problem of the source of tritium is examined. (author)

  4. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1975-11-01

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  5. On the implementation of a chain nuclear reaction of thermonuclear fusion on the basis of the p+11B process

    Science.gov (United States)

    Belyaev, V. S.; Krainov, V. P.; Zagreev, B. V.; Matafonov, A. P.

    2015-07-01

    Various theoretical and experimental schemes for implementing a thermonuclear reactor on the basis of the p+11B reaction are considered. They include beam collisions, fusion in degenerate plasmas, ignition upon plasma acceleration by ponderomotive forces, and the irradiation of a solid-state target from 11B with a proton beam under conditions of a Coulomb explosion of hydrogen microdrops. The possibility of employing ultra-short high-intensity laser pulses to initiate the p+11B reaction under conditions far from thermodynamic equilibrium is discussed. This and some other weakly radioactive thermonuclear reactions are promising owing to their ecological cleanness—there are virtually no neutrons among fusion products. Nuclear reactions that follow the p+11B reaction may generate high-energy protons, sustaining a chain reaction, and this is an advantage of the p+11B option. The approach used also makes it possible to study nuclear reactions under conditions close to those in the early Universe or in the interior of stars.

  6. Fusion energy research for ITER and beyond

    International Nuclear Information System (INIS)

    Romanelli, Francesco; Laxaaback, Martin

    2011-01-01

    The achievement in the last two decades of controlled fusion in the laboratory environment is opening the way to the realization of fusion as a source of sustainable, safe and environmentally responsible energy. The next step towards this goal is the construction of the International Thermonuclear Experimental Reactor (ITER), which aims to demonstrate net fusion energy production on the reactor scale. This paper reviews the current status of magnetic confinement fusion research in view of the ITER project and provides an overview of the main remaining challenges on the way towards the realization of commercial fusion energy production in the second half of this century. (orig.)

  7. Efforts of development on the next generation nuclear reactor in the Mitsubishi Heavy Industries, Ltd

    International Nuclear Information System (INIS)

    Mukai, Hiroshi

    2002-01-01

    At present, the Mitsubishi Heavy Industry, Ltd. (MHI) enters to development on APWR+ for a large-scale reactor, AP1000 and pebble bed modular reactor (PBMR) for middle- and small-scale one, and innovative one, under cooperation of power industries, manufacturers and institutes in and out of Japan. On APWR+, MHI occupies the most advanced position of conventional large-scale route, intends to carry out further upgrading of large capacity on a base of already developed 1500 MWe class APWR, and aims at further upgrading of economical efficiency. On the other reactor, as it becomes possible to perform value addition specific to the small-scale reactor with smaller output, it is planned to overcome its scale demerit by introducing more innovative techniques. And, on AP1000, it is intended to remove dynamic safety system by introducing a static one, to upgrade simplification of apparatus and reliability of safety system and to reduce its human factors. In addition, here was described on the next generation nuclear reactors under development. (G.K.)

  8. Multi-step chemical and radiation process for the production of gas

    International Nuclear Information System (INIS)

    O'Neal, R.D.; Leffert, C.B.; Teichmann, T.; Teitel, R.J.

    1979-01-01

    It has previously been proposed to use the radiation energy within the central reaction chamber of a thermonuclear reactor for the dissociation of water into hydrogen and oxygen in one step. However, the coefficient of recombination of pure hydrogen and oxygen at the elevated reaction chamber temperature is relatively high so that the overall yield is low. Furthermore, it is desirable to recover any unspent tritium from the reaction chamber exhaust, but separation of residual tritium from pure hydrogen in the chamber exhaust is relatively difficult. In the process provided in this patent pure carbon dioxide rather than steam is injected into the central reaction chamber of a thermonuclear reactor. Radiolysis of carbon dioxide yields carbon monoxide and pure oxygen. While the oxygen may be separated and collected at the exhaust vent of the reaction chamber, the carbon monoxide is separated and then combined with water to produce pure hydrogen and reconstituted carbon dioxide, which may be collected and recycled so that the overall closed-loop system produces pure hydrogen and oxygen at the expense of water. The efficiency of the process is high due, in large part, to the relatively low coefficient of recombination of carbon monoxide and oxygen at the reaction chamber temperature. Heat required for the reaction of carbon monoxide with water may be provided by suitable heat transfer from the heated reaction chamber. The chamber exhaust contains carbon monoxide and oxygen, so that any unburnt tritium in the exhaust stream may be easily collected and recycled to form additional pellet fuel. The carbon dioxide molecules injected into the reaction chamber provide protection for the reaction chamber walls from the deleterious effects of charged-particle and x-ray bombardment. (LL)

  9. HIGH PERFORMANCE ADVANCED TOKAMAK REGIMES FOR NEXT-STEP EXPERIMENTS

    International Nuclear Information System (INIS)

    GREENFIELD, C.M.; MURAKAMI, M.; FERRON, J.R.; WADE, M.R.; LUCE, T.C.; PETTY, C.C.; MENARD, J.E; PETRIE, T.W.; ALLEN, S.L.; BURRELL, K.H.; CASPER, T.A; DeBOO, J.C.; DOYLE, E.J.; GAROFALO, A.M; GORELOV, Y.A; GROEBNER, R.J.; HOBIRK, J.; HYATT, A.W; JAYAKUMAR, R.J; KESSEL, C.E; LA HAYE, R.J; JACKSON, G.L; LOHR, J.; MAKOWSKI, M.A.; PINSKER, R.I.; POLITZER, P.A.; PRATER, R.; STRAIT, E.J.; TAYLOR, T.S; WEST, W.P.

    2003-01-01

    OAK-B135 Advanced Tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic (MHD) stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half-radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding (ELMing) H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts

  10. Next generation reactor development activity at Hitachi, Ltd

    International Nuclear Information System (INIS)

    Yamashita, Junichi

    2005-01-01

    Developments of innovative nuclear systems in Japan have been highly requested to cope with uncertain future nuclear power generation and fuel cycle situation. Next generation reactor system shall be surely deployed earlier to be capable to provide with several options such as plutonium multi-recycle, intermediate storage of spent fuels, simplified reprocessing of spent fuels and separated storage of 'Pu+FP' and 'U', spent fuels storage after Pu LWR recycle and their combinations, while future reactor system will be targeted at ideal fuel recycle system of higher breeding gain and transmutation of radioactive wastes. Modified designs of the ABWR at large size and medium and small size have been investigated as well as a BWR based RMWR and a supercritical-pressure LWR to ensure safety and improve economics. Advanced fuel cycle technologies of a combination of fluoride volatility process and PUREX process with high decontamination (FLUOREX process) and a modified fluoride volatility process with low decontamination have been developed. (T. Tanaka)

  11. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  12. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  13. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  14. Next Steps for "Big Data" in Education: Utilizing Data-Intensive Research

    Science.gov (United States)

    Dede, Chris

    2016-01-01

    Data-informed instructional methods offer tremendous promise for increasing the effectiveness of teaching, learning, and schooling. Yet-to-be-developed data science approaches have the potential to dramatically advance instruction for every student and to enhance learning for people of all ages. Next steps that emerged from a recent National…

  15. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

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

  16. Study on dual plant concept for the next generation boiling water reactors

    International Nuclear Information System (INIS)

    Sato, Takashi; Oikawa, Hirohide

    1999-01-01

    The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

  17. Annual Report 1991. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1992-01-01

    In 1991 the operation of the High Flux Reactor was carried out as planned. The availability was more than 100% of scheduled operating time. The average utilization of the reactor was 69% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. Development activities addressed upgrading of irradiation devices, neutron capture therapy, neutron radiography and neutron transmutation doping of silicon. General activities in support of running irradiation programmes progressed in the normal way

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

    International Nuclear Information System (INIS)

    Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

    1987-09-01

    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)

  19. 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)

    1995-07-01

    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.)

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

    International Nuclear Information System (INIS)

    Linke, J; Pintsuk, G.; Rödig, M.

    2013-01-01

    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

  1. Thermonuclear Runaway model

    International Nuclear Information System (INIS)

    Sparks, W.M.; Kutter, G.S.; Starrfield, S.; Truran, J.W.

    1989-01-01

    The nova outburst requires an energy source that is energetic enough to eject material and is able to recur. The Thermonuclear Runaway (TNR) model, coupled with the binary nature of nova systems satisfies these conditions. The white dwarf/red dwarf binary nature of novae was first recognized as a necessary conditions by Kraft. The small separation characteristic of novae systems allows the cool, red secondary to overflow is Roche lobe. In the absence of strong, funneling magnetic fields, the angular momentum of this material prevents it from falling directly onto the primary, and it first forms a disk around the white dwarf. This material is eventually accreted from the disk onto the white dwarf. As the thickness of this hydrogen-rich layer increases, the degenerate matter at the base reaches a temperature that is high enough to initiate thermonuclear fusion of hydrogen. Thermonuclear energy release increases the temperature which in turn increases the energy generation rate. Because the material is degenerate, the pressure does not increase with temperature, which normally allows a star to adjust itself to a steady nuclear burning rate. Thus the temperature and nuclear energy generation increase and a TNR results. When the temperature reaches the Fermi temperature, degeneracy is lifted and the rapid pressure increase causes material expansion. The hydrogen-rich material either is ejected or consumed by nuclear burning, and the white dwarf returns to its pre-outburst state. The external source of hydrogen fuel from the secondary allows the while process to repeat. 43 refs., 8 figs

  2. Merging White Dwarfs and Thermonuclear Supernovae

    OpenAIRE

    van Kerkwijk, Marten H.

    2012-01-01

    Thermonuclear supernovae result when interaction with a companion reignites nuclear fusion in a carbon-oxygen white dwarf, causing a thermonuclear runaway, a catastrophic gain in pressure, and the disintegration of the whole white dwarf. It is usually thought that fusion is reignited in near-pycnonuclear conditions when the white dwarf approaches the Chandrasekhar mass. I briefly describe two long-standing problems faced by this scenario, and our suggestion that these supernovae instead resul...

  3. International Thermonuclear Experimental Reactor U.S. Home Team Quality Assurance Plan

    Energy Technology Data Exchange (ETDEWEB)

    Sowder, W. K.

    1998-10-01

    The International Thermonuclear Experimental Reactor (ITER) project is unique in that the work is divided among an international Joint Central Team and four Home Teams, with the overall responsibility for the quality of activities performed during the project residing with the ITER Director. The ultimate responsibility for the adequacy of work performed on tasks assigned to the U.S. Home Team resides with the U.S. Home Team Leader and the U.S. Department of Energy Office of Fusion Energy (DOE-OFE). This document constitutes the quality assurance plan for the ITER U.S. Home Team. This plan describes the controls exercised by U.S. Home Team management and the Performing Institutions to ensure the quality of tasks performed and the data developed for the Engineering Design Activities assigned to the U.S. Home Team and, in particular, the Research and Development Large Projects (7). This plan addresses the DOE quality assurance requirements of 10 CFR 830.120, "Quality Assurance." The plan also describes U.S. Home Team quality commitments to the ITER Quality Assurance Program. The ITER Quality Assurance Program is based on the principles described in the International Atomic Energy Agency Standard No. 50-C-QA, "Quality Assurance for Safety in Nuclear Power Plants and Other Nuclear Facilities." Each commitment is supported with preferred implementation methodology that will be used in evaluating the task quality plans to be submitted by the Performing Institutions. The implementing provisions of the program are based on guidance provided in American National Standards Institute/American Society of Mechanical Engineers NQA-1 1994, "Quality Assurance." The individual Performing Institutions will implement the appropriate quality program provisions through their own established quality plans that have been reviewed and found to comply with U.S. Home Team quality assurance plan commitments to the ITER Quality Assurance Program. The extent of quality program provisions

  4. Introduction to magnetic fusion reactor design

    International Nuclear Information System (INIS)

    Watanabe, Kenji

    1988-01-01

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

  5. Development of a Ground Test and Analysis Protocol to Support NASA's NextSTEP Phase 2 Habitation Concepts

    Science.gov (United States)

    Beaton, Kara H.; Chappell, Steven P.; Bekdash, Omar S.; Gernhardt, Michael L.

    2018-01-01

    The NASA Next Space Technologies for Exploration Partnerships (NextSTEP) program is a public-private partnership model that seeks commercial development of deep space exploration capabilities to support extensive human spaceflight missions around and beyond cislunar space. NASA first issued the Phase 1 NextSTEP Broad Agency Announcement to U.S. industries in 2014, which called for innovative cislunar habitation concepts that leveraged commercialization plans for low Earth orbit. These habitats will be part of the Deep Space Gateway (DSG), the cislunar space station planned by NASA for construction in the 2020s. In 2016, Phase 2 of the NextSTEP program selected five commercial partners to develop ground prototypes. A team of NASA research engineers and subject matter experts have been tasked with developing the ground test protocol that will serve as the primary means by which these Phase 2 prototype habitats will be evaluated. Since 2008, this core test team has successfully conducted multiple spaceflight analog mission evaluations utilizing a consistent set of operational products, tools, methods, and metrics to enable the iterative development, testing, analysis, and validation of evolving exploration architectures, operations concepts, and vehicle designs. The purpose of implementing a similar evaluation process for the NextSTEP Phase 2 Habitation Concepts is to consistently evaluate the different commercial partner ground prototypes to provide data-driven, actionable recommendations for Phase 3.

  6. Investigations in the area of thermonuclear structural material science in the Republic of Kazakhstan

    International Nuclear Information System (INIS)

    Tazhibayeva, I.; Shestakov, V.; Cherepnin, Yu.S.

    2001-01-01

    The investigations in the area of structural materials for fusion program initiated within the framework of ITER project in the Republic of Kazakhstan are devoted basically in the following direction: to studying the behaviour of hydrogen isotopes in structural elements of the first wall and the divertor in conditions simulating real conditions of material operation, accident situations arising during steam interaction with the beryllium armour of the first wall during accidental coolant loss, to establish an experimental facility for study aspects of tritium safety of thermonuclear installations, for example, levels of tritium accumulation and release; efficiency of barrier layers and protective coating; influence of brazing and welding zones on tritium permeation. The work on determination of tritium release from lead/lithium eutectic alloy by mass-spectrometry method and the development of permeation barriers has begun. At present, work has begun to create Kazakhstan's own tokamak type reactor for investigation of the behaviour of various first wall materials and divertor plates during normal and accident conditions. The concept of spherical tokamak will be used in the construction of KTM reactor. (author)

  7. Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

  8. Proceedings of the international workshop on engineering design of next step reversed field pinch devices

    International Nuclear Information System (INIS)

    Thomson, D.B.

    1987-11-01

    These Proceedings contain the formal contributed papers, the workshop papers and workshop summaries presented at the International Workshop on Engineering Design of Next Step RFP Devices held at Los Alamos, July 13-17, 1987. Contributed papers were presented at formal sessions on the topics: (1) physics overview (3 papers); (2) general overview (3 papers); (3) front-end (9 papers); (4) computer control and data acquisition (1 paper); (5) magnetics (5 papers); and (6) electrical design (9 papers). Informal topical workshop sessions were held on the topics: (1) RFP physics (9 papers); (2) front-end (7 papers); (3) magnetics (3 papers); and (4) electrical design (1 paper). This volume contains the summaries written by the Chairmen of each of the informal topical workshop sessions. The papers in these Proceedings represent a significant review of the status of the technical base for the engineering design of the next step RFP devices being developed in the US, Europe, and Japan, as of this date

  9. Proceedings of the international workshop on engineering design of next step reversed field pinch devices

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, D.B. (comp.)

    1987-11-01

    These Proceedings contain the formal contributed papers, the workshop papers and workshop summaries presented at the International Workshop on Engineering Design of Next Step RFP Devices held at Los Alamos, July 13-17, 1987. Contributed papers were presented at formal sessions on the topics: (1) physics overview (3 papers); (2) general overview (3 papers); (3) front-end (9 papers); (4) computer control and data acquisition (1 paper); (5) magnetics (5 papers); and (6) electrical design (9 papers). Informal topical workshop sessions were held on the topics: (1) RFP physics (9 papers); (2) front-end (7 papers); (3) magnetics (3 papers); and (4) electrical design (1 paper). This volume contains the summaries written by the Chairmen of each of the informal topical workshop sessions. The papers in these Proceedings represent a significant review of the status of the technical base for the engineering design of the next step RFP devices being developed in the US, Europe, and Japan, as of this date.

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

    International Nuclear Information System (INIS)

    Hovingh, J.

    1976-09-01

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

  11. Thermonuclear model for high energy transients

    International Nuclear Information System (INIS)

    Woosley, S.E.

    1982-01-01

    The thermonuclear model for x- and γ-ray bursts is discussed. Different regimes of nuclear burning are reviewed, each appropriate to a given range of (steady state) accretion rate. Accretion rates in the range 10 -14 to 10 -8 Msub solar y -1 all appear capable of producing x-ray transients of various durations and intervals. Modifications introduced by radiatively driven mass loss, the thermal inertia of the envelope, different burning mechanisms, and two-dimensional considerations are discussed as are difficulties encountered when the thermonuclear model is confronted with observations of rapidly recurrent bursts (less than or equal to 10 min), and super-Eddington luminosities and temperatures. Results from a numerical simulation of a combined hydrogen-helium runaway initiated at pycnonuclear density are presented for the first time. The thermonuclear model for γ-ray bursts is also reviewed and updated, particularly with regard to the breakdown of the steady state hypothesis employed in previous work. Solely on the basis of nuclear instability, γ-ray bursts of various types appear possible for a very broad variety of accretion rates (approx. 10 -17 to approx. 10 -11 Msub solar y -1 ) although other considerations may restrict this range. The thermonuclear model appears capable of yielding a great diversity of high energy transient phenomena for various accretion rates, magnetic field configurations, and neutron star envelope histories

  12. Status of the US foreign research reactor spent nuclear fuel program

    International Nuclear Information System (INIS)

    Chacey, K.A.; Zeitoun, A.; Saris, E.C.

    1997-01-01

    A significant step was made in 1996 with the establishment of a new nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel. Specifically the United States will accept over a 13-year period up to 20 tonnes of spent nuclear fuel from 41 countries. Only spent fuel containing uranium enriched in the United States is covered under this policy. Since the acceptance policy took effect on 13 May 1996, the Department of Energy has undertaken a number of steps to effectively implement the policy. An implementation strategy plan, mitigation action plan, and detailed transportation plans have been developed. Other activities include foreign research reactor assessments, and the determination of shipment priorities and schedules. The first shipment under the acceptance policy was received into the United States in September 1996. A second shipment was received from Canada in December 1996. The next shipment of foreign research reactor spent nuclear fuel is expected from Europe in early March 1997. The primary challenge for DOE is to continue to transport this material in a consistent, cost-effective manner over the 13-year duration of the program. This article covers the following topics: background; acceptance policy; implementation of the acceptance policy; next steps/closing. 6 figs

  13. The next fifty years

    International Nuclear Information System (INIS)

    Scaldwell, Reg

    1995-01-01

    The General Manager of the reactor division of Centronics, a world class manufacturer of reactor control systems and radiation detectors, describes its achievements and looks forward to speculate on what the next fifty years may mean in terms of technological development. Tables are given of the locations of Centronics Fission Chambers and of reactors controlled with Centronic Detectors. (UK)

  14. AECL's advanced CANDU reactor - the ACR

    International Nuclear Information System (INIS)

    Alizadeh, Ala; Allsop, Peter; Hedges, Ken; Hopwood, Jerry; Yu, Stephen

    2003-01-01

    The ACR, the next generation CANDU design, represents the next step in development of the CANDU family of designs. AECL has achieved significant incremental improvements to the mid-size CANDU 6 nuclear power plant through successive projects, both in design and in project delivery. Building on this knowledge base, AECL is continuing to adapt the CANDU design to develop the ACR. This paper summarizes the ACR design features, which include major improvements in economics, inherent safety characteristics, performance and construction methods. Aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs, the ACR is an evolutionary design based on the very successful CANDU 6 reactor. The new ACR product is specifically designed to produce power at a cost competitive with other forms of power generation while achieving short construction times, improved safety, international licensability, high investor returns, and low investor risk. It achieves these targets by taking advantage of the latest advances in both pressure-tube and pressure-vessel reactor technologies and experience. The flexibility and development potential of the fuel channel approach also enables designs to be developed that address priorities identified in international long-term specification programs such as the US Department of Energy (DOE) sponsored Generation IV program and IAEA hosted INPRO program. ACR-700 can be built in 36 months with a 48 month project duration, and deliver a lifetime capacity factor in excess of 90%. Overall, the ACR design represents a balance of proven design basis and innovations to give step improvements in safety, reliability and economics. The ACR development program, now well into the detail design stage, includes parallel formal licensing in the USA and Canada. Based on the status of the ACR design and AECL's on-going experience delivering reactor projects on-time and on-budget, the first ACR could be in service by

  15. Thermonuclear plasma physic: inertial confinement fusion

    International Nuclear Information System (INIS)

    Bayer, Ch.; Juraszek, D.

    2001-01-01

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

  16. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

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

  17. Cryogenic instrumentation needs in the controlled thermonuclear research program

    International Nuclear Information System (INIS)

    Walstrom, P.L.

    1976-01-01

    The magnet development effort for the controlled thermonuclear research program will require extensive testing of superconducting coils at various sizes from small-scale models to full-size prototypes. Extensive use of diagnostic instrumentation will be required and to make detailed comparisons of predicted and actual performance in magnet tests and to monitor the test facility for incipient failure modes. At later stages of the program, cryogenic instrumentation will be required to monitor magnet system performance in fusion power reactors. Measured quantities may include temperature, strain, deflection, coil resistance, helium coolant pressure and flow, current, voltages, etc. The test environment, which includes high magnetic fields (up to 8-10 T) and low temperature, makes many commercial measuring devices inoperative or at least inaccurate. In order to ensure reliable measurements, careful screening of commercial devices for performance in the test environment will be required. A survey of potentially applicable instrumentation is presented along with available information on operation in the test environment based on experimental data or on analysis of the physical characteristics of the device. Areas where further development work is needed are delineated

  18. Safety design criteria for the next generation Sodium-cooled fast reactors based on lessons learned from the Fukushima NPS accident

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2012-01-01

    In this presentation, architecture of the safety design criteria as requirements for SFR system and the activities on safety research works to establish safety evaluation methods for the next generation SFRs are summarized with the basis on lessons learned from the Fukushima NPS accident. Nuclear safety is a grovel issue which should be achieved by the international cooperation. In respect of the development for the next generation reactor, it is necessary to build the harmonized safety criteria and evaluation methods to establish the next level of safety

  19. Blankets for thermonuclear device

    International Nuclear Information System (INIS)

    Maki, Koichi; Fukumoto, Hideshi.

    1986-01-01

    Purpose: To produce tritium more than consumed, through thermonuclear reaction. Constitution: The energy spectrum of neutron generated by neutron multiplying reaction in a neutron multiplying blanket and moderated neutrons has a large ratio in a low energy section. In the low-energy absorption region of stainless steel which is a material of cooling pipes constituting a neutron multiplying blanket cooling channel, the neutrons are absorbed, lessening the neutron multiplying effect. To prevent this, the neutron multiplying blanket cooling channel is covered with tritium breeding blankets, thereby enabling the production of a substantially great amount of tritium more than the amount of tritium to be consumed by the thermonuclear reaction by preventing neutron absorption by the component materials of the cooling channel, improving the tritium breeding ratio by 20 to 25 %, and increasing the efficiency of use of neutrons for tritium generation. (Horiuchi, T.)

  20. The laser thermonuclear fusion

    International Nuclear Information System (INIS)

    Coutant, J.; Dautray, R.; Decroisette, M.; Watteau, J.P.

    1987-01-01

    Principle of the thermonuclear fusion by inertial confinement: required characteristics of the deuterium-tritium plasma and of the high power lasers to be used Development of high power lasers: active media used; amplifiers; frequency conversion; beam quality; pulse conditioning; existing large systems. The laser-matter interaction: collision and collective interaction of the laser radiation with matter; transport of the absorbed energy; heating and compression of deuterium-tritium; diagnoses and their comparison with the numerical simulation of the experiment; performances. Conclusions: difficulties to overcome; megajoule lasers; other energy source: particles beams [fr

  1. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1975-02-01

    This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

  2. Shielding wall for thermonuclear device

    International Nuclear Information System (INIS)

    Uchida, Takaho.

    1989-01-01

    This invention concerns shielding walls opposing to plasmas of a thermonuclear device and it is an object thereof to conduct reactor operation with no troubles even if a portion of shielding wall tiles should be damaged. That is, the shielding wall tiles are constituted as a dual layer structure in which the lower base tiles are connected by means of bolts to first walls. Further, the upper surface tiles are bolt-connected to the layer base tiles. In this structure, the plasma thermal loads are directly received by the surface layer tiles and heat is conducted by means of conduction and radiation to the underlying base tiles and the first walls. Even upon occurrence of destruction accidents to the surface layer tiles caused by incident heat or electromagnetic force upon elimination of plasmas, since the underlying base tiles remain as they are, the first walls constituted with stainless steels, etc. are not directly exposed to the plasmas. Accordingly, the integrity of the first walls having cooling channels can be maintained and sputtering intrusion of atoms of high atom number into the plasmas can be prevented. (I.S.)

  3. The ''Dolphin'' power laser installation for spherical thermonuclear target heating

    International Nuclear Information System (INIS)

    Basov, N.G.; Bykovskij, N.E.; Danilov, A.E.

    1978-01-01

    12-channel laser installation the ''Dolphin'' for thermonuclear target heating in the radiation spheric geometry has been developed to carry out series of physical investigations of laser-thermonuclear plasma system, optimization of target heating conditions and obtaining a comparatively large value of thermonuclear output in ratio to the energy of absorbed light radiation in the target. The description of installation main elements, consisting of the following components, is given: 1)neodymium laser with the maximum permissible radiation energy of 10kJ, with light pulse duration of 10 -10 /10 -9 c and radiation divergence of approximately 5x10 -4 rad; 2)vacuum chamber, where laser radiation interaction with plasma takes place; 3)diagnostic means of laser and plasma parameters and 4)focus system. The focus system provides a high degree of target spherical radiation symmetry at current maximum density on its surface of approximately 10 15 W/cm 2

  4. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

  5. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

  6. Deposition Diagnostics for Next-step Devices

    International Nuclear Information System (INIS)

    Skinner, C.H.; Roquemore, A.L.; Bader, A.; Wampler, W.R.

    2004-01-01

    The scale-up of deposition in next-step devices such as ITER will pose new diagnostic challenges. Codeposition of hydrogen with carbon needs to be characterized and understood in the initial hydrogen phase in order to mitigate tritium retention and qualify carbon plasma facing components for DT operations. Plasma facing diagnostic mirrors will experience deposition that is expected to rapidly degrade their reflectivity, posing a new challenge to diagnostic design. Some eroded particles will collect as dust on interior surfaces and the quantity of dust will be strictly regulated for safety reasons - however diagnostics of in-vessel dust are lacking. We report results from two diagnostics that relate to these issues. Measurements of deposition on NSTX with 4 Hz time resolution have been made using a quartz microbalance in a configuration that mimics that of a typical diagnostic mirror. Often deposition was observed immediately following the discharge suggesting that diagnostic shutters should be closed as soon as possible after the time period of interest. Material loss was observed following a few discharges. A novel diagnostic to detect surface particles on remote surfaces was commissioned on NSTX

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

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1991-06-01

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

  8. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    International Nuclear Information System (INIS)

    Beck, J.M.; Collins, J.W.; Garcia, C.B.; Pincock, L.F.

    2010-01-01

    High Temperature Gas Reactors (HTGR) have been designed and operated throughout the world over the past five decades. These seven HTGRs are varied in size, outlet temperature, primary fluid, and purpose. However, there is much the Next Generation Nuclear Plant (NGNP) has learned and can learn from these experiences. This report captures these various experiences and documents the lessons learned according to the physical NGNP hardware (i.e., systems, subsystems, and components) affected thereby.

  9. Rates of Thermonuclear Reactions in Dense Plasmas

    International Nuclear Information System (INIS)

    Tsytovich, V.N.; Bornatici, M.

    2000-01-01

    The problem of plasma screening of thermonuclear reactions has attracted considerable scientific interest ever since Salpeter's seminal paper, but it is still faced with controversial statements and without any definite conclusion. It is of relevant importance to thermonuclear reactions in dense astrophysical plasmas, for which charge screening can substantially affect the reaction rates. Whereas Salpeter and a number of subsequent investigations have dealt with static screening, Carraro, Schafer, and Koonin have drawn attention to the fact that plasma screening of thermonuclear reactions is an essentially dynamic effect. In addressing the issue of collective plasma effects on the thermonuclear reaction rates, the first critical overview of most of the work carried out so far is presented and the validity of the test particle approach is assessed. In contrast to previous investigations, we base our description on the kinetic equation for nonequilibrium plasmas, which accounts for the effects on the rates of thermonuclear reactions of both plasma fluctuations and screening and allows one to analyze explicitly the effects of the fluctuations on the reaction rates. Such a kinetic formulation is more general than both Salpeter's approach and the recently developed statistical approaches and makes it possible to obtain a more comprehensive understanding of the problem. A noticeable result of the fluctuation approach is that the static screening, which affects both the interaction and the self-energy of the reacting nuclei, does not affect the reaction rates, in contrast with the results obtained so far. Instead, a reduction of the thermonuclear reaction rates is obtained as a result of the effect of plasma fluctuations related to the free self-energy of the reacting nuclei. A simple physical explanation of the slowing down of the reaction rates is given, and the relation to the dynamically screened test particle approach is discussed. Corrections to the reaction rates

  10. Research programme on controlled thermonuclear fusion - Synthesis report 2010

    International Nuclear Information System (INIS)

    Vaucher, C.; Tran, M. Q.; Villard, L.; Marot, L.

    2011-01-01

    Since 1978, research on thermonuclear fusion in Switzerland is closely related to the research programme of the European Atomic Energy Community (EURATOM). The Swiss projects tackle aspects of plasma physics and fusion technology. Switzerland participates to the construction and operation of the Joint European Torus (JET). The International Thermonuclear Experimental Reactor (ITER) is being built; the first plasma is expected in 2019. The 'Centre de Recherches en Physique des Plasmas' (CRPP) of the EPFL participates to EURATOM scientific and technological projects in magnetic confinement physics, through an experimental contribution (the Variable Configuration Tokamak, TCV) and theoretical studies. Thanks to the large flexibility of the TCV design and operation modus, plasmas of different shapes can be created and controlled, what is a very useful option to verify numerical simulation results. Besides, the injection of millimetre waves allows directing the injected power according to specific profiles. A configuration of type 'snowflakes' could be created, reducing the power deposition at the edge of the plasma. Theoretical studies on turbulence have improved the plasma stability in the TCV. For the first time in the world, TCV could reach a stable plasma, the plasma current being generated using the so-called 'bootstrap' phenomenon. Besides turbulence, studies were focused on heat and particle transport in tokamaks, on an analysis of the equilibrium and magneto-hydrodynamic stability of tokamaks and stellarators, on the application of radiofrequency waves and on the optimization of new confinement configurations. Experiments in the JET facility confirmed the numerical results of theoretical simulations. The TORPEX facility, which is simpler than TCV, allows high space-temporal resolution measurements for the study of turbulences and plasma threads ('blobs'). At the Paul Scherrer Institute (PSI), research topics include superconductivity and materials. The Fusion

  11. Performance of cable-in-conduit conductors in ITER [International Thermonuclear Experimental Reactor] toroidal field coils with varying heat loads

    International Nuclear Information System (INIS)

    Kerns, J.A.; Wong, R.L.

    1989-01-01

    The toroidal field (TF) coils in the International Thermonuclear Experimental Reactor (ITER) will operate with varying heat loads generated by ac losses and nuclear heating. The total heat load is estimated to be 2 kW per TF coil under normal operation and can be higher for different operating scenarios. Ac losses are caused by ramping the poloidal field (PF) for plasma initiation, burn, and shutdown; nuclear heating results from neutrons that penetrate into the coil past the shield. Present methods to reduce or eliminate these losses lead to larger and more expensive machines, which are unacceptable with today's budget constraints. A suitable solution is to design superconductors that operate with high heat loads. The cable-in-conduit conductor (CICC) can operate with high heat loads. One CICC design is analyzed for its thermal performance using two computer codes developed at LLNL. One code calculates the steady state flow conditions along the flow path, while the other calculates the transient conditions in the flow. We have used these codes to analyze the superconductor performance during the burn phase of the ITER plasma. The results of these analyses give insight to the choice of flow rate on superconductor performance. 4 refs., 5 figs

  12. An assessment the severe accident equipment survivability for the Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Lee, B. C.; Moon, Y. T.; Park, J. W.; Kho, H. J.; Lee, S. W.

    1999-01-01

    One of the prominent design approaches to cope with the severe accident challenges in the Korean Next Generation Reactor is an assessment of equipment survivability in the severe accident environment at early design stage. In compliance with 10CFR50.34(f) and SECY-93-087, this work addresses that a reasonable level of assurance be provided to demonstrate that sufficient instrumentation and equipment will survive the consequences of a severe accident and will be available so that the operator may recover from and trend severe core damage sequences, including those scenarios which result in 100 percent oxidation of the active fuel cladding. An analytical and systematic approach was used to identify the equipment and instrumentation of safety-function and define severe accident environments including temperature, pressure, humidity, and radiation before and after the reactor vessel breach. As a result, it was concluded that with minor exceptions, existing design basis equipment qualification methods are sufficient to provide a reasonable level of assurance that this equipment will function during a severe accident. Furthermore, supplemental severe accident equipment and instrument procurement requirements were identified. (author)

  13. Global seismic inversion as the next standard step in the processing sequence

    Energy Technology Data Exchange (ETDEWEB)

    Maver, Kim G.; Hansen, Lars S.; Jepsen, Anne-Marie; Rasmussen, Klaus B.

    1998-12-31

    Seismic inversion of post stack seismic data has until recently been regarded as a reservoir oriented method since the standard inversion techniques rely on extensive well control and a detailed user derived input model. Most seismic inversion techniques further requires a stable wavelet. As a consequence seismic inversion is mainly utilised in mature areas focusing of specific zones only after the seismic data has been interpreted and is well understood. By using an advanced 3-D global technique, seismic inversion is presented as the next standard step in the processing sequence. The technique is robust towards noise within the seismic data, utilizes a time variant wavelet, and derives a low frequency model utilizing the stacking velocities and only limited well control. 4 figs.

  14. On fire risk/methodology for the next generation of reactors and nuclear facilities

    International Nuclear Information System (INIS)

    Majumdar, K.C.; Alesso, H.P.; Altenbach, T.J.

    1992-01-01

    Methodologies for including fire in probabilistic risk assessments (PRAs) have been evolving during the last ten years. Many of these studies show that fire risk constitutes a significant percentage of external events, as well as the total core damage frequency. This paper summarizes the methodologies used in the fire risk analysis of the next generation of reactors and existing DOE nuclear facilities. Methodologies used in other industries, as well as existing nuclear power plants, are also discussed. Results of fire risk studies for various nuclear plants and facilities are shown and compared

  15. Compact toroid refueling of reactors

    International Nuclear Information System (INIS)

    Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

    1988-04-01

    The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

  16. Research programme on controlled thermonuclear fusion. Synthesis report 2011

    International Nuclear Information System (INIS)

    Vaucher, C.; Tran, M. Q.; Villard, L.; Marot, L.

    2012-01-01

    Since 1978, research on thermonuclear fusion in Switzerland is closely related to the research programme of the European Atomic Energy Community (EURATOM). The Swiss projects tackle aspects of plasma physics and fusion technology. Switzerland participates to the construction and operation of the Joint European Torus (JET), which started operation again in 2011. The International Thermonuclear Experimental Reactor (ITER) is the last step before DEMO, a prototype fusion reactor able to deliver electricity and demonstrate the economic viability of fusion energy. The 'Centre de Recherches en Physique des Plasmas' (CRPP) of the EPFL went on with its participation to the scientific and technological programme of EURATOM. Researches are carried out essentially on 2 sites: (i) at EPFL, where topics dealt with include the physics of magnetic confinement studied using the Variable Configuration Tokamak (TCV), the basic experiment TORPEX, theory and numerical modelling, and the technology of plasma heating and current generation by hyper-frequency waves; (ii) at the Paul Scherrer Institute (PSI), where activities are devoted to superconductivity and structure materials. Thanks to the large flexibility of the TCV design and operation modus, plasmas of different shapes can be created and controlled, what is a very useful option to verify numerical simulation results. Besides, the injection of millimetre waves allows directing the injected power according to specific profiles. In the TCV it could be demonstrated for the first time that the injection of Electronic Cyclotronic Heating (ECH) waves is able to double the frequency of so-called 'Edge Localized Modes' (ELM), reducing by a factor of 2 the energy expelled by each ELM. In particular, it was possible to considerably reduce the statistical dispersion of the repetition frequency of ELM, and to avoid the appearance of gigantic ELM that are particularly harmful for reactor operation. The effect of plasma internal relaxation

  17. Inertia thermonuclear device

    International Nuclear Information System (INIS)

    Madarame, Haruki; Nakamura, Norio; Oomura, Hiroshi.

    1983-01-01

    Purpose: To enable effective recovery of the thermonuclear reaction energy and effective protection of a cylinder metal against thermal destruction by forming a uniform and stable liquid metal wall to the inside of a cylindrical member. Constitution: Cylindrical body having a lateral axis is rotatably supported so that a liquid metal wall for use in the wet wall type thermonuclear device is formed centrifugally. A liquid metal injection port for injecting the liquid metal to the cylindrical member is disposed to the lateral axis and a liquid metal exit for flowing out the injected liquid metal is disposed to the body of the cylindrical member, so as to form a moving liquid metal layer flowing from the injection port through the inner circumferential surface of the cylindrical member to the liquid metal exit port. Then, the liquid metal is centrifugally forced to the inner surface of the cylindrical body to form a uniform and stable liquid metal wall at the inner surface of the cylindrical body, whereby the reaction energy can effectively be recovered and the cylinder metal can effectively be protected against thermal destruction. (Yoshihara, H.)

  18. Thermonuclear investigation development

    International Nuclear Information System (INIS)

    Pistunovich, V.I.; Solov'ev, N.S.

    1975-01-01

    The patent situation, based mainly on a study of the situations of Great Britain, USA, France, Federal Republic of Germany and Japan from 1958 to 1974 is reviewed. Applicants have obtained around 300 patents on equipment for control of thermonuclear reactions. In the second half some decrease in the introduction of patents on high-temperature-plasma studies is noted. Multipole magnet systems for holding plasma and toroidal equipment of the takamak type have been developed recently. In the 70s, patents were published on the use of high-energy electrons for stabilization and heating of plasma in toroidal stationary systems. Starting with the mid 60s, considerable attention has been given to heating of plasma with laser radiation and to conversion of thermonuclear energy to electrical. There are 20 domestic patents on laser heating of plasma, and 75 and 45 domestic patents, respectively, on open and composite traps and 120 and 40 such patents abroad. While in the 60s equipment of different types was patented in many directions, part of which has not found further use, today work abroad is being patented basically on laser heating of plasma, toroidal magnetic systems, ion beam interference, and plasma bunching

  19. Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

    Science.gov (United States)

    Rachkov, V. I.; Arnol'dov, M. N.; Efanov, A. D.; Kalyakin, S. G.; Kozlov, F. A.; Loginov, N. I.; Orlov, Yu. I.; Sorokin, A. P.

    2014-05-01

    By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.

  20. Localized thermonuclear runaways and volcanoes on degenerate dwarf stars

    Energy Technology Data Exchange (ETDEWEB)

    Shara, M.M.

    1982-10-15

    Practically all studies to date of thermonuclear runaways on degenerate dwarf stars in binary systems have considered only spherically symmetric eruptions. We emphasize that even slightly non-spherically symmetric accretion leads to transverse temperature gradients in the dwarfs' accreted envelopes. Over a rather broad range of parameter space, thermalization time scales in accreted envelopes are much longer than thermonuclear runaway time scales. Thus localized thermonuclear runaways (i.e., runaways much smaller than the host degenerate star) rather than spherically symmetric global eruptions are likely to occur on many degenerate dwarfs. Localized runaways are more likely to occur on more massive and/or hotter dwarfs.

  1. Radiation environment of fusion experimental reactor

    International Nuclear Information System (INIS)

    Mori, Seiji; Seki, Yasushi

    1988-01-01

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

  2. Merging white dwarfs and thermonuclear supernovae.

    Science.gov (United States)

    van Kerkwijk, M H

    2013-06-13

    Thermonuclear supernovae result when interaction with a companion reignites nuclear fusion in a carbon-oxygen white dwarf, causing a thermonuclear runaway, a catastrophic gain in pressure and the disintegration of the whole white dwarf. It is usually thought that fusion is reignited in near-pycnonuclear conditions when the white dwarf approaches the Chandrasekhar mass. I briefly describe two long-standing problems faced by this scenario, and the suggestion that these supernovae instead result from mergers of carbon-oxygen white dwarfs, including those that produce sub-Chandrasekhar-mass remnants. I then turn to possible observational tests, in particular, those that test the absence or presence of electron captures during the burning.

  3. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Yu, S.; Pakan, M.; Soulard, M.

    2002-01-01

    AECL has developed a suite of technologies for Candu R reactors that enable the next step in the evolution of the Candu family of heavy-water-moderated fuel-channel reactors. These technologies have been combined in the design for the Advanced Candu Reactor TM1 (ACRTM), AECL's next generation Candu power plant. The ACR design builds extensively on the existing Candu experience base, but includes innovations, in design and in delivery technology, that provide very substantial reductions in capital cost and in project schedules. In this paper, main features of next generation design and delivery are summarized, to provide the background basis for the cost and schedule reductions that have been achieved. In particular the paper outlines the impact of the innovative design steps for ACR: - Selection of slightly enriched fuel bundle design; - Use of light water coolant in place of traditional Candu heavy water coolant; - Compact core design with unique reactor physics benefits; - Optimized coolant and turbine system conditions. In addition to the direct cost benefits arising from efficiency improvement, and from the reduction in heavy water, the next generation Candu configuration results in numerous additional indirect cost benefits, including: - Reduction in number and complexity of reactivity mechanisms; - Reduction in number of heavy water auxiliary systems; - Simplification in heat transport and its support systems; - Simplified human-machine interface. The paper also describes the ACR approach to design for constructability. The application of module assembly and open-top construction techniques, based on Candu and other worldwide experience, has been proven to generate savings in both schedule durations and overall project cost, by reducing premium on-site activities, and by improving efficiency of system and subsystem assembly. AECL's up-to-date experience in the use of 3-D CADDS and related engineering tools has also been proven to reduce both engineering and

  4. An evolutionary approach to advanced water cooled reactors

    International Nuclear Information System (INIS)

    Antariksawan, A.R.; Subki, I.

    1997-01-01

    Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP's are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR's/A-PHWR's and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complex machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative assessment of integral reactor concept. (author)

  5. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  6. Potential Next Steps for the New Orleans City Council Energy Efficiency Resolution

    Energy Technology Data Exchange (ETDEWEB)

    Doris, E.

    2011-09-01

    This document is adapted from an actual February 2008 deliverable memo and report delivered by the National Renewable Energy Laboratory (NREL) to the City Council of New Orleans, the office of the Mayor of New Orleans, the Chairperson of the Citizen Stakeholders Group (New Orleans Energy Task Force) and the Department of Energy Project Officer in February of 2008. In January 2008, the New Orleans Utility Committee requested review, commentary, and suggestions for Utility Committee next steps related to the Energy Efficiency Resolution (the Resolution) passed by the City Council in December 2007. The suggestions are reprinted here as: (1) An illustration of opportunities for other local governments for the development and implementation of effective energy efficiency ordinances and resolutions; and (2) An example of the type of policy technical assistance that DOE/NREL provides to communities. For more information on the strategy for delivering assistance, please see: www.nrel.gov/docs/fy11osti/48689.pdf. Based on experience in other communities and energy efficiency policies and programs, NREL found the Resolution to be a solid framework for increasing the responsible use of energy efficiency and reaping the associated economic and environmental benefits in the city of New Orleans. The remainder of this document provides the requested suggestions for next steps in implementing the word and spirit of the resolution. These suggestions integrate the extensive work of other entities, including the New Orleans Mayor's office, the New Orleans Energy Advisory Committee, the Energy Efficiency Initiative, and the U.S. Environmental Protection Agency's National Action Plan for Energy Efficiency. In general, three actions were suggested for funding mechanisms, two for near-term successes, and two for longer-term success.

  7. Hybrid nuclear reactors and muon catalysis

    International Nuclear Information System (INIS)

    Petrov, Yu.

    1983-01-01

    Three methods are described of the conversion of isotope 238 U to 239 Pu by neutron capture in fast breeder reactors, in the breeding blanket of hybrid thermonuclear reactors using neutrons generated by fusion and electronuclear breeding in which the target is bombarded with 1 GeV protons. Their possible use in power production is discussed. Another prospective energy source is the use of muon catalysis in the fusion of deuterium and tritium nuclei. (J.P.)

  8. Thermal insulation layer for the vacuum containers of a thermonuclear device

    International Nuclear Information System (INIS)

    Nishikawa, Masana; Yamada, Masao; Kameari, Akihisa; Niikura, Setsuo.

    1980-01-01

    Purpose: To prevent temperature rise of a thermal insulation layer for a vacuum container of a thermonuclear device higher than allowable value when irradiated by neutron by constructing the layer of a cooling unit in thermal insulation material. Constitution: A metal plate attached with cooling pipes is buried in a thermal insulation material forming a thermal insulation layer to form the layer provided between a vacuum container of a thermonuclear device and a shield. (Yoshihara, H.)

  9. Free Modal Algebras Revisited: The Step-by-Step Method

    NARCIS (Netherlands)

    Bezhanishvili, N.; Ghilardi, Silvio; Jibladze, Mamuka

    2012-01-01

    We review the step-by-step method of constructing finitely generated free modal algebras. First we discuss the global step-by-step method, which works well for rank one modal logics. Next we refine the global step-by-step method to obtain the local step-by-step method, which is applicable beyond

  10. Thermonuclear astrophysics

    International Nuclear Information System (INIS)

    Clayton, D.D.; Woosley, S.E.

    1974-01-01

    We discuss the types of thermonuclear reactions that are of importance to stellar evolution and nucleosynthesis, with particular attention to the explosive ejection of shells of He, C, O, and Si. We present tables of the reactions important in the various burning phases, including the reason for their importance and an estimate of the value of a carefully measured rate. This format is chosen for dual purpose: (1) to clarify the nuclear needs by evaluating the importance of specific reactions within the astronomical settings and (2) by assigning a value scale for cross-section measurements

  11. Agroforestry—The Next Step in Sustainable and Resilient Agriculture

    Directory of Open Access Journals (Sweden)

    Matthew Heron Wilson

    2016-06-01

    Full Text Available Agriculture faces the unprecedented task of feeding a world population of 9 billion people by 2050 while simultaneously avoiding harmful environmental and social effects. One effort to meet this challenge has been organic farming, with outcomes that are generally positive. However, a number of challenges remain. Organic yields lag behind those in conventional agriculture, and greenhouse gas emissions and nutrient leaching remain somewhat problematic. In this paper, we examine current organic and conventional agriculture systems and suggest that agroforestry, which is the intentional combination of trees and shrubs with crops or livestock, could be the next step in sustainable agriculture. By implementing systems that mimic nature’s functions, agroforestry has the potential to remain productive while supporting a range of ecosystem services. In this paper, we outline the common practices and products of agroforestry as well as beneficial environmental and social effects. We address barriers to agroforestry and explore potential options to alter policies and increase adoption by farmers. We conclude that agroforestry is one of the best land use strategies to contribute to food security while simultaneously limiting environmental degradation.

  12. Energy market reform - lessons learned and next steps

    International Nuclear Information System (INIS)

    Doucet, G.

    2004-01-01

    This presentation will be based on the World Energy Council's recently published report, Energy Market Reform: Lessons Learned and Next Steps with Special Emphasis on the Energy Access Problems of Developing Countries. The report draws on practical lessons from past studies carried out by the World Energy Council and on current experiences on the desirable architecture of market reforms in electricity and natural gas. The approach of the study was not to further deepen the analysis or to provide technical recommendations but rather, to build a debate guided by the common thread of energy security and end-user e mpowerment , highlighting the possible areas of conflict of interest and the broad solutions that might be chosen depending on the local circumstances for different parts of the energy chains. The ambition was to identify key concerns and to initiate a debate on possible answers.(author)

  13. Serious Games for Health: Features, Challenges, Next Steps.

    Science.gov (United States)

    Blumberg, Moderators Fran C; Burke, Lauren C; Hodent, Participants Celia; Evans, Michael A; Lane, H Chad; Schell, Jesse

    2014-10-01

    As articles in this journal have demonstrated over the past 3 years, serious game development continues to flourish as a vehicle for formal and informal health education. How best to characterize a "serious" game remains somewhat elusive in the literature. Many researchers and practitioners view serious games as capitalizing on computer technology and state-of-the-art video graphics as an enjoyable means by which to provide and promote instruction and training, or to facilitate attitude change among its players. We invited four distinguished researchers and practitioners to further discuss with us how they view the characteristics of serious games for health, how those characteristics differ from those for academic purposes, the challenges posed for serious game development among players of different ages, and next steps for the development and empirical examination of the effectiveness of serious games for players' psychological and physical well-being.

  14. Decommissioning of the research reactors at the Russian Research Centre Kurchatov Institute

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoy, N.N.; Ryantsev, E.P.; Kolyadin, V.I.; Kucharkin, N.E.; Melkov, E.S.; Gorlinsky, Yu.E.; Kyznetsova, T.I.; Bulkin, B.K.

    2002-01-01

    The Kurchatov Institute is the largest research center of Russia in the field of nuclear science and engineering. It comprises more than 10 research institutes and scientific-technological complexes carrying out research work in the field of safe development of atomic engineering, controlled thermonuclear fusion, and plasma physics, nuclear physics and elementary particle physics, research reactors, radiation materials technology, solid state physics and superconductivity, molecular and chemical physics, and also perspective know-how's, information science and ecology. This report is basically devoted to the decommissioning of the research reactor installations, in particular to the reactor MR because of the volume and complexity of actions involved. (author)

  15. Towards upper power levels: thermonuclear fusion

    International Nuclear Information System (INIS)

    Vedel, Jean

    1983-01-01

    This paper is a brief introduction to the use of power lasers to achieve controlled thermonuclear fusion. After shortly describing thermonuclear fusion and the conditions of temperature, density and duration required it is showed how the laser enables such conditions to be created. The neodymium-doped glass laser NOVA that is being installed at the Livermore laboratory in the USA is described; at the time of its completion in 1984, this laser will be the most powerful in the world. In comparison, the OCTAL laser in operation at the Limeil establishment ''Centre d'Etudes'' of ''Commissariat Francais a l'Energie Atomique'' (the French atomic energy authority) is more modest; it is presented here [fr

  16. Evaluation Metrics for Intermediate Heat Exchangers for Next Generation Nuclear Reactors

    International Nuclear Information System (INIS)

    Sabharwall, Piyush; Kim, Eung Soo; Anderson, Nolan

    2011-01-01

    The Department of Energy (DOE) is working with industry to develop a next generation, high-temperature gas-cooled reactor (HTGR) as a part of the effort to supply the United States with abundant, clean, and secure energy as initiated by the Energy Policy Act of 2005 (EPAct; Public Law 109-58,2005). The NGNP Project, led by the Idaho National Laboratory (INL), will demonstrate the ability of the HTGR to generate hydrogen, electricity, and/or high-quality process heat for a wide range of industrial applications.

  17. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and 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

  19. Implications of rf current drive theory for next step steady-state tokamak design

    International Nuclear Information System (INIS)

    Schultz, J.H.

    1985-06-01

    Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor

  20. The extraordinarily beautiful physical principle of thermonuclear charge design (on the occasion of the 50th anniversary of the test of RDS-37 - the first Soviet two-stage thermonuclear charge)

    Energy Technology Data Exchange (ETDEWEB)

    Goncharov, German A [Russian Federal Nuclear Center ' All-Russian Scientific Research Institute of Experimental Physics' , Sarov, Nizhnii Novgorod Region (Russian Federation)

    2005-11-30

    On 22 November 1955, the Semipalatinsk test site saw the test of the first domestic two-stage thermonuclear RDS-37 charge. The charge operation was based on the principle of radiation implosion. The kernel of the principle consists in the radiation generated in a primary A-bomb explosion and confined by the radiation-opaque casing propagating throughout the interior casing volume and flowing around the secondary thermonuclear unit. The secondary unit experiences a strong compression under the irradiation, with a resulting nuclear and thermonuclear explosion. The RDS-37 explosion was the strongest of all those ever realized at the Semipalatinsk test site. It produced an indelible impression on the participants in the test. This document-based paper describes the genesis of the ideas underlying the RDS-37 design and reflects the critical moments in its development. The advent of RDS-37 was an outstanding accomplishment of the scientists and engineers of our country. (from the history of physics)

  1. The extraordinarily beautiful physical principle of thermonuclear charge design (on the occasion of the 50th anniversary of the test of RDS-37 - the first Soviet two-stage thermonuclear charge)

    International Nuclear Information System (INIS)

    Goncharov, German A

    2005-01-01

    On 22 November 1955, the Semipalatinsk test site saw the test of the first domestic two-stage thermonuclear RDS-37 charge. The charge operation was based on the principle of radiation implosion. The kernel of the principle consists in the radiation generated in a primary A-bomb explosion and confined by the radiation-opaque casing propagating throughout the interior casing volume and flowing around the secondary thermonuclear unit. The secondary unit experiences a strong compression under the irradiation, with a resulting nuclear and thermonuclear explosion. The RDS-37 explosion was the strongest of all those ever realized at the Semipalatinsk test site. It produced an indelible impression on the participants in the test. This document-based paper describes the genesis of the ideas underlying the RDS-37 design and reflects the critical moments in its development. The advent of RDS-37 was an outstanding accomplishment of the scientists and engineers of our country. (from the history of physics)

  2. Design requirements of instrumentation and control systems for next generation reactor

    International Nuclear Information System (INIS)

    Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju

    1994-03-01

    In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator's aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs

  3. Design requirements of instrumentation and control systems for next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-03-01

    In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator`s aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs.

  4. Neutral pumping rates for a next step tokamak ignition device

    International Nuclear Information System (INIS)

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

    1985-01-01

    Neutral pumping rates are calculated for pump-limiter and divertor options of a next step tokamak ignition device using a method that accounts for the coupled effects of neutral transport and plasma transport. For both pump limiters and divertors the plasma flow into the channel surrounding the neutralizer plate is greatly reduced by the neutral recycling. The fraction of this flow that is pumped can be large (>50%) but in general is dependent on the particular geometry and plasma conditions. It is estimated that pumping speeds greater than or approximately 10 5 L/s are adequate for the exhaust requirements in the pump-limiter and the divertor cases

  5. The gas turbine-modular helium reactor (GT-MHR), high efficiency, cost competitive, nuclear energy for the next century

    International Nuclear Information System (INIS)

    Zgliczynski, J.B.; Silady, F.A.; Neylan, A.J.

    1994-04-01

    The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a small passively safe reactor with key technology developments in the US during the last decade: large industrial gas turbines, large active magnetic bearings, and compact, highly effective plate-fin heat exchangers. The GT-MHR is the only reactor concept which provides a step increase in economic performance combined with increased safety. This is accomplished through its unique utilization of the Brayton cycle to produce electricity directly with the high temperature helium primary coolant from the reactor directly driving the gas turbine electrical generator. This cannot be accomplished with another reactor concept. It retains the high levels of passive safety and the standardized modular design of the steam cycle MHTGR, while showing promise for a significant reduction in power generating costs by increasing plant net efficiency to a remarkable 47%

  6. Operating large controlled thermonuclear fusion research facilities

    International Nuclear Information System (INIS)

    Gaudreau, M.P.J.; Tarrh, J.M.; Post, R.S.; Thomas, P.

    1987-01-01

    The MIT Tara Tandem Mirror is a large, state of the art controlled thermonuclear fusion research facility. Over the six years of its design, implementation, and operation, every effort was made to minimize cost and maximize performance by using the best and latest hardware, software, and scientific and operational techniques. After reviewing all major DOE fusion facilities, an independent DOE review committee concluded that the Tara operation was the most automated and efficient of all DOE facilities. This paper includes a review of the key elements of the Tara design, construction, operation, management, physics milestones, and funding that led to this success. The authors emphasize a chronological description of how the system evolved from the proposal stage to a mature device with an emphasis on the basic philosophies behind the implementation process. This description can serve both as a qualitative and quantitative database for future large experiment planning. It includes actual final costs and manpower spent as well as actual run and maintenance schedules, number of data shots, major system failures, etc. The paper concludes with recommendations for the next generation of facilities

  7. Towards a new generation of control and data acquisition systems for thermonuclear fusion research

    International Nuclear Information System (INIS)

    Van Haren, P.C.

    1993-01-01

    Because of the complexity of thermonuclear fusion test reactors, control systems are indispensable. The physical properties of the reactor medium, i.e. the plasma, are still not well understood. Therefore, many diagnostic techniques are applied to investigate the plasma and to discover its properties. As a consequence, data acquisition systems play an important role in thermonuclear fusion research. This thesis reports on three projects that were carried out in the field of control and data acquisition. The target experiment is the Rijnhuizen Tokamak Project (RTP), a medium-sized experiment dedicated to studies of transport in the reactor medium. One of the projects is aimed at the development of a new Plasma Position and Current Control feedback System (PPCCS). This system evaluates signals of a large (about 20) number of sensors, computes the actual state of the plasma from these signals and generates command signals for the power supplies that govern the plasma position. The most ambitious project described in this thesis is the development of a data acquisition system, called TRAMP (Transient Recorders and Amoeba Multi Processor), that aims to be a testbed for smart data acquisition strategies. TRAMP attempts to acquire and store temporarily all possible data at a high sampling frequency from a single RTP pulse, and accommodates for a resampling in software prior to transferring the data to a mass storage facility. The software resampling frequency can be tuned by analysis of the acquired data and, in that way, only interesting data will be stored. In the course of the development of both the above-mentioned systems it turned out that the existing database format applied for managing experimental data provided many hurdles in the realization of efficient solutions. Consequently, a new database format was developed together with software to deal with it. This new database, called DOM4 (Data Organization and Management), is now applied at all data acquisition

  8. Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

    International Nuclear Information System (INIS)

    Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

    1978-01-01

    The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

  9. Temperature measurements in thermonuclear plasmas

    International Nuclear Information System (INIS)

    Breton, D.

    1958-01-01

    The temperatures needed to produce thermonuclear reactions are of the order of several million degrees Kelvin. Devising methods for measuring such temperatures has been the subject of research in many countries. In order to present the problem clearly and to demonstrate its importance, the author reviews the various conditions which must be fulfilled in order that reactions may be qualified as thermonuclear. The relationship between the temperature and the cross-section of the reactions is studied, and it is shown that the notion of temperature in the plasmas is complex, which leads to a consideration of the temperature of the ions and that of the electrons. None of the methods for the temperature measurements is completely satisfactory because of the hypotheses which must be made, and which are seldom fulfilled during high-intensity discharges in the plasmas. In practice it is necessary to use several methods simultaneously. (author) [fr

  10. Dose rate in the reactor room and environment during maintenance in fusion reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

    1995-01-01

    According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

  11. Alternative divertor target concepts for next step fusion devices

    Science.gov (United States)

    Mazul, I. V.

    2016-12-01

    The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

  12. Interest in 100% MOX future reactors as seen from the fuel fabrication and from the Pu manager point of view

    International Nuclear Information System (INIS)

    Golinelli, C.; Guillet, J.L.; Nigon, J.L.

    1996-01-01

    Today, plutonium recycling in PWR type reactors has reached the industrial phase. But, on a competitive market, cost reduction can be achieved by improving fuel performances and fuel management. That is why researches on MOX future reactors are still carried out in the world and particularly in France. As a matter of fact, MOX future reactors can be more competitive if the in-reactor utilization is improved. This solution should certainly be the next step to re-use the recovered plutonium from reprocessed spent fuel. (O.M.)

  13. Occupational health physics at a fusion reactor

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  14. A two-step method for developing a control rod program for boiling water reactors

    International Nuclear Information System (INIS)

    Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift

  15. Liner of a thermonuclear pulse THETA-pinch reactor

    International Nuclear Information System (INIS)

    Baranov, G.A.; Izotov, E.N.; Karasev, B.G.; Komin, A.V.; Krivosheev, M.V.; Levashov, A.D.

    1975-01-01

    Some possible constructive solutions to the problem of fabrication of the theta-pinch reactor liner by the method of centrifugal casting in a casting mould are considered. A scheme for liner manufacturing is presented, which includes the following elements: 1) a casting mould of dielectric material presenting a hollow cylinder of 4 m in diam., 3 m in length and 12 t in weight, which rotates at 8 rps in the reactor chamber; 2) a system for heat protection of the casting mould; the volume heat of the mould is suggested to remove by gaseous helium flowing under pressure along axial cooling channels of 5 mm in diam.; the channels are evenly distributed throughout the thickness of the mould shell; 3) a system for preparation and supply of a liquid metal to the casting mould, the metal is being supplied into the casting mould from its both ends at a rate of 1.7 t of the melt per second; 4) a system for rotation of the mould, which comprises two gas turbines mounted on both ends of the mould and two main stop-radial slip supports with gas lubrication

  16. Resonant thermonuclear reaction rate

    International Nuclear Information System (INIS)

    Haubold, H.J.; Mathai, A.M.

    1986-01-01

    Basic physical principles for the resonant and nonresonant thermonuclear reaction rates are applied to find their standard representations for nuclear astrophysics. Closed-form representations for the resonant reaction rate are derived in terms of Meijer's G-function. Analytic representations of the resonant and nonresonant nuclear reaction rates are compared and the appearance of Meijer's G-function is discussed in physical terms

  17. Thermonuclear reaction listing

    International Nuclear Information System (INIS)

    Fukai, Yuzo

    1993-01-01

    The following 10 elements, including T, are well known as nuclear fusion fuels: p, D, T, 3 He, 4 He, 6 Li, 7 Li, 9 Be, 10 B, 11 B, ( 12 C, 13 C), where 12 C and 13 C are considered only in the calculation of Q value. Accordingly the number of the thermonuclear reactions is 55, and 78, if including carbon elements. The reactions have some branches. For the branches having two and three reaction products, the reaction products, Q value and threshold energy are calculated by using a computer. We have investigated those of the branches having more than three products from the papers of Ajzenberg-Selove and so on. And also, by the same papers, we check whether the above mentioned branch has been observed or not. The results are as follows: (I) the number of reactions which have Q 0 branches only with γ ray production, and Q 0 and neutron production is 36(17), and (IV) that of reactions whose branch with Q > 0 does not produce neutrons is 9(3). The value in the parentheses shows the number of the case of the carbon elements. For 55 thermonuclear reactions induced by lighter nuclides than 11 B, the reaction products, the values of Q and threshold energy, and the papers with reaction cross section data are presented in the tables. (author)

  18. Thermonuclear reaction rates. III

    International Nuclear Information System (INIS)

    Harris, M.J.; Fowler, W.A.; Caughlan, G.R.; Zimmerman, B.A.

    1983-01-01

    Stellar thermonuclear reaction rates are revised and updated, adding a number of new important reaction rates. Several reactions with large negative Q-values are included, and examples of them are discussed. The importance of the decay rates for Mg-26(p,n) exp 26 Al and Al-26(n,p) exp 26 Mg for stellar studies is emphasized. 19 references

  19. A high resolution pneumatic stepping actuator for harsh reactor environments

    Science.gov (United States)

    Tippetts, Thomas B.; Evans, Paul S.; Riffle, George K.

    1993-01-01

    A reactivity control actuator for a high-power density nuclear propulsion reactor must be installed in close proximity to the reactor core. The energy input from radiation to the actuator structure could exceed hundreds of W/cc unless low-cross section, low-absorptivity materials are chosen. Also, for post-test handling and subsequent storage, materials should not be used that are activated into long half-life isotopes. Pneumatic actuators can be constructed from various reactor-compatible materials, but conventional pneumatic piston actuators generally lack the stiffness required for high resolution reactivity control unless electrical position sensors and compensated electronic control systems are used. To overcome these limitations, a pneumatic actuator is under development that positions an output shaft in response to a series of pneumatic pulses, comprising a pneumatic analog of an electrical stepping motor. The pneumatic pulses are generated remotely, beyond the strong radiation environment, and transmitted to the actuator through tubing. The mechanically simple actuator uses a nutating gear harmonic drive to convert motion of small pistons directly to high-resolution angular motion of the output shaft. The digital nature of this actuator is suitable for various reactor control algorithms but is especially compatible with the three bean salad algorithm discussed by Ball et al. (1991).

  20. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics

    International Nuclear Information System (INIS)

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  1. Helical-type device and laser fusion. Rivals for tokamak-type device at n-fusion development in Japan

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Under the current policy on the research and development of nuclear fusion in Japan, as enunciated by the Atomic Energy Commission of Japan, the type of a prototype fusion reactor will be chosen after 2020 from tokamak, helical or some other type including the inertial confinement fusion using lasers. A prototype fusion reactor is the next step following the tokamak type International Thermonuclear Experimental Reactor (ITER). With the prototype reactor, the feasibility as a power plant will be examined. At present the main research and development of nuclear fusion in Japan are on tokamak type, which have been promoted by Japan Atomic Energy Research Institute (JAERI). As for the other types of nuclear fusion, researches have been carried out on the helical type in Kyoto University and National Institute for Fusion Science (NIFS), the mirror type in Tsukuba University, the tokamak type using superconductive coils in Kyushu University, and the laser fusion in Osaka University. The features and the present state of research and development of the Large Helical Device and the laser fusion which is one step away from the break-even condition are reported. (K.I.)

  2. Next Generation Reactors in Korea

    International Nuclear Information System (INIS)

    Oh, Yongshick; Choi, Youngsang; Park, Keecheol

    1990-01-01

    In Korea, nuclear power will be continuously needed to meet the trend of steady increase in electricity demand. But in relation to the further development of nuclear energy, there are still many uncertainties to be solved such as power demand forecast, site availability, thermal energy utilization and technology enhancement for economic and safety. To cope with those uncertainties effectively and to proceed the nuclear projects uninterruptedly, KEPCO decided to initiate two research project. i. e., one is 'the outlook and developmental strategy of nuclear energy for the early 21st century in the R. O. K' and the other is 'the feasibility study on the advanced reactors in Korea. Prospects of nuclear energy in Korea was overviewed and recommendations from the industry were introduced. It is strong opinion of Korea nuclear industry that nuclear policy should be changed from the support policy to the target management policy. In the point of reactor strategy, the life of light water reactor technology might be longer than expected before in Korea and it is emphasized that good maintenance of light water reactor technology and smooth transition program to the advanced technologies should be carefully considered. There are differences in the opinions between preferences to the evolutionary and/or passive, inherently safe reactors but, in the long-term point of view, it is judged to be desirable to have alternatives

  3. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

  4. World progress toward fusion energy

    International Nuclear Information System (INIS)

    Davies, N.A.

    1989-01-01

    The author discusses international progress in fusion research during the last three years. Much of the technical progress has been achieved through international collaboration in magnetic fusion research. This progress has stimulated political interest in a multinational effort, aimed at designing and possibly constructing the world's first experimental fusion reactor. This interest was reflected in recent summit-level discussions involving President Mitterand, General Secretary Gorbachev, and President Reagan. Most recently, the European Community (EC), Japan, the United States, and the U.S.S.R. have decided to begin serious preparation for taking the next step toward practical fusion energy. These parties have agreed to begin the design and supporting R and D for an International Thermonuclear Experimental Reactor (ITER) under the auspices of the International Atomic Energy Agency (IAEA). The initiation of this international program to prepare for a fusion test reactor is discussed

  5. Materials and manufacturing for sodium cooled breeder and fusion power reactor

    International Nuclear Information System (INIS)

    Baldev Raj

    2013-01-01

    The paper narrates definitions of challenges relating to materials and manufacturing for sodium cooled fast reactors thermonuclear fusion reactors. Science and technology developed indigenously but in the context of bench marks in the world is described through examples. Solutions to challenges requires synergy among theoretical physicists, computational chemists, material scientists, metallurgists and engineers with their domains of expertise along with foresight effective management

  6. Thermonuclear device

    International Nuclear Information System (INIS)

    Inoue, Toyokazu; Murata, Toru.

    1983-01-01

    Purpose: To shield superconducting coils for use in the generation of magnetic field against neutron irradiation thereby preventing tritium contamination. Constitution: The thermonuclear device comprises, in its inside, a vacuum container for containing plasmas, superconducting coils disposed to the outside of the vacuum container and neutron absorbers disposed between the super-conducting coils and the vacuum container. since neutrons issued from the plasma are absorbed by neutron absorbers and not irradiated to the superconducting coils, generation of tritium due to the reaction between 3 He in the liquid helium as the coolants for the super-conducting coils and the neutrons is prevented. (Aizawa, K.)

  7. Reactor potential for magnetized target fusion

    International Nuclear Information System (INIS)

    Dahlin, J.E.

    2001-06-01

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

  8. 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.

  9. Fission energy: The integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements

  10. Fission energy: The integral fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  11. Review of fast reactor activities in India

    Energy Technology Data Exchange (ETDEWEB)

    Paranjpe, S R [Reactor Research Centre, Kalpakkam, Tamil Nadu (India)

    1981-05-01

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties.

  12. Review of fast reactor activities in India

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1981-01-01

    It may be recalled that In the presentation at the last meeting of the IWGFR (13th Annual meeting), a broad outline of India's nuclear energy programme and the role of fast breeders in the programme has been provided. The steps taken to enable the fast breeders to fulfil their role have also been described. In brief, fast breeder reactors are considered as an essential and integral part of the programme of nuclear energy and constitute the second step in the programme, the first being the construction of natural uranium heavy water moderated reactors which will consume natural uranium but will produce plutonium to fuel fast breeder reactors. This basic position has remained unchanged and the Government is now taking steps to build a large number of heavy water reactors, say 10 million Kw capacity in the next 20 years. This defines the time frame for developing the fast breeder technology in the country. It has therefore been decided to mobilise the efforts towards design, construction and operation of a medium sized (about 500 M We) reactor by mid-nineties. Thus, the climate for fast breeder reactors is good and there is a good deal of enthusiasm amongst the scientists and engineers working in the field although the actual implementation of the programme during the year had to face certain difficulties

  13. Comparison between a pumped-limiter and a divertor for the next step machines

    International Nuclear Information System (INIS)

    Harrison, F.F.A.

    1985-01-01

    The paper presents a simple description of the physics issues which influence the conceptual design of a pumped-limiter and single-null poloidal divertor in a next step, long burn tokamak of NET/INTOR scale. Predicted performance of the limiter and divertor are compared in regard to localised recycling, sputtering of the plasma collection surfaces, penetration of sputtered impurities into the fusion plasma, surface power loading and exhaust of helium ash. It is concluded that the performance of the divertor is superior and that it can be predicted with a reasonable degree of confidence. The viability of the limiter remains in doubt but the concept cannot be rejected at the present time

  14. Energy balance of controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Hashmi, M.; Staudenmaier, G.

    2000-01-01

    It is shown that a discrepancy and incompatibility persist between basic physics and fusion-literature regarding the radiation losses from a thermonuclear plasma. Whereas the fusion-literature neglects the excitation or line radiation completely, according to basic physics it depends upon the prevailing conditions and cannot be neglected in general. Moreover, for a magnetized plasma, while the fusion-literature assumes a self-absorption or reabsorption of cyclotron or synchrotron radiation emitted by the electrons spiraling along the magnetic field, the basic physics does not allow any effective reabsorption of cyclotron or synchrotron radiation. As is demonstrated, fallacious assumptions and notions, which somehow or other crept into the fusion-literature, are responsible for this discrepancy. In the present work, the theory is corrected. On the grounds of basic physics, a complete energy balance of magnetized and non-magnetized plasmas is presented for pulsed, stationary and self-sustaining operations by taking into account the energy release by reactions of light nuclei as well as different kinds of diffusive (conduction) and radiative (bremsstrahlung, cyclotron or synchrotron radiation and excitation radiation) energy losses. Already the energy losses by radiation make the energy balance negative. Hence, a fusion reactor-an energy producing device-seems to be beyond the realms of realization. (orig.)

  15. Photocatalytic reactors for treating water pollution with solar illumination: a simplified analysis for n-steps flow reactors with recirculation

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Universitaet Hannover (Germany). Institut fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [INTEC Universidad Nacional del Litoral and CONICET, Sante Fe (Argentina)

    2005-09-01

    The concentration of dissolved oxygen in water, in equilibrium with atmospheric air (ca. 8 ppm at 20{sup o}C), defines the limits of all practical oxidizing processes for removing pollutants in photocatalytic reactors. To solve this limitation, an alternative approach to that of a continuously aerated reactor is the use of a recirculating system with aeration performed after every cycle at the reactor entering stream. As defined by the nature of a single recirculating step (the need of a reactor operation at a rather low concentration range), this procedure results in a very low photonic efficiency (thus requiring a large photon collecting area and consequently increasing the capital cost). The design engineer will have to resort to a series of several reactors with recirculation. This solution may then lead to a very high Photonic Efficiency for the entire process (i.e., a reduced light harvesting area) at the price of an increase in the required capital cost (due to the larger number of reactors). This paper provides a very simple analysis and analytical expressions that can be used to estimate, for a desired degree of degradation, a trade-off solution between a high number of reactors and a very large surface area to collect the solar photons. (author)

  16. Analysis of quench-vent pressures for present design of ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coils

    International Nuclear Information System (INIS)

    Slack, D.S.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three-dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 3 figs., 1 tab

  17. Fusion reactor materials semiannual progress report for the period ending March 31, 1990

    International Nuclear Information System (INIS)

    1990-08-01

    This report mainly discusses topics on the physical effects of radiation on thermonuclear reactor materials. The areas discussed are: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; fundamental mechanical behavior; radiation effects; mechanistic studies, theory and modeling; development of structural alloys; solid breeding materials; and ceramics. (FI)

  18. Scoping analyses for the safety injection system configuration for Korean next generation reactor

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Song, Jin Ho; Park, Jong Kyoon

    1996-01-01

    Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are performed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSl pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SlT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA

  19. New materials in nuclear fusion reactors

    International Nuclear Information System (INIS)

    Iwata, Shuichi

    1988-01-01

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

  20. The LOFT perspective on neutron star thermonuclear bursts

    DEFF Research Database (Denmark)

    in ’t Zand, J.J.M.; Altamirano, D.; Ballantyne, D. R.

    This is a White Paper in support of the mission concept of the Large Observatory for X-ray Timing (LOFT), proposed as a medium-sized ESA mission. We discuss the potential of LOFT for the study of thermonuclear X-ray bursts on accreting neutron stars. For a summary, we refer to the paper....

  1. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  2. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

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

  3. Areva's water chemistry guidebook with chemistry guidelines for next generation plants (AREVA EPRTM reactors)

    International Nuclear Information System (INIS)

    Ryckelynck, N.; Chahma, F.; Caris, N.; Guillermier, P.; Brun, C.; Caron-Charles, M.; Lamanna, L.; Fandrich, J.; Jaeggy, M.; Stellwag, B.

    2012-09-01

    Over the years, AREVA globally has maintained a strong expertise in LWR water chemistry and has been focused on minimizing short-term and long-term detrimental effects of chemistry for startup, operation and shutdown chemistry for all key plant components (material integrity and reliability, promote optimal thermal performances, etc.) and fuel. Also AREVA is focused on minimizing contamination and equipment/plant dose rates. Current Industry Guidelines (EPRI, VGB, etc.) provide utilities with selected chemistry guidance for the current operating fleet. With the next generation of PWR plants (e.g. AREVA's EPR TM reactor), materials of construction and design have been optimized based on industry lessons learned over the last 50+ years. To support the next generation design, AREVA water chemistry experts, have subsequently developed a Chemistry Guidebook with chemistry guidelines based on an analysis of the current international practices, plant operating experience, R and D data and calculation codes now available and/or developed by AREVA. The AREVA LWR chemistry Guidebook can be used to help resolve utility and safety authority questions and addresses regulation requirement questions/issues for next generation plants. The Chemistry Guidebook provides water chemistry guidelines for primary coolant, secondary side circuit and auxiliary systems during startup, normal operation and shutdown conditions. It also includes conditioning and impurity limits, along with monitoring locations and frequency requirements. The Chemistry Guidebook Guidelines will be used as a design reference for AREVA's next generation plants (e.g. EPR TM reactor). (authors)

  4. Two Step Procedure Using a 1-D Slab Spectral Geometry in a Pebble Bed Reactor Core Analysis

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Kim, Kang Seog; Noh, Jae Man; Joo, Hyung Kook

    2005-01-01

    A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted iteration between the spectrum calculation in a spectral zone and the global core calculation. In VSOP, the whole problem domain is divided into many spectral zones in which the fine group spectrum is calculated using bucklings for fast groups and albedos for thermal groups from the global core calculation. The resulting spectrum in each spectral zone is used to generate broad group cross sections of the spectral zone for the global core calculation. In this paper, we demonstrate a two step procedure in a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. The equivalent cross sections generated in this way include the effect of the spectral interaction between the core and the reflector. In the second step, we perform a diffusion calculation using the equivalent cross sections generated in the first step. A simple benchmark problem derived from the PMBR-400 Reactor was introduced to verify this approach. We compared the two step solutions with the Monte Carlo (MC) solutions for the problem

  5. 28. Zvenigorod conference on the plasma physics and controlled thermonuclear synthesis. Theses of reports

    International Nuclear Information System (INIS)

    2001-01-01

    Theses of reports, presented at the 28th Conference on the plasma physics and controlled thermonuclear synthesis (Zvenigorod, 19-23 February 2001) are published. 246 reports were heard at the following sections: magnetic confinement, theory and experiments; inertial thermonuclear synthesis; plasma processes and physics of gas-discharge plasma; physical bases of plasma technologies. 17 reports had the summarizing character [ru

  6. A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

    International Nuclear Information System (INIS)

    Csom, Gyula; Feher, Sandor; Szieberthj, Mate

    2002-01-01

    Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutation capabilities are studied. The goal is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named 'multi-region' MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a 'conventional' MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on. (authors)

  7. Development of ceramic humidity sensor for the Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Na Young; Hwang, Il Soon; Yoo, Han Ill; Song, Chang Rock; Park, Sang Duk; Yang, Jun Seog

    1997-01-01

    For the Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside its containment to achieve cost and safety improvement. To apply LBB concept to MSL, leak sensors highly sensitive to humidity is required. In this paper, a ceramic material, MgCr 2 O 4 -TiO 2 has been developed as a humidity sensor for MSL applications. Experiments performed to characterize the electrical conductivity shows that the conductivity of MgCr 2 O 4 -TiO 2 responds sensitively to both temperature and humidity changes. At a constant temperature below 100 .deg. C, the conductivity increases as the relative humidity increases, which makes the sensor favorable for application to the outside of MSL insulation layer. But as temperature increases beyond 100 .deg. C, the sensor composition should be adjusted for the application to KNGR is to be made at temperature above 100 .deg. C

  8. Operating large controlled thermonuclear fusion research facilities

    International Nuclear Information System (INIS)

    Gaudreau, M.P.J.; Tarrh, J.M.; Post, R.S.; Thomas, P.

    1987-10-01

    The MIT Tara Tandem Mirror is a large, state of the art controlled thermonuclear fusion research facility. Over the six years of its design, implementation, and operation, every effort was made to minimize cost and maximize performance by using the best and latest hardware, software, and scientific and operational techniques. After reviewing all major DOE fusion facilities, an independent DOE review committee concluded that the Tara operation was the most automated and efficient of all DOE facilities. This paper includes a review of the key elements of the Tara design, construction, operation, management, physics milestones, and funding that led to this success. We emphasize a chronological description of how the system evolved from the proposal stage to a mature device with an emphasis on the basic philosophies behind the implementation process. This description can serve both as a qualitative and quantitative database for future large experiment planning. It includes actual final costs and manpower spent as well as actual run and maintenance schedules, number of data shots, major system failures, etc. The paper concludes with recommendations for the next generation of facilities. 13 refs., 15 figs., 3 tabs

  9. Thermal hydrodynamic modeling and simulation of hot-gas duct for next-generation nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Injun [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of); Hong, Sungdeok; Kim, Chansoo [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Bai, Cheolho; Hong, Sungyull [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of); Shim, Jaesool, E-mail: jshim@ynu.ac.kr [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of)

    2016-12-15

    Highlights: • Thermal hydrodynamic nonlinear model is presented to examine a hot gas duct (HGD) used in a fourth-generation nuclear power reactor. • Experiments and simulation were compared to validate the nonlinear porous model. • Natural convection and radiation are considered to study the effect on the surface temperature of the HGD. • Local Nusselt number is obtained for the optimum design of a possible next-generation HGD. - Abstract: A very high-temperature gas-cooled reactor (VHTR) is a fourth-generation nuclear power reactor that requires an intermediate loop that consists of a hot-gas duct (HGD), an intermediate heat exchanger (IHX), and a process heat exchanger for massive hydrogen production. In this study, a mathematical model and simulation were developed for the HGD in a small-scale nitrogen gas loop that was designed and manufactured by the Korea Atomic Energy Research Institute. These were used to investigate the effect of various important factors on the surface of the HGD. In the modeling, a porous model was considered for a Kaowool insulator inside the HGD. The natural convection and radiation are included in the model. For validation, the modeled external surface temperatures are compared with experimental results obtained while changing the inlet temperatures of the nitrogen working fluid. The simulation results show very good agreement with the experiments. The external surface temperatures of the HGD are obtained with respect to the porosity of insulator, emissivity of radiation, and pressure of the working fluid. The local Nusselt number is also obtained for the optimum design of a possible next-generation HGD.

  10. Next-generation reactors in the national energy strategy

    International Nuclear Information System (INIS)

    McGoff, D.J.

    1991-01-01

    In February 1991, the Bush Administration released the National Energy Strategy designed to provide an adequate and balanced energy supply. The strategy provides for major increases in energy efficiency and conservation. Even with these savings, however, there will be a need for substantial increases in base-load electrical generating capacity to sustain economic growth. The strategy identifies the actions required to allow nuclear power to cleanly and safely meet a substantial portion of this needed additional base-load capacity after the turn of the century. On June 27, 1991, the US Department of Energy (DOE) transmitted to Congress the Strategic Plan for Civilian Reactor Development, which reflects the initiative identified in the National Energy Strategy. The strategic plan identifies the advanced light water reactor (ALWR) as the basis for expanded use of nuclear power. The second advanced reactor concept that is being pursued is the modular high-temperature gas-cooled reactor (MHTGR)

  11. On the major DYN3D developments for fast reactor design and transient analysis

    International Nuclear Information System (INIS)

    Merk, B.; Kliem, S.

    2013-01-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  12. On the major DYN3D developments for fast reactor design and transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  13. Development of digital plant protection system for Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Suk-Joon Park

    1998-01-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author)

  14. Trends in fusion reactor safety research

    International Nuclear Information System (INIS)

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

    1991-01-01

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions. 18 refs

  15. Fusion reactor materials: Semiannual progress report for the period ending March 31, 1988

    International Nuclear Information System (INIS)

    1988-08-01

    This report contains papers on thermonuclear reactor materials. The general categories of these papers are: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; development of structural alloys; solid breeding materials; ceramics; and radiation effects. Selected papers have been processed for inclusion in the energy database

  16. Theses of the reports of the XXXI Zvenigorod conference on the plasma physics and controlled thermonuclear synthesis

    International Nuclear Information System (INIS)

    Kovrizhnykh, L.M.; Ivanov, V.A.; Nagaeva, M.L.; Aleksandrov, A.F.; Vorob'ev, V.S.; Ivanenkov, G.V.; Meshcheryakov, A.I.

    2004-01-01

    Theses of the reports of the 31th Zvenigorod Conference on the physics and controlled thermonuclear synthesis, presented by Russian and foreign scientists, are published. The total number of reports is 258, namely, summarizing ones 16, magnetic confinement of high temperature plasma - 98, inertial thermonuclear synthesis - 44, physical processes in low temperature plasma - 58, physical bases of plasma and beam technologies - 42 [ru

  17. The possible transmutation of radioactive waste from nuclear reactors

    International Nuclear Information System (INIS)

    Harries, J.R.

    1974-01-01

    A nuclear reactor power program produces high level and long lived radioactive wastes. The high level activity is associated with fission products, but beyond 400 years the principal waste hazard is from transuranic elements produced in the reactor. Several schemes have been proposed for the transmutation of the problem isotopes into more easily handled isotopes. The neutron flux in a thermal reactor is not high enough to significantly reduce the longer lived fission product isotopes 90 Sr and 132 Gs, but the transuranic elements can be reduced by recycling through power reactors. The limitation on recycling of the transuranic elements is the separation process to remove trace quantities from the waste stream. In fast reactors the transuranic elements are the principal fuel and fast reactor waste contains only half as much 90 Sr as thermal reactors. However, the overall waste hazard is similar to thermal reactors. A sufficiently intense neutron flux for fission product transmutation could perhaps be produced by a spallation reactor driven by a proton linear accelerator or a controlled thermonuclear reactor. However, both concepts are still some years in the future. Transmutation by accelerator sources of protons, electrons of gammas tend to require more energy than neutron transmutation. (author)

  18. Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

    Science.gov (United States)

    Kieffer, A. W.

    1972-01-01

    Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

  19. ITER: the Sun rises over nuclear fusion with West

    International Nuclear Information System (INIS)

    Sacco, Laurent

    2013-01-01

    The ITER project is considered as a critical step on the way to commercial production of electricity by a thermonuclear reactor based on controlled fusion. This project notably requires the development of a divertor which is the objective of the West project which will use the famous Cadarache superconductive magnet reactor, Tore Supra. After having outlined the future lack of fossil energies at the world scale, presented the operation principles of tokamaks and recalled some results obtained in their development, this article justifies the use of superconductive magnets. It presents the ITER project as a step in the production of thermonuclear electricity. ITER will be in fact a proof that such plants can be realised, and it should be followed by Demo, a demonstration power plant, by 2050. The article presents the West project, a test bench for ITER, which introduced modifications in the Tore Supra reactor to create conditions almost similar to that existing at the surface of the Sun. It notably comprises a divertor made of tungsten for the fusion with tritium. It finally outlines that the fusion will be a hot one, not a cold one

  20. Safety of next generation power reactors

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

  1. Laser thermonuclear fusion with force confinement of hot plasma

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. 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...

  3. Helium and its effects on the creep-fatigue behaviour of electron beam welds in the steel AISI-316-L

    International Nuclear Information System (INIS)

    Paulus, M.

    1992-12-01

    Within the scope of R and D work for materials development for the NET fusion experiment (Next European Torus) and the International Thermonuclear Experimental Reactor (ITER), the task reported was to examine electron beam welds in the austenitic stainless steel AISI 316 L (NET reference material) for their fatigue behaviour under creep load, and the effects of helium implantation on there mechanical properties. (orig.) [de

  4. High temperature structural integrity evaluation method and application studies by ASME-NH for the next generation reactor design

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, Jae Han

    2006-01-01

    The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500 .deg. C and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated

  5. Group Health Coaching: Strengths, Challenges, and Next Steps

    Science.gov (United States)

    Wolever, Ruth Q.; Manning, Linda; Elam, Roy; Moore, Margaret; Frates, Elizabeth Pegg; Duskey, Heidi; Anderson, Chelsea; Curtis, Rebecca L.; Masemer, Susan; Lawson, Karen

    2013-01-01

    There is great need for cost effective approaches to increase patient engagement and improve health and well-being. Health and wellness coaching has recently demonstrated great promise, but the majority of studies to date have focused on individual coaching (ie, one coach with one client). Newer initiatives are bringing a group coaching model from corporate leadership development and educational settings into the healthcare arena. A group approach potentially increases cost-effective access to a larger number of clients and brings the possible additional benefit of group support. This article highlights some of the group coaching approaches currently being conducted across the United States. The group coaching interventions included in this overview are offered by a variety of academic and private sector institutions, use both telephonic and in-person coaching, and are facilitated by professionally trained health and wellness coaches as well as trained peer coaches. Strengths and challenges experienced in these efforts are summarized, as are recommendations to address those challenges. A working definition of “Group Health and Wellness Coaching” is proposed, and important next steps for research and for the training of group coaches are presented. PMID:24416678

  6. Group health coaching: strengths, challenges, and next steps.

    Science.gov (United States)

    Armstrong, Colin; Wolever, Ruth Q; Manning, Linda; Elam, Roy; Moore, Margaret; Frates, Elizabeth Pegg; Duskey, Heidi; Anderson, Chelsea; Curtis, Rebecca L; Masemer, Susan; Lawson, Karen

    2013-05-01

    There is great need for cost effective approaches to increase patient engagement and improve health and well-being. Health and wellness coaching has recently demonstrated great promise, but the majority of studies to date have focused on individual coaching (ie, one coach with one client). Newer initiatives are bringing a group coaching model from corporate leadership development and educational settings into the healthcare arena. A group approach potentially increases cost-effective access to a larger number of clients and brings the possible additional benefit of group support. This article highlights some of the group coaching approaches currently being conducted across the United States. The group coaching interventions included in this overview are offered by a variety of academic and private sector institutions, use both telephonic and in-person coaching, and are facilitated by professionally trained health and wellness coaches as well as trained peer coaches. Strengths and challenges experienced in these efforts are summarized, as are recommendations to address those challenges. A working definition of "Group Health and Wellness Coaching" is proposed, and important next steps for research and for the training of group coaches are presented.

  7. Heating and cooling device for use in the vacuum container of a thermonuclear device

    International Nuclear Information System (INIS)

    Morita, Hiroaki; Onozuka, Masanori; Fukui, Hiroshi.

    1986-01-01

    Purpose: To prevent the generation of great temperature difference within a hollow doughnuts-shaped space of the torus vacuum container of a tokamak type thermonuclear reactor, as well as effectively eliminate the local injection of heat to the vacuum container. Constitution: A hollow doughnuts-like space is formed between the inner wall and the double outer wall of a vacuum container main body, which is divided into a plurality of regions by partition plates extended in the toroidal direction. An input/output header is disposed in adjacent with each of the partition plates for inputting/outputting heat medium. Further, heat medium inlet/outlets are disposed to define two flow channels on every one-half circumference. This enables to reduce the temperature difference of the heat medium between the inlet and the outlet by the shortening of the flow channel length and heating or cooling can be performed without causing unevenness in the temperature distribution of the vacuum container. (Horiuchi, T.)

  8. XXXII Zvenigorod conference on the plasma physics and controlled thermonuclear synthesis. Theses of reports

    International Nuclear Information System (INIS)

    2005-01-01

    Theses of the reports, presented at the XXXII International conference on the plasma physics and controlled thermonuclear synthesis (Zvenigorod, 14-18 February 2005) are published. The total number of reports is 322, including 16 summarizing ones. The other reports are distributed by the following sections: magnetic confinement of high-temperature plasma (88 reports), inertial thermonuclear fusion (65), physical processes in low-temperature plasma (99) and physical bases of the plasma and beam technologies (54) [ru

  9. Experimental and computational studies of thermal mixing in next generation nuclear reactors

    Science.gov (United States)

    Landfried, Douglas Tyler

    The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHTR utilizes helium as a coolant in the primary loop of the reactor. Helium traveling through the reactor mixes below the reactor in a region known as the lower plenum. In this region there exists large temperature and velocity gradients due to non-uniform heat generation in the reactor core. Due to these large gradients, concern should be given to reducing thermal striping in the lower plenum. Thermal striping is the phenomena by which temperature fluctuations in the fluid and transferred to and attenuated by surrounding structures. Thermal striping is a known cause of long term material failure. To better understand and predict thermal striping in the lower plenum two separate bodies of work have been conducted. First, an experimental facility capable of predictably recreating some aspects of flow in the lower plenum is designed according to scaling analysis of the VHTR. Namely the facility reproduces jets issuing into a crossflow past a tube bundle. Secondly, extensive studies investigate the mixing of a non-isothermal parallel round triple-jet at two jet-to-jet spacings was conducted. Experimental results were validation with an open source computational fluid dynamics package, OpenFOAMRTM. Additional care is given to understanding the implementation of the realizable k-a and Launder Gibson RSM turbulence Models in OpenFOAMRTM. In order to measure velocity and temperature in the triple-jet experiment a detailed investigation of temperature compensated hotwire anemometry is carried out with special concern being given to quantify the error with the measurements. Finally qualitative comparisons of trends in the experimental results and the computational results is conducted. A new and unexpected physical behavior was observed in the center jet as it appeared to spread unexpectedly for close spacings (S/Djet = 1.41).

  10. Radiation effects and tritium technology for fusion reactors. Volume I. Proceedings of the international conference, Gatlinburg, Tennessee, October 1--3, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Watson, J.S.; Wiffen, F.W.; Bishop, J.L.; Breeden, B.K. (eds.)

    1976-03-01

    Separate abstracts were prepared for the 29 included papers in Vol. I. The topics covered in this volume include swelling and microstructures in thermonuclear reactor materials. Some papers on modeling and damage analysis are included. (MOW)

  11. Development of digital plant protection system for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suk-Joon [NSSS Engineering and Development Division, Korea Power Engineering Company, Taejon (Korea, Republic of)

    1998-10-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author) 8 refs, 4 figs, 3 tabs

  12. Rates of the main thermonuclear reactions

    International Nuclear Information System (INIS)

    Abramovich, S.N.; Guzhovskii, B.Ya.; Dunaeva, S.A.; Fomushkin, E.F.

    1992-01-01

    The data on the cross sections of main thermonuclear reactions have been estimated with an account of the latest experimental results in a form of S-factor spline presentation. Based on this estimation, the reates of these reactions in 0.0001-1 MeV temperature range in the supposition of Maxwell distribution of relative velocities have been computed. The Maxwell-Boltzmann averaged -factors were calculated according to the table values of the reaction rates. Then the -factors were approximated with the 3 order spline-function. The necessity of the account of electron shielding and intramolecular movement at low temperatures is discussed (orig.)

  13. FROM THE HISTORY OF PHYSICS: The extraordinarily beautiful physical principle of thermonuclear charge design (on the occasion of the 50th anniversary of the test of RDS-37 — the first Soviet two-stage thermonuclear charge)

    Science.gov (United States)

    Goncharov, German A.

    2005-11-01

    On 22 November 1955, the Semipalatinsk test site saw the test of the first domestic two-stage thermonuclear RDS-37 charge. The charge operation was based on the principle of radiation implosion. The kernel of the principle consists in the radiation generated in a primary A-bomb explosion and confined by the radiation-opaque casing propagating throughout the interior casing volume and flowing around the secondary thermonuclear unit. The secondary unit experiences a strong compression under the irradiation, with a resulting nuclear and thermonuclear explosion. The RDS-37 explosion was the strongest of all those ever realized at the Semipalatinsk test site. It produced an indelible impression on the participants in the test. This document-based paper describes the genesis of the ideas underlying the RDS-37 design and reflects the critical moments in its development. The advent of RDS-37 was an outstanding accomplishment of the scientists and engineers of our country.

  14. Probing thermonuclear burning on accreting neutron stars

    Science.gov (United States)

    Keek, L.

    2008-12-01

    that the models need to be extended with a new heat source. Another rare phenomenon is the occurrence of bursts with recurrence times of less than 30 minutes. In a long set of observations of the source EXO 0748-676 we find for the first time triple bursts, where three bursts occur within 30 minutes. This time is too short to accrete new fuel for the next burst, which suggests that not all hydrogen and helium is burned during the first burst. Finally, using a hydrodynamic stellar evolution code we create a multi-zone numerical model of the neutron star envelope. For the first time we include mixing due to rotation and a rotationally induced magnetic field. We find that thermonuclear burning proceeds in a stable manner at a lower heat flux of the crust for models including mixing. This may explain the observed transition of stable to unstable burning at a lower mass accretion rate than models previously predicted.

  15. Synthetic report 2012. Research programme on controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Vaucher, C.; Tran, M. Q.; Villard, L.; Marot, L.

    2013-01-01

    Since 1961, Switzerland participates in the research on thermonuclear fusion thanks to the creation of the Research Centre in Plasma Physics. In 1979 it entered into partnership with the European programme on fusion through its adhesion to EURATOM. The thermonuclear fusion is an interesting energy source because the basic fuel is practically inexhaustible and its use does not release any significant CO 2 quantity and very little radioactive residues. But its working up faces enormous physical and technological difficulties. The International Thermonuclear Reactor (ITER), presently in construction, has to demonstrate the technological feasibility of the controlled fusion. Il will be followed by DEMO, foreseen for 2040-2050, which must guarantee the economical rentability. At CRPP the research projects are partitioned onto several sites: at the Swiss Federal Institute of Technology (EPFL) in Lausanne, they concern the physics of the magnetic confinement with the Variable Geometry Tokamak (TCV), the development of theoretical models and the numerical simulation, the plasma heating and the generation of hyper frequency waves; the Paul Scherrer Institute (PSI) studies the superconductivity and the materials; the interactions between the plasma and the Tokamak walls are studied at the Basel University for the structures of ITER. Thanks to its large flexibility, TCV allows the creation and the control of plasmas of very different forms. The injection system of millimetric waves allows orienting the injected power according to specific profiles. By using the asymmetry of the flow in the toroidal sense, the plasma rotation could be measured with a much better accuracy than before. In TCV, by playing on the form of the plasma, it was possible to strongly reduce the energy quantity which is expelled by the Edge Localized Modes (ELM) onto the wall of the vacuum chamber. The ‘snowflake’ configuration created in TCV allows distributing the ELM energy onto several impact

  16. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  17. Plasma engineering innovations for the ORNL TNS reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-01-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma β values up to 10 percent and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors

  18. Prospects for toroidal fusion reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.D.

    1994-01-01

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

  19. Future view of total energy system and reactor engineering and reactor physics

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

    This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

  20. Problems in the assessment of inherent safety characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Garribba, S.F.; Vivante, C.

    1988-01-01

    A number of proposals are being made for an increased RD and D effort on advanced nuclear power reactors that would display outstanding safety performance. A common characteristic of the different reactor concepts would be their limited reliance upon active engineered systems under major accident conditions. However, when submitted to a more close scrutiny reactor concept options may reveal diverging safety behaviors and also development opportunities. In this respect, three issues are explored in this paper. A first question is the meaning of non-active, i.e. inherent and passive safety features. Next, is the ranking of advanced and new reactor concepts from the viewpoint of inherent and passive safety. Multiple correspondence analysis may provide a simple tool, whose use is shown for the case of HTR-500, AP600 and PRISM. Conversely, probabilistic risk assessment would allow quantitative comparisons, although lack of information and data is an obstacle. Finally, is demonstration of safety performances as a step toward market deployment of the new reactor systems

  1. Investigation of the stationary-thermonuclear-reaction realization possibility in a tokamak device

    International Nuclear Information System (INIS)

    Kolesnichenko, Ya.I.; Reznik, S.N.; Fursa, A.D.

    1976-01-01

    The stationary (quasistationary) selfsustaining thermonuclear D-T reaction is shown to be possible in a toroidal device such as 'Tokamak' with large enough plasma radius. The stationary temperature of the plasma can be quite high. Thus when the transport processes are assumed to be neoclassical the temperature of the central part of a plasma colomn of radius approximately 10-200 cm in the stationary state is 70 keV.The stationary temperature distribution is reached spontaneously as a result of the thermal instability development if plasma is preheated to 10 keV. The stationary thermonuclear burning is also possible at lower temperatures if plasma energy balance is controlled

  2. Spin-charge-family theory is offering next step in understanding elementary particles and fields and correspondingly universe

    International Nuclear Information System (INIS)

    Mankoč Borštnik, Norma Susana

    2017-01-01

    More than 40 years ago the standard model made a successful new step in understanding properties of fermion and boson fields. Now the next step is needed, which would explain what the standard model and the cosmological models just assume: a. The origin of quantum numbers of massless one family members. b. The origin of families. c. The origin of the vector gauge fields. d. The origin of the Higgses and Yukawa couplings. e. The origin of the dark matter. f. The origin of the matter-antimatter asymmetry. g. The origin of the dark energy. h. And several other open problems. The spin-charge-family theory, a kind of the Kaluza-Klein theories in ( d = (2 n − 1) + 1)-space-time, with d = (13 + 1) and the two kinds of the spin connection fields, which are the gauge fields of the two kinds of the Clifford algebra objects anti-commuting with one another, may provide this much needed next step. The talk presents: i. A short presentation of this theory. ii. The review over the achievements of this theory so far, with some not published yet achievements included. iii. Predictions for future experiments. (paper)

  3. Spin-charge-family theory is offering next step in understanding elementary particles and fields and correspondingly universe

    Science.gov (United States)

    Mankoč Borštnik, Norma Susana

    2017-05-01

    More than 40 years ago the standard model made a successful new step in understanding properties of fermion and boson fields. Now the next step is needed, which would explain what the standard model and the cosmological models just assume: a. The origin of quantum numbers of massless one family members. b. The origin of families. c. The origin of the vector gauge fields. d. The origin of the Higgses and Yukawa couplings. e. The origin of the dark matter. f. The origin of the matter-antimatter asymmetry. g. The origin of the dark energy. h. And several other open problems. The spin-charge-family theory, a kind of the Kaluza-Klein theories in (d = (2n - 1) + 1)-space-time, with d = (13 + 1) and the two kinds of the spin connection fields, which are the gauge fields of the two kinds of the Clifford algebra objects anti-commuting with one another, may provide this much needed next step. The talk presents: i. A short presentation of this theory. ii. The review over the achievements of this theory so far, with some not published yet achievements included. iii. Predictions for future experiments.

  4. 1982 annual status report: thermonuclear fusion technology

    International Nuclear Information System (INIS)

    1982-01-01

    The objective of this programme is to study the technological problems related to ''Post Jet'' experimental machines and, in a longer range, to assess the engineering aspects of Fusion Power Reactor Plants. According to the decision taken by the Council of Ministers on the JRC multiannual programme (1980-1983), the work performed on 1982 concerns four projects, namely: The Project 1: ''Fusion Reactor Studies''concerns mainly the NET (Next European Torus) studies which have been continued in the framework of the European participation to INTOR (INternational TOkamak Reactor). This represents a collaborative effort to design a major fusion experiment beyond the-upcoming generation of large tokamaks. The Project 2: ''Blanket Technology'' has the aim to investigate the behaviour of blanket materials in fusion conditions. The Project 3: ''Materials Sorting and Development'' has the aim to assess the mechanical properties and radiation damage of standard and advanced materials suited for structures, in particular for application as first wall of the fusion reactors. The Project 4: ''Cyclotron Operation and Experiments''has the task to exploit a cyclotron to simulate radiation damages to materials in a fusion ambient

  5. Space Drive Physics: Introduction and Next Steps

    Science.gov (United States)

    Millis, M. G.

    Research toward the visionary goal of propellantless ``space drives'' is introduced, covering key physics issues and a listing of roughly 2-dozen approaches. The targeted advantage of a space drive is to circumvent the propellant constraints of rockets and the maneuvering limits of light sails by using the interactions between the spacecraft and its surrounding space for propulsion. At present, the scientific foundations from which to engineer a space drive have not been discovered and, objectively, might be impossible. Although no propulsion breakthroughs appear imminent, the subject has matured to where the relevant questions have been broached and are beginning to be answered. The critical make-break issues include; conservation of momentum, uncertain sources of reaction mass, and the net-external thrusting requirement. Note: space drives are not necessarily faster- than-light devices. Speed limits are a separate, unanswered issue. Relevant unsolved physics includes; the sources and mechanisms of inertial frames, coupling of gravitation and electromagnetism, and the nature of the quantum vacuum. The propulsion approaches span mostly stages 1 through 3 of the scientific method (defining the problem, collecting data, and articulating hypotheses), while some have matured to stage 4 (testing hypotheses). Nonviable approaches include `stiction drives,' `gyroscopic antigravity,' and `lifters.' No attempt is made to gauge the prospects of the remaining approaches. Instead, a list of next-step research questions is derived from the examination of these goals, unknowns, and concepts.

  6. Usability of an internet-based platform (Next.Step for adolescent weight management

    Directory of Open Access Journals (Sweden)

    Pedro Sousa

    2015-02-01

    Full Text Available OBJECTIVE: The current study evaluates the usability perception of an e-therapeutic platform (supported by electronic processes and communication, aiming to promote the behavior change and to improve the adolescent health status through increased and interactive contact between the adolescent and the clinical staff. METHODS: This was a correlational study with a sample of 48 adolescents (12-18 years who attended a Pediatric Obesity Clinic between January and August of 2012. Participants were invited to access, during 24 weeks, the e-therapeutic multidisciplinary platform (Next.Step in addition to the standard treatment program. A usability questionnaire was administered and the platform performance and utilization indicators were analyzed. RESULTS: The users' perception of satisfaction, efficiency, and effectiveness regarding the Next.Step platform was clearly positive. However, only 54.17% of the enrolled adolescents accessed the platform, with a mean task-completion rate of 14.55% (SD = 18.853. The higher the number of the platform consulted resources, the greater the tendency to enjoy the platform, to consider it exciting and quick, to consider that the time spent in it was useful, to consider the access to information easy, and to login easier. Post-intervention assessment revealed a significant reduction in anthropometric and behavioral variables, including body mass index z-score, waist circumference percentile, hip circumference, and weekly screen time. CONCLUSION: These results highlight the importance of information and communication technologies in the health information access and the healthcare provision. Despite the limited adherence rate, platform users expressed a positive overall perception of its usability and presented a positive anthropometric and behavioral progress.

  7. Structural materials issues for the next generation fission reactors

    Science.gov (United States)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  8. Transition period fuel cycle from current to next generation reactors for Japan

    International Nuclear Information System (INIS)

    Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

    2007-01-01

    Long-term energy security and global warming prevention can be achieved by a sustainable electricity supply with next generation fast breeder reactors (FBRs). Current light water reactors (LWRs) will be replaced by FBRs and FBR cycle will be established in the future considering the limited amount of uranium (U) resource. The introduction of FBRs requires plutonium (Pu) recovered from LWR spent fuel. The authors propose advanced system named Flexible Fuel Cycle Initiative (FFCI)' which can supply enough Pu and hold no surplus Pu, can respond flexibly the future technical and social uncertainties, and can achieve an economical FBR cycle. FFCI can simplify the 2nd LWR reprocessing facility for Japan (after Rokkasho Reprocessing Plant) which only carries out U removal from LWR spent fuel. Residual 'Recycle Material' is, according to FBRs introduction status, immediately treated in the FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of Recycle Material as a buffer material between LWR and FBR cycles. (author)

  9. EDF view on next generation reactor safety and operability issues

    International Nuclear Information System (INIS)

    Serviere, G.

    2002-01-01

    is involved at various degrees in the evaluation of next generation: - Light-water reactors; - Gas-cooled reactors; - Liquid-metal reactors. Available information is not the same for all concepts, but nevertheless adequate for identifying areas where confirmation of assumptions would be needed. After discussing some crosscut issues, this paper outlines which areas would have to be clarified for each type of reactor before they could be considered proven by the company. Issues for which R and D programs could be needed are also be identified. (author)

  10. Controlled thermonuclear fusion. Present state and prospective

    International Nuclear Information System (INIS)

    Consoli, T.

    1976-01-01

    The interest of thermonuclear fusion for energy production is underlined. The present state of the research in this field is presented, emphasis being given to Tokamak configurations. The problems concerning confinement and additional heating in these devices are presented [fr

  11. Direct conversion of nuclear energy into radiation: New direction in thermonuclear laser fusion

    International Nuclear Information System (INIS)

    Babaev, Yu.N.; Vedenov, A.A.; Filyukov, A.A.

    1995-01-01

    In investigations dealing with thermonuclear fusion, a radical new direction appeared some time ago, namely the direct conversion of nuclear and thermonuclear energy into radiation energy. This paper reviews early work on this topic in Russia and the United States and discusses some recent new directions

  12. 1D thermonuclear model for x-ray transients

    International Nuclear Information System (INIS)

    Wallace, R.K.

    1982-01-01

    The thermonuclear evolution of a 1.41 M solar mass neutron star, with a radius of 14.3 km, accreting various mixtures of hydrogen, helium, and heavy elements at rates of 10 -11 to 10 -10 M solar mass/yr is examined, in conjunction with S.E. Woosley and T.A. Weaver, using a one-dimensional numerical model. We have ignored any effects due to general relativity or magnetic fields. Two cases shall be discussed. In both models, the accretion rate is such that the hydrogen shell burns to helium in steady state, with the hydrogen burning stabilized by the β-limited CNO cycle. A thick helium shell is produced, which is eventually ignited under extremely degenerate conditions, producing a thermonuclear runaway

  13. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  14. Final repository searching in Germany - what are the next steps?; Endlagerstandortsuche in Deutschland. Wie geht's weiter?

    Energy Technology Data Exchange (ETDEWEB)

    Neles, Julia Mareike [Oeko-Institut e.V., Darmstadt (Germany). Bereich Nukleartechnik und Anlagensicherheit

    2016-07-01

    Up to now (2016) the question of final disposal of high-level radioactive wastes is still open in Germany. The Commission of radioactive waste disposal has finalized its report including recommendations for the further process. The next step will be the site selection procedure based on a ''white map''. The protest of several Federal states and communities against repository sites in their region is already developing.

  15. Development of a framework for the neutronics analysis system for next generation (3)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Development of innovative analysis methods and models in fundamental studies for next-generation nuclear reactor systems is in progress. In order to efficiently and effectively reflect the latest analysis methods and models to primary design of commercial reactor and/or in-core fuel management for power reactors, a next-generation analysis system MARBLE has been developed. The next-generation analysis system provides solutions to the following requirements: (1) flexibility, extensibility and user-friendliness that can apply new methods and models rapidly and effectively for fundamental studies, (2) quantitative proof of solution accuracy and adaptive scoping range for design studies, (3) coupling analysis among different study domains for the purpose of rationalization of plant systems and improvement of reliability, (4) maintainability and reusability for system extensions for the purpose of total quality management and development efficiency. The next-generation analysis system supports many fields, such as thermal-hydraulic analysis, structure analysis, reactor physics etc., and now we are studying reactor physics analysis system for fast reactor in advance. As for reactor physics analysis methods for fast reactor, we have established the JUPITER standard analysis methods based on the past study. But, there has been a problem of extreme inefficiency due to lack of functionality in the conventional analysis system when changing analysis targets and/or modeling levels. That is why, we have developed the next-generation analysis system for reactor physics which reproduces the JUPITER standard analysis method that has been developed so far and newly realizes burnup and design analysis for fast reactor and functions for cross section adjustment. In the present study, we examined in detail the existing design and implementation of ZPPR critical experiment analysis database followed by unification of models within the framework of the next-generation analysis system by

  16. Sonoluminescence, shock waves, and micro-thermonuclear fusion

    International Nuclear Information System (INIS)

    Moss, W.C.; Clarke, D.B.; White, J.W.; Young, D.A.

    1995-08-01

    We have performed numerical hydrodynamic simulations of the growth and collapse of a sonoluminescing bubble in a liquid. Our calculations show that spherically converging shock waves are generated during the collapse of the bubble. The combination of the shock waves and a realistic equation of state for the gas in the bubble provides an explanation for the measured picosecond optical pulse widths and indicates that the temperatures near the center of the bubble may exceed 3O eV. This leads naturally to speculation about obtaining micro-thermonuclear fusion in a bubble filled with deuterium (D 2 ) gas. Consequently, we performed numerical simulations of the collapse of a D 2 bubble in D 2 0. A pressure spike added to the periodic driving amplitude creates temperatures that may be sufficient to generate a very small, but measurable number of thermonuclear D-D fusion reactions in the bubble

  17. World must build two atomic reactors each day the next hundred years. [Summary of and commentary on book, 'Mankind at the Turning Point'

    Energy Technology Data Exchange (ETDEWEB)

    1974-07-24

    In summarizing and commenting on the ideas presented in Mesarovic and Pestel's book ''Mankind at the Turning Point'' it is pointed out that the global energy crisis makes comprehensive long-term planning a necessity. Assuming, optimistically, that nuclear power alone is able to supply the total projected energy demand in 100 years, it is stated that this will require 3000 nuclear power stations, each with 8 fast breeder reactors, totally 100 GW(t). This means a net rate of construction of four reactors per week, which again means allowing for a 30-year life, two reactors per day, every day, for the next hundred years. Fueling of these reactors will require the production and transport of 15 x 10/sup 6/ kg of /sup 239/Pu per year. It is therefore obvious that the energy crisis is not only a technological, but also a political, social, and even psychological problem.

  18. The Influence of Stellar Spin on Ignition of Thermonuclear Runaways

    Science.gov (United States)

    Galloway, Duncan K.; in ’t Zand, Jean J. M.; Chenevez, Jérôme; Keek, Laurens; Sanchez-Fernandez, Celia; Worpel, Hauke; Lampe, Nathanael; Kuulkers, Erik; Watts, Anna; Ootes, Laura; The MINBAR collaboration

    2018-04-01

    Runaway thermonuclear burning of a layer of accumulated fuel on the surface of a compact star provides a brief but intense display of stellar nuclear processes. For neutron stars accreting from a binary companion, these events manifest as thermonuclear (type-I) X-ray bursts, and recur on typical timescales of hours to days. We measured the burst rate as a function of accretion rate, from seven neutron stars with known spin rates, using a burst sample accumulated over several decades. At the highest accretion rates, the burst rate is lower for faster spinning stars. The observations imply that fast (>400 Hz) rotation encourages stabilization of nuclear burning, suggesting a dynamical dependence of nuclear ignition on the spin rate. This dependence is unexpected, because faster rotation entails less shear between the surrounding accretion disk and the star. Large-scale circulation in the fuel layer, leading to enhanced mixing of the burst ashes into the fuel layer, may explain this behavior; further numerical simulations are required to confirm this.

  19. Considerations of severe accidents in the design of Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Dong Wook Jerng; Choong Sup Byun

    1998-01-01

    The severe accident is one of the key issues in the design of Korean Next Generation Reactor (KNGR) which is an evolutionary type of pressurized water reactor. As IAEA recommends in TECDOC-801, the design objective of KNGR with regard to safety is provide a sound technical basis by which an imminent off-site emergency response to any circumstance could be practically unnecessary. To implement this design objective, probabilistic safety goals were established and design requirements were developed for systems to mitigate severe accidents. The basic approach of KNGR to address severe accidents is firstly prevent severe accidents by reinforcing its capability to cope with the design basis accidents (DBA) and further with some accidents beyond DBAs caused by multiple failures, and secondly mitigate severe accidents to ensure the retention of radioactive materials in the containment by providing mean to maintain the containment integrity. For severe accident mitigation, KNGR principally takes the concept of ex-vessel corium cooling. To implement this concept, KNGR is equipped with a large cavity and cavity flooding system connected to the in-containment refueling water storage tank. Other major systems incorporated in KNGR are hydrogen igniters and safety depressurization systems. In addition, the KNGR containment is designed to withstand the pressure and temperature conditions expected during the course of severe accidents. In this paper, the design features and status of system designs related with severe accidents will be presented. Also, R and D activities related to severe accident mitigation system design will be briefly described

  20. Next Steps: Water Technology Advances (Research)

    Science.gov (United States)

    This project will focus on contaminants and their impact on health, adequate removal of contaminants from various water systems, and water and resource recovery within treatment systems. It will develop the next generation of technological advances to provide guidance in support ...

  1. Decommissioning: demonstrating the solution to a problem for the next century

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Tables are presented showing present decommissioning experience with shut-down reactors, and listing the oldest operating reactors to illustrate the likely extent of decommissioning work early in the next century. (author)

  2. First wall of thermonuclear device

    International Nuclear Information System (INIS)

    Kizawa, Makoto; Koizumi, Makoto; Nishihara, Yoshihiro.

    1990-01-01

    The first wall of a thermonuclear device is constituted with inner wall tiles, e.g. made of graphite and metal substrates for fixing them. However, since the heat expansion coefficient is different between the metal substrates and intermediate metal members, thermal stresses are caused to deteriorate the endurance of the inner wall tiles. In view of the above, low melting metals are disposed at the portion of contact between the inner wall tiles and the metal substrates and, further, a heat pipe structure is incorporated into the metal substrates. Under the thermal load, for example, during operation of the thermonuclear device, the low melting metals at the portion of contact are melted into liquid metals to enhance the state of contact between the inner wall tiles and the metal substrate to reduce the heat resistance and improve the heat conductivity. Even if there is a difference in the heat expansion coefficient between the inner wall tiles and the metal substrates, neither sharing stresses not thermal stresses are caused. Further, since the heat pipe structure is incorporated into the metal substrates, the lateral unevenness of the temperature in the metal substrates can be eliminated. Thus, the durability of the inner wall tiles can be improved. (N.H.)

  3. Elise - the next step in development of induction heavy ion drivers for inertial fusion energy

    International Nuclear Information System (INIS)

    Lee, E.; Bangerter, R.O.; Celata, C.; Faltens, A.; Fessenden, T.; Peters, C.; Pickrell, J.; Reginato, L.; Seidl, P.; Yu, S.

    1994-11-01

    LBL, with the participation of LLNL and industry, proposes to build Elise, an electric-focused accelerator as the next logical step towards the eventual goal of a heavy-ion induction linac powerful enough to implode or open-quotes driveclose quotes inertial-confinement fusion targets. Elise will be at full driver scale in several important parameters-most notably line charge density (a function of beam size), which was not explored in earlier experiments. Elise will be capable of accelerating and electrostatically focusing four parallel, full-scale ion beams and will be designed to be extendible, by successive future construction projects, to meet the goal of the USA DOE Inertial Fusion Energy program (IFE). This goal is to address all remaining issues in heavy-ion IFE except target physics, which is currently the responsibility of DOE Defense Programs, and the target chamber. Thus Elise is the first step of a program that will provide a solid foundation of data for further progress toward a driver, as called for in the National Energy Strategy and National Energy Policy Act

  4. Insulation structure of thermonuclear device

    International Nuclear Information System (INIS)

    Suzuki, Takayuki; Usami, Saburo; Tsukamoto, Hideo; Kikuchi, Mitsuru

    1998-01-01

    The present invention provides an insulating structure of a thermonuclear device, in which insulation materials between toroidal coils are not broken even if superconductive toroidal coils are used. Namely, a tokamak type thermonuclear device of an insulating structure type comprises superconductive toroidal coils for confining plasmas arranged in a circular shape directing the center each at a predetermined angle, and the toroidal coils are insulated from each other. The insulation materials are formed by using a biaxially oriented fiber reinforced plastics. The contact surface of the toroidal coils and the insulating materials are arranged so that they are contact at a woven surface of the fiber reinforced plastics. Either or both of the contact surfaces of the fiber reinforced plastics and the toroidal coils are coated with a high molecular compound having a low friction coefficient. With such a constitution, since the interlayer shearing strength of the biaxially oriented fiber reinforced plastics is about 1/10 of the compression strength, the shearing stress exerted on the insulation material is reduced. Since a static friction coefficient on the contact surface is reduced to provide a structure causing slipping, shearing stress does not exceeds a predetermined limit. As a result, breakage of the insulation materials between the toroidal coils can be prevented. (I.S.)

  5. What is the Plasma Focus Thermonuclear Pulsors Technology?

    International Nuclear Information System (INIS)

    Ramos, R.; Gonzalez, J.; Moreno, C.; Clausse, A.

    2003-01-01

    In this paper we describe a type of neutron generators, called Plasma Focus, which is suitable to several applications, where traditional generators are non-applicable.The main characteristics are its transportability and to be non-contaminating, which would allow in-situ tests.The Plasma Focus, produces neutron pulses by thermonuclear fusion reactions, satisfy these requirements and it is comparatively non expensive.This last feature would assure competitivity in the neutron sources market

  6. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Duffey, R.B.; Torgerson, D.F.

    2001-01-01

    We describe the development philosophy, direction and concepts that are being utilized by AECL to refine the CANDU reactor to meet the needs of current and future competitive energy markets. The technology development path for CANDU reactors is based on the optimization of the pressure tube concept. Because of the inherent modularity and flexibility of this basis for the core design, it is possible to provide a seamless and continuous evolution of the reactor design and performance. There is no need for a drastic shift in concept, in technology or in fuel. By continual refinement of the flow and materials conditions in the channels, the basic reactor can be thermally and operationally efficient, highly competitive and economic, and highly flexible in application. Thus, the design can build on the successful construction and operating experience of the existing plants, and no step changes in development direction are needed. This approach minimizes investor, operator and development risk but still provides technological, safety and performance advances. In today's world energy markets, major drivers for the technology development are: (a) reduced capital cost; (b) improved operation; (c) enhanced safety; and (d) fuel cycle flexibility. The drivers provide specific numerical targets. Meeting these drivers ensures that the concept meets and exceeds the customer economic, performance, safety and resource use goals and requirements, including the suitable national and international standards. This logical development of the CANDU concept leads naturally to the 'Next Generation' of CANDU reactors. The major features under development include an optimized lattice for SEU (slightly enriched uranium) fuel, light water cooling coupled with heavy water moderation, advanced fuel channels and CANFLEX fuel, optimization of plant performance, enhanced thermal and BOP (balance of plant) efficiency, and the adoption of layout and construction technology adapted from successful on

  7. Thermonuclear device

    International Nuclear Information System (INIS)

    Oosaki, Osamu; Masuda, Kenju.

    1980-01-01

    Purpose: To provide excellent electric properties and high reliability in a thermonuclear device by improving a current collecting board connected to a coil device. Constitution: A current collecting board element perforated with an opening for enserting a connecting terminal is sized to be inserted into a plating tank, and is surface treated in the plating tank. Only the current collecting board element preferably surface treated is picked up. A plurality of such current collecting board elements are connected and welded to form a large current collecting board. In this manner, the current collecting board having several m 2 to several ten order m 2 in area can be obtained as preferably surface treated at the connecting terminal hole. The current collecting board element can be determined in shape with the existing facility without increasing the size of a surface treating tank. (Kamimura, M.)

  8. The problem of helium in structural materials for fusion reactor

    International Nuclear Information System (INIS)

    Nikiforov, A.S.; Zakharov, A.P.; Chuev, V.I.

    1982-01-01

    The processes of helium buildup in some metals and alloys at different energy neutron flux irradiation under thermonuclear reactor conditions are considered. The data on high temperature helium embrittlement of a number of stainless steels, titanium and aluminium alloys etc. are given A review of experiments concerning the implanted helium behaviour is presented. Possible reactions between helium atoms and point defects or their clusters are discussed. Analysed are material structure variations upon buildup in them up to 1 at % of helium

  9. Novel cancer immunotherapy agents with survival benefit: recent successes and next steps

    Science.gov (United States)

    Sharma, Padmanee; Wagner, Klaus; Wolchok, Jedd D.; Allison, James P.

    2012-01-01

    The US Food and Drug Administration (FDA) recently approved two novel immunotherapy agents, sipuleucel-T and ipilimumab, which showed a survival benefit for patients with metastatic prostate cancer and melanoma, respectively. The mechanisms by which these agents provide clinical benefit are not completely understood. However, knowledge of these mechanisms will be crucial for probing human immune responses and tumour biology in order to understand what distinguishes responders from non-responders. The following next steps are necessary: first, the development of immune-monitoring strategies for the identification of relevant biomarkers; second, the establishment of guidelines for the assessment of clinical end points; and third, the evaluation of combination therapy strategies to improve clinical benefit. PMID:22020206

  10. Design optimization of general arrangement in Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, S. H.; Jung, D. W.; Choi, Y. B.; Cho, S. J.

    1999-01-01

    In order to optimize the general arrangement(GA) of Korean Next Generation Reactor (KNGR), field opinions in domestic nuclear power plants have been collected, and the bench marking on UCN No.1,2 which were estimated to be the most excellent in view of operability and maintenance has been accomplished. Through this work, design optimization items for GA were reviewed. Major items to be selected for optimization are summarized as follows; 'Expanding the compound building function and the mezzanine floor concept in the auxiliary building', 'Including the diesel generator building to the auxiliary building', 'Change of the equipment removal method in the auxiliary building'. With these GA design optimization, the auxiliary building boundary will be improved as a complete rectangular type. The power block volume except the changing effect to the single containment structure will be reduced to about 10% in comparison with that of in KNGR phase II

  11. Fast fission assisted ignition of thermonuclear microexplosions

    International Nuclear Information System (INIS)

    Winterberg, F.

    2006-01-01

    It is shown that the requirements for fast ignition of thermonuclear microexplosions can be substantially relaxed if the deuterium-tritium (DT) hot spot is placed inside a shell of U-238 (Th-232). An intense laser - or particle beam-projected into the shell leads to a large temperature gradient between the hot DT and the cold U-238 (Th-232), driving thermomagnetic currents by the Nernst effect, with magnetic fields large enough to entrap within the hot spot the α-particles of the DT fusion reaction. The fast fission reactions in the U-238 (Th-232) shell implode about 1/2 of the shell onto the DT, increasing its density and reaction rate. With the magnetic field generated by the Nernst effect, there is no need to connect the target to a large current carrying transmission line, as it is required for magnetized target fusion, solving the so-called ''stand off'' problem for thermonuclear microexplosions. (orig.)

  12. Research works at the Physics Institute nuclear reactor for the nuclear power engineering

    International Nuclear Information System (INIS)

    Gavars, V.V.; Kalnin'sh, D.O.; Lapenas, A.A.; Tomsons, E.Ya.; Ulmanis, U.A.

    1985-01-01

    Methods for neutron spectra determination in the nuclear reactor core and vessel have been developed. On their base the neutron spectra at the Novo-Voronezh and kola NPPs have been measured. Such measurements are necessary for the determination of the nuclear fuel reprocessing coefficients, for the evaluation of the construction radiation-damage stability and the NPP economical efficiency on the whole. A new type of the reactor regulator - a liquid metal one - has been created. Such regulators are promising in respect of their use at the NPPs. The base for studying new radiation-damage-stable insulators has been founded. The materials obtained are now applied to designing the reactors of the second (fast) and the third (thermonuclear H) generations. There have developed and by a long-time exploitation checked a hot loop, used for materials irradiation. the nuclear reactor in Salaspils provides training of students being the new brain-power for the nuclear power engineering

  13. Design reliability assurance program for Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Beom-Su; Han, Jin-Kyu; Na, Jang Hwan; Yoo, Kyung Yeong

    1997-01-01

    The Korean Next Generation Reactor (KNGR) project is to develop standardized nuclear power plant design for the construction of future nuclear power plants in Korea. The main purpose of the KNGR project is to develop the advanced nuclear power plants, which enhance safety and economics significantly through the incorporation of design concepts for severe accident prevention and mitigation, supplementary passive safety concept, simplification and application of modularization and so on. For those, Probabilistic Safety Assessment (PSA) and availability study will be performed at the early stage of the design, and the Design Reliability Assurance Program (D-RAP) is applied in the development of the KNGR to ensure that the safety and availability evaluated in the PSA and availability study at the early phase of the design is maintained through the detailed design, construction, procurement and operation of the plants. This paper presents the D-RAP concept that could be applied at the stage of the basic design of the nuclear power plants, based on the models for the reference plants and/or similar plants. 4 refs., 1 fig

  14. Conceptual design of the Fast-Liner Reactor (FLR) for fusion power

    International Nuclear Information System (INIS)

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

    1979-02-01

    The generation of fusion power from the Fast-Liner Reactor (FLR) concept envisages the implosion of a thin (3-mm) metallic cylinder (0.2-m radius by 0.2-m length) onto a preinjected plasma. This plasma would be heated to thermonuclear temperatures by adiabatic compression, pressure confinement would be provided by the liner inertia, and thermal insulation of the wall-confined plasma would be established by an embedded azimuthal magnetic field. A 2- to 3-mu s burn would follow the approx. 10 4 m/s radial implosion and would result in a thermonuclear yield equal to 10 to 15 times the energy initially invested into the liner kinetic energy. For implosions occurring once every 10 s a gross thermal power of 430 MWt would be generated. The results of a comprehensive systems study of both physics and technology (economics) optima are presented. Despite unresolved problems associated with both the physics and technology of the FLR, a conceptual power plant design is presented

  15. The TDF System for Thermonuclear Plasma Reaction Rates, Mean Energies and Two-Body Final State Particle Spectra

    International Nuclear Information System (INIS)

    Warshaw, S I

    2001-01-01

    The rate of thermonuclear reactions in hot plasmas as a function of local plasma temperature determines the way in which thermonuclear ignition and burning proceeds in the plasma. The conventional model approach to calculating these rates is to assume that the reacting nuclei in the plasma are in Maxwellian equilibrium at some well-defined plasma temperature, over which the statistical average of the reaction rate quantity σv is calculated, where σ is the cross-section for the reaction to proceed at the relative velocity v between the reacting particles. This approach is well-understood and is the basis for much nuclear fusion and astrophysical nuclear reaction rate data. The Thermonuclear Data File (TDF) system developed at the Lawrence Livermore National Laboratory (Warshaw 1991), which is the topic of this report, contains data on the Maxwellian-averaged thermonuclear reaction rates for various light nuclear reactions and the correspondingly Maxwellian-averaged energy spectra of the particles in the final state of those reactions as well. This spectral information closely models the output particle and energy distributions in a burning plasma, and therefore leads to more accurate computational treatments of thermonuclear burn, output particle energy deposition and diagnostics, in various contexts. In this report we review and derive the theoretical basis for calculating Maxwellian-averaged thermonuclear reaction rates, mean particle energies, and output particle spectral energy distributions for these reactions in the TDF system. The treatment of the kinematics is non-relativistic. The current version of the TDF system provides exit particle energy spectrum distributions for two-body final state reactions only. In a future report we will discuss and describe how output particle energy spectra for three- and four-body final states can be developed for the TDF system. We also include in this report a description of the algorithmic implementation of the TDF

  16. Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

    International Nuclear Information System (INIS)

    Culbert, M.E.

    1978-07-01

    The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

  17. Simulation of the dynamics of sausage development in a z pinch with a high rate of thermonuclear heat production

    International Nuclear Information System (INIS)

    Vikhrev, V.V.; Rozanova, G.A.

    1993-01-01

    The development of the sausage instability in a z pinch is accompanied by the formulation of a high-temperature plasma. This high-temperature region initiates a wave of thermonuclear burning propagating along the pinch. A numerical solution of the MHD equations has been carried out, taking into account plasma energy losses through radiation and thermonuclear heating. Results of calculations on the growth of the sausage instability are presented for ρr = 0.23 g/cm 2 . It is accompanied by the development of a stable wave of thermonuclear burning. 12 refs., 4 figs

  18. Thermonuclear device

    International Nuclear Information System (INIS)

    Suzuki, Shohei

    1988-01-01

    Purpose: To obtain high voltage withstanding current introduction terminals not suffering from the effects of the reduction in the creeping voltage withstanding property by the application of magnetic fields. Constitution: This invention concerns a current introduction terminal for supplying electric current to coils for use in a thermonuclear device, etc. The conductor of the current introduction terminal on the side of vacuum is completely covered with solid insulator. This can eliminate the portion of securing the creeping withstanding voltage. The voltage withstanding characteristics of the solid insulator covering the portion of the conductor on the side of vacuum has a constant value irrespective of the atmosphere or the absence or presence of magnetic fields. Accordingly, the voltage withstanding characteristics of the current introduction terminal on the side of vacuum are determined by the property of the solid insulator, which is not reduced by the application of magnetic fields. (Ikeda, J.)

  19. Thermonuclear device

    International Nuclear Information System (INIS)

    Kajiura, Soji.

    1984-01-01

    Purpose: To suppress the generation of electromagnetic forces and improve the strength of a vacuum container for sealing plasmas and of a support frame for covering the coils disposed around the periphery of the vacuum container. Constitution: Either one of the vacuum container or the support frame is made of a composite material, whose first material has low radioactivatability and the second has low radioactivatability and stronger electrical resistance than that of the first; therein, with the first material being disposed on the surface. The damage caused by neutrons resulted from thermonuclear reaction can be extremely small since the constituent is made of the material having the low radioactivatability. Further, eddy current does not occurs in the second material, but in the first material only in case magnetic fields change rapidly, whereby the electromagnetic force resulted in this portion is decreased as a whole. (Moriyama, K.)

  20. Isotopic hydrology of Berrocal area (Toledo, Spain): I: Tritium in springs with thermonuclear resources

    International Nuclear Information System (INIS)

    1995-01-01

    This book the study on isotopic hydrology in El Berrocal, Toledo (Spain). The special topic was the study about the tritium of springs with thermonuclear source. The study are articulated in 3 chapter: 1.- Chemical analysis of wastes 2.- Tritium with thermonuclear source 3.- Human resources

  1. Integrated concept development of next-step helical-axis advanced stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Warmer, Felix

    2016-04-13

    With the increasing energy demand of mankind and the transformation of our society towards sustainability, nuclear fusion by magnetic confinement is a promising option for the sustainable electricity supply in the future. In view of these prospects this thesis focuses on the concept development of next-step helical-axis advanced stellarator (HELIAS) burning-plasma devices. The HELIAS-line is the continued development of the prototype optimised stellarator Wendelstein 7-X which started operation in 2015. For the integrated concept development of such devices, the approach taken in this work encompasses detailed physics and engineering considerations while also including economic aspects. Starting with physics considerations, the properties of plasma transport and confinement of 3D stellarator configurations are discussed due to their critical importance for the device design. It becomes clear that current empirical confinement time scalings are not sufficient to predict the confinement in future stellarator devices. Therefore, detailed 1D transport simulations are carried out to reduce the uncertainties regarding confinement. Beyond the well-validated neoclassical approach, first attempts are made to include results from state-of-the-art turbulence simulations into the 1D transport simulations to further enhance the predictive capabilities. Next, for the systematic development of consistent design points, stellarator-specific models are developed and implemented in the well-established European systems code PROCESS. This allows a consistent description of an entire HELIAS fusion power plant including physics, engineering, and economic considerations. With the confidence obtained from a verification study, systems studies are for the first time applied for a HELIAS power-plant which shows that the available design window is constrained by the beta-limit. Furthermore, an economic comparison of an exemplary design point to an ''equivalent'' tokamak

  2. Recent DIII-D results

    International Nuclear Information System (INIS)

    Petersen, P.I.

    1994-07-01

    This paper summarizes the recent DIII-D experimental results and the development of the relevant hardware systems. The DIII-D program focuses on divertor solutions for next generation tokamaks such as International Thermo-nuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX), and on developing configurations with enhanced confinement and stability properties that will lead to a more compact and economical fusion reactor. The DIII-D program carries out this research in an integrated fashion

  3. Reports on BMBF-sponsored research projects in the field of reactor safety. Reporting period 1 July - 31 December 1995

    International Nuclear Information System (INIS)

    1996-01-01

    The Gesellschaft fuer Anlagen- und Reaktorsicherheit informs of the status of LWR tasks and projects on the safety of advanced reactors. Each progress report represents a compilation of individual reports about objectives, the work performed, the results, and the next steps of the works. The individual reports of quality assurance, safety of reactor component, emergency core cooling, lors of coolant, meltdown, fission product release, risk and reliability, are classified according to projects to the reactor safety research program. Another table uses the same classification system as applied in the nuclear safety index of the CEC. (DG)

  4. ITER: the first experimental fusion reactor

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1995-01-01

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

  5. Fusion reactor materials semiannual progress report for period ending September 30, 1992

    International Nuclear Information System (INIS)

    1992-01-01

    This report contains papers on the following topics on thermonuclear reactor materials: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters,and activation calculations; radiation effects, mechanistic studies, theory and modeling; development of structural alloys; solid breeding materials and beryllium; and ceramics. These reports have been index separately elsewhere

  6. Development of a Ground Test and Analysis Protocol for NASA's NextSTEP Phase 2 Habitation Concepts

    Science.gov (United States)

    Gernhardt, Michael L.; Beaton, Kara H.; Chappell, Steven P.; Bekdash, Omar S.; Abercromby, Andrew F. J.

    2018-01-01

    The NASA Next Space Technologies for Exploration Partnerships (NextSTEP) program is a public-private partnership model that seeks commercial development of deep space exploration capabilities to support human spaceflight missions around and beyond cislunar space. NASA first issued the Phase 1 NextSTEP Broad Agency Announcement to U.S. industries in 2014, which called for innovative cislunar habitation concepts that leveraged commercialization plans for low-Earth orbit. These habitats will be part of the Deep Space Gateway (DSG), the cislunar space station planned by NASA for construction in the 2020s. In 2016, Phase 2 of the NextSTEP program selected five commercial partners to develop ground prototypes. A team of NASA research engineers and subject matter experts (SMEs) have been tasked with developing the ground-test protocol that will serve as the primary means by which these Phase 2 prototypes will be evaluated. Since 2008, this core test team has successfully conducted multiple spaceflight analog mission evaluations utilizing a consistent set of operational tools, methods, and metrics to enable the iterative development, testing, analysis, and validation of evolving exploration architectures, operations concepts, and vehicle designs. The purpose of implementing a similar evaluation process for the Phase 2 Habitation Concepts is to consistently evaluate different commercial partner ground prototypes to provide data-driven, actionable recommendations for Phase 3. This paper describes the process by which the ground test protocol was developed and the objectives, methods, and metrics by which the NextSTEP Phase 2 Habitation Concepts will be rigorously and systematically evaluated. The protocol has been developed using both a top-down and bottom-up approach. Top-down development began with the Human Exploration and Operations Mission Directorate (HEOMD) exploration objectives and ISS Exploration Capability Study Team (IECST) candidate flight objectives. Strategic

  7. Fuel, structural material and coolant for an advanced fast micro-reactor

    International Nuclear Information System (INIS)

    Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

    2011-01-01

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

  8. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

    1992-09-01

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  9. Remote maintenance for fusion experimental reactor

    International Nuclear Information System (INIS)

    Koizumi, Koichi; Takeda, Nobukazu

    2000-01-01

    Here was introduced on maintenance of reactor core portion operated by remote control among maintenance of the International Thermonuclear Experimental Reactor (ITER) begun on its design since 1988 under international cooperation of U.S.A., Europe, Russia and Japan. Every appliances constructing the reactor core portion is necessary to carry out all of their inspection and maintenance by using remote controlled apparatus because of their radiation due to neutron generated by DT combustion of plasma. For engineering design activity (EDA) in ITER, not only design and development of the remote control appliances but also design under consideration of remote maintenance for from structural design of maintained objective appliances to access method to appliances, transportation and preservation method of radiated matters, and out-reactor maintenance in a hot cell, is now under progress. Here were also reported on basic concept on maintenance and conservation of ITER, maintenance design of diverter and blanket with high maintenance frequency and present state on development of maintenance appliances. (G.K.)

  10. History of the development of zirconium alloys for use in nuclear reactors

    International Nuclear Information System (INIS)

    Rickover, H.G.; Geiger, L.D.; Lustman, B.

    1975-01-01

    The technical problems and the major decisions made during the early development of zirconium alloys for use in naval reactors are outlined. A summary is given of the development of commercial sources of supply for zirconium and hafnium metal over the period 1950 to 1965, and the problems encountered in obtaining zirconium needed for early naval prototype and shipboard reactors are identified. Steps taken in the Government procurement process are described and statistics on production amounts, prices, and inventory are included. Also included are the technical aspects associated with the development of zirconium for water-cooled nuclear reactors, beginning in early 1949 when the Bettis Atomic Power Laboratory was established as a part of the Naval Reactors Program. While in the course of the next 25 years, small-scale investigations were performed on other potential core structural materials such as stainless steel, niobium, aluminum, and beryllium, the pressure for continual development, improvement, and application of zirconium was predominant and unrelenting. (U.S.)

  11. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  12. Multi-step Monte Carlo calculations applied to nuclear reactor instrumentation - source definition and renormalization to physical values

    Energy Technology Data Exchange (ETDEWEB)

    Radulovic, Vladimir; Barbot, Loic; Fourmentel, Damien; Villard, Jean-Francois [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Snoj, Luka; Zerovnik, Gasper [Jozef Stefan Institute, Reactor Physics Department, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Trkov, Andrej [IAEA, Vienna International Centre, PO Box 100, A-1400 Vienna, (Austria)

    2015-07-01

    Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. These data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm{sup 3}). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which

  13. Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

    Directory of Open Access Journals (Sweden)

    Rodrigues Davide

    2015-12-01

    Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

  14. Response of beryllium to severe thermal shocks -simulation of disruption and vertical displacement events in future thermonuclear devices

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Duwe, R.; Roedig, M.; Schuster, A. [Association Euratom-Forschungszentrum Juelich GmbH (Germany); Merola, M.; Qian, R.H.

    1998-01-01

    Beryllium will play an important role for plasma facing components in next step thermonuclear fusion devices such as ITER. In particular for the first wall beryllium will be used with an armor thickness of several millimeters. However, during plasma instabilities they will experience severe thermal shocks. Here plasma disruptions with deposited energy densities of several ten MJm{sup -2} are the most essential damaging mechanism. However, a signifant fraction of the incident energy will be absorbed by a dense cloud of ablation vapor, hence reducing the effective energy density at the beryllium surface to values in the order of 10 MJm{sup -2}. To investigate the material response to all these plasma instabilities thermal shock tests on small scale test coupons (disruption effects) and on actively cooled divertor modules (VDEs) have been performed in the electron beam test facility JUDITH at ITER relevant surface heat loads. These tests have been performed on different bulk beryllium grades and on plasma sprayed coatings; the influence of pulse duration, power density, and temperature effects has been investigated experimentally. Detailed in-situ diagnostics (for beam characterization, optical pyrometry etc.) and post mortem analyses (profilometry, metallography, optical and electron microscopy) have been applied to quantify the resulting material damage. 1D- and 2D models have developed to verify the experimental results obtained in the electron beam simulation experiments. (J.P.N.)

  15. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  16. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

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

    International Nuclear Information System (INIS)

    1988-05-01

    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)

  18. Jaderná bezpečnost fúzních elektráren a jejich vliv na životní prostředí.

    Czech Academy of Sciences Publication Activity Database

    Entler, Slavomír; Dostál, V.

    2017-01-01

    Roč. 25, 9-10 (2017), s. 262-268 ISSN 1210-7085 Grant - others:AV ČR(CZ) StrategieAV21/2 Program:StrategieAV Institutional support: RVO:61389021 Keywords : design * reactor safety * thermonuclear fuel s * thermonuclear reactions * thermonuclear reactor materials * thermonuclear reactors * tokamak devices Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering

  19. Repairing method and device for thermonuclear device

    International Nuclear Information System (INIS)

    Sakurai, Akiko; Masumoto, Hiroshi; Tachikawa, Nobuo.

    1995-01-01

    The present invention provides a method of and a device for repairing a first wall and a divertor disposed in a vacuum vessel of a thermonuclear device. Namely, an armour tile of the divertor secured, by a brazing material, in a vacuum vessel of the thermonuclear device in which high temperature plasmas of deuterium and tritium are confined to cause fusion reaction is induction-heated or heated by microwaves to melt the brazing material. Only the armour tile is thus exchanged by its attachment/detachment. This device comprises, in the vacuum vessel, an armour tile attaching/detaching manipulator and a repairing manipulator comprising a heating manipulator having induction heating coils at the top end thereof. Induction heating coils are connected to an AC power source. According to the present invention, the armour tile is exchanged without taking the divertor out of the vacuum vessel. Therefore, cutting of a divertor cooling tube for taking the divertor out of the vacuum vessel and re-welding of the divertor for attaching it to the vacuum vessel again are no more necessary. (I.S.)

  20. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  1. Developing maintainability in controlled thermonuclear reactors. Progress report, October 1, 1977--April 30, 1978

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-05-01

    During the period 1 October 1977 through 30 April 1978 the study has completed work on Task 6, Candidate Reference Systems. Four candidate reference systems have been defined. These are based on the conceptual designs of the UWMAK-III, the General Atomic Company Demonstration Power Reactor, the Oak Ridge National Laboratory Cassette defined in the Demonstration Power Study and the Culham laboratory Mark II Reactors. These reactor concepts are normalized to 3000 MW/sub th/ and near minimum cost of electricity. In addition, designs of four major subsystems have been selected and defined for application to these reactors. These include a primary coolant system, primary and secondary vacuum zone systems, the neutral beam injection system and the magnetic field system. These magnet systems are unique to each reactor. The cases for which maintenance plans are being developed in Task 7 have been selected to allow evaluation of design features, particularly the vacuum wall locations, and the impacts of unscheduled and contact maintenance of subsystems on the cost of electricity

  2. Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

    International Nuclear Information System (INIS)

    Rose, R.P.

    1979-05-01

    This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

  3. Estimating heterotrophic respiration at large scales: Challenges, approaches, and next steps

    Science.gov (United States)

    Bond-Lamberty, Ben; Epron, Daniel; Harden, Jennifer W.; Harmon, Mark E.; Hoffman, Forrest; Kumar, Jitendra; McGuire, Anthony David; Vargas, Rodrigo

    2016-01-01

    Heterotrophic respiration (HR), the aerobic and anaerobic processes mineralizing organic matter, is a key carbon flux but one impossible to measure at scales significantly larger than small experimental plots. This impedes our ability to understand carbon and nutrient cycles, benchmark models, or reliably upscale point measurements. Given that a new generation of highly mechanistic, genomic-specific global models is not imminent, we suggest that a useful step to improve this situation would be the development of “Decomposition Functional Types” (DFTs). Analogous to plant functional types (PFTs), DFTs would abstract and capture important differences in HR metabolism and flux dynamics, allowing modelers and experimentalists to efficiently group and vary these characteristics across space and time. We argue that DFTs should be initially informed by top-down expert opinion, but ultimately developed using bottom-up, data-driven analyses, and provide specific examples of potential dependent and independent variables that could be used. We present an example clustering analysis to show how annual HR can be broken into distinct groups associated with global variability in biotic and abiotic factors, and demonstrate that these groups are distinct from (but complementary to) already-existing PFTs. A similar analysis incorporating observational data could form the basis for future DFTs. Finally, we suggest next steps and critical priorities: collection and synthesis of existing data; more in-depth analyses combining open data with rigorous testing of analytical results; using point measurements and realistic forcing variables to constrain process-based models; and planning by the global modeling community for decoupling decomposition from fixed site data. These are all critical steps to build a foundation for DFTs in global models, thus providing the ecological and climate change communities with robust, scalable estimates of HR.

  4. Two-dimensional nucleonics calculations for a ''FIRST STEP'' conceptual ICF reactor

    International Nuclear Information System (INIS)

    Davidson, J.W.; Battat, M.E.; Saylor, W.W.; Pendergrass, J.H.; Dudziak, D.J.

    1985-01-01

    A detailed two-dimensional nucleonic analysis has been performed for the FIRST STEP conceptual ICF reactor blanket design. The reactor concept incorporated in this design is a modified wetted-wall cavity with target illumination geometry left as a design variable. The 2-m radius spherical cavity is surrounded by a blanket containing lithium and 238 U as fertile species and also as energy multipliers. The blanket is configured as 0.6-m-thick cylindrical annuli containing modified LMFBR-type fuel elements with 0.5-m-thick fuel-bearing axial end plugs. Liquid lithium surrounds the inner blanket regions and serves as the coolant for both the blanket and the first wall. The two-dimensional analysis of the blanket performance was made using the 2-D discrete-ordinates code TRISM, and benchmarked with the 3-D Monte Carlo code MCNP. Integral responses including the tritium breeding ratio (TBR), plutonium breeding ratio (PUBR), and blanket energy multiplication were calculated for axial and radial blanket regions. Spatial distributions were calculated for steady-state rates of fission, neutron heating, prompt gamma-ray heating, and fuel breeding

  5. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  6. Critical density and disruptions in α-heated thermonuclear Tokamak discharges

    International Nuclear Information System (INIS)

    Cotsaftis, M.; Firestone, M.; Wang, P.K.C.

    1985-02-01

    The study of existence of a critical density limit has been extended to the case of thermonuclear α-particle heated regime. To proceed, a 0-D model including sources and sinks affecting the evolution of ion and electron temperatures and of electron and α-particle densities with auxiliary neutral injected power has been developed. It is mainly shown when considering a Tokamak machine adapted for thermonuclear performances that, like in previous case, there is a critical density above which no other equilibrium point than 0 does exist. Temperatures then drop down the 0 past this critical value, leading to disruption. Analytic expression for critical density is given in terme of auxiliary projected power Psup(N). For Psup(N)=0, critical density value is low, but it increases fast enough for small Psup(N) to give a large safety margin once Psup(N) is moderate, much below the power required for reaching thermonuclear regime. So it is only at shutdown power periods that critical density can be crossed. But in this case, the heat content of particles in the discharge can significantly contribute to smooth out the temperature drop off. This typically operates up to the point where, due to change in magnetic islands configuration resulting from profile modification due to energy release at critical density crossing, heat transport doubles. Then on a fast thermal diffusion time scale, temperature drops now to a new equilibrium value, which can be made above the limiting value for which position control system of the plasma cannot forbid the plasma current to drop off itself, which is the important phenomenon of disruption. So on top of controls previously discussed, it is possible to use the α-particles themselves as a new preventive control against disruptions, making this phenomenon less dangerous for thermonuclear regime operation

  7. Italy, EURATOM and Early Research on Controlled Thermonuclear Fusion (1957-1962)

    International Nuclear Information System (INIS)

    Curli, Barbara

    2017-01-01

    This chapter traces the early origins of European collaboration in controlled thermonuclear fusion research, within the larger picture of Cold War nuclear policy in the late 1950s-early 1960s, and as a consequence of the signing of the EURATOM treaty in 1957. It then presents some preliminary findings on the Association contract which was signed in 1960 between EURATOM and Italy, in order to carry out research in controlled thermonuclear fusion at the then newly created 'Laboratori nazionali di Frascati', near Rome, within the framework of the Comitato Nazionale Energia Nucleare (CNEN), the Italian civilian nuclear energy agency.

  8. Project of law, adopted by the Senate, giving permission to the approval of the agreement between the French government and the international organization for thermonuclear fusion energy ITER, relative to the head office of ITER organization and to the privileges and immunities of ITER organization in the French territory

    International Nuclear Information System (INIS)

    2008-01-01

    The will of building up an international thermonuclear experimental reactor (ITER) gathers since several years the European community of atomic energy (Euratom), Japan, the USA, and Russia, next followed by China, South Korea and, since 2005, by India. The agreement signed in Paris between these seven parties on November 21, 2006 entrusted the international organization ITER with the realization of this project. The implications of the ITER project are enormous both in their scientific and in their economical aspects. France has a particular position in this project since the head office of ITER organisation is sited at Saint-Paul-lez-Durance and the tokamak will be built at Cadarache. Therefore, an agreement has been signed between ITER organization and the French government. The approval of this agreement is the object of this project of law made of a single article. The agreement between the French government and the international organization ITER is attached to the document. It defines the juridical status, the privileges and immunities of the organization itself and of its personnel inside the French territory. An appendix to the agreement precises the cooperation modalities between the French authorities and ITER organization. (J.S.)

  9. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

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

  10. Joining of SiCf/SiC composites for thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Ferraris, M.; Badini, C.; Montorsi, M.; Appendino, P.; Scholz, H.W.

    1994-01-01

    Due to their favourable radiological behaviour, SiC f /SiC composites are promising structural materials for future use in fusion reactors. A problem to cope with is the joining of the ceramic composite material (CMC) to itself for more complex structures. Maintenance concepts for a reactor made of SiC f /SiC will demand a method of joining. The joining agents should comply with the low-activation approach of the base material. With the acceptable elements Si and Mg, sandwich structures of composite/metal/composite were prepared in Ar atmosphere at temperatures just above the melting points of the metals. Another promising route is the use of joining agents of boro-silicate glasses: their composition can be tailored to obtain softening temperatures of interest for fusion applications. The glassy joint can be easily ceramised to improve thermomechanical properties. The joining interfaces were investigated by SEM-EDS, XRD and mechanical tests. ((orig.))

  11. Atypical Thermonuclear Supernovae from Tidally Crushed White Dwarfs

    International Nuclear Information System (INIS)

    Rosswog, S.; Ramirez-Ruiz, E.; Hix, William Raphael

    2008-01-01

    Suggestive evidence has accumulated that intermediate mass black holes (IMBHs) exist in some globular clusters. Some stars will inevitably wander sufficiently close to the hole to suffer a tidal disruption. IMBHs can disrupt not only solar-type stars but also compact white dwarf stars. We investigate the fate of white dwarfs that approach the hole close enough to be disrupted and compressed to such an extent that explosive nuclear burning is triggered. Based on a precise modeling of the gas dynamics together with the nuclear reactions, it is argued that thermonuclear ignition is a natural outcome for white dwarfs of all masses passing well within the tidal radius. A good fraction of the star is accreted, yielding high luminosities that persist for up to a year. A peculiar, underluminous thermonuclear explosion accompanied by a soft X-ray transient signal would, if detected, be a compelling testimony for the presence of an IMBH

  12. Brazilian programme for plasma physics and controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Chian, A.C.L.; Reusch, M.F.; Nascimento, I.C.; Pantuso-Sudano, J.

    1992-01-01

    A proposal for a National Programme of Plasma Physics and Controlled Thermonuclear Fusion in Brazil is presented, aimimg the dissemination of the researchers thought in plasma physics for the national authorities and the scientific community. (E.O.)

  13. Fusion Reactor Materials semiannual progress report for period ending September 30, 1991

    International Nuclear Information System (INIS)

    1992-04-01

    This report contains papers on topic in the following areas of thermonuclear 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 beryllium; and ceramics. These paper have been index separately elsewhere. (LSP)

  14. World must build two atomic reactors each day the next hundred years. [Summary of and commentary on book, 'Mankind at the Turning Point'

    Energy Technology Data Exchange (ETDEWEB)

    1974-07-24

    In summarizing and commenting on the ideas presented in Mesarovic and Pestel's book ''Mankind at the Turning Point'' it is pointed out that the global energy crisis makes comprehensive long-term planning a necessity. Assuming, optimistically, that nuclear power alone is able to supply the total projected energy demand in 100 years, it is stated that this will require 3000 nuclear power stations, each with 8 fast breeder reactors, totally 100 GW(t). This means a net rate of construction of four reactors per week, which again means allowing for a 30-year life, two reactors per day, every day, for the next hundred years. Fueling of these reactors will require the production and transport of 15 x 10/sup 6/ kg of /sup 239/Pu per year. It is therefore obvious that the energy crisis is not only a technological, but also a political, social, and even psychological problem.

  15. Analysis of induction phenomena in thermonuclear experiments

    International Nuclear Information System (INIS)

    Deeds, W.E.; Dodd, C.V.

    1976-01-01

    Many of the problems involving transients induced by changing currents in the large coils of thermonuclear machines are identical to those arising in nondestructive testing by eddy currents. There are three chief methods used for calculating such induction phenomena: analytical boundary-value solutions, relaxation or iteration techniques, and model experiments. Some of the results obtained by each of these methods are described below

  16. Design innovations of the next-step spherical torus experiment and spherical torus development path

    International Nuclear Information System (INIS)

    Ono, M.; Kessel, C.; Peng, M.

    2003-01-01

    The spherical torus (ST) fusion energy development path is complementary to the tokamak burning plasma experiment such as ITER as it focuses toward the compact Component Test Facility (CTF) and higher toroidal beta regimes to improve the design of DEMO and a Power Plant. To support the ST development path, one option of a Next Step Spherical Torus (NSST) device is examined. NSST is a 'performance extension' (PE) stage ST with a plasma current of 5 - 10 MA, R = 1.5, B T ≤ 2.7 T with flexible physics capability to 1) Provide a sufficient physics basis for the design of the CTF, 2) Explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, 3) Contribute to the general plasma/fusion science of high β toroidal plasmas. The NSST facility is designed to utilize the TFTR site to minimize the cost and time required for the construction. (author)

  17. Shock Ignition of Thermonuclear Fuel with High Areal Density

    International Nuclear Information System (INIS)

    Betti, R.; Zhou, C. D.; Anderson, K. S.; Theobald, W.; Solodov, A. A.; Perkins, L. J.

    2007-01-01

    A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy

  18. Shock ignition of thermonuclear fuel with high areal density.

    Science.gov (United States)

    Betti, R; Zhou, C D; Anderson, K S; Perkins, L J; Theobald, W; Solodov, A A

    2007-04-13

    A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy.

  19. A method for carrying out radiolysis and chemical reactions by means of the radiations resulting from a thermonuclear reaction

    International Nuclear Information System (INIS)

    Gomberg, H.J.

    1974-01-01

    The invention relates to the use of the radiations resulting from thermonuclear reactions. It deals with a method comprising a combination of thermo-chemical and radiolytic reactions for treating a molecule having a high absorption rate, by the radiations of a thermonuclear reaction. This is applicable to the dissociation of water into oxygen and hydrogen [fr

  20. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-01-01

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified

  1. International research co-operation in the field of controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    2004-01-01

    This 26th report by the Swiss Federal Office for Education and Science presents a review of work done in Swiss institutes in 2003 as part of international research into thermonuclear fusion. A broad outline of the project and of its significance within the wider field of thermonuclear fusion research is given. This is followed by a review of the significant events in the world of fusion research, with emphasis placed on ITER and on the EURATOM fusion programme. A further chapter summarises events in Switzerland in 2003 and the report closes with a list of contacts for more information. Three annexes provide information on the current situation in fusion research, as well as scientific and technical highlights of the work performed in 2003 at the Plasma Physics Research Centre CRPP at the Federal Institute of Technology EPFL in Lausanne, Switzerland. Annex 3 reports on results obtained at the Physics Institute of the University of Basle. The annexes are for the benefit of the technically and scientifically versed reader, and brief summaries of them are given in the main body of the report

  2. Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

    International Nuclear Information System (INIS)

    Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

    2002-01-01

    Kazakhstan takes part in ITER project in partnership with Russian Federation since the year of 1994. At present the technical stage of the project is completed and ITER Council should take a decision on the site for international reactor. Four countries such as Canada, Japan, Spain and France have offered their territories for being used as site for launching ITER construction. ITER partners started preparing new international agreement that will cover activities on construction, operation and decommissioning of ITER. It will also include the list of research and experimental work that is conducted in support of ITER project. Kazakhstan has already made an important contribution into technical stage realization of ITER project due to scientific and technical researches conducted by National Nuclear Center, by Institute of Experimental and Theoretical Physics and by JSC 'Ulba Metallurgical plant' ('UMP'). Research activity carried out for the support of ITER project is performed in accordance with the following main trends: Tritium safety (permeability and retentin of hydrogen isotopes during in-pile irradiation in various structural materials, co-deposed layers and protective coatings); Verification of computer codes (LOCA type) loss of coolant accidents modeling in ITER reactor; Investigation of liquid metal blanket of thermonuclear reactor (tritium production in lithium containing eutectics Li17Pb83 and ceramics Li 2 TiO 3 , study of tritium permeability). At present the working group of ITER project participants started introducing proposals for cost distribution and for placing the orders on reactor construction. Further Kazakhstan participation in ITER project may be in manufacturing high-tech parts and assemblies from commercial grades of beryllium. They will be used for armouring the reactor first wall, for its thermal protection and for protection of superconductor's components for magnetic systems that are at JSC U MP'. Scientific and technical support of

  3. Thermonuclear device

    International Nuclear Information System (INIS)

    Takano, Hirohisa; Nakamoto, Kazunari; Hanai, Satoshi.

    1984-01-01

    Purpose: To provide coils of high mechanical strength for use at the center of a torus type thermonuclear device. Constitution: A plurality of copper plates having cooling holes and bolt holes and insulation paper sheets of the same shape are prepared. The copper plate is different from the insulation paper sheet only in that the position-phase angle of the opening portion is larger by 15 - 30 0 . The copper plates and the insulation paper sheets are alternately stacked by a required number of turns while displacing the angle, and then clamped by bolts to form a mechanically strong coil with no metallurgical joining. Further, since the insulation paper sheets are not present in the radial direction and only one insulation paper sheet is inserted for each turn in the direction of the coil height, the space occupied by the coil can be decreased. According to this invention, the magnetic flux density at the center of the device can be increased as compared with the conventional case to thereby apply a higher voltage on the side of plasmas. (Moriyama, K.)

  4. Approximated neutronic calculation for the tritium breeding ratio in fusion reactor blankets

    International Nuclear Information System (INIS)

    Santos, Raul dos

    1983-01-01

    An approximated model for the calculation of the tritium breeding ratio in conceptual thermonuclear fusion reactor blankets is presented. This model makes use of the exponential absorption concept due to the Li 6 (n, He 4 )T and Li 7 (n, n'He 4 )T reactions. The results of this approximated method are compared with reference benchmarks which were generated by the nuclear codes ANISN (discrete ordinates) and MORSE (Monte Carlo method). The maximum deviation among the results have been around 10%. (Author) [pt

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

    International Nuclear Information System (INIS)

    Vitela, E. Javier; Martinell, J. Julio

    2000-01-01

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

  6. HIA, the next step: Defining models and roles

    International Nuclear Information System (INIS)

    Putters, Kim

    2005-01-01

    If HIA is to be an effective instrument for optimising health interests in the policy making process it has to recognise the different contests in which policy is made and the relevance of both technical rationality and political rationality. Policy making may adopt a rational perspective in which there is a systematic and orderly progression from problem formulation to solution or a network perspective in which there are multiple interdependencies, extensive negotiation and compromise, and the steps from problem to formulation are not followed sequentially or in any particular order. Policy problems may be simple with clear causal pathways and responsibilities or complex with unclear causal pathways and disputed responsibilities. Network analysis is required to show which stakeholders are involved, their support for health issues and the degree of consensus. From this analysis three models of HIA emerge. The first is the phases model which is fitted to simple problems and a rational perspective of policymaking. This model involves following structured steps. The second model is the rounds (Echternach) model that is fitted to complex problems and a network perspective of policymaking. This model is dynamic and concentrates on network solutions taking these steps in no particular order. The final model is the 'garbage can' model fitted to contexts which combine simple and complex problems. In this model HIA functions as a problem solver and signpost keeping all possible solutions and stakeholders in play and allowing solutions to emerge over time. HIA models should be the beginning rather than the conclusion of discussion the worlds of HIA and policymaking

  7. Combined development of international nuclear fusion test reactors

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Ambassadors of the four most important partners (Common Market, Japan, USA and USSR) in the IAEA sponsored INTOR project, met on the 15 and 16 March 1987 in Vienna under the auspices of the IAEA. A press release was issued acknowledging the considerable technical progress made in magnetic nuclear fusion research. Future design concepts, assistance in research and development work and other activities towards the provision of an international experimental thermonuclear reactor were discussed. (G.T.H.)

  8. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J. [KAERI, Taejon (Korea, Republic of); Vien, Luong Ba; Dien, Nguyen Nhi [Vietnam Atomic Energy Commission, Hanoi (Viet Nam)

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon.

  9. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J.; Vien, Luong Ba; Dien, Nguyen Nhi

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon

  10. The fusion evaluated data library (FENDL) its processing and related benchmark calculations

    International Nuclear Information System (INIS)

    Muir, D.W.

    1989-01-01

    FENDL is a nuclear data library being assembled by the IAEA Nuclear Data Section, in support of a variety of national and international fusion research projects. Notable examples of such projects are the International Experimental Thermonuclear Reactor (ITER), Fusion Engineering Reactor (FER, Japan), and the Next European Torus (NET). The development of the FENDL library is an approved program of the IAEA and is supported by several IAEA Coordinated Research Programs. It appears to me that the planned FENDL data processing and data testing efforts will be a shared effort, with significant contributions coming from the IAEA itself and from the participating research laboratories and data centers

  11. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  12. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  13. Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

    International Nuclear Information System (INIS)

    Huang Xueqing; Zhang Senru

    1999-01-01

    Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

  14. Comparison of nuclear reactor types of the next generation; Komparativni prikaz novih tipova reaktorskih komercijalnih postrojenja slijedece generacije

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, Z; Kastelan, M [NPP Krsko (Slovenia)

    1992-07-01

    The paper presents a comparison for a selected relevant set of parameters for different commercial nuclear reactor types at the next generation. This parameters overview could serve as the base for the semi-quantitative decision bases for the selection of the future nuclear strategy. The number of advanced reactor designs of the LWR, HWR, GCR and LMR type are presented. Even currently many of them are still on the drawing boards, the concepts and designs should be assessed in the sense of sensible approach for planning the possible future nuclear strategy. (author) Clanek predstavlja usporedbu odabranih bitnih parametara karakteristicnih za razlicite tipove energetskih nuklearnih postrojenja slijedece generacije. Prikazani pregled parametara omogucava osnov za polu kvantitativnu osnovu za odlucivanje u svrhu donosenja odluke oko odrednica buduce strategije uporabe nuklearne energije. Brojni koncepti naprednih nuklearnih reaktora tipa LWR, HWR, GCR i LMR su prezentirani. S obzirom na cinjenicu da se mnogi of prezentiranih nalaze jos uvijek na crtacim daskama projektanata, koncepti i projekti koji su iz njih proizasli zahtijevaju analizu u smislu kvalitativnog pristupa planiranja moguce buduce nuklearne startegije. (author)

  15. Power supply for magnetic coils in thermonuclear devices

    International Nuclear Information System (INIS)

    Shimada, Ryuichi; Tamura, Sanae; Kishimoto, Hiroshi.

    1981-01-01

    Purpose: To decrease the load fluctuations in an external power supply, as well as to increase the operation efficiency capacity of thermonuclear devices. Constitution: Electrical power with the same frequency as that of a dynamo generator is supplied by a power supply-driving power source including a frequency converter and the like to DC converters for driving plasma-exciting and -controlling coils. At the same time, the electrical power from the frequency converter is supplied to the dynamo generator with flywheel to add accumulate energies to the EC converters. Accordingly, the energy for the great power pulses in a short time comprises the sum of the energy supplied from the dynamo generator with flywheel and the energy supplied continuously from the outside to eliminate the need of providing a stand-by period for the re-acceleration of the dynamo generator with flywheel even if the scale of the thermonuclear device is enlarged and energy consumed in one cycle is increased, whereby the decrease in the operation efficiency can be prevented and the capacity of the flywheel can be reduced. (Yoshino, Y.)

  16. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  17. Atomic and molecular physics of controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Joachain, C.J.; Post, D.E.

    1983-01-01

    This book attempts to provide a comprehensive introduction to the atomic and molecular physics of controlled thermonuclear fusion, and also a self-contained source from which to start a systematic study of the field. Presents an overview of fusion energy research, general principles of magnetic confinement, and general principles of inertial confinement. Discusses the calculation and measurement of atomic and molecular processes relevant to fusion, and the atomic and molecular physics of controlled thermonuclear research devices. Topics include recent progress in theoretical methods for atomic collisions; current theoretical techniques for electron-atom and electronion scattering; experimental aspects of electron impact ionization and excitation of positive ions; the theory of charge exchange and ionization by heavy particles; experiments on electron capture and ionization by multiply charged ions; Rydberg states; atomic and molecular processes in high temperature, low-density magnetically confined plasmas; atomic processes in high-density plasmas; the plasma boundary region and the role of atomic and molecular processes; neutral particle beam production and injection; spectroscopic plasma diagnostics; and particle diagnostics for magnetic fusion experiments

  18. Forecasting parameters of a the monuclear power plant with a torsatron reactor

    International Nuclear Information System (INIS)

    Artyugina, I.M.; Semenov, A.A.; Smirnov, A.N.

    1982-01-01

    A number of problems related to forecasting technical economical factors of thermonuclear electric plant (TNPP) based on the torsatron reactor is considered. Possible methodic approaches to the estimation of TNPP nonstandard equipment construction-mounting works and the results of forecasting the investment structure in TNPP are analysed. The influence of TP basic systems on the total investment value depending on accepted price level is shown. Quantitative estimations of specific investments and electric energy production cost permit to estimate rather optimistically the considered TNPP type and to draw a conclusion on advisability of the further study

  19. Pulsating Instability of Turbulent Thermonuclear Flames in Type Ia Supernovae

    Science.gov (United States)

    Poludnenko, Alexei Y.

    2014-01-01

    Presently, one of the main explosion scenarios of type Ia supernovae (SNIa), aimed at explaining both "normal" and subluminous events, is the thermonuclear incineration of a white-dwarf in a single-degenerate system. The underlying engine of such explosions is the turbulent thermonuclear flame. Modern, large-scale, multidimensional simulations of SNIa cannot resolve the internal flame structure, and instead must include a subgrid-scale prescription for the turbulent-flame properties. As a result, development of robust, parameter-free, large-scale models of SNIa crucially relies on the detailed understanding of the turbulent flame properties during each stage of the flame evolution. Due to the complexity of the flame dynamics, such understanding must be validated by the first-principles direct numerical simulations (DNS). In our previous work, we showed that sufficiently fast turbulent flames are inherently susceptible to the development of detonations, which may provide the mechanism for the deflagration-to-detonation transition (DDT) in the delayed-detonation model of SNIa. Here we extend this study by performing detailed analysis of the turbulent flame properties at turbulent intensities below the critical threshold for DDT. We carried out a suite of 3D DNS of turbulent flames for a broad range of turbulent intensities and system sizes using a simplified, single-step, Arrhenius-type reaction kinetics. Our results show that at the later stages of the explosion, as the turbulence intensity increases prior to the possible onset of DDT, the flame front will become violently unstable. We find that the burning rate exhibits periodic pulsations with the energy release rate varying by almost an order of magnitude. Furthermore, such flame pulsations can produce pressure waves and shocks as the flame speed approaches the critical Chapman-Jouguet deflagration speed. Finally, in contrast with the current theoretical understanding, such fast turbulent flames can propagate at

  20. Research and development on next generation reactor (phase I)

    International Nuclear Information System (INIS)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author)