WorldWideScience

Sample records for experimental reactor iter

  1. 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

  2. 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

  3. 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.

  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. 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

  6. 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)

  7. 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

  8. 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

  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. 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)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. Cryogenic structures of superconducting coils for fusion experimental reactor 'ITER'

    International Nuclear Information System (INIS)

    Nakajima, Hideo; Iguchi, Masahide; Hamada, Kazuya; Okuno, Kiyoshi; Takahashi, Yoshikazu; Shimamoto, Susumu

    2013-01-01

    This paper describes both structural materials and structural design of the Toroidal Field (TF) coil and Central Solenoid (CS) for the International Thermonuclear Experimental Reactor (ITER). All the structural materials used in the superconducting coil system of the ITER are austenitic stainless steels. Although 316LN is used in the most parts of the superconducting coil system, the cryogenic stainless steels, JJ1 and JK2LB, which were newly developed by the Japan Atomic Energy Agency (JAEA) and Japanese steel companies, are used in the highest stress area of the TF coil case and the whole CS conductor jackets, respectively. These two materials became commercially available based on demonstration of productivity and weldability of materials, and evaluations of 4 K mechanical properties of trial products including welded parts. Structural materials are classified into five grades depending on stress distribution in the TF coil case. JAEA made an industrial specification for mass production based on the ITER requirements. In order to simplify quality control in mass production, JAEA has used materials specified in the material section of 'Codes for Fusion Facilities - Rules on Superconducting Magnet Structure (2008)' issued by the Japan Society of Mechanical Engineers (JSME) in October 2008, which was established using an extrapolation method of 4 K material strengths from room temperature strength and chemical compositions developed by JAEA. It enables steel suppliers to easily control the quality of products at room temperature. JAEA has already started actual production with several manufacturing companies. The first JJ1 product to be used in the TF coil case and the first JK2LB jackets for CS were completed in October and September 2013, respectively. (author)

  3. 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

  4. 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

  5. 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.)

  6. [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.''

  7. 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)

  8. Conceptual design Fusion Experimental Reactor (FER/ITER)

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Nagashima, Takashi; Ikeda, Yoshitaka

    1991-11-01

    This report describes a conceptual design of Lower Hybrid Wave (LH) system for FER and ITER. In JAERI, the conceptual design of LH system for FER has been performed in these 3 years in parallel to that of ITER. There must be a common design part with ITER and FER. The physical requirement of LH system is the saving of volt · sec in the current start-up phase, and the current drive at the boundary region. The frequency of 5GHz is mainly chosen for avoidance of the α particle absorption and for the availability of electron tube development. Seventy-two klystrons (FER) and one hundred klystrons (ITER) are necessary to inject the 30 MW (FER) and 45-50 MW (ITER) rf power into plasma using 0.7 - 0.8 MW klystron per one tube. The launching system is the multi-junction type and the rf spectrum must be as sharp as possible with high directivity to improve the current drive efficiency. One port (FER) and two ports (ITER) are used and the injection direction is in horizontal, in which the analysis of the ray-tracing code and the better coupling of LH wave is considered. The transmission line is over-sized waveguide with low rf loss. (author)

  9. 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

  10. 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

  11. 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

  12. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  13. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  14. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Science.gov (United States)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  15. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    International Nuclear Information System (INIS)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-01-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed

  16. ITER at the international conference on fusion reactor materials

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.; Matera, R.

    1998-01-01

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

  17. 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.

  18. ITER...ation

    International Nuclear Information System (INIS)

    Troyon, F.

    1997-01-01

    Recurrent attacks against ITER, the new generation of tokamak are a mix of political and scientific arguments. This short article draws a historical review of the European fusion program. This program has allowed to build and manage several installations in the aim of getting experimental results necessary to lead the program forwards. ITER will bring together a fusion reactor core with technologies such as materials, superconductive coils, heating devices and instrumentation in order to validate and delimit the operating range. ITER will be a logical and decisive step towards the use of controlled fusion. (A.C.)

  19. 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)

  20. ITER: a technology test bed for a fusion reactor

    International Nuclear Information System (INIS)

    Huguet, M.; Green, B.J.

    1996-01-01

    The ITER Project aims to establish nuclear fusion as an energy source that has potential safety and environmental advantages, and to develop the technologies required for a fusion reactor. ITER is a collaborative project between the European Union, Japan, the Russian Federation and the United States of America. During the current phase of the Project, an R and D programme of about 850 million dollars is underway to develop the technologies required for ITER. This technological effort should culminate in the construction of the components and systems of the ITER machine and its auxiliaries. The main areas of technological development include the first wall and divertor technology, the blanket technology and tritium breeding, superconducting magnet technology, pulsed power technology and remote handling. ITER is a test bed and an essential step to establish the technology of future fusion reactors. Many of the ITER technologies are of potential interest to other fields and their development is expected to benefit the industries involved. (author)

  1. 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

  2. New technology and neo-science on the nuclear fusion reactor engineering. ITER and super high speed phenomena

    International Nuclear Information System (INIS)

    1996-12-01

    This research meeting has been held under cooperation of the ''nuclear fusion reactor engineering research group'' and ''nuclear fusion reactor materials research group'' of the Yayoi Research Group. This meeting was planned and conducted for 2 days under the following predominant thema: Present status of research on thermo-nuclear fusion experimental reactor engineering design (ITER/EDA) and its promoting method in Japan, and a new scientific side in the research and development of nuclear fusion reactor materials or the super high speed phenomena. In the former item, the following reports were published: Creative period of R and D on the nuclear fusion reactor, present statue and future development of ITER/EDA, meanings of ITER under realization of the nuclear fusion energy, and others. And in the latter item, the following reports were published: Nuclear fusion materials engineering and system quantum engineering, dynamic imagination of atom and molecule using pulse snap shot method, laser wake field acceleration and ultra short x-ray pulse generation, development of T-cube laser in JAERI, and others. (G.K.)

  3. 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.

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

    International Nuclear Information System (INIS)

    Janeschitz, G.; Bahm, W.

    2007-01-01

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

  5. 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

  6. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

  7. 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

  8. Conceptual design of SC magnet system for ITER, (6)

    International Nuclear Information System (INIS)

    Yoshida, Kiyoshi; Sugimoto, Makoto; Tsuji, Hiroshi

    1991-08-01

    The International Thermonuclear Experimental Reactor (ITER) is an experimental thermonuclear tokamak reactor in order to test the basic physics performance and technologies. The conceptual design activity (CDA) of ITER required the joint work at a technical site at the Max Plank Institute for Plasma Physics in the Garching, Germany from 1988 to 1990. The technical proposals from Japan were summarized by the Fusion Experimental Reactor (FER) Team and the Superconducting Magnet Laboratory of the Japan Atomic Energy Research Institute (JAERI). This paper describes the Japanese contributions of the R and D proposals to the magnet system for the ITER. These proposals were discussed in ITER CDA design team and summarized in ITER Technical report No. 20. The development program of Toroidal Field Coil is basically proposed from Japan with the design and analysis reports. The Japanese proposals are almost adopted in the ITER Long-Term R and D program. (author)

  9. India's participation in the ITER (International Thermonuclear Experimental Reactor) collaboration

    International Nuclear Information System (INIS)

    Deshpande, Shishir

    2012-01-01

    Keeping its vision of developing fusion energy as a viable source, India joined the ITER collaboration in December 2005. ITER is a seven party collaboration with China, EU, India, Japan, S. Korea, Russia and the USA. ITER has a challenging mission of achieving Q=10 figure of merit at 500 MW fusion power output. The construction of ITER is structured as a set of 'in-kind' procurement packages to be executed by the partners. This involves all activities like design, prototyping, testing, shipping and assembly with commissioning at the ITER site at Cadarache, France. Currently, ITER presents the only opportunity to carry out novel experiments with burning plasmas and the new realms of fusion physics. It is important to participate in such experiments with a view for their exploitation in future. This talk summarizes the ITER device, its key challenges, role played by India and how these enmesh with the future of domestic program in fusion research. (author)

  10. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  11. ETR/ITER systems code

    International Nuclear Information System (INIS)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

  12. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  13. ITER concept definition. V.2

    International Nuclear Information System (INIS)

    1989-01-01

    Volume II of the two volumes describing the concept definition of the International Thermonuclear Experimental Reactor deals with the ITER concept in technical depth, and covers all areas of design of the ITER tokamak. Included are an assessment of the current database for design, scoping studies, rationale for concepts selection, performance flexibility, the ITER concept, the operations and experimental/testing program, ITER parameters and design phase schedule, and research and development specific to ITER. This latter includes a definition of specific research and development tasks, a division of tasks among members, specific milestones, required results, and schedules. Figs and tabs

  14. The ITER activity

    International Nuclear Information System (INIS)

    Glass, A.J.

    1991-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a collaboration among four parties, the United States, the Soviet Union, Japan, and the European Communities, to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes. ITER will demonstrate this through the construction of a tokamak fusion reactor capable of generating 1000 megawatts of fusion power. The ITER project has three missions, as follows: (1) Physics mission -- to demonstrate ignition and controlled burn, with pulse durations from 200 to 1000 S; (2) Technology mission -- to demonstrate the technologies essential to a reactor in an integrated system, operating with high reliability and availability in pulsed operation, with steady-state operation as the ultimate goal; and (3) Testing mission -- to test nuclear and high-heat-flux components at flux levels for 1 mw/m 2 , and fluences of order 1 mw-yr/m 2

  15. Conceptual design of fusion experimental reactor (FER/ITER)

    International Nuclear Information System (INIS)

    Kimura, Haruyuki; Saigusa, Mikio; Saitoh, Yasushi

    1991-06-01

    Conceptual design of the Ion Cyclotron Wave (ICW) system for FER and Japanese contribution to the conceptual design of the ITER ICW system are presented. A frequency range of the FER ICW system is 50-85 MHz, which covers 2ω cT heating, current drive by transit time magnetic pumping (TTMP) and 2ω cD heating. Physics analyses show that the FER and the ITER ICW systems are suitable for the central ion heating and the burn control. The launching systems of the FER ICW system and the ITER high frequency ICW system are characterized by in-port plug and ridged-waveguide-fed 5x4 phased loop array. Merits of those systems are (1) a ceramic support is not necessary inside the cryostat and (2) remote maintenance of the front end part of the launcher is relatively easy. Overall structure of the launching system is consistent with radiation shielding, cooling, pumping, tritium safety and remote maintenance. The launcher has injection capability of 20 MW in the frequency range of 50-85 MHz with the separatrix-antenna distance of 15 cm and steep scrape-off density profile of H-mode. The shape of the ridged waveguide is optimized to provide desired frequency range and power handling capability with a finite element method. Matching between the current strap and the ridged waveguide is satisfactorily good. Thermal analysis of the Faraday shield shows that high electric conductivity low Z material such as beryllium should be chosen for a protection tile of the Faraday shield. Thick Faraday shield is necessary to tolerate electromagnetic force during disruptions. R and D needs for the ITER/FER ICW systems are identified and gain from JT-60/60U ICRF experiments and operations are indicated in connection with them. (author)

  16. International Thermonuclear Experimental Reactor (ITER) neutral beam design

    International Nuclear Information System (INIS)

    Myers, T.J.; Brook, J.W.; Spampinato, P.T.; Mueller, J.P.; Luzzi, T.E.; Sedgley, D.W.

    1990-10-01

    This report discusses the following topics on ITER neutral beam design: ion dump; neutralizer and module gas flow analysis; vacuum system; cryogenic system; maintainability; power distribution; and system cost

  17. 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

  18. ITER power electrical networks

    International Nuclear Information System (INIS)

    Sejas Portela, S.

    2011-01-01

    The ITER project (International Thermonuclear Experimental Reactor) is an international effort to research and development to design, build and operate an experimental facility to demonstrate the scientific and technological possibility of obtaining useful energy from the physical phenomenon known as nuclear fusion.

  19. ITER power electrical networks; Sistemas electricos de alimentacion a los consumidores del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Sejas Portela, S.

    2011-07-01

    The ITER project (International Thermonuclear Experimental Reactor) is an international effort to research and development to design, build and operate an experimental facility to demonstrate the scientific and technological possibility of obtaining useful energy from the physical phenomenon known as nuclear fusion.

  20. Installation of the ITER committee industry. Participants guide; Installation du Comite industrie ITER. Dossier des participants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    ITER is an international project to design and build an experimental fusion reactor based on the tokamak concept. This guide presents the ITER project and objectives and the associated organizations in France, the recommendations and actions for ITER, the industrial mobilization, the industrial committee and its members, technological sheets for the enterprises and the statistical document of the SESSI. (A.L.B.)

  1. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  2. A cryogenic system design for the international thermonuclear experimental reactor (ITER)

    International Nuclear Information System (INIS)

    Slack, D.S.

    1991-01-01

    A conceptual design for ITER was completed last year. The author developed a suitable cryogenic system for ITER as part of this conceptual design effort. An overview of the design is reported. Emphasis is on the fact that cryogenics is a mature science, and a system supporting ITER needs can be made from time-proven components without loss of efficiency or reliability. Because of the large size of the ITER cryogenic system, large numbers of compressors and expanders must be used. Very high reliability is assured by arranging these components in parallel banks where servicing of individual components can be done without interruption of operations. This and other ideas based on the author's experience with Mirror Fusion Test Facility (MFTF) operations are described. 5 refs., 3 figs

  3. Nuclear analysis for ITER

    International Nuclear Information System (INIS)

    Santoro, R.T.; Iida, H.; Khripunov, V.; Petrizzi, L.; Sato, S.; Sawan, M.; Shatalov, G.; Schipakin, O.

    2001-01-01

    This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shut-down. (author)

  4. 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)

  5. US power outage won't dim ITER

    International Nuclear Information System (INIS)

    Lawler, A.

    1996-01-01

    The $8 billion International Thermonuclear Experimental Reactor (ITER) is moving ahead, without definite support of the USA. However, still undecided are where it will be built and how much each partner will pay. This article discusses the international political aspects of building the ITER, with a particular emphasis on the Japanese approach to landing the ITER. Also discussed are possible cost-saving solutions

  6. Stress analyses of ITER toroidal field coils under fault conditions

    International Nuclear Information System (INIS)

    Jong, C.T.J.

    1990-02-01

    The International Thermonuclear Experimental Reactor (ITER) is intended as an experimental thermonuclear tokamak reactor for testing the basic physics, performance and technologies essential to future fusion reactors. The ITER design will be based on extensive new design work, supported by new physical and technological results, and on the great body of experience built up over several years from previous national and international reactor studies. Conversely, the ITER design process should provide the fusion community with valuable insights into what key areas need further development or clarification as we move forward towards practical fusion power. As part of the design process of the ITER toroidal field coils the mechanical behaviour of the magnetic system under fault conditions has to be analysed in more detail. This paper describes the work carried out to create a detailed finite element model of two toroidal field coils as well as some results of linear elastic analyses with fault conditions. The analyses have been performed with the finite element code ANSYS. (author). 5 refs.; 8 figs.; 2 tabs

  7. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  8. ITER EDA and technology

    International Nuclear Information System (INIS)

    Baker, C.C.

    2001-01-01

    The year 1998 was the culmination of the six-year Engineering Design Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project. The EDA results in design and validating technology R and D, plus the associated effort in voluntary physics research, is a significant achievement and major milestone in the history of magnetic fusion energy development. Consequently, the ITER EDA was a major theme at this Conference, contributing almost 40 papers

  9. 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

  10. ITER and world chaos; Iter ou le bouleversement du monde

    Energy Technology Data Exchange (ETDEWEB)

    Pourcel, Eric

    2012-02-15

    ITER is the International Thermonuclear Experimental Reactor: the author here develops three scenarios linked to the control of nuclear fusion as a method of producing electrical energy that could take over from fossil fuels in the twenty-First century. His expose shows the likely strategic disarray that might result

  11. Remote maintenance challenges presented in the ITER engineering design

    International Nuclear Information System (INIS)

    Burgess, T.W.; Herndon, J.N.; Schrock, S.L.; Lousteau, D.C.

    1995-01-01

    Leading fusion energy research institutions are currently engaged in the Engineering Design Activity for the International Thermonuclear Experimental Reactor (ITER). A tokamak reactor design is evolving which emphasizes high system performance in a minimum overall reactor and building size. The resulting high component density dictates careful attention to ITER remote maintenance considerations in the development of the configuration. The complexity and scale of ITER remote maintenance tasks are well beyond the scope of today's experience and technology. This paper discusses the remote maintenance philosophy, describes the basic configuration as it relates to maintenance, and describes the basic procedures and equipment required. Key enabling technology research and development needs are also addressed

  12. Development of tritium fuel processing system using electrolytic reactor for ITER

    International Nuclear Information System (INIS)

    Yamanishi, T.; Kawamura, Y.; Iwai, Y.

    2001-01-01

    The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

  13. Development of tritium fuel processing system using electrolytic reactor for ITER

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Kawamura, Y.; Iwai, Y.

    1999-01-01

    The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

  14. JET contributions to ITER R and D programme

    International Nuclear Information System (INIS)

    Gambier, D.J.; Tubbing, B.J.D.

    1992-08-01

    This report contains the Joint European Torus Project (JET) contributions to the International Thermonuclear Experimental Reactor (ITER) related research and development programme 1991-1992. The contributions, from many JET authors, were gathered in May/June 1992, so that the results of the 1991/92 experimental campaign could be fully incorporated. The contributions are ordered according to the description of tasks of the ITER-related Physics Research and Development programme, described in document ITER-TN-PH-0-7, issued April 30, 1991. (Author)

  15. 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...

  16. 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)

  17. ITER definition phase

    International Nuclear Information System (INIS)

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned as a fusion device which would demonstrate the scientific and technological feasibility of fusion power. As a first step towards achieving this goal, the European Community, Japan, the Soviet Union, and the United States of America have entered into joint conceptual design activities under the auspices of the International Atomic Energy Agency. A brief summary of the Definition Phase of ITER activities is contained in this report. Included in this report are the background, objectives, organization, definition phase activities, and research and development plan of this endeavor in international scientific collaboration. A more extended technical summary is contained in the two-volume report, ''ITER Concept Definition,'' IAEA/ITER/DS/3. 2 figs, 2 tabs

  18. installation of the ITER committee industry. Participants guide

    International Nuclear Information System (INIS)

    2006-01-01

    ITER is an international project to design and build an experimental fusion reactor based on the tokamak concept. This guide presents the ITER project and objectives and the associated organizations in France, the recommendations and actions for ITER, the industrial mobilization, the industrial committee and its members, technological sheets for the enterprises and the statistical document of the SESSI. (A.L.B.)

  19. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  20. Power converters for ITER

    CERN Document Server

    Benfatto, I

    2006-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

  1. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    International Nuclear Information System (INIS)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A; Constans, S; Merola, M; Riccardi, B

    2009-01-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  2. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Science.gov (United States)

    Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

    2009-12-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  3. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

    2009-12-15

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  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. Tritium supply assessment for ITER and DEMOnstration power plant

    International Nuclear Information System (INIS)

    Ni, Muyi; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-01-01

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource

  6. Tritium supply assessment for ITER and DEMOnstration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi, E-mail: muyi.ni@fds.org.cn; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-10-15

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

  7. 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

  8. 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

  9. The physics role of ITER

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1997-04-01

    Experimental research on the International Thermonuclear Experimental Reactor (ITER) will go far beyond what is possible on present-day tokamaks to address new and challenging issues in the physics of reactor-like plasmas. First and foremost, experiments in ITER will explore the physics issues of burning plasmas--plasmas that are dominantly self-heated by alpha-particles created by the fusion reactions themselves. Such issues will include (i) new plasma-physical effects introduced by the presence within the plasma of an intense population of energetic alpha particles; (ii) the physics of magnetic confinement for a burning plasma, which will involve a complex interplay of transport, stability and an internal self-generated heat source; and (iii) the physics of very-long-pulse/steady-state burning plasmas, in which much of the plasma current is also self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achieving and sustaining burning plasma regimes in a tokamak necessarily requires plasmas that are larger than those in present experiments and have higher energy content and power flow, as well as much longer pulse length. Accordingly, the experimental program on ITER will embrace the study of issues of plasma physics and plasma-materials interactions that are specific to a reactor-scale fusion experiment. Such issues will include (i) confinement physics for a tokamak in which, for the first time, the core-plasma and the edge-plasma are simultaneously in a reactor-like regime; (ii) phenomena arising during plasma transients, including so-called disruptions, in regimes of high plasma current and thermal energy; and (iii) physics of a radiative divertor designed for handling high power flow for long pulses, including novel plasma and atomic-physics effects as well as materials science of surfaces subject to intense plasma interaction. Experiments on ITER will be conducted by researchers in control rooms situated at major

  10. Structural design of the superconducting toroidal field coils for ITER

    International Nuclear Information System (INIS)

    Wong, F.M.G.; Sborchia, C.; Thome, R.J.; Malkov, A.; Titus, P.H.

    1995-01-01

    Structural design issues and features of the superconducting toroidal field (TF) coils for the International Thermonuclear Experimental Reactor (ITER) will be discussed. Selected analyses of the structural and mechanical behavior of the ITER TF coils will also be presented. (orig.)

  11. Status of the Design Tool Development for ITER TBM and Fusion Reactor System in Korea

    International Nuclear Information System (INIS)

    Jin, H. G.; Lee, D. W.; Shin, K. I.; Lee, E. H.; Yoon, J. S.; Kim, S. K.; Ahn, M. Y.; Cho, S.

    2013-01-01

    Korea has developed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) and Helium Cooled Ceramic Reflector (HCCR) TBM to be tested in the ITER. The main purpose for developing the TBM is to develop the design technology for the DEMO and fusion reactor, and it should be proved experimentally in the ITER. Therefore, we have developed the design scheme and codes including the safety analysis capability for obtaining the license for testing in the ITER. In this study, the current status of the design tool development is summarized. For developing the design scheme and system codes of the ITER TBM program in Korea, the developed system codes such as MARS and GAMMA+ from Gen. IV projects were modified and verified considering the fusion application. For He coolant, 3D analysis and a McEligot correlation as the heat transfer model were proposed and validated considering the high heat from the plasma side and extreme temperature difference between the wall and fluid. For tritium behavior in the He coolant, the TBEC+GAMMA code was developed, and the oxidation layer growth and its permeation rate change were considered in this development. For a liquid metal breeder such as PbLi and Li, GAMMA-FR was developed including physical properties of the generation model and basic heat transfer model in them. For MHD simulation, the Miyazaki model was implemented in GAMMA, and it was validated successfully with the experimental data. Extending the capability of GAMMA-FR, a fusion system design code (SUPERCODE) is going to be coupled with a 3D neutronics code (MCNP)

  12. Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

    International Nuclear Information System (INIS)

    Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

    1991-01-01

    An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

  13. 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

  14. RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

  15. ITER review team takes bullish stance

    International Nuclear Information System (INIS)

    Lawler, A.

    1997-01-01

    A large team of U.S. fusion researchers last week began poring over the latest blueprints for a massive international machine designed to demonstrate fusion power and provide plasma physicists with an exciting new facility. The review of the $10 billion International Thermonuclear Experimental Reactor (ITER) was prompted by controversy over the reactor's design and the shrinking U.S. fusion budget

  16. Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

    Science.gov (United States)

    Borisov, A. A.; Deryabina, N. A.; Markovskij, D. V.

    2017-12-01

    Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing "expensive" experiments, is used for numerical integration of 235U and 238U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with 235U chambers and calibrate 238U chambers by responses of the 235U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor

  17. ITER conceptual design report

    International Nuclear Information System (INIS)

    1991-01-01

    Results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity (CDA) are reported. This report covers the Terms of Reference for the project: defining the technical specifications, defining future research needs, define site requirements, and carrying out a coordinated research effort coincident with the CDA. Refs, figs and tabs

  18. Fusion Power measurement at ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bertalot, L.; Barnsley, R.; Krasilnikov, V.; Stott, P.; Suarez, A.; Vayakis, G.; Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-07-01

    Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also to the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)

  19. ITER Conceptual design: Interim report

    International Nuclear Information System (INIS)

    1990-01-01

    This interim report describes the results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activities after the first year of design following the selection of the ITER concept in the autumn of 1988. Using the concept definition as the basis for conceptual design, the Design Phase has been underway since October 1988, and will be completed at the end of 1990, at which time a final report will be issued. This interim report includes an executive summary of ITER activities, a description of the ITER device and facility, an operation and research program summary, and a description of the physics and engineering design bases. Included are preliminary cost estimates and schedule for completion of the project

  20. ITER at Cadarache

    International Nuclear Information System (INIS)

    2005-06-01

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  1. 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

  2. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  3. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  4. ITER safety and operational scenario

    International Nuclear Information System (INIS)

    Shimomura, Y.; Saji, G.

    1998-01-01

    The safety and environmental characteristics of ITER and its operational scenario are described. Fusion has built-in safety characteristics without depending on layers of safety protection systems. Safety considerations are integrated in the design by making use of the intrinsic safety characteristics of fusion adequate to the moderate hazard inventories. In addition to this, a systematic nuclear safety approach has been applied to the design of ITER. The safety assessment of the design shows how ITER will safely accommodate uncertainties, flexibility of plasma operations, and experimental components, which is fundamental in ITER, the first experimental fusion reactor. The operation of ITER will progress step by step from hydrogen plasma operation with low plasma current, low magnetic field, short pulse and low duty factor without fusion power to deuterium-tritium plasma operation with full plasma current, full magnetic field, long pulse and high duty factor with full fusion power. In each step, characteristics of plasma and optimization of plasma operation will be studied which will significantly reduce uncertainties and frequency/severity of plasma transient events in the next step. This approach enhances reliability of ITER operation. (orig.)

  5. Construction Safety Forecast for ITER

    Energy Technology Data Exchange (ETDEWEB)

    cadwallader, lee charles

    2006-11-01

    The International Thermonuclear Experimental Reactor (ITER) project is poised to begin its construction activity. This paper gives an estimate of construction safety as if the experiment was being built in the United States. This estimate of construction injuries and potential fatalities serves as a useful forecast of what can be expected for construction of such a major facility in any country. These data should be considered by the ITER International Team as it plans for safety during the construction phase. Based on average U.S. construction rates, ITER may expect a lost workday case rate of < 4.0 and a fatality count of 0.5 to 0.9 persons per year.

  6. Experimental evaluation of tritium permeation through stainless steel tubes of heat exchanger from primary to secondary water in ITER

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Nishi, Masataka

    2004-01-01

    Tritium permeation through heat exchanger from primary cooling water to secondary cooling water has been investigated experimentally with SS316L heat exchanger under simulated ITER (international thermonuclear experimental reactor) operation condition in order to establish the tritium permeation evaluation method through the heat exchanger. As the result, the permeation rate of aqueous tritium was found to be about three orders smaller than that of the gaseous tritium. Tritium permeation through the heat exchanger in ITER was then evaluated, and it was revealed that total tritium permeation amount based on obtained aqueous permeability was about one order less than that with the former method with the gaseous permeability and putting the permeation reduction factor as 1000. Evaluated tritium permeation amount into secondary water during 20 years was quite small, which could be considered as negligible from the safety viewpoint

  7. 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.)

  8. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  9. 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

  10. 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

  11. 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.

  12. Design considerations for ITER magnet systems

    International Nuclear Information System (INIS)

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

    1989-01-01

    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 authors present here preliminary ITER magnet systems design parameters taken from trade studies, design, 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

  13. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huguet, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.R.; Shimada, M.; Aymar, R.; Chuyanov, V.A.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties. (author)

  14. Household energy consumption: the future is in our hands. ITER, an experimental fusion reactor. Do CO2-free energies exist? Liquefied natural gas, king of the gas market

    International Nuclear Information System (INIS)

    Anon.

    2008-01-01

    This issue of Alternatives newsletter features 4 main articles dealing with: 1 - Household energy consumption - the future is in our hands: With energy resources growing scarcer and more expensive, everyone has a duty to conserve energy. Because combating global warming also means adopting simple habits and using the right equipment - with help from our governments to lead us to change. A practical look at what we can do. 2 - ITER, an experimental fusion reactor: The entire international community is trying to reproduce here on Earth the fusion of hydrogen atoms occurring naturally in the Sun, lured by the promise of a virtually inexhaustible source of energy. More on ITER from the project's Director General. 3 - Do CO 2 -free energies exist?: As nations struggle to reduce greenhouse gas emissions, the question is moot. Environmental engineer Jean-Marc Jancovici gives us his point of view. 4 - Liquefied natural gas, king of the gas market: LNG's many advantages are enticing industry to develop supply routes and infrastructure to meet strong demand. But the race for LNG is not without its limits

  15. Physics fundamentals for ITER

    International Nuclear Information System (INIS)

    Rosenbluth, M.N.

    1999-01-01

    The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an 'empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should

  16. Present activities for the preparation of a Japanese draft of structural design guidelines for the experimental fusion reactor

    International Nuclear Information System (INIS)

    Miya, K.; Muto, Y.; Takatsu, H.; Hada, K.; Koizumi, K.; Jitsukawa, S.; Arai, T.; Ohkawa, Y.; Shimakawa, T.; Aoto, K.; Shiraishi, H.; Takagi, T.; Miki, N.; Takahashi, S.; Sato, K.; Takemasa, F.; Kasaba, M.; Kudough, F.; Fujita, J.; Kajiura, S.; Kinoshita, S.

    1996-01-01

    Since November 1990, systematic research has been carried out in preparation for a Japanese draft of structural design guidelines for the experimental fusion reactor. This report summarizes the major results of the work and the status of these efforts. A classification of components and definition of operating conditions are proposed on the basis of the ITER-CDA design, in the light of the safety characteristics of the fusion reactor and relevant conventions for the existing fission reactor design code. Specific issues regarding the structural design of the experimental fusion reactor are discussed based on the experimental and analytical work. The validity of the existing structural design method is confirmed for the use of irradiated 316 SS, irrespective of the significant reduction in uniform elongation capability caused by heavy neutron irradiation. Further important phenomena are treated such as magnetic damping, magnetic stiffness and fracture due to electromagnetic forces. Finally, the issues concerned with welding and non-destructive examinations are discussed with relevance to component classification. (orig.)

  17. ITER neutral beam system

    International Nuclear Information System (INIS)

    Mondino, P.L.; Di Pietro, E.; Bayetti, P.

    1999-01-01

    The Neutral Beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration with the tokamak and with the rest of the plant. Operational requirements and maintainability have been considered in the design. The paper considers the integration with the tokamak, discusses design improvements which appear necessary and finally notes R and D progress in key areas. (author)

  18. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

  19. ITER must make its case

    International Nuclear Information System (INIS)

    1998-01-01

    Last month, as expected, the four partners in the International Thermonuclear Experimental Reactor (ITER) project announced a three-year extension of the ITER engineering design activity. Detailed design work on the next-generation fusion-energy device started in 1992 and has cost about $1 bn so far. A decision to build the device, once scheduled to be taken this year, will now be made in 2001 at the earliest. The ITER council said that the extension would ''provide the framework for undertaking jointly site(s)-specific and other activities with the aim of enabling future decision on construction and operation of ITER''. What the project is really doing is buying time as it tries to find a cheaper option that the partners will find acceptable. The US is keen to cut the project's cost by two-thirds. (author)

  20. Involvement of the EU industry in ITER EDA

    International Nuclear Information System (INIS)

    Bogusch, E.

    2001-01-01

    Since the fifties, European industry has been involved in research and development in the field of nuclear fusion as a potential future source of energy. Early contributions mainly included deliveries of plant components and services to experimental facilities. In the Engineering Design Activities (EDA) phase of the planned multinational International Thermonuclear Experimental Reactor (ITER) in 1993 to 2001 this commitment of industry was intensified. Industries from seven European countries participated in the project with various contributions, e.g., in the development, design, and manufacture of components, and in the development of methods of planning and executing the complex ITER project. These activities were accompanied by an extensive R and D program. e.g., about materials and methods of manufacturing ITER components. In this way, European industry made an important contribution to the further development of nuclear fusion within the framework of ITER EDA activities, and will be able to continue this work intensively in the expected ITER construction phase to follow. (orig.) [de

  1. Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

    2013-07-01

    The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

  2. Neutron spectrometer for DD/DT burning ratio measurement in fusion experimental reactor

    International Nuclear Information System (INIS)

    Asai, Keisuke; Naoi, Norihiro; Iguchi, Tetsuo; Watanabe, Kenichi; Kawarabayashi, Jun; Nishitani, Takeo

    2006-01-01

    The most feasible fuels for a fusion reactor are D (Deuterium) and T (Tritium). DD and/or DT fusion reaction or nuclear burning reaction provides two kinds of neutrons, DD neutron and DT neutron, respectively. DD/DT burning ratio, which can be estimated by DD/DT neutron ratio in the burning plasma, is essential for burn control, alpha particle emission rate monitoring and tritium fuel cycle estimation. Here we propose a new neutron spectrometer for the absolute DD/DT burning ratio measurement. The system consists of a Proton Recoil Telescope (PRT) and a Time-of-Flight (TOF) technique. We have conducted preliminary experiments with a prototype detector and a DT neutron beam (φ20 mm) at the Fusion Neutronics Source, Japan Atomic Energy Agency (JAEA), to assess its basic performance. The detection efficiency obtained by the experiment is consistent with the calculation results in PRT, and sufficient energy resolution for the DD/DT neutron discrimination has been achieved in PRT and TOF. The validity of the Monte Carlo calculation has also been confirmed by comparing the experimental results with the calculation results. The design consideration of this system for use in ITER (International Thermonuclear Experimental Reactor) has shown that this system is capable of monitoring the line-integrated DD/DT burning ratio for the plasma core line of sight with the required measurement accuracy of 20% in the upper 4 decades of the ITER operation (fusion power: 100 kW-700 MW). (author)

  3. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huget, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.; Shimada, M.; Aymar, R.; Chuyanov, V.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first 10 years' operation will be devoted primarily to physics issues at low neutron fluence and the following 10 years' operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes such as inductive high Q modes, long pulse hybrid modes, non-inductive steady-state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours per day but also in involving the world-wide fusion communities and in promoting scientific competition among the Parties. (author)

  4. 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

  5. ITER concept definition. V.1

    International Nuclear Information System (INIS)

    1989-01-01

    Under the auspices of the International Atomic Energy Agency (IAEA), an agreement among the four parties representing the world's major fusion programs resulted in a program for conceptual design of the next logical step in the fusion program, the International Thermonuclear Experimental Reactor (ITER). The definition phase, which ended in November, 1989, is summarized in two reports: a brief summary is contained in the ITER Definition Phase Report (IAEA/ITER/DS/2); the extended technical summary and technical details of ITER are contained in this two-volume report. The first volume of this report contains the Introduction and Summary, and the remainder will appear in Volume II. In the Conceptual Design Activities phase, ITER has been defined as being a tokamak device. The basic performance parameters of ITER are given in Volume I of this report. In addition, the rationale for selection of this concept, the performance flexibility, technical issues, operations, safety, reliability, cost, and research and development needed to proceed with the design are discussed. Figs and tabs

  6. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  7. Neoclassical tearing mode stabilization by ECCD in ITER

    International Nuclear Information System (INIS)

    Giruzzi, G.; Zabiego, M.

    2001-01-01

    A dynamic model, based on a 3-D Fokker-Planck code coupled to the island evolution equations, is used to evaluate the feasibility of active control of Neoclassical Tearing modes by Electron Cyclotron Current Drive (ECCD) in International Thermonuclear Experimental Reactor (ITER). The parameters of the present version of ITER, i.e., RTO/RC ITER (IAM option) are used. Both m=3, n=2 and m=2, n=1 modes are considered. It is shown that an Electron Cyclotron wave system at 140 GHz, with toroidally steerable antennas, can stabilize both modes simultaneously if a power ≥30 MW is available

  8. 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

  9. The ITER reduced cost design

    International Nuclear Information System (INIS)

    Aymar, R.

    2000-01-01

    Six years of joint work under the international thermonuclear experimental reactor (ITER) EDA agreement yielded a mature design for ITER which met the objectives set for it (ITER final design report (FDR)), together with a corpus of scientific and technological data, large/full scale models or prototypes of key components/systems and progress in understanding which both validated the specific design and are generally applicable to a next step, reactor-oriented tokamak on the road to the development of fusion as an energy source. In response to requests from the parties to explore the scope for addressing ITER's programmatic objective at reduced cost, the study of options for cost reduction has been the main feature of ITER work since summer 1998, using the advances in physics and technology databases, understandings, and tools arising out of the ITER collaboration to date. A joint concept improvement task force drawn from the joint central team and home teams has overseen and co-ordinated studies of the key issues in physics and technology which control the possibility of reducing the overall investment and simultaneously achieving the required objectives. The aim of this task force is to achieve common understandings of these issues and their consequences so as to inform and to influence the best cost-benefit choice, which will attract consensus between the ITER partners. A report to be submitted to the parties by the end of 1999 will present key elements of a specific design of minimum capital investment, with a target cost saving of about 50% the cost of the ITER FDR design, and a restricted number of design variants. Outline conclusions from the work of the task force are presented in terms of physics, operations, and design of the main tokamak systems. Possible implications for the way forward are discussed

  10. Establishment of ITER: Relevant documents

    International Nuclear Information System (INIS)

    1988-01-01

    At the Geneva Summit Meeting in November, 1985, a proposal was made by the Soviet Union to build a next-generation tokamak experiment on a collaborative basis involving the world's four major fusion blocks. In October, 1986, after consulting with Japan and the European Community, the United States responded with a proposal on how to implement such an activity. Ensuing diplomatic and technical discussions resulted in the establishment, under the auspices of the IAEA, of the International Thermonuclear Experimental Reactor Conceptual Design Activities. This tome represents a collection of all documents relating to the establishment of ITER, beginning with the initial meeting of the ITER Quadripartite Initiative Committee in Vienna on 15-16 March, 1987, through the meeting of the Provisional ITER Council, also in Vienna, on 8-9 February, 1988

  11. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  12. 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)

  13. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    2001-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  14. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    1999-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  15. Poloidal field system for the ITER hard design option

    International Nuclear Information System (INIS)

    Schultz, J.H.; Pillsbury, R.D.

    1992-01-01

    This paper reports on ITER, the International Thermonuclear Experimental Reactor, a collaborative design by the US, EC, Japan, and the USSR of a tokamak fusion reactor that will demonstrate the physics and test the technology needed for commercial fusion reactors. In 1990, the ITER team completed a Conceptual Design Activity (CDA) in which a candidate design was shown to meet the specified goals of the ITER activity at a conceptual level. The four parties have agreed to an Engineering Design Activity (EDA) that includes the necessary additional design and analysis, along with the R and D needed to construct ITER with confidence. The CDA design includes a toroidal field (TF) magnet system that provides the main containment field and a poloidal field (PF) system used to control plasma current and position. The PF system is also used as transformer primary to induce and sustain current in the plasma. Since the volt-seconds available for full-current plasma burn are less than 10% of the total available volt-seconds from the PF system, an area of concern in the CDA design is that unfavorable plasma conditions could compromise the ability of the physics base case design to achieve long pulse burns. A High Aspect Ratio Design (HARD) was conceived as an alternative design option with a much larger bore in the central solenoid to enhance ITER's capabilities for long-burn operation

  16. Non-iterative method to calculate the periodical distribution of temperature in reactors with thermal regeneration

    International Nuclear Information System (INIS)

    Sanchez de Alsina, O.L.; Scaricabarozzi, R.A.

    1982-01-01

    A matrix non-iterative method to calculate the periodical distribution in reactors with thermal regeneration is presented. In case of exothermic reaction, a source term will be included. A computer code was developed to calculate the final temperature distribution in solids and in the outlet temperatures of the gases. The results obtained from ethane oxidation calculation in air, using the Dietrich kinetic data are presented. This method is more advantageous than iterative methods. (E.G.) [pt

  17. Materials challenges for ITER - Current status and future activities

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V. [ITER International Team, Boltsmannstrasse 2, 85748 Garching (Germany)]. E-mail: valdimir.barabash@iter.org; Peacock, A. [EFDA Close Support Unit, 85748 Garching (Germany); Fabritsiev, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Kalinin, G. [ENES, P.O. Box 788, 101000 Moscow (Russian Federation); Zinkle, S. [Metals and Ceramics Division, ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138 (United States); Rowcliffe, A. [Metals and Ceramics Division, ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138 (United States); Rensman, J.-W. [NRG, P.O. Box 25, 1755 ZG Petten (Netherlands); Tavassoli, A.A. [Commissariat a l' Energie Atomique, CEA/Saclay, 91191 Gif sur Yvette cedex (France); Marmy, P. [CRPP, EPFL, Association EURATOM-Confederation Suisse, 5232, Villigen PSI (Switzerland); Karditsas, P.J. [EURATOM/UKAEA Fusion Association, Abingdon, Oxon OX14 3DB (United Kingdom); Gillemot, F. [AEKI Atomic Research Institute, 1121 Budapest, (Hungary); Akiba, M. [JAEA, Naka-machi, Naka-gun, Ibaraki-ken 311-0193 (Japan)

    2007-08-01

    ITER will be the first experimental fusion facility, which brings together the key physical, material and technological issues related to development of fusion reactors. The design of ITER is complete and the construction will start soon. This paper discusses the main directions of the project oriented materials activity and main challenges related to selection of materials for the ITER components. For each application in ITER the main materials issues were identified and these issues were addressed in the dedicated ITER R and D program. The justification of materials performance was fully documented, which allows traceability and reliability of design data. Several examples are given to illustrate the main achievements and recommendations from the recently updated ITER Materials Properties Handbook. The main ongoing and future materials activities are described.

  18. JET ITER-like wall—overview and experimental programme

    International Nuclear Information System (INIS)

    Matthews, G F; Beurskens, M; Loving, A; Kear, M; Mayoral, M-L; Prior, P; Riccardo, V; Watkins, M L; Brezinsek, S; Groth, M; Joffrin, E; Villedieu, E; Neu, R; Rimini, F; Sips, G; Rubel, M; De Vries, P

    2011-01-01

    This paper reports the successful installation of the JET ITER-like wall and the realization of its technical objectives. It also presents an overview of the planned experimental programme which has been optimized to exploit the new wall and other JET enhancements in 2011/12.

  19. ITER EDA newsletter. V. 3, no. 1. (International Thermonuclear Experimental Reactor Engineering Design Activities)

    International Nuclear Information System (INIS)

    1994-01-01

    This issues of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the Fourth ITER Management Advisory Committee Meeting (MAC) held at San Diego, USA, 13-14 January, 1994, a Technical Committee Meeting on Plasma Equilibrium and Control held at Naka, Japan, 9-12 November 1993, and a Technical Committee Meeting on Radio-Frequency Heating and Current Drive held in Garching, Germany, 21-26 October 1993

  20. Final Report on ITER Task Agreement 81-18

    Energy Technology Data Exchange (ETDEWEB)

    Brad J. Merrill

    2008-02-01

    During 2007, the US International Thermonuclear Experimental Reactor (ITER) Project Office (USIPO) entered into a Task Agreement (TA) with the ITER International Organization (IO) to conduct Research and Development activity and/or Design activity in the area of Safety Analyses. There were four tasks within this TA, which were to provide the ITER IO with: 1) Quality Assurance (QA) documentation for the MELCOR 1.8.2 Fusion code, 2) a pedigreed version of MELCOR 1.8.2, 3) assistance in MELCOR input deck development and accident analyses, and 4) support and assistance in the operation of the MELCOR 1.8.2. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-18.

  1. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  2. ITER merges energies in Provence

    International Nuclear Information System (INIS)

    Barla, J.Ch.

    2009-01-01

    The works around the Cadarache site where the experimental nuclear fusion reactor ITER is to be built have already generated about 366 million euros of contracts and provisions with French companies by September 30, 2009. The advance of the project should bring 3000 to 4000 persons more around the site but the Provence region suffers from the lack of a real projected management of employment and skills. (J.S.)

  3. Outline and status of ITER program

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi; Shimomura, Yasuo

    2002-01-01

    ITER is an international joint program for the next-step fusion experimental reactor which aims to demonstrate extended/steady-state fusion burn of deuterium-tritium plasmas and to demonstrate the fusion technologies in an integrated manner as well as to perform integrated testing of components required to utilize fusion energy for practical purposes. On the basis of the recent scientific and engineering achievements in the world-wide tokamak research, the Engineering Design Activities for nine years were fully completed in July 2001. The so-called compact ITER with a finite Q≥10 was proposed and its detailed engineering design was developed along the line of world fusion research. Large scale engineering research and development were completed for superconducting coils, remote-maintenance technology, etc.. The four ITER Parties (Japan, the European Union, the Soviet Federation, and Canada) have initiated the governmental negotiations for the joint implementation of ITER. (author)

  4. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

  5. Ceramics radiation effects issues for ITER

    International Nuclear Information System (INIS)

    Zinkle, S.J.

    1993-01-01

    The key radiation effects issues associated with the successful operation of ceramic materials in components of the planned International Thermonuclear Experimental Reactor (ITER) are discussed. Radiation-induced volume changes and degradation of the mechanical properties should not be a serious issue for the fluences planned for ITER. On the other hand, radiation-induced electrical degradation effects may severely limit the allowable exposure of ceramic insulators. Degradation of the loss tangent and thermal conductivity may also restrict the location of some components such as ICRH feedthrough insulators to positions far away from the first wall. In-situ measurements suggest that the degradation of physical properties in ceramics during irradiation is greater than that measured in postirradiation tests. Additional in-situ data during neutron irradiation are needed before engineering designs for ITER can be finalized

  6. Plan of ITER remote experimentation center

    Energy Technology Data Exchange (ETDEWEB)

    Ozeki, T., E-mail: ozeki.takahisa@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Obuchi Rokkasho, Kitakami-gun, Aomori 039-3212 (Japan); Clement, S.L. [Fusion for Energy, Torres Diagonal Litoral, B3, 13/03, 08019 Barcelona (Spain); Nakajima, N. [National Institute for Fusion Science and Project Leader of IFERC, 2-166 Obuchi, Rokkasho, Kamikita-gun, Aomori 039-3212 (Japan)

    2014-05-15

    Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.

  7. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  8. ITER: Fusion research at the dawn of a new era

    International Nuclear Information System (INIS)

    Aymar, R.; Chuyanov, V.; Shimomura, Y.; Huguet, M.

    2003-01-01

    Given the expected success of on-going negotiations on the Joint Implementing Agreement for ITER construction and operation, a new era is opening for experimentation with reactor-relevant physics integrated with key reactor technologies in a licensed nuclear environment. The ITER design, cost estimate and safety analysis are supported by a large body of validating physics and technology R and D. The main features of the design, and analysis of its performance, give confidence that it will fulfil its technical objectives and demonstrate the environmental attractiveness of fusion. This paper gives illustrative confirmation of these expectations and an update on the technical preparations for construction, as well as the status of negotiations. (author)

  9. Development of pellet injection systems for ITER

    International Nuclear Information System (INIS)

    Combs, S.K.; Gouge, M.J.; Baylor, L.R.

    1995-01-01

    Oak Ridge National Laboratory (ORNL) has been developing innovative pellet injection systems for plasma fueling experiments on magnetic fusion confinement devices for about 20 years. Recently, the ORNL development has focused on meeting the complex fueling needs of the International Thermonuclear Experimental Reactor (ITER). In this paper, we describe the ongoing research and development activities that will lead to a ITER prototype pellet injector test stand. The present effort addresses three main areas: (1) an improved pellet feed and delivery system for centrifuge injectors, (2) a long-pulse (up to steady-state) hydrogen extruder system, and (3) tritium extruder technology. The final prototype system must be fully tritium compatible and will be used to demonstrate the operating parameters and the reliability required for the ITER fueling application

  10. ITER central solenoid manufacturing R and D

    International Nuclear Information System (INIS)

    Jay Jayakumar, R.; Tsuji, H.; Ohsaki, O.

    2001-01-01

    The International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) includes the development of high performance superconductors, high current joints between superconducting cables and insulating materials. Also in the EDA, the resulting products of this R and D are incorporated in a Central Solenoid Model Coil which utilizes full size conductors. The manufacturing of the model coil and components has led to the development of the design, materials, tooling and process which are fully applicable to the manufacture of the ITER relevant CS coil. The R and D is essentially complete and final stages of the CS Model Coil manufacturing are underway. (author)

  11. ITER central solenoid manufacturing R and D

    International Nuclear Information System (INIS)

    Jayakumar, R.J.; Tsuji, H.; Ohsaki, O.

    1999-01-01

    The International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) includes the development of high performance superconductors, high current joints between superconducting cables and insulating materials. Also in the EDA, the resulting products of this R and D are incorporated in a Central Solenoid Model Coil which utilizes full size conductors. The manufacturing of the model coil and components has led to the development of the design, materials, tooling and process which are fully applicable to the manufacture of the ITER relevant CS coil. The R and D is essentially complete and final stages of the CS Model Coil manufacturing are underway. (author)

  12. Thermal and mechanical design of the plasma core CXRS diagnostics for the fusion reactor ITER; Thermische und mechanische Auslegung der Plasma Core CXRS Diagnostik des ITER Kernfusionsreaktors

    Energy Technology Data Exchange (ETDEWEB)

    Greza, H. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Neubauer, O.; Wolters, J. [Forschungszentrum Juelich GmbH (Germany)

    2009-07-01

    In the frame of the research project ITER (international thermonuclear experimental reactor) the plasma state is monitored using the plasma core diagnostics CXRS (charge exchange recombination spectroscopy).The authors describe the thermal and mechanical design of the first mirror of the CXRS diagnostics. The components of the first mirror are exposed to high heat and neutron irradiation. The surface temperature will be 300 to 400 deg C. The misalignment tolerance is plus or minus 0.1 degree. The maximum mechanical stresses in the mirror have to be minimized. The design calculations use the finite element code ANSYS. The results indicate that the heat input from the plasma can be removed by the coolant flow. Further calculation shave to concern the brazed joints between mirror and cooling block.

  13. Thermal and mechanical design of the plasma core CXRS diagnostics for the fusion reactor ITER; Thermische und mechanische Auslegung der Plasma Core CXRS Diagnostik des ITER Kernfusionsreaktors

    Energy Technology Data Exchange (ETDEWEB)

    Greza, H.; Knauff, R. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany); Neubauer, O.; Wolters, J.; Offermanns, G.; Biel, W. [Forschungszentrum Juelich GmbH (Germany)

    2011-07-01

    In the frame of the research project ITER (international thermonuclear experimental reactor) the plasma state is monitored using the plasma core diagnostics CXRS (charge exchange recombination spectroscopy).The authors describe the thermal and mechanical design of the first mirror of the CXRS diagnostics. The components of the first mirror are exposed to high heat and neutron irradiation. The surface temperature will be 300 to 400 deg C. The misalignment tolerance is plus or minus 0.1 degree. The maximum mechanical stresses in the mirror have to be minimized. The design calculations use the finite element code ANSYS. The results indicate that the heat input from the plasma can be removed by the coolant flow. Further calculation shave to concern the brazed joints between mirror and cooling block.

  14. Progress on ITER remote experimentation centre

    International Nuclear Information System (INIS)

    Ozeki, Takahisa; Clement-Lorenzo, Susana; Nakajima, Noriyoshi

    2016-01-01

    Construction of ITER remote experimentation centre (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is progressing. In order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, development of a remote experiment system by using the Satellite Tokamak (JT-60SA) facilities was planned and the development of software for the remote experiment is ongoing, including the systems for the remote connection and the communication between the remote site and the on-site facility. The network system from REC in Rokkasho-site of Japan to the network in EU was established in collaboration with the National Institute of Informatics (NII). Effective data transfer method that is capable of fast transfer speeds in the gigabit range is investigated. Data transfer at the rate of several Gbps was successfully obtained between the institutes in Japan. The preliminary versions of the software for data analysis are developed, such as for visualization of time dependent experimental data and transport simulations, visualization of plasma boundary/equilibrium and spatial profiles of diagnostic data. The remote data access program and an integrated platform for Documentation and Experiment Management are also being developed. A remote experiment room in the Rokkasho-site in Japan was designed and the construction started. The function of REC will be tested and the total system will be demonstrated by the middle of 2017.

  15. Progress on ITER remote experimentation centre

    Energy Technology Data Exchange (ETDEWEB)

    Ozeki, Takahisa, E-mail: ozeki.takahisa@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Obuchi Rokkasho, Kitakami-gun, Aomori 039-3212 (Japan); Clement-Lorenzo, Susana [Fusion for Energy, Torres Diagonal Litoral, B3, 13/03, Barcelona 08019 (Spain); Nakajima, Noriyoshi [National institute for Fusion Science and Project leader of IFERC, 2-166 Obuchi, Rokkasho, Kamikita-gun, Aomori 039-3212 (Japan)

    2016-11-15

    Construction of ITER remote experimentation centre (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is progressing. In order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, development of a remote experiment system by using the Satellite Tokamak (JT-60SA) facilities was planned and the development of software for the remote experiment is ongoing, including the systems for the remote connection and the communication between the remote site and the on-site facility. The network system from REC in Rokkasho-site of Japan to the network in EU was established in collaboration with the National Institute of Informatics (NII). Effective data transfer method that is capable of fast transfer speeds in the gigabit range is investigated. Data transfer at the rate of several Gbps was successfully obtained between the institutes in Japan. The preliminary versions of the software for data analysis are developed, such as for visualization of time dependent experimental data and transport simulations, visualization of plasma boundary/equilibrium and spatial profiles of diagnostic data. The remote data access program and an integrated platform for Documentation and Experiment Management are also being developed. A remote experiment room in the Rokkasho-site in Japan was designed and the construction started. The function of REC will be tested and the total system will be demonstrated by the middle of 2017.

  16. ITER status, design and material objectives

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

    During the ITER Engineering Design Activities (EDA), completed in July 2001, the Joint Central Team and Home Teams developed a robust design of ITER, summarised in this paper, with parameters which fully meet the required scientific and technological objectives, construction costs and safety requirements, with appropriate margins. The design is backed by R and D to qualify the technology, including materials R and D. Materials for ITER components have been selected largely because of their availability and well-established manufacturing technologies, taking account of the low fluence experienced during neutron irradiation, and the experimental nature of the device. Nevertheless, for specific needs relevant to a future fusion reactor, improved materials, in particular for magnet structures, in-vessel components, and joints between the different materials needed for plasma facing components, have been successfully developed. Now, with the technical readiness to decide on ITER construction, negotiations, supported by coordinated technical activities of an international team and teams from participant countries, are underway on joint construction of ITER with a view to the signature/ratification of an agreement in 2003

  17. 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

  18. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  19. Multiphase flow problems on thermofluid safety for fusion reactors

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    2003-01-01

    As the thermofluid safety study for the International Thermonuclear Experimental Reactor (ITER), thermal-hydraulic characteristics of Tokamak fusion reactors under transient events were investigated experimentally and analyzed numerically. As severe transient events an ingress-of-coolant event (ICE) and a loss-of-vacuum event (LOVA) were considered. An integrated ICE test facility was constructed to demonstrate that the ITER safety design approach and parameters are adequate. Water-vapor two-phase flow behavior and performance of the ITER pressure suppression system during the ICE were clarified by the integrated ICE experiments. The TRAC was modified to specify the two-phase flow behavior under the ICE. The ICE experimental results were verified using the modified TRAC code. On the other hand, activated dust mobilization and air ingress characteristics in the ITER vacuum vessel during the LOVA were analyzed using a newly developed analysis code. Some physical models on the motion of dust were considered. The rate of dust released from the vacuum vessel through breaches to the outside was characterized quantitatively. The predicted average pressures in the vacuum vessel during the LOVA were in good agreement with the experimental results. Moreover, direct-contact condensation characteristics between water and vapor inside the ITER suppression tank were observed visually and simulated by the direct two-phase flow analysis. Furthermore, chemical reaction characteristics between vapor and ITER plasma-facing component materials were predicted numerically in order to obtain qualitative estimation on generation of inflammable gases such as hydrogen and methane. The experimental and numerical results of the present studies were reflected in the ITER thermofluid safety design. (author)

  20. Qualification of the US Made Conductors for ITER TF Magnet System

    International Nuclear Information System (INIS)

    Martovetsky, Nicolai N.; Hatfield, Daniel R.; Miller, John R.; Bruzzone, P.; Stepanov, B.; Seber, B.

    2010-01-01

    The US Domestic Agency (USDA) is one of the six suppliers of the Toroidal Field (TF) conductor for the International Thermonuclear Experimental Reactor (ITER). In order to qualify conductors according to ITER requirements we prepared several lengths of the CICC and short samples for testing in the SULTAN facility in CRPP, Switzerland. We also fully characterized the strands that were used in these SULTAN samples. Fabrication experience and test results are presented and discussed.

  1. ITER site selection studies in Spain

    International Nuclear Information System (INIS)

    Medrano, M.; Alejaldre, C.; Doncel, J.; Garcia, A.; Ibarra, A.; Jimenez, J.A.; Sanchez de Mora, M.A.; Alcala, F.; Diez, J.E.; Dominguez, M.; Albisu, F.

    2003-01-01

    The studies carried out to evaluate and select a candidate site for International Thermonuclear Experimental Reactor (ITER) construction in Spain are presented in this paper. The ITER design, completed in July 2001, considered a number of technical requirements that must be fulfilled by the selected site. Several assumptions concerning the ITER site were made in order to carry on the design before final site selection. In the studies undertaken for ITER site selection in Spain, the referred technical requirements and assumptions were applied across the whole of Spain and two areas were identified as being preferential. These areas are on the Mediterranean coast and are situated in the Catalan and Valencian regions. A comparative evaluation based on technical characteristics for the concrete plots, proposed within the preferential areas, has been done. The result of these studies was the selection of a site that was deemed to be the most competitive--Vandellos (Tarragona)--and it was proposed to the European Commission for detailed studies in order to be considered as a possible European site for ITER construction. Another key factor for hosting ITER in Spain, is the licensing process. The present status is summarised in this paper

  2. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

    1997-01-01

    Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

  3. 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

  4. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1992-01-01

    This paper reports on results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ITER is capable of meeting anticipated regulatory dose limits, but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. Much research and development (R ampersand D) and design analysis is needed to establish that ITER meets regulatory requirements. There is a further oportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, three programmatic challenges and three technical challenges must be overcome. The first step is to fund a comprehensive safety and environmental ITER R ampersand D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of three key technical challenges is plasma engineering - burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost

  5. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    1995-01-01

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers ampersand Constructors and Chicago Bridge ampersand Iron (Raytheon/CB ampersand I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB ampersand I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB ampersand I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB ampersand I and documented accordingly

  6. Dose rate evaluation of body phantom behind ITER bio-shield wall using Monte Carlo method

    International Nuclear Information System (INIS)

    Beheshti, A.; Jabbari, I.; Karimian, A.; Abdi, M.

    2012-01-01

    One of the most critical risks to humans in reactors environment is radiation exposure. Around the tokamak hall personnel are exposed to a wide range of particles, including neutrons and photons. International Thermonuclear Experimental Reactor (ITER) is a nuclear fusion research and engineering project, which is the most advanced experimental tokamak nuclear fusion reactor. Dose rates assessment and photon radiation due to the neutron activation of the solid structures in ITER is important from the radiological point of view. Therefore, the dosimetry considered in this case is based on the Deuterium-Tritium (DT) plasma burning with neutrons production rate at 14.1 MeV. The aim of this study is assessment the amount of radiation behind bio-shield wall that a human received during normal operation of ITER by considering neutron activation and delay gammas. To achieve the aim, the ITER system and its components were simulated by Monte Carlo method. Also to increase the accuracy and precision of the absorbed dose assessment a body phantom were considered in the simulation. The results of this research showed that total dose rates level near the outside of bio-shield wall of the tokamak hall is less than ten percent of the annual occupational dose limits during normal operation of ITER and It is possible to learn how long human beings can remain in that environment before the body absorbs dangerous levels of radiation. (authors)

  7. Multi-level iteration optimization for diffusive critical calculation

    International Nuclear Information System (INIS)

    Li Yunzhao; Wu Hongchun; Cao Liangzhi; Zheng Youqi

    2013-01-01

    In nuclear reactor core neutron diffusion calculation, there are usually at least three levels of iterations, namely the fission source iteration, the multi-group scattering source iteration and the within-group iteration. Unnecessary calculations occur if the inner iterations are converged extremely tight. But the convergence of the outer iteration may be affected if the inner ones are converged insufficiently tight. Thus, a common scheme suit for most of the problems was proposed in this work to automatically find the optimized settings. The basic idea is to optimize the relative error tolerance of the inner iteration based on the corresponding convergence rate of the outer iteration. Numerical results of a typical thermal neutron reactor core problem and a fast neutron reactor core problem demonstrate the effectiveness of this algorithm in the variational nodal method code NODAL with the Gauss-Seidel left preconditioned multi-group GMRES algorithm. The multi-level iteration optimization scheme reduces the number of multi-group and within-group iterations respectively by a factor of about 1-2 and 5-21. (authors)

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

    International Nuclear Information System (INIS)

    Rimminen, H.; Kyynäräinen, J.

    2013-01-01

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

  9. The experimental and technological developments reactor

    International Nuclear Information System (INIS)

    Carbonnier, J.L.

    2003-01-01

    THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

  10. Remote maintenance development for ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Shibanuma, Kiyoshi

    1998-01-01

    This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

  11. Remote maintenance development for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shibanuma, Kiyoshi

    1998-04-01

    This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

  12. Disruptions, loads, and dynamic response of ITER

    International Nuclear Information System (INIS)

    Nelson, B.; Riemer, B.; Sayer, R.; Strickler, D.; Barabaschi, P.; Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

    1995-01-01

    Plasma disruptions and the resulting electromagnetic loads are critical to the design of the vacuum vessel and in-vessel components of the International Thermonuclear Experimental Reactor (ITER). This paper describes the status of plasma disruption simulations and related analysis, including the dynamic response of the vacuum vessel and in-vessel components, stresses and deflections in the vacuum vessel, and reaction loads in the support structures

  13. ITER EDA technical activities

    International Nuclear Information System (INIS)

    Aymar, R.

    1998-01-01

    Six years of technical work under the ITER EDA Agreement have resulted in a design which constitutes a complete description of the ITER device and of its auxiliary systems and facilities. The ITER Council commented that the Final Design Report provides the first comprehensive design of a fusion reactor based on well established physics and technology

  14. US ITER limiter module design

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.; Hassanein, A.

    1996-08-01

    The recent U.S. effort on the ITER (International Thermonuclear Experimental Reactor) shield has been focused on the limiter module design. This is a multi-disciplinary effort that covers design layout, fabrication, thermal hydraulics, materials evaluation, thermo- mechanical response, and predicted response during off-normal events. The results of design analyses are presented. Conclusions and recommendations are also presented concerning, the capability of the limiter modules to meet performance goals and to be fabricated within design specifications using existing technology

  15. Device for rearranging control rods of experimental reactors

    International Nuclear Information System (INIS)

    Louda, J.

    1975-01-01

    The invention claims a means for the adjustment of control rods in experimental reactors with a continuously variable pitch of the fuel element spacer. The proposed device permits obtaining maximum variability in the physical modelling of nuclear power reactor cores in experimental reactors. (F.M.)

  16. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R S [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  17. 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

  18. Accuracy analysis of hybrid parallel robot for the assembling of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Wang Yongbo [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland); The State Key Laboratory of Mechanical Transmission, Chongqing University (China); Pessi, Pekka [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland); Wu Huapeng [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland)], E-mail: huapeng@lut.fi; Handroos, Heikki [Institute of Mechatronics and Virtual Engineering, Lappeenranta University of Technology, Skinnarilankatu 34, 53850 Lappeenranta (Finland)

    2009-06-15

    This paper presents a novel mobile parallel robot, which is able to carry welding and machining processes from inside the international thermonuclear experimental reactor (ITER) vacuum vessel (VV). The kinematics design of the robot has been optimized for ITER access. To improve the accuracy of the parallel robot, the errors caused by the stiffness and manufacture process have to be compensated or limited to a minimum value. In this paper kinematics errors and stiffness modeling are given. The simulation results are presented.

  19. Accuracy analysis of hybrid parallel robot for the assembling of ITER

    International Nuclear Information System (INIS)

    Wang Yongbo; Pessi, Pekka; Wu Huapeng; Handroos, Heikki

    2009-01-01

    This paper presents a novel mobile parallel robot, which is able to carry welding and machining processes from inside the international thermonuclear experimental reactor (ITER) vacuum vessel (VV). The kinematics design of the robot has been optimized for ITER access. To improve the accuracy of the parallel robot, the errors caused by the stiffness and manufacture process have to be compensated or limited to a minimum value. In this paper kinematics errors and stiffness modeling are given. The simulation results are presented.

  20. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  1. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  2. Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

    International Nuclear Information System (INIS)

    Onozuka, M.; Takeda, N.; Nakahira, M.; Shimizu, K.; Nakamura, T.

    2003-01-01

    The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods

  3. Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: masanori_onozuka@mhi.co.jp; Takeda, N.; Nakahira, M.; Shimizu, K.; Nakamura, T

    2003-09-01

    The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods.

  4. Two-dimensional over-all neutronics analysis of the ITER device

    Science.gov (United States)

    Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

    1993-07-01

    The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

  5. Two-dimensional over-all neutronics analysis of the ITER device

    International Nuclear Information System (INIS)

    Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi.

    1993-07-01

    The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR) and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li 2 O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No.5 and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design. (author)

  6. 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...

  7. Conceptual design of SC magnet system for ITER, (2)

    International Nuclear Information System (INIS)

    Koizumi, Koichi; Hasegawa, Mitsuru; Yoshida, Kiyoshi

    1991-08-01

    The International Thermonuclear Experimental Reactor (ITER) is an experimental tokamak machine testing the basic plasma performance and technologies required for future tokamak reactor. The design proposals for the Superconducting (SC) Magnet System from Japan were summarized by the Fusion Experimental Reactor (FER) Design Team and the Superconducting Magnet Laboratory of the Japan Atomic Energy Research Institute (JAERI). This report is one of the series reports on 'Conceptual design of superconducting magnet system for ITER', and describes the major results of the stress analysis regarding the Toroidal Field (TF) coil, the Center Solenoid (CS) coil and the Equilibrium Field (EF) coil and their support structures. Among the design issues, the mechanical design of the coil system was one of the most critical items, not only because of the huge electromagnetic loads due to large size and high magnetic field, but also because of the demand of high reliability under neutron irradiation. In order to satisfy both the coil performance and the mechanical reliability, different types of conductors were employed for each coils. The mechanical behaviors and the safety margin of each coil were analyzed by using finite element method (FEM) of MSC/NASTRAN. The procedure to obtain the equivalent winding stiffness employed for the each FEM analysis is also described in this report. The details on the coil specifications, conductor design and mechanical design for each coils are described in other report of the series reports. (J.P.N.)

  8. ITER-FEAT safety

    International Nuclear Information System (INIS)

    Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

    2001-01-01

    Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

  9. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  10. The ITER CODAC conceptual design

    International Nuclear Information System (INIS)

    Lister, J.B.; Farthing, J.W.; Greenwald, M.; Yonekawa, I.

    2007-01-01

    CODAC orchestrates the activity of 60-90 Plant Systems in normal ITER operation. Interlock Systems protect ITER from potentially damaging operating off-normal conditions. Safety Systems protect the personnel and the environment and will be subject to licensing. The principal challenges to be met in the design and implementation of CODAC include: complexity, reliability, transparent access respecting security, a high experiment data rate and data volume since ITER is an experimental reactor, scientific exploitation from multiple Participant Team Experiment Sites and the long 35-year period for construction and operation. Complexity is addressed by prescribing the communication interfaces to the Plant Systems and prescribing the technical implementation within the Plant Systems. Plant Systems export to CODAC all the information on their construction and operation as 'self-description'. Complexity is also addressed by automating the operation of ITER and of the plasma, using a structured data description of 'Operation Schedules' which encompass all non-manual control, including Plasma Control. Reliability is addressed by maximising code reuse and maximising the use of existing products thereby minimising in-house development. The design is hierarchical and modular in both hardware and software. The latter facilitates evolution of methods during the project lifetime. Guaranteeing security while maximising access is addressed by flow separation. Out-flowing data, including experimental signals and the status of ITER plant is risk-free. In-flowing commands and data originate from Experiment Sites. The Cadarache Experiment Site is equated with the Remote Experiment Sites and a rigorous 'Operation Request Gatekeeper' is provided. The high data rates and data volumes are handled with high performance networks. Global Area Networks allow Participant Teams to access all CODAC data and applications. Scientific exploitation of ITER will remain a human as well as technical

  11. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  12. Environmental dose rate assessment of ITER using the Monte Carlo method

    Directory of Open Access Journals (Sweden)

    Karimian Alireza

    2014-01-01

    Full Text Available Exposure to radiation is one of the main sources of risk to staff employed in reactor facilities. The staff of a tokamak is exposed to a wide range of neutrons and photons around the tokamak hall. The International Thermonuclear Experimental Reactor (ITER is a nuclear fusion engineering project and the most advanced experimental tokamak in the world. From the radiobiological point of view, ITER dose rates assessment is particularly important. The aim of this study is the assessment of the amount of radiation in ITER during its normal operation in a radial direction from the plasma chamber to the tokamak hall. To achieve this goal, the ITER system and its components were simulated by the Monte Carlo method using the MCNPX 2.6.0 code. Furthermore, the equivalent dose rates of some radiosensitive organs of the human body were calculated by using the medical internal radiation dose phantom. Our study is based on the deuterium-tritium plasma burning by 14.1 MeV neutron production and also photon radiation due to neutron activation. As our results show, the total equivalent dose rate on the outside of the bioshield wall of the tokamak hall is about 1 mSv per year, which is less than the annual occupational dose rate limit during the normal operation of ITER. Also, equivalent dose rates of radiosensitive organs have shown that the maximum dose rate belongs to the kidney. The data may help calculate how long the staff can stay in such an environment, before the equivalent dose rates reach the whole-body dose limits.

  13. Implementation of technical stage of ITER project in Kazakstan

    International Nuclear Information System (INIS)

    Shkol'nik, V.S.; Shemanskij, V.A.; Cherepnin, Yu.S.; Skoz, E.Ya.; Nazarenko, L.I.; Shestakov, V.P.; Tazhibaeva, I.L.

    1996-01-01

    The main executors of ITER project in Kazakstan are: National nuclear Centre; National Joint-stock Co. of the enterprises on atomic engineering and industry (KATEP) and Science and Research Inst. of Experimental and Theoretical Physics of the Kazakh National Univ. Principle directions of Kazakhstan Republic works in 1996/97 under the ITER project is possible to formulate as follows: - irradiation test of materials on the BN-350 reactor; -continuation of in-pile experiments on determination of deuterium interaction parameters with beryllium samples, manufactured by Ulba Metallurgical Plant (KATEP), using Research Water Heterogeneous Reactor; - conducting of in-pile experiments on determination of hydrogen isotope permeation parameters through vanadium alloys using Research Water Heterogeneous Reactor. Conducting of out-pile experiments with beryllium, and vanadium alloys; - investigation of beryllium structure and properties in respect to hydrogen in condition, simulating gas impurities inventory, under thermal and neutron influence; -simulation of plasma disruption on divertor; -simulation of water steam interaction with beryllium; - determination of hydrogen permeation parameters for various materials with 'in situ' surface by electronic spectroscopy; -development of manufacturing technology of beryllium connections with copper alloys

  14. ITER EDA newsletter. V. 7, no. 1

    International Nuclear Information System (INIS)

    1998-01-01

    This issue of the ITER Newsletter contains a summary report on the Thirteenth meeting of the ITER Management Advisory Committee (MAC), a report on ITER at the International Conference on Fusion Reactor Materials and a report of a Russian scientist working at ITER Garching JWS

  15. Nuclear aspects of diagnostics in RTO/RC ITER

    International Nuclear Information System (INIS)

    Walker, C.I.; Yamamoto, S.; Costley, A.; Kock, L. de; Ebisawa, K.; Janeschitz, G.; Khripunov, V.; Martin, E.; Vayakis, G.

    2000-01-01

    ITER (international thermonuclear experimental reactor) will be the first fusion device where the design of the plasma diagnostic systems will make extensive use of the materials and techniques developed in the nuclear technology field. The designs have to satisfy stringent requirements for tritium confinement, nuclear shielding and vacuum integrity. This paper introduces the requirements for diagnostics in the ITER long pulse, burning plasma environment, and addresses the impact of the reactor environment on the diagnostics and ancillary equipment. These systems necessarily require access to the plasma or first wall, which generally conflicts with the requirements of the basic machine. Holes are required through the first wall, primary shielding, containment boundaries and biological shielding. Components have a limited life and require maintenance. This paper describes the effect of the radiation environment on diagnostic design at different locations. Ex-vessel and in-vessel remote handling, hot cell refurbishment and tritium confinement are also described

  16. Japanese contributions to ITER testing program of solid breeder blankets for DEMO

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

    1991-04-01

    ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

  17. Remote handling maintenance of ITER

    International Nuclear Information System (INIS)

    Haange, R.

    1999-01-01

    The remote maintenance strategy and the associated component design of the International Thermonuclear Experimental Reactor (ITER) have reached a high degree of completeness, especially with respect to those components that are expected to require frequent or occasional remote maintenance. Large-scale test stands, to demonstrate the principle feasibility of the remote maintenance procedures and to develop the required equipment and tools, were operational at the end of the Engineering Design Activities (EDA) phase. The initial results are highly encouraging: major remote equipment deployment and component replacement operations have been successfully demonstrated. (author)

  18. ITER conceptual design

    International Nuclear Information System (INIS)

    Tomabechi, K.; Gilleland, J.R.; Sokolov, Yu.A.; Toschi, R.

    1991-01-01

    The Conceptual Design Activities of the International Thermonuclear Experimental Reactor (ITER) were carried out jointly by the European Community, Japan, the Soviet Union and the United States of America, under the auspices of the International Atomic Energy Agency. The European Community provided the site for joint work sessions at the Max-Planck-Institut fuer Plasmaphysik in Garching, Germany. The Conceptual Design Activities began in the spring of 1988 and ended in December 1990. The objectives of the activities were to develop the design of ITER, to perform a safety and environmental analysis, to define the site requirements as well as the future research and development needs, to estimate the cost and manpower, and to prepare a schedule for detailed engineering design, construction and operation. On the basis of the investigation and analysis performed, a concept of ITER was developed which incorporated maximum flexibility of the performance of the device and allowed a variety of operating scenarios to be adopted. The heart of the machine is a tokamak having a plasma major radius of 6 m, a plasma minor radius of 2.15 m, a nominal plasma current of 22 MA and a nominal fusion power of 1 GW. The conceptual design can meet the technical objectives of the ITER programme. Because of the success of the Conceptual Design Activities, the Parties are now considering the implementation of the next phase, called the Engineering Design Activities. (author). Refs, figs and tabs

  19. Fuel cycle design evolution from FDR-ITER to RTO/RC-ITER

    International Nuclear Information System (INIS)

    Murdoch, D.K.; Glugla, M.; Kveton, O.

    2000-01-01

    Instantaneous fuelling and plasma exhaust flow rates for the reduced technical objective/reduced cost version of International Thermonuclear Experimental Reactor (RTO/RC-ITER) are similar to those described in the Final Design Report (FDR) of ITER, despite the reduction in fusion power by a factor of about two. However, the reduced pulse length and the lower fraction of campaign time spent in burn mode, together with the lower integrated operating lifetime proposed, will generate cost savings in several systems of the fuel cycle. As the quantity of tritium handled per pulse is now smaller, this could be buffered, allowing systems in the tritium plant still to operate in steady state mode as in the FDR design, thereby increasing the potential for downsizing of system capacities. The lower operating time fraction will increase performance margins for some systems, for example, the Torus Exhaust Gas Processing System (TEGPS) which was designed to meet a specified daily release rate for the FDR design conditions which were more onerous than RTO/RC-ITER. As no break through of tritium into cooling water is now expected, the duties of the Water and Atmosphere Detritiation Systems are considerably reduced, and design concepts which are simpler, cheaper and more amenable to modular implementation can be adopted

  20. MELCOR 1.8.2 Analyses in Support of ITER's RPrS

    International Nuclear Information System (INIS)

    Brad J Merrill

    2008-01-01

    The International Thermonuclear Experimental Reactor (ITER) Program is performing accident analyses for ITER's 'Rapport Preliminaire de Surete' (Report Preliminary on Safety - RPrS) with a modified version of the MELCOR 1.8.2 code. The RPrS is an ITER safety document required in the ITER licensing process to obtain a 'Decret Autorisation de Construction' (a Decree Authorizing Construction - DAC) for the ITER device. This report documents the accident analyses performed by the US with the MELCOR 1.8.2 code in support of the ITER RPrS effort. This work was funded through an ITER Task Agreement for MELCOR Quality Assurance and Safety Analyses. Under this agreement, the US was tasked with performing analyses for three accident scenarios in the ITER facility. Contained within the text of this report are discussions that identify the cause of these accidents, descriptions of how these accidents are likely to proceed, the method used to analyze the consequences of these accidents, and discussions of the transient thermal hydraulic and radiological release results for these accidents

  1. Iterative solutions of finite difference diffusion equations

    International Nuclear Information System (INIS)

    Menon, S.V.G.; Khandekar, D.C.; Trasi, M.S.

    1981-01-01

    The heterogeneous arrangement of materials and the three-dimensional character of the reactor physics problems encountered in the design and operation of nuclear reactors makes it necessary to use numerical methods for solution of the neutron diffusion equations which are based on the linear Boltzmann equation. The commonly used numerical method for this purpose is the finite difference method. It converts the diffusion equations to a system of algebraic equations. In practice, the size of this resulting algebraic system is so large that the iterative methods have to be used. Most frequently used iterative methods are discussed. They include : (1) basic iterative methods for one-group problems, (2) iterative methods for eigenvalue problems, and (3) iterative methods which use variable acceleration parameters. Application of Chebyshev theorem to iterative methods is discussed. The extension of the above iterative methods to multigroup neutron diffusion equations is also considered. These methods are applicable to elliptic boundary value problems in reactor design studies in particular, and to elliptic partial differential equations in general. Solution of sample problems is included to illustrate their applications. The subject matter is presented in as simple a manner as possible. However, a working knowledge of matrix theory is presupposed. (M.G.B.)

  2. Status and plans for US ITER studies

    International Nuclear Information System (INIS)

    Doggett, J.N.

    1992-01-01

    The United States' participation in the International Thermonuclear Experimental Reactor (ITER) began in later 1987 when the initiative to start a cooperative program among the four Parties -- the Soviet Union, Japan, the European Community, and the United States -- was initiated. Participation then continued through the start of Joint Work in May 1988 until the conclusion of the Conceptual Design Activities (CDA) in December 1990. In the period between the conclusion of the CDA and the agreement to execute the Engineering Design Activities (EDA), the US ITER Home Team continued to do work on the design, executed additional research and development, and participated in the preparations for the EDA. Activities included one major design study on a High-Aspect-Ratio Design and input to the National ITER Technical Review, the ITER Steering Committee -- US, Special Working Group 1, and the Fusion Energy Advisory Committee's Panel 1. Research and development was continued in areas of work that were identified as critical-path elements by an international panel chartered by the four ITER Parties near the end of the CDA. I will describe the conclusion of the CDA and the interim US ITER activities and will give an indication of our involvement in the EDA

  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. Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

    International Nuclear Information System (INIS)

    Jammes, Ch.

    2010-07-01

    After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

  5. A mobile robot with parallel kinematics to meet the requirements for assembling and machining the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Pessi, Pekka [Lappeenranta University of Technology, Lappeenranta (Finland)], E-mail: pessi@lut.fi; Wu, Huapeng; Handroos, Heikki [Lappeenranta University of Technology, Lappeenranta (Finland); Jones, Lawrence [EFDA Close Support Unit, Boltzmannstrasse 2, Garching D-85748 (Germany)

    2007-10-15

    The present paper introduces a mobile parallel robot developed for International Thermonuclear Experimental Reactor (ITER). The task of the robot is to carry out welding and machining processes inside the ITER vacuum vessel. The kinematic design of the robot has been optimized for the ITER access. The kinematic analysis is given in the paper. A virtual prototype of the parallel robot is built. A dynamic behavior of the whole robot is studied by the multi-body system simulation (MBS)

  6. A mobile robot with parallel kinematics to meet the requirements for assembling and machining the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Pessi, Pekka; Wu, Huapeng; Handroos, Heikki; Jones, Lawrence

    2007-01-01

    The present paper introduces a mobile parallel robot developed for International Thermonuclear Experimental Reactor (ITER). The task of the robot is to carry out welding and machining processes inside the ITER vacuum vessel. The kinematic design of the robot has been optimized for the ITER access. The kinematic analysis is given in the paper. A virtual prototype of the parallel robot is built. A dynamic behavior of the whole robot is studied by the multi-body system simulation (MBS)

  7. The experimental nuclear reactor: AQUILON

    International Nuclear Information System (INIS)

    Girard, Y.; Koechlin, J.C.; Moreau, J.M.

    1958-01-01

    'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [fr

  8. ITER Neutral Beam Injection System

    International Nuclear Information System (INIS)

    Ohara, Yoshihiro; Tanaka, Shigeru; Akiba, Masato

    1991-03-01

    A Japanese design proposal of the ITER Neutral Beam Injection System (NBS) which is consistent with the ITER common design requirements is described. The injection system is required to deliver a neutral deuterium beam of 75MW at 1.3MeV to the reactor plasma and utilized not only for plasma heating but also for current drive and current profile control. The injection system is composed of 9 modules, each of which is designed so as to inject a 1.3MeV, 10MW neutral beam. The most important point in the design is that the injection system is based on the utilization of a cesium-seeded volume negative ion source which can produce an intense negative ion beam with high current density at a low source operating pressure. The design value of the source is based on the experimental values achieved at JAERI. The utilization of the cesium-seeded volume source is essential to the design of an efficient and compact neutral beam injection system which satisfies the ITER common design requirements. The critical components to realize this design are the 1.3MeV, 17A electrostatic accelerator and the high voltage DC acceleration power supply, whose performances must be demonstrated prior to the construction of ITER NBI system. (author)

  9. Status of international collaborative efforts on selected ITER materials

    International Nuclear Information System (INIS)

    Belyakov, V.A.; Fabritsiev, S.A.; Mazul, I.V.; Rowcliffe, A.F.

    2000-01-01

    The paper presents an overview of the performance of refractory metals, beryllium, and copper alloys, for the international thermonuclear experimental reactor (ITER) high heat flux structures. High temperature brazing, hot isostatic pressing (HIP), friction welding, explosive bonding, and other methods were explored to join copper alloys to 316 stainless steel for first wall and limiter applications. It is concluded that the main material problems for the ITER high heat flux components are: (a) degradation of properties after the manufacturing cycle (especially for Be/Cu and Cu/stainless steel (SS) joints); (b) helium embrittlement of Be, and Cu, and; (c) radiation-induced loss of fracture toughness for Be, W, and Cu alloys

  10. Orphee reactor experimental equipment

    International Nuclear Information System (INIS)

    1987-01-01

    Experimental equipment around the ORPHEE reactor is presented. The neutron source; and the spectrometers and sample environment (inelastic and quasi-elastic scattering, elastic scattering, spread scattering, small angle scattering) are described. An experiment proposal and reports guide is supplied [fr

  11. International Thermonuclear Experimental Reactor (ITER) divertor plate performance and lifetime considerations

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1990-03-01

    The ITER divertor plate performance during the technology phase of operation has been analyzed. High-Z materials, such as tungsten and tantalum, have been considered as plasma side materials, and refractory metal alloys, Ta-10W, TZM, Nb-1Zr, and V-15Cr-5Ti, plus copper alloys have been considered as the structural materials. The fatigue lifetime have been predicted for structural plates and for duplex plates with the plasma side material bonded to the structure. The results indicate that refractory alloys have a comparable or improved performance to copper alloys. Peak allowable heat fluxes for these analyses are in the range of 15--20 MW/m 2 for 2 mm thick structural plates and 7--11 MW/m 2 for 4 mm thick duplex plates. 4 refs., 55 figs., 6 tabs

  12. Quantification and disposal of radioactive waste from ITER operation

    International Nuclear Information System (INIS)

    Olsson, G.; Devell, L.; Johnsson, B.; Gulden, W.

    1991-01-01

    The work on the safety and environment for the Next European Torus (NET) is being performed within the European Fusion Technology Safety and Environment Programme by the NET team and under NET contracts. In the area of NET-oriented investigations concerning waste management and disposal, Studsvik is concentrating on the operational waste from both NET and ITER (International Thermonuclear Experimental Reactor). This paper gives a characterization and quantification of the radioactive waste generated from the operation of ITER during the Physics Phase, and from the replacement of all blanket segments (European shielding blanket option) at the end of the Physics Phase after an integrated first-wall loading of 0.03 MWy/m 2 . The total activity contents and volumes of packaged waste from the Physics Phase operation and from the blanket replacement are estimated. The waste volume from replacement of the shielding blanket segments of ITER is considerably larger than estimated in earlier calculations for NET due to the fact that the ITER conceptual design includes more of the stell shielding in the removable segments. The waste handling and disposal are described using existing Swedish and German concepts for similar waste categories from nuclear fission reactors. This includes the choice of suitable packagings, intermediate storage time for cooling, and type of repository for final disposal. Some typical cost figures for waste handling are also presented. (orig.)

  13. The European fusion research and development programme and the ITER Project

    International Nuclear Information System (INIS)

    Green, B.J.

    2004-01-01

    The EURATOM fusion R and D programme is a well integrated and co-ordinated programme a good example of a European Research Area. Its goal is 'the joint creation of prototype reactors for power stations to meet the needs of society: operational safety, environmental compatibility, economic viability'. The programme is focussed on the magnetic confinement approach to fusion energy and supports 21 associated laboratories and a range of experimental and fusion technology facilities. The paper will briefly describe this programme and how it is organised and implemented. Its success and that of other national programmes has defined the international ITER Project, which is the next logical step in fusion R and D. The paper will describe ITER, its aims, its design, and the supporting manufacture of prototype components. The European contribution to ITER, as well as the exploitation of the Joint European Torus (JET) and long-term fusion reactor technology R and D are carried out under the European Fusion Development Agreement (EFDA). Finally, the potential advantages of fusion as an energy source will be presented. (author)

  14. Introduction of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Kawahara, Hirotaka; Aoyama, Takafumi

    2006-01-01

    The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO. (author)

  15. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  16. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  17. Structural analyses of ITER toroidal field coils under fault conditions

    International Nuclear Information System (INIS)

    Jong, C.T.J.

    1992-04-01

    ITER (International Thermonuclear Experimental Reactor) is intended to be an experimental thermonuclear tokamak reactor testing the basic physics performance and technologies essential to future fusion reactors. The magnet system of ITER consists essentially of 4 sub-systems, i.e. toroidal field coils (TFCs), poloidal field coils (PFCs), power supplies, and cryogenic supplies. These subsystems do not contain significant radioactivity inventories, but the large energy inventory is a potential accident initiator. The aim of the structural analyses is to prevent accidents from propagating into vacuum vessel, tritium system and cooling system, which all contain significant amounts of radioactivity. As part of design process 3 conditions are defined for PF and TF coils, at which mechanical behaviour has to be analyzed in some detail, viz: normal operating conditions, upset conditions and fault conditions. This paper describes the work carried out by ECN to create a detailed finite element model of 16 TFCs as well as results of some fault condition analyses made with the model. Due to fault conditions, either electrical or mechanical, magnetic loading of TFCs becomes abnormal and further mechanical failure of parts of the overall structure might occur (e.g. failure of coil, gravitational supports, intercoil structure). The analyses performed consist of linear elastic stress analyses and electro-magneto-structural analyses (coupled field analyses). 8 refs.; 5 figs.; 5 tabs

  18. US--ITER activation analysis

    International Nuclear Information System (INIS)

    Attaya, H.; Gohar, Y.; Smith, D.

    1990-09-01

    Activation analysis has been made for the US ITER design. The radioactivity and the decay heat have been calculated, during operation and after shutdown for the two ITER phases, the Physics Phase and the Technology Phase. The Physics Phase operates about 24 full power days (FPDs) at fusion power level of 1100 MW and the Technology Phase has 860 MW fusion power and operates for about 1360 FPDs. The point-wise gamma sources have been calculated everywhere in the reactor at several times after shutdown of the two phases and are then used to calculate the biological dose everywhere in the reactor. Activation calculations have been made also for ITER divertor. The results are presented for different continuous operation times and for only one pulse. The effect of the pulsed operation on the radioactivity is analyzed. 6 refs., 12 figs., 1 tab

  19. ITER ITA newsletter. No. 11, December 2003

    International Nuclear Information System (INIS)

    2003-12-01

    This issue of the ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER including information from the editor about ITER update, about progress in ITER magnet design and preparation of procurement packages and about 25th anniversary of the First Steering Committee Meeting of the International Tokamak Reactor (INTOR) Workshop, organized under the auspices of the IAEA, took place at the IAEA Headquarters in Vienna

  20. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

  1. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1991-01-01

    Results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ''ITER is capable of meeting anticipated regulatory dose limits,'' but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. We need much research and development (R ampersand D) and design analysis to establish that ITER meets regulatory requirements. We have a further opportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, we need to overcome three programmatic challenges and three technical challenges. The first programmatic challenge is to fund a comprehensive safety and environmental ITER R ampersand D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of the three key technical challenges is plasma engineering -- burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost. Some design requirements are now too strict; some are too lax. Fuel cycle design requirements are presently too strict, mandating inappropriate T separation from H and D. Heat sink requirements are presently too lax; they should be strengthened to ensure that maximum loss of coolant accident temperatures drop

  2. Status of experimental data related to Be in ITER materials R and D data bank

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Shigeru [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    To keep traceability of many valuable raw data that were experimentally obtained in the ITER Technology R and D Tasks related to materials for In-Vessel components (divertor, first wall, blanket, vacuum vessel, etc.) and to easily make the best use of these data in the ITER design activities, the `ITER Materials R and D Data Bank` has been built up, with the use of Excel{sup TM} spread sheets. The paper describes status of experimental data collected in this data bank on thermo-mechanical properties of unirradiated and neutron irradiated Be, on plasma-material interactions of Be, on mechanical properties of various kinds of Be/Cu joints (including plasma sprayed Be), and on thermal fatigue tests of Be/Cu mock-ups. (author)

  3. Final ITER CTA project board meeting

    International Nuclear Information System (INIS)

    Vlasenkov, V.

    2003-01-01

    The final ITER CTA Project Board Meeting (PB) took place in Barcelona, Spain on 8 December 2002. The PB took notes of the comments concerning the status of the International Team and the Participants Teams, including Dr. Aymar's report 'From ITER to a FUSION Power Reactor' and the assessment of the ITER project cost estimate

  4. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  5. Opportunities for reactor scale experimental physics

    International Nuclear Information System (INIS)

    1999-01-01

    A reactor scale tokamak plasma will exhibit three areas of physics phenomenology not accessible by contemporary experimental facilities. These are: (1) instabilities generated by energetic alpha particles; (2) self-heating phenomena; and (3) reactor scale physics, which includes integration of diverse physics phenomena, each with its own scaling properties. In each area, selected examples are presented that demonstrate the importance and uniqueness of physics results from reactor scale facilities for both inductive and steady state reactor options. It is concluded that the physics learned in such investigations will be original physics not attainable with contemporary facilities. In principle, a reactor scale facility could have a good measure of flexibility to optimize the tokamak approach to magnetic fusion energy. (author)

  6. The legacy of Artismovich and the lessons of ITER

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    This talk was made during 17th IAEA fusion energy conference at the Artsimovich Memorial Session and Dr. Artsimovich's contributions to the development of plasma physics, tokamak devices were highly acknowledged. His main elaborations are widely used in the design of ITER, the International Tokamak Experimental Reactor, whose successful implementation will lead most efficiently towards the realisation of the promise of fusion energy for the benefit of all mankind

  7. Current status of the European contribution to the Remote Data Access System of the ITER Remote Experimentation Centre

    International Nuclear Information System (INIS)

    De Tommasi, G.; Manduchi, G.; Muir, D.G.; Ide, S.; Naito, O.; Urano, H.; Clement-Lorenzo, S.; Nakajima, N.; Ozeki, T.; Sartori, F.

    2015-01-01

    The ITER Remote Experimentation Centre (REC) is one of the projects under implementation within the BA agreement. The final objective of the REC is to allow researchers to take part in the experimentation on ITER from a remote location. Before ITER first operations, the REC will be used to evaluate ITER-relevant technologies for remote participation. Among the different software tools needed for remote participation, an important one is the Remote Data Access System (RDA), which provides a single software infrastructure to access data stored at the remotely participating experiment, regardless of the geographical location of the users. This paper introduces the European contribution to the RDA system for the REC.

  8. The International Thermonuclear Experimental Reactor (ITER) international organisation: which laws apply to this international nuclear operator?

    International Nuclear Information System (INIS)

    Grammatico-Vidal, L.

    2009-01-01

    ITER is being carried out by way of international collaboration between seven partners (the European atomic energy community -EURATOM-, China, India, Japan, Russia, south korea and the United states) which together represent more than half the world population. From a project organisation point of view, it is supported by both financial and in-kind contributions provided by each of the partner; each member makes its contribution through a special legal entity called a 'domestic agency' to an international organisation which was set up by the Agreement on the Establishment of an International Fusion Energy Organization for the joint Implementation of the ITER project signed in Paris on 21. november 2006 and which entered into force on 24. october 2007 after ratification by each of the partners. The international agreement is to remain in effect for a period of 35 years and may be renewed for a period of 10 years without any change to its content. It is supplemented by an agreement of the same date on the privileges and immunities of the organisation and of its staff. The function of the ITER organisation is to construct, commission, operate and permanently shutdown the ITER facility, to encourage their exploitation by laboratories, other institutions and personnel participating in the fusion energy research and development programmes of its members and to promote public understanding and acceptance of fusion energy. The unique institutional structure for this project will be described briefly in the introduction before analysing the law applicable to this international organisation, a French nuclear operator, unique in France today. (N.C.)

  9. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

  10. Design considerations for ITER 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 (U.S.S.R.), and the United States. This paper describes a magnetic and mechanical design methodology for toroidal field (TF) coils that employs Nb 3 Sn superconductor technology. Coil winding is sized by using conductor concepts developed for the U.S. TIBER concept. Manifold concepts are presented for the complete 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

  11. Design of the ITER Tokamak Assembly Tools

    International Nuclear Information System (INIS)

    Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

    2006-01-01

    ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

  12. Potential for Australian involvement in ITER

    International Nuclear Information System (INIS)

    O'Connor, D. J.; Collins, G. A.; Hole, M. J.

    2006-01-01

    Full text: Full text: Fusion, the process that powers the sun and stars, offers a solution to the world's long-term energy needs: providing large scale energy production with zero greenhouse gas emissions, short-lived radio-active waste compared to conventional nuclear fission cycles, and a virtually limitless supply of fuel. Almost three decades of fusion research has produced spectacular progress. Present-day experiments have a power gain ratio of approximately 1 (ratio of power out to power in), with a power output in the 10's of megawatts. The world's next major fusion experiment, the International Thermonuclear Experimental Reactor (ITER), will be a pre-prototype power plant. Since announcement of the ITER site in June 2005, the ITER project, has gained momentum and political support. Despite Australia's foundation role in the field of fusion science, through the pioneering work of Sir Mark Oliphant, and significant contributions to the international fusion program over the succeeding years, Australia is not involved in the ITER project. In this talk, the activities of a recently formed consortium of scientists and engineers, the Australian ITER Forum will be outlined. The Forum is drawn from five Universities, ANSTO (the Australian Nuclear Science and Technology Organisation) and AINSE (the Australian Institute for Nuclear Science and Engineering), and seeks to promote fusion energy in the Australian community and negotiate a role for Australia in the ITER project. As part of this activity, the Australian government recently funded a workshop that discussed the ways and means of engaging Australia in ITER. The workshop brought the research, industrial, government and general public communities, together with the ITER partners, and forged an opportunity for ITER engagement; with scientific, industrial, and energy security rewards for Australia. We will report on the emerging scope for Australian involvement

  13. Design standard issues for ITER in-vessel components

    International Nuclear Information System (INIS)

    Majumdar, S.

    1994-01-01

    Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

  14. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  15. Carbon fiber composites application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Akiba, M.; Nakamura, K.; Bonal, J.P.; Pacher, H.D.; Roedig, M.; Vieider, G.; Wu, C.H.

    1998-01-01

    Carbon fiber composites (CFCs) are one of the candidate armour materials for the plasma facing components of the international thermonuclear experimental reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R and D needs are critically discussed. (orig.)

  16. Benchmarking the cad-based attila discrete ordinates code with experimental data of fusion experiments and to the results of MCNP code in simulating ITER

    International Nuclear Information System (INIS)

    Youssef, M. Z.

    2007-01-01

    Attila is a newly developed finite element code based on Sn neutron, gamma, and charged particle transport in 3-D geometry in which unstructured tetrahedral meshes are generated to describe complex geometry that is based on CAD input (Solid Works, Pro/Engineer, etc). In the present work we benchmark its calculation accuracy by comparing its prediction to the measured data inside two experimental mock-ups bombarded with 14 MeV neutrons. The results are also compared to those based on MCNP calculations. The experimental mock-ups simulate parts of the International Thermonuclear Experimental Reactor (ITER) in-vessel components, namely: (1) the Tungsten mockup configuration (54.3 cm x 46.8 cm x 45 cm), and (2) the ITER shielding blanket followed by the SCM region (simulated by alternating layers of SS316 and copper). In the latter configuration, a high aspect ratio rectangular streaming channel was introduced (to simulate steaming paths between ITER blanket modules) which ends with a rectangular cavity. The experiments on these two fusion-oriented integral experiments were performed at the Fusion Neutron Generator (FNG) facility, Frascati, Italy. In addition, the nuclear performance of the ITER MCNP 'Benchmark' CAD model has been performed with Attila to compare its results to those obtained with CAD-based MCNP approach developed by several ITER participants. The objective of this paper is to compare results based on two distinctive 3-D calculation tools using the same nuclear data, FENDL2.1, and the same response functions of several reaction rates measured in ITER mock-ups and to enhance confidence from the international neutronics community in the Attila code and how it can precisely quantify the nuclear field in large and complex systems, such as ITER. Attila has the advantage of providing a full flux mapping visualization everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. In addition, the

  17. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  18. Development of liquid type TBM technology for ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, S. K.; Yoon, J. S.

    2012-03-01

    The final objectives of this project are as follows; Development of the key techniques for the liquid type TBM for ITER: Developing plan for leading and participating liquid TBM concepts; Estimating cost and schedule according to development schedule and managing technologies; Developing integrated design system and completing the engineering design for liquid TBM; Developing the key technologies for the liquid TBM; Construction of performance test systems for liquid TBM and verification of the performance. We are technically surveying the ITER system design data, the insufficient part of ITER design, and required R and D items and so on. In Korea, HCML TBM, liquid type breeder with lithium or lead lithium, has been studied during the past years to develop a tritium breeding technology for tritium self-sufficiency of nuclear fusion reactor and the TBM was proposed to be tested in ITER. In this study, we can obtain the key technology of nuclear fusion reactor especially on the TBM design, analysis and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea

  19. Remote maintenance development for ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Shibanuma, Kiyoshi

    1997-01-01

    This paper both describes the overall design concept of the ITER remote maintenance system, which has been developed mainly for use with in-vessel components such as divertor and blanket, and outlines of the ITER R and D program, which has been established to develop remote handling equipment/tools and radiation hard components. In ITER, the reactor structures inside cryostat have to be maintained remotely because of activation due to DT operation. Therefore, remote-handling technology is fundamental, and the reactor-structure design must be made consistent with remote maintainability. The overall maintenance scenario and design concepts of the required remote handling equipment/tools have been developed according to their maintenance classification. Technologies are also being developed to verify the feasibility of the maintenance design and include fabrication and testing of a fullscale remote-handling equipment/tools for in-vessel maintenance. (author)

  20. ITER EDA newsletter. V. 2, no. 9

    International Nuclear Information System (INIS)

    1993-09-01

    This ITER EDA (Engineering Design Activities) Newsletter issue contains a report on the third meeting of the ITER Technical Advisory Committee, a summary report for the ITER Magnetic Technical Meeting, a brief account of the International Workshop on Nuclear Data for Fusion Reactor Technology, and a description of approved arrangements for visiting home team personnel

  1. Current Status on the Korean Test Blanket Module Development for testing in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Jung, Ki Sok

    2010-01-01

    Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is used as the structural material and helium (He) is used as a coolant to cool the first wall (FW) and breeding zone. Liquid lithium (Li) is circulated for a tritium breeding, not for a cooling purpose. Main purpose for developing the TBM is to develop the design technology for DEMO and fusion reactor and it should be proved through the experiment in the ITER with TBM. Therefore, we have developed the design scheme and related codes including the safety analysis for obtain the license to be tested in the ITER. In order to develop and install at the ITER, several technologies were developed in parallel; fabrication, breeder, He cooling, tritium extraction and so on. Figure 1 shows the overall TBM development scheme. In Korea, official strategy for developing the TBM is to participate to other parties' concept such as US and EU ones, in which PbLi (lead lithium eutectic), He, and FM steel were used for liquid breeder, coolant, and structural material, respectively

  2. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  3. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-02-01

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

  4. Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

    After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

  5. In-vessel tritium retention and removal in ITER

    International Nuclear Information System (INIS)

    Federici, G.; Anderl, R.A.

    1998-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  6. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  7. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

    1983-03-01

    Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

  8. 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.)

  9. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

  10. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

  11. Studies related to plasma-wall interactions in ITER - Final scientific report 2009

    International Nuclear Information System (INIS)

    Marot, L.

    2009-09-01

    In this final scientific report made by the University of Basel, Switzerland, on-going work on plasma-wall interactions in the International Thermonuclear Experimental Reactor ITER is reported on. The growing interest concerning the use of rhodium (Rh) as a material for the first mirrors in ITER and the necessity of using it as a thin film deposited on a polished substrate has necessitated the development of a robust deposition technique for the preparation of high-reflectivity mirrors. The realisation and tests of high-quality rhodium coated mirrors using magnetron sputtering is reported on. Also, the exposure of the rhodium and molybdenum coated mirrors in the Tokamak fusion reactor system is reported on and the role of carbon and tungsten impurities in the optical degradation of metallic mirrors is looked at. Optical measurements made at the Joint European Torus (JET) are also reported on

  12. Catalytic membrane reactors for tritium recovery from tritiated water in the ITER fuel cycle

    International Nuclear Information System (INIS)

    Tosti, S.; Violante, V.; Basile, A.; Chiappetta, G.; Castelli, S.; De Francesco, M.; Scaglione, S.; Sarto, F.

    2000-01-01

    Palladium and palladium-silver permeators have been obtained by coating porous ceramic tubes with a thin metal layer. Three coating techniques have been studied and characterized: chemical electroless deposition (PdAg film thickness of 10 μm), ion sputtering (about 1 μm) and rolling of thin metal sheets (50 μm). The Pd-ceramic membranes have been used for manufacturing catalytic membrane reactors (CMR) for hydrogen and its isotopes recovering and purifying. These composite membranes and the CMR have been studied and developed for a closed-loop process with reference to the design requirements of the international thermonuclear experimental reactor (ITER) blanket tritium recovery system in the enhanced performance phase of operation. The membranes and CMR have been tested in a pilot plant equipped with temperature, pressure and flow-rate on-line measuring and controlling devices. The conversion value for the water gas shift reaction in the CMR has been measured close to 100% (always above the equilibrium one, 80% at 350 deg. C): the effect of the membrane is very clear since the reaction is moved towards the products because of the continuous hydrogen separation. The rolled thin film membranes have separated the hydrogen from other gases with a complete selectivity and exhibited a slightly larger mass transfer resistance with respect to the electroless membranes. Preliminary tests on the sputtered membranes have also been carried out with a promising performance. Considerations on the use of different palladium alloy in order to improve the performances of the membranes in terms of permeation flux and mechanical strength, such as palladium/yttrium, are also reported

  13. The First Benchmarking of ITER BR Nb3Sn Strand of CNDA

    International Nuclear Information System (INIS)

    Long Feng; Liu Fang; Wu Yu; Ni Zhipeng

    2012-01-01

    According to the International Thermonuclear Experimental Reactor (ITER) Procurement Arrangement (PA) of Cable-In-Conduit Conductor (CICC) unit lengths for the Toroidal Field (TF) and Poloidal Field (PF) magnet systems of ITER, at the start of process qualification, the Domestic Agency (DA) shall be required to conduct a benchmarking of the room and low temperature acceptance tests carried out at the Strand Suppliers and/or at its Reference Laboratories designated by the ITER Organization (IO). The first benchmarking was carried out successfully in 2009. Nineteen participants from six DAs (China, European Union, Japan, South Korea, Russia, and the United States) participated in the first benchmarking. Bronze-route (BR) Nb 3 Sn strand and samples prepared by the ITER reference lab (CERN) were sent out to each participant by CERN. In this paper, the test facility and test results of the first benchmarking by the Chinese DA (CNDA) are presented.

  14. Safety characteristics of options for plasma-facing components for ITER and beyond

    International Nuclear Information System (INIS)

    Piet, S.J.; McCarthy, K.A.; Holland, D.F.; Longhurst, G.R.; Merrill, B.J.

    1991-01-01

    Plasma-facing components (PFC) likely dominate the safety hazards of the International Thermonuclear Experimental Reactor (ITER) and post-ITER machines. To gain regulatory approval and for fusion energy to fulfill its ultimate attractive safety and environmental potential, safety must be considered when selecting among PFC options. This paper summarizes current PFC safety information. PFC safety issues fall into seven areas: disruption tolerance, disruption severity, tritium inventory and permeation, accidental energy release, activation/toxin hazards, cooling disturbances, and system issues. RFC options include current ITER mainline options (Be or W coating, C tiles), variants on current ITER options, and liquid metal (LM) divertors. No PFC option that we have examined is free of critical safety concerns. There are also innovative ideas that may improve any PFC's performance -- super-permeable vacuum ducts, helium self-pumping, and gaseous divertors. We conclude with recommendations and a future strategy. 17 refs., 1 fig., 3 tabs

  15. ECE diagnostics for RTO/RC ITER

    International Nuclear Information System (INIS)

    Vayakis, G.; Bartlett, D.V.; Costley, A.E.

    2001-01-01

    This paper presents the current status of the Electron Cyclotron Emission (ECE) diagnostic on the Reduced Technical Objectives/Reduced Cost International Thermonuclear Experimental Reactor (RTO/RC ITER). It discusses the implications of the new machine design on the measurement requirements, the ability of the diagnostic technique to meet these, and the changes in the implementation imposed by the new layout. Finally, it outlines the physics studies, design and R and D work required prior to the detailed design and construction of the diagnostic. Key results are: (i) that the localisation of the measurement is similar to that in ITER-FDR (40-100 mm in X-mode, 60-200 mm in O-mode for the reference scenario), so that the relative spatial resolution degrades in this, smaller, machine, and (ii) the expected effect of transport barriers on the temperature profile in the high temperature region will be poorly resolved, because the effect of the temperature gradient on the outboard side is to degrade the resolution to (∼250 mm in X-mode, ∼350 mm in O-mode). Nevertheless ECE will be able to make a unique and useful contribution to the RTO/RC ITER measurement set

  16. Experimental facility of innovative types as the laboratory analog of research reactor experimental device

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Zabud'ko, A.N.; Kremenetskij, A.K.; Nikolaev, A.N.; Trykov, L.A.

    1991-01-01

    The paper analyses capability of creating laboratory analogs of complex experimental facilities at research reactors utilizing power radionuclide neutron sources fabricated in industrial conditions. Some experimental and calculational investigations of neutron-physical characteristics are presented, which have been attained at the RIZ research reactor laboratory analog. Experimental results are supplemented by calculational investigations, fulfilled by means of the BRAND three-dimensional computational complex and the ROZ-6 one-dimensional program. 4 refs.; 3 figs

  17. Spirit and prospects of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Velikhov, E.P. [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

    2002-10-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  18. Spirit and prospects of ITER

    International Nuclear Information System (INIS)

    Velikhov, E.P.

    2002-01-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  19. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  20. Stress analysis of a double-wall vacuum vessel for ITER

    International Nuclear Information System (INIS)

    Conner, D.L.; Williamson, D.E.; Nelson, B.E.

    1991-01-01

    The preliminary structural analyses performed in support of the design of the vacuum vessel for the International Thermonuclear Experimental Reactor (ITER) are described. A thin, double-wall, all-welded structure is the proposed design concept analyzed. The results of the static stress analysis indicate the adequacy of such a structure. The effects of the proposed high-aspect-ratio design configuration on loading and stresses are also discussed. 4 refs., 6 figs., 1 tab

  1. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    Najzer, M.; Pauko, M.; Glumac, B.; Acquah, I.N.; Moskon, F.

    1977-01-01

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  2. Use of MCAM in creating 3D neutronics model for ITER building

    International Nuclear Information System (INIS)

    Zeng Qin; Wang Guozhong; Dang Tongqiang; Long Pengcheng; Loughlin, Michael

    2012-01-01

    Highlights: ► We created a 3D neutronics model of the ITER building. ► The model was produced from the engineering CAD model by MCAM software. ► The neutron flux map in the ITER building was calculated. - Abstract: The three dimensional (3D) neutronics reference model of International Thermonuclear Experimental Reactor (ITER) only defines the tokamak machine and extends to the bio-shield. In order to meet further 3D neutronics analysis needs, it is necessary to create a 3D reference model of the ITER building. Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM) was developed as a computer aided design (CAD) based bi-directional interface program between general CAD systems and Monte Carlo radiation transport simulation codes. With the help of MCAM version 4.8, the 3D neutronics model of ITER building was created based on the engineering CAD model. The calculation of the neutron flux map in ITER building during operation showed the correctness and usability of the model. This model is the first detailed ITER building 3D neutronics model and it will be made available to all international organization collaborators as a reference model.

  3. ITER EDA newsletter. V. 8, no. 11

    International Nuclear Information System (INIS)

    1999-11-01

    This ITER EDA Newsletter contains summary reports on the eleventh meeting of the ITER diagnostic expert group in Cadarache, France, on the ITER JCT presentation at the international conference on fusion reactor materials in Colorado Springs, USA and on the seventh workshop on plasma edge theory in fusion devices in Tajimi, Japan. Individual abstracts are prepared for the three contributions

  4. ITER Council proceedings: April 1988 - August 1989

    International Nuclear Information System (INIS)

    1990-01-01

    The ITER documentation series, of which this is the sixth report, began with a concise record of the decisions and actions taken in establishing ITER. The contents of that first report include the Terms of Reference Concerning Conceptual Design Activities for an International Thermonuclear Experimental Reactor. The Terms of Reference, which were part of the IAEA's invitation to the prospective parties, formally describe how the co-operative work of the four Parties in the specified activities is directed and managed. The first report in the series also covered activities from the initial meeting of the ITER Quadripartite Initiative Committee in March 1987 through March 1988. The present report is intended to make available in convenient form the essential information on ''landmark'' events in the direction of the ITER activities from the first meeting of the ITER Council (IC), in April 1988, through the letter report by the Council following their fourth meeting in July 1989. This report therefore covers approximately the first half of the Conceptual Design Activities, which are to be concluded in December 1990. THe next section of this report provides, for convenient reference, an overview of the organization and schedule that were adopted for the ITER activities through 1990. The sections that follow contain, for each of the four IC meetings during the period covered by this report, a copy of the official record of the transactions. The written reports of the ITER Management Committee (IMC) and the ITER Scientific and Technical Advisory Committee (ISTAC) to the IC in connection with the meetings are represented by summaries, prepared by the ITER Secretariat. These summaries include direct quotations of especially significant statements in the reports. In general, other supporting material is included only if it is of more than transitory significance. 1 fig and tabs

  5. 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

  6. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    International Nuclear Information System (INIS)

    El-Morshedy, Salah El-Din; Hassanein, Ahmed

    2009-01-01

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  7. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  8. Design of water detritiation system for fusion reactor

    International Nuclear Information System (INIS)

    Xie Bo; Wang Heyi; Liu Yunnu; Guan Rui

    2006-01-01

    The water detritiation system (WDS) of tritium plant for the International Thermonuclear Experimental Reactor (ITER) was designed. The concept of the Combined Electrolysis Catalytic Exchange and Gas Chromatography (CECE-GC) process was selected for the system and subsystems' descriptions of the WDS. ITER-WDS is characterised from the present demonstration system by rejecting the use of a recombiner and alkali electrolyzer, but a solid polymer electrolyzer (SPE) and a Pd/Ag membrane permeator system are adopted to recover tritium. (authors)

  9. The 'Reacteur Jules Horowitz': a new experimental reactor project

    International Nuclear Information System (INIS)

    Frachet, S.; Ballagny, A.

    1999-01-01

    The Jules Horowitz Reactor (RJH) is a new research reactor project dedicated to materials and nuclear fuel testing, the location of which is foreseen at the CEA-CADARACHE site, and the start-up in 2006. The launching of this project originated from a double finding: The development of nuclear power plants aimed at satisfying the energy needs of the next century, cannot be envisaged without experimental reactors which are unrivaled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. The existing experimental reactors are 30 to 40 years old and it is advisable to examine henceforth the necessity for and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a study on the mid and long term irradiation needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfill the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The reactor project selected among several different concepts, is finally a light water pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows for many irradiations in and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and the French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the following most important items: neutronic and thermalhydraulic studies on alternative core designs, with or without added pressurization

  10. EU contribution to ITER CTA fuel cycle design and R and D

    International Nuclear Information System (INIS)

    Murdoch, D.; Bekris, N.; Boissin, J.-C.; Day, C.; Gilroy, J.; Glugla, M.; Laesser, R.; Mack, A.

    2003-01-01

    Design and supporting R and D are being carried out in the EU in preparation for the construction of International Thermonuclear Experimental Reactor (ITER). As the Fuel Cycle systems are located in 'time-critical' buildings the detailed design of piping, ducting, cabling and services is being done to ensure compatibility with the building. R and D is focussed on improvements to the reference designs for ITER, extension of parametric studies to cover all operating modes and long term experiments to test component reliability. In addition emphasis is being placed on design and R and D for the Water Detritiation System, an important barrier against routine release of tritium

  11. Helium cooling of fusion reactors

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Wall conditioning for ITER: Current experimental and modeling activities

    Energy Technology Data Exchange (ETDEWEB)

    Douai, D., E-mail: david.douai@cea.fr [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Kogut, D. [CEA, IRFM, Association Euratom-CEA, 13108 St. Paul lez Durance (France); Wauters, T. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Brezinsek, S. [FZJ, Institut für Energie- und Klimaforschung Plasmaphysik, 52441 Jülich (Germany); Hagelaar, G.J.M. [Laboratoire Plasma et Conversion d’Energie, UMR5213, Toulouse (France); Hong, S.H. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Lomas, P.J. [CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Lyssoivan, A. [LPP-ERM/KMS, Association Belgian State, 1000 Brussels (Belgium); Nunes, I. [Associação EURATOM-IST, Instituto de Plasmas e Fusão Nuclear, 1049-001 Lisboa (Portugal); Pitts, R.A. [ITER International Organization, F-13067 St. Paul lez Durance (France); Rohde, V. [Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany); Vries, P.C. de [ITER International Organization, F-13067 St. Paul lez Durance (France)

    2015-08-15

    Wall conditioning will be required in ITER to control fuel and impurity recycling, as well as tritium (T) inventory. Analysis of conditioning cycle on the JET, with its ITER-Like Wall is presented, evidencing reduced need for wall cleaning in ITER compared to JET–CFC. Using a novel 2D multi-fluid model, current density during Glow Discharge Conditioning (GDC) on the in-vessel plasma-facing components (PFC) of ITER is predicted to approach the simple expectation of total anode current divided by wall surface area. Baking of the divertor to 350 °C should desorb the majority of the co-deposited T. ITER foresees the use of low temperature plasma based techniques compatible with the permanent toroidal magnetic field, such as Ion (ICWC) or Electron Cyclotron Wall Conditioning (ECWC), for tritium removal between ITER plasma pulses. Extrapolation of JET ICWC results to ITER indicates removal comparable to estimated T-retention in nominal ITER D:T shots, whereas GDC may be unattractive for that purpose.

  13. Magnetic fusion and project ITER

    International Nuclear Information System (INIS)

    Park, H.K.

    1992-01-01

    It has already been demonstrated that our economics and international relationship are impacted by an energy crisis. For the continuing prosperity of the human race, a new and viable energy source must be developed within the next century. It is evident that the cost will be high and will require a long term commitment to achieve this goal due to a high degree of technological and scientific knowledge. Energy from the controlled nuclear fusion is a safe, competitive, and environmentally attractive but has not yet been completely conquered. Magnetic fusion is one of the most difficult technological challenges. In modem magnetic fusion devices, temperatures that are significantly higher than the temperatures of the sun have been achieved routinely and the successful generation of tens of million watts as a result of scientific break-even is expected from the deuterium and tritium experiment within the next few years. For the practical future fusion reactor, we need to develop reactor relevant materials and technologies. The international project called ''International Thermonuclear Experimental Reactor (ITER)'' will fulfill this need and the success of this project will provide the most attractive long-term energy source for mankind

  14. ITER Blanket First Wall (WBS 1.6{sub 1}A)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kim, H. G.; Kim, J. H. (and others)

    2008-03-15

    International Thermonuclear Experimental Reactor (ITER) project is the international collaboration one for the commercialization of nuclear fusion energy through the technical and engineering verification. In ITER project, we plan to procure the blanket systems which has the risk of technology and cost when it is newly developed. We are developing the manufacturing process and joining technology for the ITER blanket to complete the procurement with qualified blanket system. To evaluate the soundness of manufacturing process, specimen and mock-up tests are being prepared. Finally, we can obtain the key technology of nuclear fusion reactor especially on the blanket design, joining and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea. In 1st year, through the fabrication of the Cu/SS and Be/Cu joint specimen, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The optimized HIP conditions (1050 .deg. C, 150 MPa, 2 hr for Cu/SS and 580 - 620 .deg. C, 100-150 MPa, 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint and NDT such as UT (10 MHz, 0.25 inch D, flat type) and ECT. Several mock-ups were fabricated for confirming the joint integrity and NDT. specimens fabricated with these mock-ups were used in mechanical tests including microstructure observation. The mock-ups were used in the HHF test after the developed NDT. In 2nd year, PHHT of Cu was investigated in order to recover its mechanical properties, and the pre-qualification mock-up were fabricated against the Qualification Program and sent to RF for HHF testing in TSEFEY. FW fabrication and joining procedure were documented in the form of the TSD. Qualification mock

  15. The status of the ITER CODAC conceptual design

    International Nuclear Information System (INIS)

    Lister, J.B.; Farthing, J.W.; Greenwald, M.; Yonekawa, I.

    2008-01-01

    CODAC orchestrates the activity of 80-120 Plant Systems in normal ITER operation using a set of dedicated networks. Interlock systems protect ITER from potentially damaging operating off-normal conditions, using a segregated network. Safety systems protect the personnel and the environment and the nuclear safety systems will be subject to licensing. The principal challenges to be met in the design and implementation of CODAC include: complexity, reliability, transparent access respecting security, a high experiment data rate and data volume since ITER is an experimental reactor, scientific exploitation from multiple participant team experiment sites and the long 35-year-period for construction and operation. Two non-technical challenges drive the CODAC design. Firstly all technical plant systems follow an 'in-kind' procurement model, as opposed to a direct commercial contractual relationship with suppliers. Secondly, the period of procurement will, itself, be long compared with the evolution of technologies in the fields of control and data acquisition.This paper addresses the various phases from definition of procurement agreements through to on-site commissioning of the project

  16. The structure and thermal properties of plasma-sprayed beryllium for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Castro, R.G.; Bartlett, A.; Elliott, K.E.; Hollis, K.J.

    1996-01-01

    Plasma spraying is being studied for in situ repair of damaged Be and W plasma facing surfaces for ITER, the next generation magnetic fusion energy device, and is also being considered for fabricating Be and W plasma-facing components for the first wall of ITER. Investigators at LANL's Beryllium Atomization and Thermal Spray Facility have concentrated on investigating the structure-property relation between as-deposited microstructures of plasma sprayed Be coatings and resulting thermal properties. In this study, the effect of initial substrate temperature on resulting thermal diffusivity of Be coatings and the thermal diffusivity at the coating/Be substrate interface (interface thermal resistance) was investigated. Results show that initial Be substrate temperatures above 600 C can improve the thermal diffusivity of the Be coatings and minimize any thermal resistance at the interface between the Be coating and Be substrate

  17. Iterative nonlinear unfolding code: TWOGO

    International Nuclear Information System (INIS)

    Hajnal, F.

    1981-03-01

    a new iterative unfolding code, TWOGO, was developed to analyze Bonner sphere neutron measurements. The code includes two different unfolding schemes which alternate on successive iterations. The iterative process can be terminated either when the ratio of the coefficient of variations in terms of the measured and calculated responses is unity, or when the percentage difference between the measured and evaluated sphere responses is less than the average measurement error. The code was extensively tested with various known spectra and real multisphere neutron measurements which were performed inside the containments of pressurized water reactors

  18. In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, Jan [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Entler, Slavomir, E-mail: slavomir.entler@cvrez.cz [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Vsolak, Rudolf; Klabik, Tomas [Research Centre Rez, Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Zlamal, Ondrej [CEZ, Duhova 2/1444, 140 53 Praha 4 (Czech Republic); Bellin, Boris; Zacchia, Francesco [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • Irradiated thermal fatigue testing of the ITER primary first wall mock-ups. • Cyclic heat flux of 0.5 MW/m{sup 2} in the neutron field of the nuclear reactor core. • 17,040 thermal cycles. • Radiation damage in the range of 0.41–1.17 dpa depending on the material. - Abstract: The TW3 in-pile rig enabled the thermal fatigue testing of ITER primary first wall mock-ups in the core of the nuclear reactor. This experiment investigated the neutron irradiation influence on the design performance under high heat flux testing. A thermal flux of 0.5 MW/m{sup 2} in the neutron field of the core of the LVR-15 nuclear reactor was applied. Within the scope of the tests with simultaneous neutron irradiation, the TW3 rig reached a record of 17,040 thermal cycles with the radiation damage in the range of 0.41–1.17 dpa depending on the material. Even after a high number of thermal cycles, while being irradiated by neutrons, no damage of the tested mock-ups was visually observed. Further testing and analysis will follow in the Forschungszentrum Juelich.

  19. 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.)

  20. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

    2001-01-01

    A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

  1. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  2. The experimental nuclear reactor: AQUILON; Le reacteur nucleaire experimental: AQUILON

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Y; Koechlin, J C; Moreau, J M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [French] 'Aquilon' est un reacteur experimental specialement concu pour l'etude neutronique de milieux multiplicateurs heterogenes a combustible solide et ralentisseur liquide. Cette etude etant en general incompatible avec la production d'energie, on a limite au minimum la puissance du reacteur pour pouvoir obtenir une structure simple et peu encombrante, un acces facile, une bonne maniabilite et une grande souplesse de fonctionnement et d'utilisation. (auteur)

  3. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-01-01

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

  4. Operation and control of ITER plasmas

    International Nuclear Information System (INIS)

    2001-01-01

    Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

  5. Operation and control of ITER plasmas

    International Nuclear Information System (INIS)

    1999-01-01

    Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

  6. The experimental program of neutronphysics for advanced water reactors

    International Nuclear Information System (INIS)

    Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

    1985-11-01

    The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

  7. The Orphee reactor current status and proposed enhancement of experimental capabilities

    International Nuclear Information System (INIS)

    Breant, P.

    1990-01-01

    This report provides a description of the Orphee reactor, together with a rapid assessment of its experimental and research capabilities. The plans for enhancing the reactor's experimental capabilities are also presented. (author)

  8. Conceptual design of SC magnet system for ITER, (5)

    International Nuclear Information System (INIS)

    Nakajima, Hideo; Nishi, Masataka; Yoshida, Kiyoshi; Tsuji, Hiroshi; Egusa, Shigenori; Seguchi, Tadao; Hagiwara, Miyuki; Kirk, M.A.; Birtcher, R.C.

    1991-08-01

    Japan Atomic Energy Research Institute (JAERI) has been developing a superconducting magnet system for a fusion reactor. One of the key items in developing the superconducting magnets is material development and evaluation. The data of superconducting materials, structural alloys, and non-metallic materials are generated to establish a material data base at JAERI. This report is prepared to provide available data generated by JAERI to designers of superconducting magnets throughout the world. The following review papers written for the International Thermonuclear Experimental Reactor (ITER) report on conceptual design of magnet system are combined here. I. Superconducting Material Data II. Mechanical Properties of the Japanese Cryogenic Steels (JCS) at Cryogenic Temperature III. Review of Radiation Degradation Studies at JAERI on Composite Organic Insulators Used in Fusion Magnets (author)

  9. Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tarchalski, M.; Pytel, K.; Wroblewska, M.; Marcinkowska, Z.; Boettcher, A.; Prokopowicz, R. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Sireta, P.; Gonnier, C.; Bignan, G. [CEA, DEN, Reactor Studies Department, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Fourmentel, D.; Barbot, L.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C.; Brun, J. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Jagielski, J. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Institute of Electronic Materials Technolgy, Wolczynska 133, 01-919 Warszawa (Poland); Luks, A. [Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland)

    2015-07-01

    Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from

  10. 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; Projet de loi adopte par le Senat, autorisant l'approbation de l'accord entre le Gouvernement de la Republique francaise et l'Organisation internationale ITER pour l'energie de fusion relatif au siege de l'Organisation ITER et aux privileges et immunites de l'Organisation ITER sur le territoire francais

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-01-15

    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.)

  11. Preliminary consideration of CFETR ITER-like case diagnostic system.

    Science.gov (United States)

    Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

    2016-11-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  12. Preliminary consideration of CFETR ITER-like case diagnostic system

    International Nuclear Information System (INIS)

    Li, G. S.; Liu, Y. K.; Gao, X.; Yang, Y.; Wang, Y. M.; Ming, T. F.; Han, X.; Liu, S. C.; Wang, E. H.; Yang, W. J.; Li, G. Q.; Hu, Q. S.

    2016-01-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  13. Experimental use of iteratively designed rotation invariant correlation filters

    International Nuclear Information System (INIS)

    Sweeney, D.W.; Ochoa, E.; Schils, G.F.

    1987-01-01

    Iteratively designed filters are incorporated into an optical correlator for position, rotation, and intensity invariant recognition of target images. The filters exhibit excellent discrimination because they are designed to contain full information about the target image. Numerical simulations and experiments demonstrate detection of targets that are corrupted with random noise (SNR≅0.5) and also partially obscured by other objects. The complex valued filters are encoded in a computer generated hologram and fabricated directly using an electron-beam system. Experimental results using a liquid crystal spatial light modulator for real-time input show excellent agreement with analytical and numerical computations

  14. ITER and the fusion reactor: status and challenge to technology

    International Nuclear Information System (INIS)

    Lackner, K.

    2001-01-01

    Fusion has a high potential, but requires an integrated physics and technology effort without precedence in non-military R and D, the basic physics feasibility demonstration will be concluded with ITER, although R and D for efficiency improvement will continue. The essential technological issues remaining at the start of ITER operation concern materials questions: first wall components and radiation tolerant (low activation materials). This paper comprised just the copy of the slides presentation with the following subjects: magnetic confinement fusion, the Tokamak, progress in Tokamak performance, ITER: its geneology, physics basis-critical issues, cutaway of ITER-FEAT, R and D - divertor cassette (L-5), differences power plant-ITER, challenges for ITER and fusion plants, main technological problems (plasma facing materials), structural and functional materials for fusion power plants, ferritic steels, EUROFER development, improvements beyond ferritic steels, costing among others. (nevyjel)

  15. Fusion technology: The Iter fusion experiment

    International Nuclear Information System (INIS)

    Dietz, K.J.

    1994-01-01

    Plans for the Iter international fusion experiment, in which the European Union, Japan, Canada, Russia, Sweden, Switzerland, and the USA cooperate, were begun in 1985, and construction work started in early 1994. These activities serve for the preparation of the design and construction documents for a research reactor in which a stable fusion plasma is to be generated. This is to be the basis for the construction of a fusion reactor for electricity generation. Preparatory work was performed in the Tokamak experiments with JET and TFTR. The fusion power of 1.5 GW will be attained, thus enabling Iter to keep a deuterium-tritium plasma burning. (orig.) [de

  16. Real-time numerical simulation with high efficiency for an experimental reactor system

    International Nuclear Information System (INIS)

    Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan

    2006-01-01

    The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)

  17. Use of MCAM in creating 3D neutronics model for ITER building

    Energy Technology Data Exchange (ETDEWEB)

    Zeng Qin [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Wang Guozhong, E-mail: mango33@mail.ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Dang Tongqiang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Long Pengcheng [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Loughlin, Michael [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lz-Durance (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer We created a 3D neutronics model of the ITER building. Black-Right-Pointing-Pointer The model was produced from the engineering CAD model by MCAM software. Black-Right-Pointing-Pointer The neutron flux map in the ITER building was calculated. - Abstract: The three dimensional (3D) neutronics reference model of International Thermonuclear Experimental Reactor (ITER) only defines the tokamak machine and extends to the bio-shield. In order to meet further 3D neutronics analysis needs, it is necessary to create a 3D reference model of the ITER building. Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM) was developed as a computer aided design (CAD) based bi-directional interface program between general CAD systems and Monte Carlo radiation transport simulation codes. With the help of MCAM version 4.8, the 3D neutronics model of ITER building was created based on the engineering CAD model. The calculation of the neutron flux map in ITER building during operation showed the correctness and usability of the model. This model is the first detailed ITER building 3D neutronics model and it will be made available to all international organization collaborators as a reference model.

  18. ITER Status and Plans

    Science.gov (United States)

    Greenfield, Charles M.

    2017-10-01

    The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

  19. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  20. Experimental investigation on streaming due to a gap between blanket modules in ITER

    International Nuclear Information System (INIS)

    Konno, Chikara; Maekawa, Fujio; Oyama, Yukio; Uno, Yoshitomo; Kasugai, Yoshimi; Maekawa, Hiroshi; Ikeda, Yujiro; Wada, Masayuki

    2000-01-01

    A gap streaming experiment was performed by using a D-T neutron source at FNS/JAERI as the ITER/EDA R and D Task T-218, in order to examine the streaming effects due to gap between shield blanket modules in ITER. The experiment had three phases. The first one defined neutron source characteristics (Source Characterization Experiment), the second (Experiment-l ) aimed at shield for welding part between shield blanket and back plate and the third (Experiment-2) focused on the influence that the gap between shield blanket modules would have on superconducting magnet. The effects of gap streaming were examined in detail experimentally. (author)

  1. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  2. The High Aspect Ratio Design (HARD): A candidate ITER concept with improved technology phase performance

    International Nuclear Information System (INIS)

    Nevins, W.M.; Perkins, L.J.; Wesley, J.C.

    1992-10-01

    The High Aspect Ratio Design (HARD) International Thermonuclear Experimental Reactor (ITER) concept developed by the US ITER team is an alternate to the low-aspect-ratio ITER design developed by the ITER participants during the Conceptual Design Activity (CDA). The CDA design, referred to hereafter as ITER CDA, has an aspect ratio, A, of 2.79, a toroidal magnetic field, B T , of 4.85 T, and a plasma current, I p , of 22 MA for operation with an ignited plasma. In contrast, HARD employs higher aspect ratio, A = 4.0, higher toroidal field, B T = 7.11 T, and lower plasma current, I p = 14.8 MA for ignition operation. The cross sections of the two designs are compared in. The parameters and performance of HARD and ITER CDA for inductively driven ignition operation are compared in. The HARD parameters provide the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) as ITER CDA in a device with comparable size and cost. However, the reason for advancing HARD rather than ITER CDA as the ITER design concept is not inductively driven ignition performance but HARD's significantly enhanced potential to achieve the technology testing and steady-state operation goals of the ITER objectives with non-inductive current drive

  3. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

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

  4. Progressing state of design and R and D of NBI for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-02-01

    In the International Thermal Nuclear Fusion Experimental Reactor (ITER), Neutral Beam Injection (NBI) apparatus is thought to be a powerful means of electric current drive for heating and stabilizing of plasma, and of controlling the plasma stably. Then, design and development of 1 MeV class negative ion NBI apparatus with compactness and better consistency with reactor have been conducted. On its engineering design, numbers of ports for NBI apparatus were changed form 3 to 4 on a stage of intermediate design completion on June, 1995, and then incident power per unit port was increased from 12.5 MW to 16.7 MW, which brought severer characteristics required for negative ion source. At present, designs of beam deflector, magnetic shield, neutron shielding, remote maintenance and so forth as well as negative ion source and accelerator have been progressed. On its engineering R and D, for development of negative ion source, both deuterium negative ion current and its density established about 1/3 of the characteristics in ITER actural apparatus at an aimed operational gas pressure. And, for development of negative ion accerelator, over 80% of the negative ion acceleration energy which corresponds to an aim of ITER could be established. (G.K.)

  5. Safety methodology implementation in the conceptual design phase of a fusion reactor

    International Nuclear Information System (INIS)

    Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

    2007-01-01

    The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

  6. A proposal for the ITER remote participation system in Japan

    International Nuclear Information System (INIS)

    Nagayama, Y.; Emoto, M.; Kozaki, Y.; Nakanishi, H.; Sudo, S.; Yamamoto, T.; Hiraki, K.; Urushidani, S.

    2010-01-01

    This paper presents a proposal of the remote participation system for the international thermonuclear experimental reactor (ITER). The object of this paper is to clarify technical issues to analyze the ITER data safely and conveniently. The Japanese case is considered as an example, but technologies presented here can be used worldwide. Major technical issues are as follows: (1) the long distance data transfer; (2) the massive data server; (3) the secure network; (4) the convenient and fast data analysis system. Raw data of ITER can be transferred from France to Japan in a short time by optimizing TCP/IP parameters. The virtual private network (VPN) technology provides a secure environment of the data mirroring and the distributed computation. The analysis server with the WEB user interface enables physicists to analyze the ITER data from the Internet. Streaming data, such as plasma parameters in the steady state, video and sound of the ITER plasma and the status of experiment, which provides feeling of reality, are delivered by using the multi-cast technology. These technologies are being developed in SNET, which is a virtual laboratory for Japanese fusion community. International collaboration is required to develop a global distributed file system and a data analysis system further.

  7. A proposal for the ITER remote participation system in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nagayama, Y., E-mail: nagayama.yoshio@nifs.ac.j [National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292 (Japan); Emoto, M.; Kozaki, Y.; Nakanishi, H.; Sudo, S.; Yamamoto, T. [National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292 (Japan); Hiraki, K. [Graduate School of Information Science and Technology, University of Tokyo, 7-3-1 Hongo, Tokyo 113-8656 (Japan); Urushidani, S. [National Institute of Informatics, 2-1-2 Hitotsubashi, Chiyoda-ku, Tokyo 101-8430 (Japan)

    2010-07-15

    This paper presents a proposal of the remote participation system for the international thermonuclear experimental reactor (ITER). The object of this paper is to clarify technical issues to analyze the ITER data safely and conveniently. The Japanese case is considered as an example, but technologies presented here can be used worldwide. Major technical issues are as follows: (1) the long distance data transfer; (2) the massive data server; (3) the secure network; (4) the convenient and fast data analysis system. Raw data of ITER can be transferred from France to Japan in a short time by optimizing TCP/IP parameters. The virtual private network (VPN) technology provides a secure environment of the data mirroring and the distributed computation. The analysis server with the WEB user interface enables physicists to analyze the ITER data from the Internet. Streaming data, such as plasma parameters in the steady state, video and sound of the ITER plasma and the status of experiment, which provides feeling of reality, are delivered by using the multi-cast technology. These technologies are being developed in SNET, which is a virtual laboratory for Japanese fusion community. International collaboration is required to develop a global distributed file system and a data analysis system further.

  8. Organization of the ITER Project - Sharing of information and procurements

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1990-01-01

    Cooperation among the four countries to build a single experimental reactor would reduce the cost for each country and provide an international pool of scientific and engineering resources. Countries and Organizations participating in international agreements with the USA in fusion energy development are: Australia, Canada, European Community, IAEA, Soviet Union, Peoples Republic of China, Israel, Spain, Japan. ITER 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. The goal of the Large Coil Task (LCT) was successfully accomplished its intended purpose (September 1987) and the world-wide participants in fusion research are reaping benefits from the multinational effort. ITER organization for conceptual design activity can be summarized as follows: Technical site for design effort at is placed in Garching, West Germany; ITER Council is responsible to IAEA for overall direction of activities; ITER Management Committee is responsible for execution of activities and manages work at Garching; ITER Scientific and Technical Advisory Committee consists of eminent scientists and engineers and advises the ITER council. Conceptual project schedule and cost are presented, with the basic principles related to task sharing and handling of intellectual property

  9. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10 6 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  10. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  11. Safety analysis results for cryostat ingress accidents in ITER

    International Nuclear Information System (INIS)

    Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

    1996-01-01

    Accidents involving the ingress of air or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits

  12. Failure Modes and Effects Analysis on ITER DFLL-TBM system

    International Nuclear Information System (INIS)

    Hu Liqin; Yuan Run; Chen Hongli; Bai Yunqing

    2012-01-01

    As required for licensing process, accident analyses of International Thermonuclear Experimental Reactor (ITER) accounting for site specifications and design changes will be updated. Chinese Dual-Functional Lithium-Lead-Test Blanket Module (DFLL-TBM) system is a key safety-related component of ITER, its detailed safety analysis, which was designated to demonstrate the integrated technologies of both Helium single coolant (SLL) blanket and Helium-LiPb dual coolant (DLL) blanket, was performed. Failure Modes and Effects Analysis (FMEA) was applied to perform the safety analysis of DFLL-TBM. This study described the process of FMEA studies on DFLL-TBM system. All safety-related Postulated Initiating Events (PIEs) was identified. And a set of PIEs recommended to be taken into account in the further deterministic transient analyses were defined for both SLL and DLL blanket concepts separately.

  13. Tasks related to increase of RA reactor exploitation and experimental potential, 03. Crane for handling the vertical experimental channels of the RA reactor - design project

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    Within the work related to improvement of experimental potential of the RA reactor, this document describes the design project of the new crane for handling the vertical experimental channels of the RA reactor, engineering drawings of the crane main elements, mechanical part, design project of the electrical part of the crane and cost estimation

  14. Status of the Japanese ITER Home Team: January 1993

    International Nuclear Information System (INIS)

    Matsuda, Shinzaburo

    1994-01-01

    In June, 1992, Atomic Energy Commission of Japan determined the Third Phase Basic Program of Fusion Research and Development. It directs national policy for the experimental reactor phase of fusion research and development. As a government committee, the promotion and the planning of the entire fusion program will be continually carried out by the Fusion Council of Atomic Energy Commission. The Fusion Council has recently established an ITER Technical Committee which will give advice on technical matters of the ITER program to the Fusion Council. Thus, the government is ready to be fully supportive of ITER for the execution of this unprecedented international collaboration. There will be some other units to be organized in the fear future, in pace with the evolution of ITER activities. The involvement of other research institutes is open as a future possibility. The number of persons nominated as Home Team members is about 100 at present and will be increased depending upon the tasks assigned to the Japanese Home Team. The participation of industries in the EDA is of significant importance for the success of ITER. Firstly, innovative concepts or proposals owing to the technical expertise in other fields can be expected. Secondly, experience in production, fabrication or assembly is valuable in the integral review of the design. Thirdly, development and integration of production technologies are essential to realize future construction

  15. Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-05-01

    This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

  16. Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-04-01

    This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

  17. Status and plans for U.S. ITER studies

    International Nuclear Information System (INIS)

    Doggett, J.N.

    1992-01-01

    The United States' participation in the International Thermonuclear Experimental Reactor (ITER) began in late 1987 when the initiative to start a cooperative program among the four Parties-the Soviet Union, Japan, the European Community, and the United States-was initiated. Participation then continued through the start of joint Work in May 1988 until the conclusion of the Conceptual Design Activities (CDA) in December 1990. In the period between the conclusion of the CDA and the agreement to execute the Engineering Design Activities (EDA), the U.S. ITER Home Team continued to do work on the design, executed additional research and development (R and D) and participated in the preparations for the EDA. Activities included one major design study on a High-Aspect-Ratio Design (HARD) and input to the National ITER Technical Review, the ITER Steering Committee-U.S. (ISCUS), Special Working Group 1 (SWG-1), and the Fusion Energy Advisory Committee's Panel 1 (FEAC-1). Research and development was continued in areas of work that were identified as critical-path elements by an international panel chartered by the four ITER Parties near the end of the CDA. During the interim period, the U.S. Home Team Management (HTM) was in the process of organizing to support the EDA both at home and in the central design sites. The major efforts have been in producing a management plan, establishing memorandums of agreement with the performing institutions for ITER tasks, establishing an industrial council, and producing a list of candidates who are qualified, willing, and available to serve on the joint Central Team or to participate in ITER home tasks. The author describes the conclusion of the CDA and the interim U.S. ITER activities and will give an indication of US involvement in the EDA

  18. Geometric interpretation of optimal iteration strategies

    International Nuclear Information System (INIS)

    Jones, R.B.

    1977-01-01

    The relationship between inner and outer iteration errors is extremely complex, and even formal description of total error behavior is difficult. Inner and outer iteration error propagation is analyzed in a variational formalism for a reactor model describing multidimensional, one-group theory. In a generalization the work of Akimov and Sabek, the number of inner iterations performed during each outer serial that minimizes the total computation time is determined. The generalized analysis admits a geometric interpretation of total error behavior. The results can be applied to both transport and diffusion theory computer methods. 1 figure

  19. Review of past experiments at the FELIX facility and future plans for ITER applications

    International Nuclear Information System (INIS)

    Hua, T.Q.; Turner, L.R.

    1993-01-01

    FELIX is an experimental test facility at Argonne National Laboratory (ANL) for the study of electromagnetic effects in first wall, blanket, shield systems of fusion reactors. From 1983 to 1986 five major test series, including static and dynamic tests, were conducted and are reviewed in this paper. The dynamic tests demonstrated an important coupling effect between eddy currents and motion in a conducting structure. Recently the U.S. has proposed to the ITER Joint Central Team to use FELIX for testing mock-up components to study electromagnetic effects encountered during plasma disruptions and other off-normal events. The near and long term plans for ITER applications are discussed. (author)

  20. R&D on high-power dc reactor prototype for ITER poloidal field converter

    Energy Technology Data Exchange (ETDEWEB)

    Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)

    2015-10-15

    Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

  1. Experimental Equipment for Physics Studies in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernander, G; Blomberg, P E; Dubois, P O

    1967-03-15

    Comprehensive physics measurements were carried out in connection with the start up of the Agesta reactor. For this purpose special experimental equipment was constructed and installed in the reactor. Parts of this were indispensable and/or time-saving for the reactivity control during the core build-up period and during the first criticality studies. This report gives mainly a detailed description of the experimental equipment used, but also the relevant physics background and the experience gained during the performance.

  2. Investigation of materials for fusion power reactors

    Science.gov (United States)

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

    2014-06-01

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

  3. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  4. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

  5. Assessment and selection of materials for ITER in-vessel components

    Science.gov (United States)

    Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

    2000-12-01

    During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

  6. Armour Materials for the ITER Plasma Facing Components

    Science.gov (United States)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

    The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

  7. Armour materials for the ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

    1999-01-01

    The selection of the armour materials for the plasma facing components (PFCs) of the international thermonuclear experimental reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-a-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R and D. (orig.)

  8. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

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

  9. Results of ITER/EDA and Japan's approach in the prolonged research period

    International Nuclear Information System (INIS)

    Tsunematsu, Toshihide

    1998-01-01

    The Engineering and Design Activity (EDA) of the International Thermonuclear Experimental Reactor (ITER) started operation with the cooperation of Japan, EU, Russia and the US in July 1992 with a plan to do work on the design of the ITER and necessary R and D. Six years passed and the final report was completed in February 1998, reporting that the initial targets were almost achieved except for design work depending on construction site. It was also decided to prolong the EDA for more three years, though the US did not participate. The new EDA items include: Design to meet site requirements and cost estimation; preparation for application for construction approval; and preparation of technical documents for equipment procurement in the future. Among them, the main EDA item is to design a low-cost ITER to be ready for construction when the three-year joint research is ended. For this, work was done to prepare technical guidelines. The EDA is now making the conceptual design of a new ITER 50% less expensive than the current ITER described in the final design report. Japan has proposed site requirements that are needed for design alteration when the ITER is constructed in Japan. The site requirements are concerned with earthquake resistance, ground, power supply and cooling conditions. (N.H.)

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

    International Nuclear Information System (INIS)

    Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

    1979-06-01

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

  11. 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

  12. ITER EDA newsletter. V. 4, no. 11

    International Nuclear Information System (INIS)

    1995-11-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a report on the Ninth Meeting of the ITER Management Advisory Committee held in St. Petersburg, Russia, on November 3, 1995; a report on the Seventh International Conference on Fusion Reactor Materials held at Obninsk, Russia, 25-29 September, 1995; on the presentation of the ITER Project during a symposium on fusion energy held at Champaign, Illinois, USA, October 1-5, 1995; and on two meetings on ITER diagnostics, i.e., an international workshop on diagnostics for ITER held in Varenna, Italy, 28 August - 1 September, 1995; followed by the Third Diagnostics Expert Group Workshop held September 4-5 in the same location

  13. The Physics Basis of ITER Confinement

    International Nuclear Information System (INIS)

    Wagner, F.

    2009-01-01

    ITER will be the first fusion reactor and the 50 year old dream of fusion scientists will become reality. The quality of magnetic confinement will decide about the success of ITER, directly in the form of the confinement time and indirectly because it decides about the plasma parameters and the fluxes, which cross the separatrix and have to be handled externally by technical means. This lecture portrays some of the basic principles which govern plasma confinement, uses dimensionless scaling to set the limits for the predictions for ITER, an approach which also shows the limitations of the predictions, and describes briefly the major characteristics and physics behind the H-mode--the preferred confinement regime of ITER.

  14. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, I.; Linke, J.; Nickel, H.; Kny, E.; Reheis, N.; Kneringer, G.; Bolt, H.

    1995-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr, 90Ni-10Ti, 90Cu- 10Ti, and 70Ag-27Cu-3Ti (compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  15. Experimental evaluation of brazed molybdenum-graphite bonds for the divertor of the NET/ITER nuclear fusion device

    International Nuclear Information System (INIS)

    Smid, Ivica; Linke, Jochen; Nickel, Hubertus; Kny, Erich; Reheis, Nikolaus; Kneringer, Guenther; Bolt, Harald

    1990-01-01

    Composites consisting of plasma-facing carbon material brazed to molybdenum (TZM) substrates are a promising system for the divertor of the Next European Torus (NET) and the International Thermonuclear Experimental Reactor (ITER). Isotropic graphite and a refractory metal (molybdenum or TZM, a high temperature alloy of molybdenum), two dissimilar substrate materials, yet closely matched in their thermal expansivities, were joined with the use of four different high-temperature brazes: Zr,90Ni-10Ti,90Cu-10Ti, and 70Ag-27Cu-3Ti(compositions in wt%). A summary is given of experiments on mechanical strength, heat transfer capability, structural changes, and failure modes under high heat loads of brazed bonds. Tensile-strength tests on the brazing interface prove the suitability of the brazes up to their melting point. The expected enhancement in thermal contact compared with graphite is confirmed. Passively cooled tiles of dimensions 25 mm x 25 mm were subjected to thermal cycling in electron-beam simulations. Heat fluxes of up to 10 MW m -2 were applied. (author)

  16. Conceptual design and related R and D on ITER mechanical based primary pumping system

    International Nuclear Information System (INIS)

    Tanzawa, Sadamitsu; Hiroki, Seiji; Abe, Tetsuya; Shimizu, Katsusuke; Inoue, Masahiko; Watanabe, Mitsunori; Iguchi, Masashi; Sugimoto, Tomoko; Inohara, Takashi; Nakamura, Jun-ichi

    2008-12-01

    The primary vacuum pumping system of the International Thermonuclear Experimental Reactor (ITER) exhausts a helium (He) ash resulting from the DT-burn with excess DT fueling gas, as well as performing a variety of functions such as pump-down, leak testing and wall conditioning. A mechanical based vacuum pumping system has some merits of a continuous pumping, a much lower tritium inventory, a lower operational cost and easy maintenance, comparing with a cryopump system, although demerits of an indispensable magnetic shield and insufficient performance for hydrogen (H 2 ) pumping is are well recognized. To overcome the demerits, we newly fabricated and tested a helical grooved pump (HGP) unit suitable for H 2 pumping at the ITER divertor pressure of 0.1-10 Pa. Through this R and D, we successfully established many design and manufacturing databases of large HGP units for the lightweight gas pumping. Based on the databases, we conceptually designed the ITER vacuum pumping system mainly comprising the HGP with an optimal pump unit layout and a magnetic shield. We also designed conceptually the reduced cost (RC)-ITER pumping system, where a compound molecular pump combining turbine bladed rotors and helical grooved ones was mainly used. The ITER mechanical based primary pumping system proposed has eventually been a back-up solution, whereas a cryopump based one was formally selected to the ITER for construction. The mechanical pumps are increasingly used in many areas with well sophisticated performance, so we believe that fusion reactors of subsequent prototype ones will select the mechanical based pumping system due to primarily a high operational reliability and a cost melt. (author)

  17. A numerical model for the simulation of quench in the ITER magnets

    International Nuclear Information System (INIS)

    Bottura, L.

    1996-01-01

    A computational model describing the initiation and evolution of normal zones in the cable-in-conduit superconductors designed for the international thermonuclear experimental reactor (ITER) is presented. Because of the particular geometry of the ITER cables, the model treats separately the helium momenta in the two cooling channels and the temperatures of the cable constituents. The numerical implementation of the model is discussed in conjunction with the selection of a well-suited solution algorithm. In particular, the solution procedure chosen is based on an implicit upwind finite element technique with adaptive time step and mesh size adjustment possibilities. The time step and mesh adaption procedures are described. Examples of application of the model are also reported. 39 refs., 6 figs., 2 tabs

  18. Experimental utilization of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Bitelli, U. d'Utra; Santos, A. dos; Jerez, R.; Diniz, R.; Fanaro, L.C.C.B.; Abe, A.Y.; Moreira, J.M.L.; Fer, N.; Giada, M.R.; Fuga, R.

    2003-01-01

    This paper aims to show the experimental utilization of the IPEN/MB-01 nuclear reactor during the last fourteen years. The IPEN/MB-01 is a zero-power critical assembly specially designed to measure integral and differential reactor physics parameters to validate calculational methodologies and related nuclear data libraries. Experiments involving determination of spectral indices, critical mass, relative abundance of delayed neutrons, the inversion point of the isothermal reactivity coefficient and burnable poison are considered the most important experiments. Current experiments at IPEN/MB-01 reactor are also commented. (author)

  19. ITER parametric analysis and operational performance

    International Nuclear Information System (INIS)

    Perkins, L.J.; Spears, W.R.; Galambos, J.D.

    1991-01-01

    One of the key components of the ITER Conceptual Design Activities (CDA) is the determination of optimum design, investigation of operation in various modes, recommendation of baseline performance specifications, studies of sensitivity of ITER design to uncertainties in physics, investigation of operational flexibility, assessment of alternative designs, and determination of implications for extrapolation to prospective DEMO reactors. These terms of reference are reported in this document. Refs, figs and tabs

  20. A study of reactor neutrino monitoring at the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Furuta, H.; Fukuda, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Ishitsuka, M.; Ito, C.; Katsumata, M.; Kawasaki, T.; Konno, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Miyata, H.; Nagasaka, Y.; Nitta, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.

    2012-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  1. Experimental measurement of zero power reactor transfer function

    International Nuclear Information System (INIS)

    Liang Shuhong

    2011-01-01

    In order to study the zero power reactor (ZPR) transfer function, the ZPR transfer function expression was deduced with the point reactor kinetics equation, which was disturbed by reactivity input response. Based on the Fourier analysis for the input of triangular wave, the relation between the transfer function and reactivity was got. Validating research experiment was made on the DF-VI fast ZPR. After the disturbed reactivity was measured, the experimental value of the transfer function was got. According to the experimental value and the calculated value, the expression of the ZPR transfer function is proved, whereas the disturbed reactivity is got from the transfer function. (authors)

  2. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

  3. Experimental Simulation of Beryllium Armour Damage Under ITER-like Intense Transient Plasma Loads

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.; Basaleev, E.; Nikolaev, G.; Kurbatova, L., E-mail: igkupr@gmail.com [A.A. Bochvar High Technology Research Institute of Inorganic Material, Moscow (Russian Federation); Podkovyrov, V.; Zhitlukhin, A. [SRC RF TRINITI, Troitsk (Russian Federation); Khimchenko, L. L. [Project Centre of ITER, Moscow (Russian Federation)

    2012-09-15

    Full text: Beryllium will be used as a plasma facing material in the next generation of tokamaks such as ITER. During plasma operation in ITER, the plasma facing materials and components will be suffered by different kinds of loading which may affect their surface or their joint to the heat sink. In addition to quasi-stationary loadings which are caused by the normal cycling operation, the plasma facing components and materials may also be exposed to the intense short transient loads like disruptions, ELMs. All these events may lead to beryllium surface melting, cracking, evaporation and erosion. It is expected that the erosion of beryllium under transient plasma loads such as ELMs and disruptions will mainly determine a lifetime of ITER first wall. To obtain the experimental data for the evaluation of the beryllium armor lifetime and dust production under ITER-relevant transient loads, the advanced plasma gun QSPA-Be facility has been constructed in Bochvar Institute. This paper presents recent results of the experiments with Russian beryllium of TGP-56FW ITER grade. The mock-ups of a special design armored with two beryllium targets (80 x 80 x 10 mm{sup 3}) were tested by hydrogen plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat load of 0.5 and 1.0 MJ/m{sup 2}. Experiments were performed at RT temperature. The evolution of surface microstructure and profile, cracks morphology and mass loss/gain under erosion process on the beryllium surface exposed to up to 250 shots will be presented and discussed. (author)

  4. Combined application of Product Lifecycle and Software Configuration Management systems for ITER remote handling

    International Nuclear Information System (INIS)

    Muhammad, Ali; Esque, Salvador; Aha, Liisa; Mattila, Jouni; Siuko, Mikko; Vilenius, Matti; Jaervenpaeae, Jorma; Irving, Mike; Damiani, Carlo; Semeraro, Luigi

    2009-01-01

    The advantages of Product Lifecycle Management (PLM) systems are widely understood among the industry and hence a PLM system is already in use by International Thermonuclear Experimental Reactor (ITER) Organization (IO). However, with the increasing involvement of software in the development, the role of Software Configuration Management (SCM) systems have become equally important. The SCM systems can be useful to meet the higher demands on Safety Engineering (SE), Quality Assurance (QA), Validation and Verification (V and V) and Requirements Management (RM) of the developed software tools. In an experimental environment, such as ITER, the new remote handling requirements emerge frequently. This means the development of new tools or the modification of existing tools and the development of new remote handling procedures or the modification of existing remote handling procedures. PLM and SCM systems together can be of great advantage in the development and maintenance of such remote handling system. In this paper, we discuss how PLM and SCM systems can be integrated together and play their role during the development and maintenance of ITER remote handling system. We discuss the possibility to investigate such setup at DTP2 (Divertor Test Platform 2), which is the full scale mock-up facility to verify the ITER divertor remote handling and maintenance concepts.

  5. ITER containment design-assist analysis

    International Nuclear Information System (INIS)

    Nguyen, T.H.

    1992-03-01

    In this report, the analysis methods, models and assumptions used to predict the pressure and temperature transients in the ITER containment following a loss of coolant accident are presented. The ITER reactor building is divided into 10 different volumes (zones) based on their functional design. The base model presented in this report will be modified in volume 2 in order to determine the peak pressure, the required size of openings between various functional zones and the differential pressures on walls separating these zones

  6. Robot vision system R and D for ITER blanket remote-handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Tesini, Alessandro

    2014-01-01

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system

  7. Robot vision system R and D for ITER blanket remote-handling system

    Energy Technology Data Exchange (ETDEWEB)

    Maruyama, Takahito, E-mail: maruyama.takahito@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

    2014-10-15

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system.

  8. ITER EDA Newsletter. V.4, no.1

    International Nuclear Information System (INIS)

    1995-01-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on (i) the seventh ITER Council Meeting held at the Naka Joint Work Site on 14-15 December 1994, (ii) the ''Confinement Modelling and Database Expert Group Workshop'' held in Seville, Spain, 3-4 October 1994, and (iii) the first meeting of the International Organizing Committee for the Seventh International Fusion Reactor Materials Conference

  9. The ITER remote maintenance system

    International Nuclear Information System (INIS)

    Tesini, A.; Palmer, J.

    2008-01-01

    The aim of this paper is to summarize the ITER approach to machine components maintenance. A major objective of the ITER project is to demonstrate that a future power producing fusion device can be maintained effectively and offer practical levels of plant availability. During its operational lifetime, many systems of the ITER machine will require maintenance and modification; this can be achieved using remote handling methods. The need for timely, safe and effective remote operations on a machine as complex as ITER and within one of the world's most hostile remote handling environments represents a major challenge at every level of the ITER Project organization, engineering and technology. The basic principles of fusion reactor maintenance are presented. An updated description of the ITER remote maintenance system is provided. This includes the maintenance equipment used inside the vacuum vessel, inside the hot cell and the hot cell itself. The correlation between the functions of the remote handling equipment, of the hot cell and of the radwaste processing system is also described. The paper concludes that ITER has equipped itself with a good platform to tackle the challenges presented by its own maintenance and upgrade needs

  10. CFTSIM-ITER dynamic fuel cycle model

    International Nuclear Information System (INIS)

    Busigin, A.; Gierszewski, P.

    1998-01-01

    Dynamic system models have been developed for specific tritium systems with considerable detail and for integrated fuel cycles with lesser detail (e.g. D. Holland, B. Merrill, Analysis of tritium migration and deposition in fusion reactor systems, Proceedings of the Ninth Symposium Eng. Problems of Fusion Research (1981); M.A. Abdou, E. Vold, C. Gung, M. Youssef, K. Shin, DT fuel self-sufficiency in fusion reactors, Fusion Technol. (1986); G. Spannagel, P. Gierszewski, Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket, Fusion Eng. Des. (1991); W. Kuan, M.A. Abdou, R.S. Willms, Dynamic simulation of a proposed ITER tritium processing system, Fusion Technol. (1995)). In order to provide a tool to understand and optimize the behavior of the ITER fuel cycle, a dynamic fuel cycle model called CFTSIM is under development. The CFTSIM code incorporates more detailed ITER models, specifically for the important isotope separation system, and also has an easier-to-use graphical interface. This paper provides an overview of CFTSIM Version 1.0. The models included are those with significant and varying tritium inventories over a test campaign: fueling, plasma and first wall, pumping, fuel cleanup, isotope separation and storage. An illustration of the results is shown. (orig.)

  11. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  12. Compressive pre-strain in Nb3Sn strand by steel tube and effect on the critical current measured on standard ITER barrel

    NARCIS (Netherlands)

    Nijhuis, Arend; Wessel, Wilhelm A.J.; Knoopers, H.G.; Ilyin, Y.; ten Kate, Herman H.J.; Della Corte, A.

    2005-01-01

    The large Cable-In-Conduit Conductors (CICC) designed for the magnet coils in the International Thermonuclear Experimental Reactor (ITER), are composed of Nb/sub 3/Sn strand bundles in a Stainless Steel (SS) or Incoloy conduit. Both, the thermal contraction of the strand composite and the conduit

  13. Construction schedule management of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Wang Yue

    2012-01-01

    China Experimental Fast Reactor (CEFR) in the first Fast Reactor in China, which is one of large project of the National High Technology Research and Development Program ('863' Program). On 21 st July 2011, CEFR had succeeded to connect to power grid, the target of construction had come true. To a large item, schedule management is one of the most important management, this paper a overall discussion about CEFR item. It has proved that the management of CEFR project is scientific, normative and high-efficiency, it will be valuable for lager Fast Reactor item and designers in interrelated field. (author)

  14. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  15. Research reactor RB, technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1978-01-01

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) [sr

  16. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  17. Experimental confirmation of the ITER cryopump high temperature regeneration scheme

    International Nuclear Information System (INIS)

    Day, C.; Haas, H.

    2007-01-01

    therefore essential for tritium inventory control. In the TIMO test bed at FZK, a half scale pump model of the torus exhaust cryopump with fully ITER relevant cryosorbent coating has been under detailed investigation over the last years, in order to determine the required high temperature regeneration conditions (times, pressures, temperatures). To replicate the ITER conditions most neatly, multi-cycle tests have been performed, aiming to identify any poisoning effects on cryopumping that may arise in the region of high accumulated gas loads of water-likes. Furthermore, the regeneration behaviour of representative water-likes has been investigated by high resolution gas analysis. The regeneration efficiency has been assessed by comparing pumping speeds before and after the contamination of the pump with the high molecular species. This paper summarizes the experimental results and draws conclusions with respect to ITER and the regeneration frequency to be considered for the ITER operational plan. (orig.)

  18. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  19. Concept of magnet systems for LHD-type reactor

    International Nuclear Information System (INIS)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2008-10-01

    Heliotron reactors have attractive features for fusion power plants, such as no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, the major radius of plasma of 14 to 17 m with a central toroidal field of 6 to 4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120 to 140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress can be comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than 1,000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is longer than 150 m that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with small extension of the ITER technology. (author)

  20. 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.

  1. Plasma cleaning of ITER first mirrors

    Science.gov (United States)

    Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

    2017-12-01

    Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

  2. Modeling a nuclear reactor for experimental purposes

    International Nuclear Information System (INIS)

    Berta, V.T.

    1980-01-01

    The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended functions during occurrences of off-normal conditions in a commercial nuclear reactor. Off-normal conditions arising from large and small break loss-of-coolant accidents (LOCA), operational transients, and anticipated transients without scram (ATWS) were to be investigated. This paper describes the LOFT model of the generic PWR and summarizes the experiments that have been conducted in the context of the significant findings involving the complex transient thermal-hydraulics and the consequent effects on the commercial reactor analytical licensing techniques. Through these techniques the validity of the LOFT model as a scaled counterpart of the generic PWR is shown

  3. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  4. ITER SAFETY TASK NID-10A:CANDU occupational exposure experience: ORE for ITER fuel cycle and cooling systems

    International Nuclear Information System (INIS)

    Lee, D.

    1995-02-01

    This report contains information on TRITIUM Occupational Exposure (Internal Dose) from typical CANDU Nuclear Generating Stations. In addition to dose, airborne tritium levels are provided, as these strongly influence operational exposure. The exposure dose data presented in this report cover a period of five years of operation and maintenance experience from four CANDU Reactors and are considered representative of other CANDU reactors. The data are broken down according to occupational function ( Operators, Maintenance and Support Service etc.). The referenced systems are mainly centered on CANDU Hear Transport System, Moderator System, Tritium Removal Facility and Heavy Water (D20) Upgrading System. These systems contain the bulk part of tritium contamination in the CANDU Reactor. Because of certain similarities between ITER and CANDU systems, this data can be used as the most relevant TRITIUM OCCUPATIONAL DOSE information for ITER COOLING and FUEL CYCLE systems dose assessment purpose, if similar design and operation principles as described in the report are adopted. (author). 16 refs., 8 tabs., 13 figs

  5. Experimental results and validation of a method to reconstruct forces on the ITER test blanket modules

    International Nuclear Information System (INIS)

    Zeile, Christian; Maione, Ivan A.

    2015-01-01

    Highlights: • An in operation force measurement system for the ITER EU HCPB TBM has been developed. • The force reconstruction methods are based on strain measurements on the attachment system. • An experimental setup and a corresponding mock-up have been built. • A set of test cases representing ITER relevant excitations has been used for validation. • The influence of modeling errors on the force reconstruction has been investigated. - Abstract: In order to reconstruct forces on the test blanket modules in ITER, two force reconstruction methods, the augmented Kalman filter and a model predictive controller, have been selected and developed to estimate the forces based on strain measurements on the attachment system. A dedicated experimental setup with a corresponding mock-up has been designed and built to validate these methods. A set of test cases has been defined to represent possible excitation of the system. It has been shown that the errors in the estimated forces mainly depend on the accuracy of the identified model used by the algorithms. Furthermore, it has been found that a minimum of 10 strain gauges is necessary to allow for a low error in the reconstructed forces.

  6. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  7. Fusion Technology for ITER, the ITER Project. Further Development Towards a DEMO Fusion Power Plant (3/4)

    CERN Multimedia

    CERN. Geneva

    2011-01-01

    This is the second half of a lecture series on fusion and will concentrate on fusion technology. The early phase of fusion development was concentrated on physics. However, during the 1980s it was realized that if one wanted to enter the area of fusion reactor plasmas, even in an experimental machine, a significant advance in fusion technologies would be needed. After several conceptual studies of reactor class fusion devices in the 1980s the engineering design phase of ITER started in earnest during the 1990s. The design team was in the beginning confronted with many challenges in the fusion technology area as well as in physics for which no readily available solution existed and in a few cases it was thought that solutions may be impossible to find. However, after the initial 3 years of intensive design and R&D work in an international framework utilizing basic fusion technology R&D from the previous decade it became clear that for all problems a conceptual solution could be found and further devel...

  8. Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

    International Nuclear Information System (INIS)

    Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

    1974-01-01

    In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

  9. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    2001-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

  10. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    1999-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L'Air Liquide in the form of a NET contract. (author)

  11. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  12. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  13. Mechanical design and first experimental results of an upgraded technical PERMCAT reactor for tritium recovery in the fuel cycle of a fusion machine

    Energy Technology Data Exchange (ETDEWEB)

    Welte, S., E-mail: stefan.welte@kit.edu [Karlsruhe Institute of Technology (KIT), Forschungszentrum Karlsruhe, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann v. Helmholtz Platz 1, 76344 Eggenstein Leopoldshafen (Germany); Demange, D.; Wagner, R. [Karlsruhe Institute of Technology (KIT), Forschungszentrum Karlsruhe, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann v. Helmholtz Platz 1, 76344 Eggenstein Leopoldshafen (Germany)

    2010-12-15

    The PERMCAT process developed for the final clean-up stage of the Tokamak Exhaust Processing systems of the ITER tritium plant combines a catalytic reactor and a Pd/Ag permeator in a single component. A first generation technical PERMCAT has been successfully operated as part of the CAPER experiment at the Tritium Laboratory Karlsruhe for several years. Various alternative PERMCAT mechanical designs were proposed and studied on small-scale prototypes. An upgraded technical PERMCAT reactor was designed, manufactured and commissioned with deuterium. A parallel arrangement of finger-type membranes inserted in a single catalyst bed design was chosen to simplify the geometry and the manufacturing while improving the robustness of the reactor. The component has been designed and manufactured to be fully tritium compatible and also fully compatible with both process and electrical connections of the previous PERMCAT to be replaced. The new PERMCAT mechanical design is more compact and easy to manufacture. This PERMCAT reactor was submitted to functional tests and experiments based on isotopic exchanges between H{sub 2}O and D{sub 2} to measure the processing performances. The first experimental results show decontamination factors versus flow rates better than all previously measured.

  14. Mechanical design and first experimental results of an upgraded technical PERMCAT reactor for tritium recovery in the fuel cycle of a fusion machine

    International Nuclear Information System (INIS)

    Welte, S.; Demange, D.; Wagner, R.

    2010-01-01

    The PERMCAT process developed for the final clean-up stage of the Tokamak Exhaust Processing systems of the ITER tritium plant combines a catalytic reactor and a Pd/Ag permeator in a single component. A first generation technical PERMCAT has been successfully operated as part of the CAPER experiment at the Tritium Laboratory Karlsruhe for several years. Various alternative PERMCAT mechanical designs were proposed and studied on small-scale prototypes. An upgraded technical PERMCAT reactor was designed, manufactured and commissioned with deuterium. A parallel arrangement of finger-type membranes inserted in a single catalyst bed design was chosen to simplify the geometry and the manufacturing while improving the robustness of the reactor. The component has been designed and manufactured to be fully tritium compatible and also fully compatible with both process and electrical connections of the previous PERMCAT to be replaced. The new PERMCAT mechanical design is more compact and easy to manufacture. This PERMCAT reactor was submitted to functional tests and experiments based on isotopic exchanges between H 2 O and D 2 to measure the processing performances. The first experimental results show decontamination factors versus flow rates better than all previously measured.

  15. The European Fusion Research and Development Programme and the ITER Project

    International Nuclear Information System (INIS)

    Green, B J

    2006-01-01

    The EURATOM fusion research and development programme is a well integrated and coordinated programme. It has the objective of ''developing the technology for a safe, sustainable, environmentally responsible and economically viable energy source.'' The programme is focussed on the magnetic confinement approach and supports 23 Associations which involve research entities (many with experimental and technology facilities) each having a bilateral contractual relationship with the European Commission. The paper will describe fusion reactions and present their potential advantages as an energy source. Further, it will describe the EURATOM programme and how it is organised and implemented. The success of the European programme and that of other national programmes, have provided the basis for the international ITER Project, which is the next logical step in the development of fusion energy. The paper will describe ITER, its aims, its design, and the supporting manufacture of prototype components. The European contribution to ITER, the exploitation of the Joint European Torus (JET), and the long-term reactor technology R and D are carried out under the multilateral European Fusion Development Agreement (EFDA)

  16. Experimental determination of neutron temperature distribution in reactor cell

    International Nuclear Information System (INIS)

    Bosevski, T.

    1965-12-01

    This paper describes theoretical preparation of the experiment for measuring neutron temperature distribution at the RB reactor by activation foils. Due to rather low neutron flux Cu and Lu foil were irradiated for 4 days. Special natural uranium fuel element was prepared to enable easy removal of foils after irradiation. Experimental device was placed in the reactor core at half height in order to measure directly the mean neutron density. Experimental data of neutron temperature distribution for square lattice pitch 16 cm are presented with mean values of neutron temperature in the moderator, in the fuel and on the fuel element surface

  17. Dielectric properties of the ITER TFMC insulation after low temperature reactor irradiation

    International Nuclear Information System (INIS)

    Humer, K.; Weber, H.W.; Hastik, R.; Hauser, H.; Gerstenberg, H.

    2001-01-01

    The insulation system for the Toroidal Field Model Coil of ITER is a fiber reinforced plastic (FRP) laminate, which consists of a combined Kapton/R-glass-fiber reinforcement tape, vacuum-impregnated with an epoxy DGEBA system. Pure disk shaped laminates, disk shaped FRP/stainless-steel sandwiches, and conductor insulation prototypes were irradiated at 5 K in a fission reactor up to a fast neutron fluence of 10 22 m -2 (E>0.1MeV) to investigate the radiation induced degradation of the dielectric strength of the insulation system. After warm-up to room temperature, swelling, weight loss, and the breakdown strength were measured at 77 K. The sandwich swells by 4% at a fluence of 5x10 21 m -2 and by 9% at 1x10 22 m -2 . The weight loss of the FRP is 2% at 1x10 22 m -2 . The dielectric strength remained unchanged over the whole dose range. (author)

  18. Experimental developments towards an ITER thermography diagnostic

    International Nuclear Information System (INIS)

    Reichle, R.; Brichard, B.; Escourbiac, F.; Gardarein, J.L.; Hernandez, D.; Le Niliot, C.; Rigollet, F.; Serra, J.J.; Badie, J.M.; van Ierschot, S.; Jouve, M.; Martinez, S.; Ooms, H.; Pocheau, C.; Rauber, X.; Sans, J.L.; Scheer, E.; Berghmans, F.; Decreton, M.

    2007-01-01

    In the course of the development of a concept for a spectrally resolving thermography diagnostic for the ITER divertor using optical fibres experimental development work has been carried out in three different areas. Firstly ZrF 4 fibres and hollow fibres (silica capillaries with internal AG/AgJ coating) were tested in a Co 60 irradiation facility under γ irradiation up to doses of 5 kGy and 27 kGy, respectively. The ZrF 4 fibres suffered more radiation induced degradation (>1 db/m) then the hollow fibres (0-0.4 db/m). Secondly multi-colour pyroreflectometry is being developed towards tokamak applicability. The emissivity and temperature of tungsten samples were measured in the range of 700-1500 o C. The angular working range for off normal observation of the method was 20-30 o . The working distance of the method has been be increased from cm to the m range. Finally, encouraging preliminary results have been obtained concerning the application of pulsed and modulated active thermography

  19. The integrated design of the ITER magnets and their auxiliary systems

    International Nuclear Information System (INIS)

    Huget, M.

    1999-01-01

    The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. This paper identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field (PF) system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems. (author)

  20. The integrated design of the ITER magnets and their auxiliary systems

    International Nuclear Information System (INIS)

    Huguet, M.

    2001-01-01

    The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. This paper identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field (PF) system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems. (author)

  1. ITER plasma facing components

    International Nuclear Information System (INIS)

    Kuroda, T.; Vieider, G.; Akiba, M.

    1991-01-01

    This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

  2. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

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

  3. Instrumentation and control improvements at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

  4. ITER Building Design (D230-B), Task No. 28. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The International Thermonuclear Experimental Reactor (ITER) Project requires a set of buildings, each with its own distinct function, to support ITER`s mission. The Joint Central Team (JCT) has identified all the buildings in the set and has placed them in an efficient arrangement on the site. The JCT has developed a conceptual layout of each individual building. The buildings have been categorized into two main groups: (1) {open_quotes}Level 1 Buildings{close_quotes} which are on the construction schedule critical path and (2) {open_quotes}Level 2 Buildings{close_quotes} which, while important, are not on the critical path. The buildings are further categorized according to construction material, that is, {open_quotes}reinforced concrete{close_quotes} or {open_quotes}steel-frame on concrete slab{close_quotes}. This Report responds to the Project`s request to perform the initial structural steel design for all the {open_quotes}steel-frame on concrete slab{close_quotes} buildings. Of the twelve (12) {open_quotes}steel-frame on concrete slab{close_quotes} buildings, four (4) are Level 1 and eight (8) are Level 2 Buildings. This Report is a deliverable for the ITER Task Assignment entitled {open_quotes}ITER Buildings Design (D230-B){close_quotes}, also designated as Task No. 28. ITER U.S. Home Team Industrial Consortium members, Raytheon Engineers & Constructors (RE&C) and Stone & Webster Engineering Corporation (SWEC), teamed to perform Task 28. This task commenced in May 1995. It was performed in accordance with the design criteria specified by the ITER-JCT, San Diego Joint Work Site.

  5. Assessment of martensitic steels for advanced fusion reactors

    International Nuclear Information System (INIS)

    Wareing, J.; Tavassoli, A.A.

    1995-01-01

    Martensitic steels are currently considered in Europe to be prime structural candidate materials for the first wall and breeding blanket of the DEMO fusion reactor. In this design, reactor power and wall loading will be significantly higher than those of an experimental reactor. ITER and will give rise to component operating temperatures in the range 250 to 550 0 C with neutron doses higher than 70 dpa. These conditions render austenitic stainless steel, which will be used in ITER, less favourable. Factors contributing to the promotion of martensitic steels are their excellent resistance to irradiation induced swelling, low thermal expansion and high thermal conductivity allied to advanced industrial maturity, compared to other candidate materials vanadium alloys. This paper described the development and optimisation of the steel and weld metal. Using data design rules generated on modified 9 Cr 1 Mo steel during its qualification as a steam generator material for the European Fast Reactor (EFR), interim design guidelines are formulated. Whilst the merits of the steel are validated, it is shown that irradiation embrittlement at low temperature, allied to the need for prolonged post-weld hat treatment and the long term creep response of welds remain areas of some concern. (author). 18 refs., 6 figs., 2 tabs

  6. Research and development needs for ITER engineering design

    International Nuclear Information System (INIS)

    Flanagan, C.; Alikaev, V.; Baker, C.

    1991-01-01

    In the series of documents that summarize the results of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the research and development (R and D) plans for 1991 - 1995. Part A describes the physics R and D, part B the technology R and D. The Physics R and D needs are presented in terms of task descriptions of an ITER-related R and D programme for 1991/1992 and beyond, while diagnostics R and D needs, although covered in Appendix A, are described in Part B. In Chapter II of Part A, ''ITER-related Physics R and D Needs for 91/92 and Beyond'', the following tasks are described as most crucial: (1) demonstration that (i) operation with a cold divertor plasma is possible, (ii) the peak heat flux onto the divertor plate can be kept below about 10 MW per square meter, (iii) and helium exhaust conditions allow a fractional burnup of about 3 percent or more; (2) a characterisation of disruptions that allows to specify their consequences for the plasma-facing-components, and that provides evidence that the number of disruptions expected allows acceptable plasma-facing-component lifetimes; (3) demonstration that steady-state operation in an enhanced-confinement regime and satisfactory plasma purity is possible, and provision of energy confinement scaling allowing the prediction of ITER performance; and (4) ensurance that the presence of a fast ion population does not jeopardize plasma performance in ITER. Part B, ''ITER Technology Research and Development Needs'', describes planning R and D for magnets, containment structure, assembly and maintenance, current drive and heating, plasma facing components, blanket, fuel cycle, structural materials, and diagnostics. A table of key milestones for Technology R and D is included, as well as cost estimates. Figs and tabs

  7. An implicit iterative scheme for solving large systems of linear equations

    International Nuclear Information System (INIS)

    Barry, J.M.; Pollard, J.P.

    1986-12-01

    An implicit iterative scheme for the solution of large systems of linear equations arising from neutron diffusion studies is presented. The method is applied to three-dimensional reactor studies and its performance is compared with alternative iterative approaches

  8. Development of pipe welding, cutting and inspection tools for the ITER blanket

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke

    1999-07-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

  9. Earthly sun called ITER

    International Nuclear Information System (INIS)

    Pozdeyev, Mikhail

    2002-01-01

    Full text: Participating in the film are Academicians Velikhov and Glukhikh, Mr. Filatof, ITER Director from Russia, Mr. Sannikov from Kurchatov Institute. The film tells about the starting point of the project (Mr. Lavrentyev), the pioneers of the project (Academicians Tamme, Sakharov, Artsimovich) and about the situation the project is standing now. Participating in [ITER now are the US, Russia, Japan and the European Union. There are two associated members as well - Kazakhstan and Canada. By now the engineering design phase has been finished. Computer animation used in the video gives us the idea how the first thermonuclear reactor based on famous Russian TOKOMAK works. (author)

  10. International ITER fusion energy organization. Paving the way to power generation from nuclear fusion

    International Nuclear Information System (INIS)

    Preuschen-Liebenstein, R. von

    2006-01-01

    ITER (Latin: the way) is the acronym of a new international large research facility gradually taking shape after the meeting of Gorbachev and Reagan in Reykjavik in 1985. Under the auspices of the IAEA, worldwide scientific and industrial cooperation with 'home teams' of each of the ITER partners began at that time which were commissioned to accumulate the knowledge and the technology of nuclear fusion in the participating countries. At the end of the preparation and decisionmaking process, the design draft of the ITER reactor was elaborated in international cooperation as the basis of the ITER Convention. After lengthy negotiations among the international ITER partners, a European site for the ITER organization and its reactor was found at Cadarache, France. As the first ITER member, Europe now initiated worldwide cooperation in research and development, seeking to demonstrate the technical and scientific feasibility of tapping fusion power for peaceful purposes. The Council of the European Union (competitiveness), at its meeting on September 25, 2006, decided to sign the ITER Convention about the establishment of the International ITER Fusion Energy Organization ('ITER Organization') and about the mutual obligation to make the necessary contributions towards the construction of ITER. (orig.)

  11. The reactor safety study of experimental multi-purpose VHTR design

    International Nuclear Information System (INIS)

    Yasuno, T.; Mitake, S.; Ezaki, M.; Suzuki, K.

    1981-01-01

    Over the past years, the design works of the Experimental Very High Temperature Reactor (VHTR) plant have been conducted at Japan Atomic Energy Research Institute. The conceptual design has been completed and the more detailed design works and the safety analysis of the experimental VHTR plant are continued. The purposes of design studies are to show the feasibility of the experimental VHTR program, to specify the characteristics and functions of the plant components, to point out the R and D items necessary for the experimental VHTR plant construction, and to analyze the feature of the plant safety. In this paper the summary of system design and safety features of the experimental reactor are indicated. Main issues are the safety philosophy for the design basis accident, the accidents assumed and the engineered safety systems adopted in the design works

  12. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  13. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  14. Compilation of reactor-physical data of the AVR experimental reactor for 1982

    International Nuclear Information System (INIS)

    Werner, H.; Wawrzik, U.; Grotkamp, T.; Buettgen, I.

    1983-12-01

    Since the end of 1981 the calculation model AVR-80 has been taken as a basis for compiling reactor-physical data of the AVR experimental reactor. A brief outline of the operation history of 1982 is given, including the beginning of a large-scale experiment dealing with change-over from high enriched uranium to low enriched uranium. Calculations relative to spectral shift, diffusion, temperature, burnup, and recirculation of the fuel elements are described in brief. The essential results of neutron-physical and thermodynamic calculations and the characteristical data of the various types of fuel used are shown in tables and illustrations. (RF) [de

  15. ITER, an artificial sun at the hand? Myths and realities on researches in the field of nuclear fusion

    International Nuclear Information System (INIS)

    2007-01-01

    This document, published by the French association 'Sortir du nucleaire' (Get out of nuclear), gives some information on what is the nuclear fusion, the obstacles to produce nuclear fusion on Earth (why reinventing the sun?), the false promises of the experimental reactor ITER (it will not produce electricity, it will cost 10 billion euros, it will use tritium and will not be safe, it will produce 30 000 tonnes of radioactive wastes), the installation of ITER in Provence (it will cause new damages to the environment, will create jobs at a high cost and to the detriment of clean energies), the vision of ITER as a prestigious project for France with essentially military goals, a research program that is contested by several high profile physicists and is extremely expensive, etc. The paper reviews 20 years of negotiations for the ITER project, and recommends other ways to take up the energy challenge

  16. The SCARABEE experimental fast reactor safety programme already completed

    International Nuclear Information System (INIS)

    Schmitt, A.P.; Teague, H.; Heusener, G.

    1979-08-01

    The SCARABEE in-pile experimental programme comprised a series of tests on unirradiated fuel pins, either single or in seven-pin clusters. The main objective was to obtain information on the mode and consequences of fast reactor fuel pin failure in conditions representative of loss of cooling in a LMFBR. The application of such programmes in full scale reactors leads to the great importance of the interpretation of experimental observations. The interpretation of that programme was carried out jointly by CEA, KFK and UKAEA; this international collaboration led to a sharper focusing on essential features to be modelled in experiments and computer codes and to a valuable convergence of views

  17. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  18. Liquid metal reactor deactivation as applied to the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

    1999-01-01

    The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF

  19. Dynamic Performance of the ITER Reactive Power Compensation System

    International Nuclear Information System (INIS)

    Sheng Zhicai; Fu Peng; Xu Liuwei

    2011-01-01

    Dynamic performance of a reactive power compensation (RPC) system for the international thermonuclear experimental reactor (ITER) power supply is presented. Static var compensators (SVCs) are adopted to mitigate voltage fluctuation and reduce the reactive power down to a level acceptable for the French/European 400 kV grid. A voltage feedback and load power feedforward controller for SVC is proposed, with the feedforward loop intended to guarantee short response time and the feedback loop ensuring good dynamics and steady characteristics of SVC. A mean filter was chosen to measure the control signals to improve the dynamic response. The dynamic performance of the SVC is verified by simulations using PSCAD/EMTDC codes.

  20. ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

    International Nuclear Information System (INIS)

    Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

    2008-01-01

    In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

  1. Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

    CERN Document Server

    Cambi, G; Parozzi, F; Porfiri, MT

    2003-01-01

    A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

  2. Simulation test of PIUS-type reactor with large scale experimental apparatus

    International Nuclear Information System (INIS)

    Tamaki, M.; Tsuji, Y.; Ito, T.; Tasaka, K.; Kukita, Yutaka

    1995-01-01

    A large scale experimental apparatus for simulating the PIUS-type reactor has been constructed keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were performed. Experimental results were compared with those obtained by the small scale apparatus in JAERI. We have already reported the effectiveness of the feedback control for the primary loop pump speed (PI control) for the stable operation. In this paper this feedback system is modified and the PID control is introduced. This new system worked well for the operation of the PIUS-type reactor even in a rapid transient condition. (author)

  3. Summary of the experimental multi-purpose very high temperature gas cooled reactor design

    International Nuclear Information System (INIS)

    1984-12-01

    The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 950 0 C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

  4. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D. [Laboratoire PROMES-CNRS, Laboratoire Procedes, Materiaux et Energie Solaire, 7 rue du Four Solaire, 66120 Font-Romeu Odeillo (France)

    2007-07-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H{sup +} ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  5. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    International Nuclear Information System (INIS)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D.

    2007-01-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H + ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  6. Manufacturing and testing of reference samples for the definition of acceptance criteria for the ITER divertor

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, E.; Libera, S.; Mancini, A.; Pizzuto, A.; Roccella, S.; Riccardi, B.; Escourbiac, F.; Sanguinetti, G.P.

    2010-01-01

    The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

  7. Manufacturing and testing of reference samples for the definition of acceptance criteria for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: visca@frascati.enea.i [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Cacciotti, E.; Libera, S.; Mancini, A.; Pizzuto, A.; Roccella, S. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.e [Fusion For Energy, Barcelona (Spain); Escourbiac, F., E-mail: frederic.escourbiac@iter.or [ITER Organization, Cadarache (France); Sanguinetti, G.P., E-mail: gianpaolo.sanguinetti@aen.ansaldo.i [Ansaldo Energia S.p.A., Genova (Italy)

    2010-12-15

    The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint. An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs. ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R and D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies. According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study. After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects. In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor. This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results.

  8. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    International Nuclear Information System (INIS)

    Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

    2005-01-01

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

  9. Radiation induced currents in mineral-insulated cables and in pick-up coils: model calculations and experimental verification in the BR1 reactor

    Science.gov (United States)

    Vermeeren, Ludo; Leysen, Willem; Brichard, Benoit

    2018-01-01

    Mineral-insulated (MI) cables and Low-Temperature Co-fired Ceramic (LTCC) magnetic pick-up coils are intended to be installed in various position in ITER. The severe ITER nuclear radiation field is expected to lead to induced currents that could perturb diagnostic measurements. In order to assess this problem and to find mitigation strategies models were developed for the calculation of neutron-and gamma-induced currents in MI cables and in LTCC coils. The models are based on calculations with the MCNPX code, combined with a dedicated model for the drift of electrons stopped in the insulator. The gamma induced currents can be easily calculated with a single coupled photon-electron MCNPX calculation. The prompt neutron induced currents requires only a single coupled neutron-photon-electron MCNPX run. The various delayed neutron contributions require a careful analysis of all possibly relevant neutron-induced reaction paths and a combination of different types of MCNPX calculations. The models were applied for a specific twin-core copper MI cable, for one quad-core copper cable and for silver conductor LTCC coils (one with silver ground plates in order to reduce the currents and one without such silver ground plates). Calculations were performed for irradiation conditions (neutron and gamma spectra and fluxes) in relevant positions in ITER and in the Y3 irradiation channel of the BR1 reactor at SCK•CEN, in which an irradiation test of these four test devices was carried out afterwards. We will present the basic elements of the models and show the results of all relevant partial currents (gamma and neutron induced, prompt and various delayed currents) in BR1-Y3 conditions. Experimental data will be shown and analysed in terms of the respective contributions. The tests were performed at reactor powers of 350 kW and 1 MW, leading to thermal neutron fluxes of 1E11 n/cm2s and 3E11 n/cm2s, respectively. The corresponding total radiation induced currents are ranging from

  10. Hydrogen/hydrocarbon explosions in the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Goranson, P.L.

    1992-01-01

    The consequences of H 2 /hydrocarbon detonations in the vacuum vessel (torus) of the International Thermonuclear Experimental Reactor (ITER) have been studied. The most likely scenario for such a detonation involves a water leak into the torus and a vent of the torus to atmosphere, permitting the formation of an explosive fuel-air mixture. The generation of fuel gases and possible sources of air or oxygen are reviewed, and the severity and effects of specific fuel-air mixture explosions are evaluated. Detonation or deflagration of an explosive mixture could result in pressures exceeding the maximum allowable torus pressure. Further studies to examine the design details and develop an event-tree study of events following a gas detonation are recommended

  11. NbTi Strands Verification for ITER PF CICC Process Qualification of CNDA

    Science.gov (United States)

    Liu, F.; Liu, H.; Liu, S.; Liu, B.; Lei, L.; Wu, Y.

    2014-05-01

    China is in charge of most of Poloidal Field (PF) conductors production for the International Thermonuclear Experimental Reactor (ITER). The execution for PF conductors shall be in three main phases. According to ITER Procurement Arrangement (PA), the Domestic Agency (DA) shall be required to verify the room and low temperature acceptance tests carried out by the strand suppliers. As the reference laboratory of Chinese DA (CNDA), the superconducting strands test laboratory of Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) was undertaking the task of strands verification for ITER conductors. The verification test includes: diameter, Nickel plating thickness, copper-to-non-copper volume ratio, twist pitch direction and length, standard critical current (IC) and resistive transition index (n), residual resistance ration (RRR), and hysteresis loss. 48 NbTi strands with 7 billets were supplied for the PF Cable-In-Conduit Conductor (CICC) process qualification. In total, 54 samples were measured. The verification level for PF CICC process qualification was 100%. The test method, facility and results of each item are described in detail in this publication.

  12. Development and performance test results of 50 kA vapor-cooled current lead for ITER

    International Nuclear Information System (INIS)

    Sugimoto, Makoto; Isono, Takaaki; Hamada, Kazuya

    1998-01-01

    JAERI has developed 50 kA current leads for the International Thermonuclear Experimental Reactor (ITER). The Center Solenoid (CS) Model Coil program is under way by means of international collaboration in ITER-EDA. The CS model coil is being developed for ITER CS coil and is for demonstration for ITER construction. The performance test of the CS model coil will be carried out at the JAERI ITER common test facility (CTF). The current lead consists of the vertical lead and the connection lead. The vertical lead is a vapor-cooled type and has a cable-in-conduit geometry. The vertical lead was designed with a heat leak of 1.2 W/kA at a helium flow rate of 0.06 g/(s·kA). The measured heat leak satisfied the designed value. The connection lead was made from copper pipe soldered with NbTi superconducting wires. The pipe was stabilized by the superconducting wires and a cooling channel. The current lead was operated up to 60 kA. The design and the performance test results of 50 kA current leads are reported. The results meet the specifications for the CS model coil test and for the ITER real machine. (author)

  13. Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

    Science.gov (United States)

    Hoogenboom, J. Eduard; Dufek, Jan

    2014-06-01

    This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

  14. Optimized iteration in coupled Monte-Carlo - Thermal-hydraulics calculations

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.; Dufek, J.

    2013-01-01

    This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration methods are also tested and it is concluded that the presented iteration method is near optimal. (authors)

  15. Development of thick wall welding and cutting tools for ITER

    International Nuclear Information System (INIS)

    Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi

    1998-01-01

    The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

  16. Development of thick wall welding and cutting tools for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

  17. Source term evaluation for accident transients in the experimental fusion facility ITER

    Energy Technology Data Exchange (ETDEWEB)

    Virot, F.; Barrachin, M.; Cousin, F. [IRSN, BP3-13115, Saint Paul lez Durance (France)

    2015-03-15

    We have studied the transport and chemical speciation of radio-toxic and toxic species for an event of water ingress in the vacuum vessel of experimental fusion facility ITER with the ASTEC code. In particular our evaluation takes into account an assessed thermodynamic data for the beryllium gaseous species. This study shows that deposited beryllium dusts of atomic Be and Be(OH){sub 2} are formed. It also shows that Be(OT){sub 2} could exist in some conditions in the drain tank. (authors)

  18. Aspects of 238Pu production in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

    2005-01-01

    Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

  19. Few-Group Transport Analysis of the Core-Reflector Problem in Fast Reactor Cores via Equivalent Group Condensation and Local/Global Iteration

    International Nuclear Information System (INIS)

    Won, Jong Hyuck; Cho, Nam Zin

    2011-01-01

    In deterministic neutron transport methods, a process called fine-group to few-group condensation is used to reduce the computational burden. However, recent results on the core-reflector problem in fast reactor cores show that use of a small number of energy groups has limitation to describe neutron flux around core reflector interface. Therefore, researches are still ongoing to overcome this limitation. Recently, the authors proposed I) direct application of equivalently condensed angle-dependent total cross section to discrete ordinates method to overcome the limitation of conventional multi-group approximations, and II) local/global iteration framework in which fine-group discrete ordinates calculation is used in local problems while few-group transport calculation is used in the global problem iteratively. In this paper, an analysis of the core-reflector problem is performed in few-group structure using equivalent angle-dependent total cross section with local/global iteration. Numerical results are obtained under S 12 discrete ordinates-like transport method with scattering cross section up to P1 Legendre expansion

  20. 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)

  1. The JET ITER-like wall experiment: First results and lessons for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.efda.org [EFDA-CSU Culham, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium)

    2013-10-15

    Highlights: ► JET has recently completed the installation of an ITER-like wall. ► Important operational aspects have changed with the new wall. ► Initial experiments have confirmed the expected low fuel retention. ► Disruption dynamics have change dramatically. ► Development of wall-compatible, ITER-relevant regimes of operation has begun. -- Abstract: The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.

  2. 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

  3. ITER Building Design (D230-B), Task No. 28. Final report

    International Nuclear Information System (INIS)

    1995-12-01

    The International Thermonuclear Experimental Reactor (ITER) Project requires a set of buildings, each with its own distinct function, to support ITER's mission. The Joint Central Team (JCT) has identified all the buildings in the set and has placed them in an efficient arrangement on the site. The JCT has developed a conceptual layout of each individual building. The buildings have been categorized into two main groups: (1) open-quotes Level 1 Buildingsclose quotes which are on the construction schedule critical path and (2) open-quotes Level 2 Buildingsclose quotes which, while important, are not on the critical path. The buildings are further categorized according to construction material, that is, open-quotes reinforced concreteclose quotes or open-quotes steel-frame on concrete slabclose quotes. This Report responds to the Project's request to perform the initial structural steel design for all the open-quotes steel-frame on concrete slabclose quotes buildings. Of the twelve (12) open-quotes steel-frame on concrete slabclose quotes buildings, four (4) are Level 1 and eight (8) are Level 2 Buildings. This Report is a deliverable for the ITER Task Assignment entitled open-quotes ITER Buildings Design (D230-B)close quotes, also designated as Task No. 28. ITER U.S. Home Team Industrial Consortium members, Raytheon Engineers ampersand Constructors (RE ampersand C) and Stone ampersand Webster Engineering Corporation (SWEC), teamed to perform Task 28. This task commenced in May 1995. It was performed in accordance with the design criteria specified by the ITER-JCT, San Diego Joint Work Site

  4. 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

  5. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  6. Zero energy reactor RB technical characteristics and experimental possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

  7. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

  8. 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

  9. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  10. Assessment of alkali metal coolants for the ITER blanket

    International Nuclear Information System (INIS)

    Natesan, K.; Reed, C.B.; Mattas, R.F.

    1994-01-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

  11. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

    An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

  12. Advances in iterative methods

    International Nuclear Information System (INIS)

    Beauwens, B.; Arkuszewski, J.; Boryszewicz, M.

    1981-01-01

    Results obtained in the field of linear iterative methods within the Coordinated Research Program on Transport Theory and Advanced Reactor Calculations are summarized. The general convergence theory of linear iterative methods is essentially based on the properties of nonnegative operators on ordered normed spaces. The following aspects of this theory have been improved: new comparison theorems for regular splittings, generalization of the notions of M- and H-matrices, new interpretations of classical convergence theorems for positive-definite operators. The estimation of asymptotic convergence rates was developed with two purposes: the analysis of model problems and the optimization of relaxation parameters. In the framework of factorization iterative methods, model problem analysis is needed to investigate whether the increased computational complexity of higher-order methods does not offset their increased asymptotic convergence rates, as well as to appreciate the effect of standard relaxation techniques (polynomial relaxation). On the other hand, the optimal use of factorization iterative methods requires the development of adequate relaxation techniques and their optimization. The relative performances of a few possibilities have been explored for model problems. Presently, the best results have been obtained with optimal diagonal-Chebyshev relaxation

  13. Neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) MCNP ''Benchmark CAD Model'' with the ATTILA discrete ordinance code

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Feder, R.; Davis, I.

    2007-01-01

    The ITER IT has adopted the newly developed FEM, 3-D, and CAD-based Discrete Ordinates code, ATTILA for the neutronics studies contingent on its success in predicting key neutronics parameters and nuclear field according to the stringent QA requirements set forth by the Management and Quality Program (MQP). ATTILA has the advantage of providing a full flux and response functions mapping everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. The ITER neutronics community had agreed to use a standard CAD model of ITER (40 degree sector, denoted ''Benchmark CAD Model'') to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD-MCNP approach developed by each party. Since ATTILA seems to lend itself as a powerful design tool with minimal turnaround time, it was decided to benchmark this model with ATTILA as well and compare the results to those obtained with the CAD MCNP calculations. In this paper we report such comparison for five responses, namely: (1) Neutron wall load on the surface of the 18 shield blanket module (SBM), (2) Neutron flux and nuclear heating rate in the divertor cassette, (3) nuclear heating rate in the winding pack of the inner leg of the TF coil, (4) Radial flux profile across dummy port plug and shield plug placed in the equatorial port, and (5) Flux at seven point locations situated behind the equatorial port plug. (orig.)

  14. Disruption modeling in support of ITER

    International Nuclear Information System (INIS)

    Bandyopadhyay, I.

    2015-01-01

    Plasma current disruptions and Vertical Displacement Events (VDEs) are one of the major concerns in any tokamak as they lead to large electromagnetic forces to tokamak first wall components and vacuum vessel. Their occurrence also means disruption to steady state operations of tokamaks. Thus future fusion reactors like ITER must ensure that disruptions and VDEs are minimized. However, since there is still finite probability of their occurrence, one must be able to characterize disruptions and VDEs and able to predict, for example, the plasma current quench time and halo current amplitude, which mainly determine the magnitude of the electromagnetic forces. There is a concerted effort globally to understand and predict plasma and halo current evolution during disruption in tokamaks through MHD simulations. Even though Disruption and VDEs are often 3D MHD perturbations in nature, presently they are mostly simulated using 2D axisymmetric MHD codes like the Tokamak Simulation Code (TSC) and DINA. These codes are also extensively benchmarked against experimental data in present day tokamaks to improve these models and their ability to predict these events in ITER. More detailed 3D models like M3D are only recently being developed, but they are yet to be benchmarked against experiments, as also they are massively computationally exhaustive

  15. Summary report for ITER Task -- D4: Activation calculations for the stainless steel ITER design

    International Nuclear Information System (INIS)

    Attaya, H.

    1995-02-01

    Detailed activation analysis for ITER has been performed as a part of ITER Task D4. The calculations have been performed for the shielding blanket (SS/water) and for the breeding blanket (LiN) options. The activation code RACC-P, which has been modified under IFER Task-D-10 for pulsed operation, has been used in this analysis. The spatial distributions of the radioactive inventory, decay heat, biological hazard potential, and the contact dose were calculated for the two designs for different operation modes and targeted fluences. A one-dimensional toroidal geometrical model has been utilized to determine the neutron fluxes in the two designs. The results are normalized for an inboard and outboard neutron wall loadings of 0.91 and 1.2 MW/M 2 , respectively. The point-wise distributions of the decay gamma sources have been calculated everywhere in the reactor at several times after the shutdown of the two designs and are then used in the transport code ONEDANT to calculate the biological dose everywhere in the reactor. The point-wise distributions of all the responses have also been calculated. These calculations have been performed for neutron fluences of 3.0 MWa/M 2 , which corresponds to the target fluence of ITER, and 0.1 MWa/M 2 , which is anticipated to correspond to the beginning of an extended maintenance period

  16. Simulation of LLCB TBM in-vessel first wall coolant break into ITER vacuum vessel by using RELAP/MOD3.4

    International Nuclear Information System (INIS)

    Tony Sandeep, K.; Chaudhari, Vilas; Rajendra Kumar, E.; Dutta, Anu; Singh, R.K.

    2013-06-01

    To prove Test Blanket Module (TBM) safety in International Thermonuclear Experimental Reactor (ITER), various accident scenarios are postulated . One of the postulated initiating events to be analysed is TBM First wall (FW) coolant leak in ITER Vacuum vessel (VV). This accident has been classified as a reference event for the TBM (probability of occurrence >1 E -06 /a). The postulated accident occurs as a result of small leak of TBM FW helium into ITER vacuum vessel (VV), caused by the TBM weld failure. The ingress of this TBM FW helium into ITER plasma induces intense plasma disruption that deposits 1.8 MJ/m 2 of plasma stored thermal energy onto the TBM FW over a period of 1 sec in duration (assumption). Runaway electrons in this process are lost from plasma current channel and cause multiple TBM and ITER FW cooling tube failures within 10 cm torriodal strip. The size of the break is identified as double ended rupture of all coolant channels within this strip around the reactor. For LLCB TBM this represents failure of 4 FW channels. The size of ITER FW break is 0.02 m 2 . Consequently, a simultaneous blow down of TBM FW helium and ITER FW water occurs, injecting helium and water into VV. This pressurisation causes the activation of VV pressure suppressions system and ingress of water into VV. This pressurisation causes the VV pressure suppressions system (VVPSS) to open in an attempt to contain the pressure below the safety limit of 0.2 MPa. This report is intended to do the above accident analysis and assessment of active components of TBM using RELAP code and hence prove its safety in ITER environment. (author)

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

    International Nuclear Information System (INIS)

    Tachikawa, K.; Iida, H.; Nishio, S.; Tone, T.; Aota, T.; Iwamoto, T.; Niikura, S.; Nishizawa, H.

    1984-01-01

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

  18. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P; Ribeiro, M I; Aparicio, P [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  19. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

    Lima, P.; Ribeiro, M.I.; Aparicio, P.

    1998-01-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  20. Experimental investigation of the MSFR molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2014-11-15

    In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

  1. Thermal stress analysis of gravity support system for ITER based on ANSYS

    International Nuclear Information System (INIS)

    Liang Shangming; Yan Xijiang; Huang Yufeng; Wang Xianzhou; Hou Binglin; Li Pengyuan; Jian Guangde; Liu Dequan; Zhou Caipin

    2009-01-01

    A method for building the finite element model of the gravity support system for International Thermonuclear Experimental Reactor (ITER) was proposed according to the characteristics of the gravity support system with the cyclic symmetry. A mesh dividing method, which has high precision and an acceptable calculating scale, was used, and a three dimensional finite element model for the toroidal 20 degree sector of the gravity support system was built by using ANSYS. Meantime, the steady-state thermal analysis and thermal-structural coupling analysis of the gravity support system were performed. The thermal stress distributions and the maximal thermal stress values of all parts of the gravity support system were obtained, and the stress intensity of parts of the gravity support system was analyzed. The results of thermal stress analysis lay the solid foundation for design and improvement for gravity supports system for ITER. (authors)

  2. Assessment of database for interaction of tritium with ITER plasma facing materials

    International Nuclear Information System (INIS)

    Dolan, T.J.; Anderl, R.A.

    1994-09-01

    The present work surveys recent literature on hydrogen isotope interactions with Be, SS and Inconels, Cu, C, and V, and alloys of Cu and V. The goals are (1) to provide input to the International Thermonuclear Experimental Reactor (ITER) team to help with tritium source term estimates for the Early Safety and Environmental Characterization Study and (2) to provide guidance for planning additional research that will be needed to fill gaps in the present materials database. Properties of diffusivity, solubility, permeability, chemical reactions, Soret effect, recombination coefficient, surface effects, trapping, porosity, layered structures, interfaces, and oxides are considered. Various materials data are tabulated, and a matrix display shows an assessment of the quality of the data available for each main property of each material. Recommendations are made for interim values of diffusivity and solubility to be used, pending further discussion by the ITER community

  3. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  4. Primary system thermal-hydraulic simulation of a experimental pool type research fast reactor

    International Nuclear Information System (INIS)

    Borges, E.M.; Braz Filho, F.A.

    1993-01-01

    The first step of the Fast Reactor Program (REARA) is the design of an experimental reactor. To this end a 5 MW t pool type reactor was adapted. The objective of this work is to evaluate the reactor behaviour at the on set protected accidents. The program NALAP was used in this study and the results showed the outstanding safety margins that this reactor type presents inherently. (author)

  5. Primary design and operation analysis of ITER air transfer system

    International Nuclear Information System (INIS)

    Wang Haitian; Li Ge; Qin Shijun

    2010-01-01

    Air transfer system (ATS) is a remote handling transfer, which can work in the nuclear radiation environment and can be driven by the electricity fully. Its motion power is provided by several servo motors. The remote control technology of ATS, which is China taking part in the plan of international Tokamak experimental reactor (ITER) and grasping this technology, is one of key technologies of ITER. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Because there is gamma irradiation and hazard material in these ITER parts, all required maintenance of port plugs and inner components are been transmitted by ATS. The pick-up or drop-off these components are completed by means of a remotely controlled TCS system between the Vacuum Vessel and the Hot Cell through the bridge-gallery. Tokamak building includes three floors, including upper port, equatorial port and lower port, linked by a lift. According to each port level configuration and safety requirement, the radius of curvature with ATS trajectory is optimized, and a trajectory of each level is determined by positioned guidance beacons. At last, the results of computer aided design (CAD) show single trajectory guidance of ATS in each level is available. (authors)

  6. Potential for use of high-temperature superconductors in fusion reactors

    International Nuclear Information System (INIS)

    Hull, J.R.

    1991-01-01

    The present rate of development of high-temperature superconductors (HTSs) is sufficiently rapid that there may be opportunities for their use in contemporary fusion devices such as the International Thermonuclear Experimental Reactor (ITER). The most likely 1application is for delivering power to the superconducting magnets, especially in substituting for the current leads between the temperatures of 4 K and 77K. A second possible application of HTSs is as a liquid-nitrogen-cooled power bus, connecting the power supplies to the magnets, thus reducing the ohmic heating losses over these relatively long cables. A third potential application of HTSs is as an inner high-field winding of the toroidal field coils that would operate at ∼20 K. While the use of higher temperature magnets offers significant advantages to the reactor system, it is unlikely that tested conductors of this type will be available within the ITER time frame. 23 refs., 2 figs

  7. Latest status of manufacturing activity of ITER divertor and engineering issues on tungsten divertor

    International Nuclear Information System (INIS)

    Suzuki, Satoshi

    2011-01-01

    Divertors for ITER are now in construction. In the present chapter, the specification and the latest status of manufacturing of ITER divertors are presented. In addition, issues in the development of divertors for the fusion demo reactor are given on the basis of experiences on the ITER divertor development. (J.P.N.)

  8. Reducing the effects of acoustic heterogeneity with an iterative reconstruction method from experimental data in microwave induced thermoacoustic tomography

    International Nuclear Information System (INIS)

    Wang, Jinguo; Zhao, Zhiqin; Song, Jian; Chen, Guoping; Nie, Zaiping; Liu, Qing-Huo

    2015-01-01

    Purpose: An iterative reconstruction method has been previously reported by the authors of this paper. However, the iterative reconstruction method was demonstrated by solely using the numerical simulations. It is essential to apply the iterative reconstruction method to practice conditions. The objective of this work is to validate the capability of the iterative reconstruction method for reducing the effects of acoustic heterogeneity with the experimental data in microwave induced thermoacoustic tomography. Methods: Most existing reconstruction methods need to combine the ultrasonic measurement technology to quantitatively measure the velocity distribution of heterogeneity, which increases the system complexity. Different to existing reconstruction methods, the iterative reconstruction method combines time reversal mirror technique, fast marching method, and simultaneous algebraic reconstruction technique to iteratively estimate the velocity distribution of heterogeneous tissue by solely using the measured data. Then, the estimated velocity distribution is used subsequently to reconstruct the highly accurate image of microwave absorption distribution. Experiments that a target placed in an acoustic heterogeneous environment are performed to validate the iterative reconstruction method. Results: By using the estimated velocity distribution, the target in an acoustic heterogeneous environment can be reconstructed with better shape and higher image contrast than targets that are reconstructed with a homogeneous velocity distribution. Conclusions: The distortions caused by the acoustic heterogeneity can be efficiently corrected by utilizing the velocity distribution estimated by the iterative reconstruction method. The advantage of the iterative reconstruction method over the existing correction methods is that it is successful in improving the quality of the image of microwave absorption distribution without increasing the system complexity

  9. 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)

  10. Study of a compact reversed shear Tokamak reactor

    International Nuclear Information System (INIS)

    Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

    1998-01-01

    A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

  11. Dust removal system for fusion experimental reactors

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. Dust removal system for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  13. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

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

    International Nuclear Information System (INIS)

    1986-11-01

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

  15. Development of laser cutting/welding system for remote maintenance of ITER manifold

    Energy Technology Data Exchange (ETDEWEB)

    Yamaoka, Hiroto; Tsuchiya, Kazuyuki; Awano, Toshihiko [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan); Oka, Kiyoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-09-01

    A remote pipe cutting/welding system using a YAG laser was designed and fabricated for the maintenance of the main structural parts of ITER (International Thermonuclear Experimental Reactor), and a mock-up test carried out. The functions of this system are to cut 100A x Sch 40 pipes of SUS316L by internal access, to adjust the core gap between the as-cut pipe and new pipe, and to weld the pipes automatically. The core gap of the pipes could be decreased within the proper welding conditions by the mock-up test, and sound beads were obtained. (author)

  16. WWER reactor fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Stefanova, S.; Chantoin, P.; Kolev, I.

    1994-01-01

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

  17. WWER reactor fuel performance, modelling and experimental support. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Chantoin, P; Kolev, I [eds.

    1994-12-31

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

  18. Design issues and cost implications of RTO/RC-ITER divertor

    International Nuclear Information System (INIS)

    Ibbott, C.; Antipenkov, A.; Chiocchio, S.; Federici, G.; Heidl, H.; Janeschitz, G.; Martin, E.; Tivey, R.

    2000-01-01

    This paper reports on the conceptual divertor design developed for the reduced technical objectives/reduced cost-international thermonuclear experimental reactor (RTO/RC-ITER). The cost drivers are discussed and a number of cost-reducing measures identified. Scaled costs, based on industrial estimates of the 1998 ITER design (Technical Basis for the ITER Final Design Report, Cost Review and Safety Analysis (FDR). ITER EDA Documentation Series No. 16. IAEA, Vienna, 1998), give for the RTO/RC-ITER ∼60% of the FDR costs. Plasma facing components (PFCs) account for 75% of the total divertor costs. Hence, PFC design simplifications are outlined in the paper showing the possibility of achieving a cost reduction of 50%. The design proposals, outlined in the paper, focus on minimising the number of sub-components and simplifying the manufacturing cycle. These changes contribute to improved reliability based on a more robust coolant design layout. The reduced space allocated to the divertor (G. Janeschitz, A. Antipenkov, V. Barabash, S. Chiocchio, G. Federici, C. Ibbott, E. Martin, R. Tivey, Overview of the Divertor Design and its Integration into RTO/RC-ITER, this conference) requires changes to the design that minimise the cassette body thickness, relocate the cassette attachments and revise the remote handling philosophy. Results of supporting electro-magnetic, neutron shielding, thermo-hydraulic and pumping conductance analyses are reported, qualifying the cassette design. A reduction in the coolant inlet temperature to 100-120 deg. C is discussed in terms of thermal-hydraulic performance and fatigue life of the heat sink. Finally, an R and D plan sets out the work needed: (1) to develop the cost saving measures of the new design; and (2) to demonstrate the reliability of the chosen technologies

  19. Preparation of the ITER Poloidal Field Conductor Insert (PFCI) test

    International Nuclear Information System (INIS)

    Zanino, R.; Egorov, S.; Kim, K.; Martovetsky, N.; Nunoya, Y.; Okuno, K.; Salpietro, E.; Sborchia, C.; Takahashi, Y.; Weng, P.; Bangasco, M.; Savoldi Richard, L.; Polak, M.; Formisano, A.; Zapretilina, E.; Shikov, A.; Vedernikov, G.; Ciazynski, D.; Zani, L.; Muzzi, L.; Ricci, M.; Deela Corte, A.; Sugimoto, M.; Hamada, K.; Portone, A.; Hurd, F.; Mitchell, N.; Nijhuis, A.; Ilyin, Y.

    2004-01-01

    The Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in Europe and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, about 45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of about 6 T, and it will be cooled by supercritical He at around 4.5 K and 0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure taps, strain and displacement sensors. The test program shall be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermalhydraulic characterization. Here we give an overview of the preparatory work towards the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far. (authors)

  20. 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)

  1. Experimental and analytical studies of high heat flux components for fusion experimental reactor

    International Nuclear Information System (INIS)

    Araki, Masanori

    1993-03-01

    In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 ± 1 MW/m 2 was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate has been analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads. (J.P.N.) 62 refs

  2. Advanced scenarios for ITER operation

    Energy Technology Data Exchange (ETDEWEB)

    Sips, A.C.C. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2004-07-01

    In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of {alpha}-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in

  3. Application of Iterative Robust Model-based Optimal Experimental Design for the Calibration of Biocatalytic Models

    DEFF Research Database (Denmark)

    Van Daele, Timothy; Gernaey, Krist V.; Ringborg, Rolf Hoffmeyer

    2017-01-01

    The aim of model calibration is to estimate unique parameter values from available experimental data, here applied to a biocatalytic process. The traditional approach of first gathering data followed by performing a model calibration is inefficient, since the information gathered during...... experimentation is not actively used to optimise the experimental design. By applying an iterative robust model-based optimal experimental design, the limited amount of data collected is used to design additional informative experiments. The algorithm is used here to calibrate the initial reaction rate of an ω......-transaminase catalysed reaction in a more accurate way. The parameter confidence region estimated from the Fisher Information Matrix is compared with the likelihood confidence region, which is a more accurate, but also a computationally more expensive method. As a result, an important deviation between both approaches...

  4. Experimental analysis of flowrates distribution features in double-loop reactor channels

    International Nuclear Information System (INIS)

    Avdeev, E.F.; Chusov, I.A.

    2013-01-01

    Experimental data on the flowrate distribution in working channels dummies of a research reactor model with double-loop configuration are presented in the paper. The procedures of experiments and received experimental data processing are provided in details [ru

  5. Assembly of the sectors and ports of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Corino, S.; Moreno, R.

    2014-01-01

    The International Thermonuclear Experimental Reactor, ITER is a very complex Project that aims to prove the technical reliability of nuclear fusion. ITER has been Ensa's commitment to the future to strengthen as one of the main manufacturers of big equipment and services internationally in the nuclear field. Ensa started working on the qualification process to be able to bid for the 'Assembly of the ITER vacuum vessel' in June 2010, after two and a half years of pre-qualification, offers, clarifications and long technical meetings, that were followed by commercial meetings Ensa achieved its goal. The 30 of November, Ensa signed what at that time was the biggest of the supplies signed by IO (ITER Organization). A lot of efforts and hard work had been done in order to achieve this goal, but the hardest of all was yet to come, after the signature of the contract, Ensa has 7 years ahead to achieve the final goal, the assembly and welding of the 9 sectors that put together the ITER vacuum vessel and the 54 ports that will allow the assembly of the different auxiliary systems. The scope of the works to be performed can generally be divided into the following areas: - Welding of the sectors and ports; - Non-destructive tests; - Machining; - Dimensional Controls. In order to achieve this goal, the project has been divided into 3 different phases. - Development phase: January 2013 - July 2015; - Pre-production phase: July 2015 - February 2016; - Production phase: February 2016 - February 2020

  6. Studies of the ultrasonic testing scheme on bonding quality in shield blanket of ITER

    International Nuclear Information System (INIS)

    Shi Sichao; Shen Jingling; He Fengqi; Jin Wanping

    2007-01-01

    International Thermonuclear Experimental Reactor (ITER) is an international cooperative item. One of its components, the First Wall (FW) functioning as neutron shielding and cooling, is an important part. According to the component materials, structural features, testing requirements of the FW, and the ultrasonic propagation characteristics, it is suggested that Broad-band ultrasonic can be used to test the bonding quality of the FW. According to the case mentioned above, the Broad-band Ultrasonic Testing scheme was presented, and the ultrasonic testing feasibility was analyzed theoretically in this paper. (authors)

  7. Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1999-01-01

    The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

  8. General description of preliminary design of an experimental fusion reactor and the future problems

    International Nuclear Information System (INIS)

    Sako, Kiyoshi

    1976-01-01

    Recently, the studies on plasma physics has progressed rapidly, and promising experimental data emerged successively. Especially expectation mounts high that Tokamak will develop into power reactors. In Japan, the construction of large plasma devices such as JT-60 of JAERI is going to start, and after several years, the studies on plasma physics will come to the end of first stage, then the main research and development will be directed to power reactors. The studies on the design of practical fusion reactors have been in progress since 1973 in JAERI, and the preliminary design is being carried out. The purposes of the preliminary design are the clarification of the concept of the experimental reactor and the requirements for the studies on core plasma, the examination of the problems for developing main components and systems of the reactor, and the development of design technology. The experimental reactor is the quasi-steady reactor of 100 MW fusion reaction output, and the conditions set for the design and the basis of their setting are explained. The outline of the design, namely core plasma, blankets, superconductive magnets and the shielding with them, vacuum wall, neutral particle injection heating device, core fuel supply and exhaust system, and others, is described. In case of scale-up the reactor structural material which can withstand neutron damage must be developed. (Kako, I.)

  9. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  10. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

  11. Parallel computation of multigroup reactivity coefficient using iterative method

    Science.gov (United States)

    Susmikanti, Mike; Dewayatna, Winter

    2013-09-01

    One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

  12. Comparative study of low and high aspect ratio devices for ITER design options

    International Nuclear Information System (INIS)

    Sugihara, Masayoshi; Tada, Eisuke; Shimomura, Yasuo; Tsunematsu, Toshihide; Nishio, Satoshi; Nakazato, Toshiko; Murakami, Yoshiki; Koizumi, Koichi

    1992-09-01

    Comparative study on the plasma performance and the engineering characteristics of low and high aspect ratio devices for ITER (International Thermonuclear Experimental Reactor) design option is done to examine quantitatively the expected merit and demerit of high aspect ratio device on steady state operation. Device parameters of aspect ratio A=3 and 4 are chosen based on ITER-power scaling law. Improvement of steady state operation with A=4 is found only moderate. Reduction of stability margin in vertical instability is about 20% and plasma elongation must be decreased from 2 down to about 1.8 to recover this reduction of stability margin with A=4. If such lower elongation is employed, single null divertor configuration should be employed to reduce the capacity of poloidal field system. Detailed 2D divertor code calculation shows that peak heat load per unit area of A=4 device with SN configuration increases compared with A=3 device with DN configuration, contrary to the predictions so far made. Preliminary engineering studies indicate that A=4 device would have less space for handling the in-vessel components and doubled toroidal field magnet weight and winding length, and hence is less desirable when compared with the present ITER design (A=3). Based on these examinations, it is concluded that high aspect ratio device does not have remarkable advantage than low aspect ratio device, and the latter device has similar capability for the prospect of future commercial reactor to the former device. (J.P.N.)

  13. ITER JCT presentation at the International Conference on Fusion Reactor Materials (ICFRM-9)

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.; Ioki, K.

    1999-01-01

    During this conference four invited papers and one poster paper were presented on behalf of the ITER Joint Central Team with the review of latest achievements. The results of the comprehensive materials R and D program in support of the ITER design were extensively reported the ITER Home Teams

  14. Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

    A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

  15. A Conceptual Design and Structural Analysis for ITER Mid-plane Brace Tools

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Jae Hyuk; Kim, Kyung kyu [SFA Engineering Corp., Changwon (Korea, Republic of); Im, Kihak; Robert, Shaw [ITER Organization, St Paul lez Durance Cedex (France)

    2010-10-15

    The ITER, International Thermonuclear Experimental Reactor, Tokamak machine is mainly composed of 9 vacuum vessel (VV)/toroidal field coils (TFCs)/vacuum vessel thermal shields (VVTS) 40 .deg. sectors. Each VV/TFCs/VVTS 40 .deg. sector is made up of one 40 .deg. VV, two 20 .deg. TFCs and associated VVTS segments. The ITER Tokamak assembly tools are purpose-built tools to assemble the ITER Tokamak machine which includes the cryostat and the components contained therein. Based on the design description document prepared by the IO (ITER international organization), Korea has carried out the conceptual design of assembly tools with IO cooperation. The 40 .deg. sector assemblies attached mid-plane brace tools sub-assembled at assembly hall are transferred to Tokamak hall using the lifting tool operated by Tokamak main cranes. The sector sub-assembly tools are composed of the upending tool, the sector sub-assembly tool, the sector lifting tool and the vacuum vessel support and mid-plane brace tools. The mid-plane brace tool is assembled to inner surface of VV and TFCs in phase of sector sub-assembly after completion of all sector components. VV, TFC and VVTS are separated fully before completion of 9 sectors at Tokamak in-pit. In this paper the mid-plane brace tools is introduced about function, structure and status of research and development are also described

  16. ITER: Promises unkept ? (1/2)

    CERN Multimedia

    CERN. Geneva

    2008-01-01

    Fusion power as the source of energy on Earth has been the dream of mankind ever since the principles were understood. ITER, the Latin word for “the way”, is the world’s largest Fusion device presently under construction in Cadarache, France. Supported by the People’s Republic of China, the European Atomic Energy Community, India, Japan, the Republic of Korea, the Russian Federation, and the United States of America, an international organization was founded after the signature of the Joint ITER Agreement in October of 2006. The goal is to build a Fusion reactor with a power amplification of 10, a total fusion power of 500 MW or more operating at extended burn times of 400-3000 seconds, with Deuterium and Tritium as its basic fuel. Following a short introduction into fusion science principles, the history of thermo nuclear fusion will be covered. Finally more recent construction projects around the world, their latest achievements and the path to ITER will be described. Technological and scientific c...

  17. ITER: Promises unkept ? (2/2)

    CERN Multimedia

    CERN. Geneva

    2008-01-01

    Fusion power as the source of energy on Earth has been the dream of mankind ever since the principles were understood. ITER, the Latin word for “the way”, is the world’s largest Fusion device presently under construction in Cadarache, France. Supported by the People’s Republic of China, the European Atomic Energy Community, India, Japan, the Republic of Korea, the Russian Federation, and the United States of America, an international organization was founded after the signature of the Joint ITER Agreement in October of 2006. The goal is to build a Fusion reactor with a power amplification of 10, a total fusion power of 500 MW or more operating at extended burn times of 400-3000 seconds, with Deuterium and Tritium as its basic fuel. Following a short introduction into fusion science principles, the history of thermo nuclear fusion will be covered. Finally more recent construction projects around the world, their latest achievements and the path to ITER will be described. Technological and scientific c...

  18. Mechanical and Electrical Modeling of Strands in Two ITER CS Cable Designs

    CERN Document Server

    Torre, A; Ciazynski, D

    2014-01-01

    Following the test of the first Central Solenoid (CS) conductor short samples for the International Thermonuclear Experimental Reactor (ITER) in the SULTAN facility, Iter Organization (IO) decided to manufacture and test two alternate samples using four different cable designs. These samples, while using the same Nb$_{3}$Sn strand, were meant to assess the influence of various cable design parameters on the conductor performance and behavior under mechanical cycling. In particular, the second of these samples, CSIO2, aimed at comparing designs with modified cabling twist pitches sequences. This sample has been tested, and the two legs exhibited very different behaviors. To help understand what could lead to such a difference, these two cables were mechanically modeled using the MULTIFIL code, and the resulting strain map was used as an input into the CEA electrical code CARMEN. This article presents the main data extracted from the mechanical simulation and its use into the electrical modeling of individual s...

  19. Inverse kinetics technique for reactor shutdown measurement: an experimental assessment. [AGR

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, T. A.; McDonald, D.

    1975-09-15

    It is proposed to use the Inverse Kinetics Technique to measure the subcritical reactivity as a function of time during the testing of the nitrogen injection systems on AGRs. A description is given of an experimental assessment of the technique by investigating known transients created by control rod movements on a small experimental reactor, (2m high, 1m radius). Spatial effects were observed close to the moving rods but otherwise derived reactivities were independent of detector position and agreed well with the existing calibrations. This prompted the suggestion that data from installed reactor instrumentation could be used to calibrate CAGR control rods.

  20. Radioactive waste produced by DEMO and commerical fusion reactors extrapolated from ITER and advanced data bases

    International Nuclear Information System (INIS)

    Stacey, W.M.; Hertel, N.E.; Hoffman, E.A.

    1994-01-01

    The potential for providing energy with minimal environmental impact is a powerful motivation for the development of fusion and is the long-term objective of most fusion programs. However, the societal acceptability of magnetic fusion may well be decided in the near-term when decisions are taken on the construction of DEMO to follow ITER (if not when the construction decision is taken on ITER). Component wastes were calculated for DEMOs based on each data base by first calculating reactor sizes needed to satisfy the physics, stress and radiation attenuation requirements, and then calculating component replacement rates based on radiation damage and erosion limits. Then, radioactive inventories were calculated and compared to a number of international criteria for open-quote near-surface close-quote burial. None of the components in either type of design would meet the Japanese LLW criterion ( 3 ) within 10 years of shutdown, although the advanced (V/Li) blanket would do so soon afterwards. The vanadium first wall, divertor and blanket would satisfy the IAEA LLW criterion (<2 mSv/h contact dose) within about 10 years after shutdown, but none of the stainless steel or copper components would. All the components in the advanced data base designs except the stainless steel vacuum vessel and shield readily satisfy the US extended 10CFR61 intruder dose criterion, but none of the components in the open-quotes ITER data baseclose quotes designs do so. It seems unlikely that a stainless steel first wall or a copper divertor plate could satisfy the US (class C) criterion for near surface burial, much less the more stringent international, criteria. On the other hand, the first wall, divertor and blanket of the V/Li system would still satisfy the intruder dose concentration limits even if the dose criterion was reduced by two orders of magnitude

  1. Development of liquid lithium divertor for fusion reactor

    International Nuclear Information System (INIS)

    Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

    2000-01-01

    Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

  2. Oscillation experiments techniques in CEA Minerve experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Antony, M.; Di-Salvo, J.; Pepino, A.; Bosq, J. C.; Bernard, D.; Leconte, P.; Hudelot, J. P.; Lyoussi, A. [CEA CADARACHE, DEN/DER/SPEx, 13108 Saint Paul-lez-Durance (France)

    2009-07-01

    This paper deals with experiments in the Minerve pool Zero Power Reactor. Minerve is mainly devoted to neutronics studies, in view to improve the calculation routes by reducing the uncertainties of the experimental databases for nuclides arising in plutonium and wastes management. Minerve experimental measurement programs are performed by using the oscillation technique. This experimental technique consists in a periodic insertion and extraction of samples containing the nuclide of interest in a well characterized neutron spectrum. The reactivity variation of the sample is compensated by a calibrated rotary automatic pilot using cadmium sectors. The normal accuracy for measurements of small-worth samples in Minerve by using such a technique is about 3% for absolute reactivity worth, including the uncertainties on the material balance and on the calibration step. Reactivity effects of less than 1.5 cent can be measured. The OSMOSE and the OCEAN programs have been carried out since 2005 and will last until 2011. These programs aim at improving, in different neutron spectra, the absorption cross sections of respectively a majority of the separated heavy nuclides from {sup 232}Th to {sup 245}Cm appearing during the reactor and the fuel cycle physics, and of current and future types of absorbers as Gd, Hf, Er, Dy and Eu. (authors)

  3. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  4. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  5. Primary design and operation analysis of the ITER air transfer system

    International Nuclear Information System (INIS)

    Wang Haitian; Li Ge; Qin Shijun

    2010-01-01

    Air transfer system (ATS) is a remote handling transfer, which can work in the nuclear radiation environment and can be driven by the electricity fully. Its motion power is provided by several servo motors. The remote control technology of ATS, which is China taking part in the plan of international Tokamak experimental reactor (ITER) and grasping this technology, is one of key technologies of ITER. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Because there is gamma irradiation and hazard material in these ITER parts, all required maintenance of port plugs and inner components are been transmitted by ATS. The pick-up or drop-off these components are completed by means of a remotely controlled TCS system between the Vacuum Vessel and the Hot Cell through the bridge-gallery. Tokamak building includes three floors, including upper port, equatorial port and lower port, linked by a lift. According to each port level configuration and safety requirement, the radius of curvature with ATS trajectory is optimized, and a trajectory of each level is determined by positioned guidance beacons. At last, the results of computer aided design (CAD) show single trajectory guidance of ATS in each level is available. (authors)

  6. Experimental Methods Related to Coupled Fast-Thermal Systems at the RB Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    In addition to the review of RB reactor characteristics this presentation is focused on the coupled fast-thermal systems achieved at the reactor. The following experimental methods are presented: neutron spectra measurements; steady state experiments and kinetic measurements ( β eff ) related to the coupled fast-thermal cores

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

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

  8. Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Onozuka, M.; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G.

    2001-01-01

    Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible

  9. Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G

    2001-09-01

    Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible.

  10. Progress on the Fabrication Methods Development for the Korean Test Blanket Module First Wall in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Cho, Seung Yon

    2010-01-01

    A Korean helium cooled molten lithium (HCML) test blanket module (TBM) has been designed to be tested in the International Thermonuclear Experimental Reactor (ITER) TBM and related fabrication methods have been developed especially for the purpose of joining. Since the first wall (FW) of the HCML TBM is composed of a beryllium (Be) as an armor material and a FMS as a structural one, joining with Be to FMS and FMS to FMS should be developed in order to fabricate it

  11. Experimental verification of integrated pressure suppression systems in fusion reactors at in-vessel loss-of-coolant events

    International Nuclear Information System (INIS)

    Takase, K.; Akimoto, H.

    2001-01-01

    An integrated ICE (Ingress-of-Coolant Event) test facility was constructed to demonstrate that the ITER safety design approach and design parameters for the ICE events are adequate. Major objectives of the integrated ICE test facility are: to estimate the performance of an integrated pressure suppression system; to obtain the validation data for safety analysis codes; and to clarify the effects of two-phase pressure drop at a divertor and the direct-contact condensation in a suppression tank. A scaling factor between the test facility and ITER-FEAT is around 1/1600. The integrated ICE test facility simulates the ITER pressure suppression system and mainly consists of a plasma chamber, vacuum vessel, simulated divertor, relief pipe and suppression tank. From the experimental results it was found quantitatively that the ITER pressure suppression system is very effective to reduce the pressurization due to the ICE event. Furthermore, it was confirmed that the analytical results of the TRAC-PF1 code can simulate the experimental results with high accuracy. (author)

  12. Analytical prediction and experimental verification of reactor safety system injection transient

    International Nuclear Information System (INIS)

    Roy, B.N.; Nomm, E.

    1991-01-01

    This paper describes the computer code that was developed for thermal hydraulic transient analysis of mixed phase fluid system and the flow tests that were carried out to validate the Code. A full scale test facility was designed to duplicate the Supplementary Shutdown System (SSS) of Savannah River Production Reactors. Several steady state and dynamic flow tests were conducted simulating the actual reactor injection transients. A dynamic multiphase fluid flow code was developed and validated with experimental results and utilized for system performance predictions and development of technical specifications for reactors. 3 refs

  13. A dense Pd/Ag membrane reactor for methanol steam reforming: Experimental study

    NARCIS (Netherlands)

    Basile, A.; Gallucci, F.; Paturzo, L.

    2005-01-01

    This paper focuses on an experimental study of the methanol steam reforming (MSR) reaction. A dense Pd/Ag membrane reactor (MR) has been used, and its behaviour has been compared to the performance of a traditional reactor (TR) packed with the same catalyst type and amount. The parameters

  14. Report made on behalf of the commission of foreign affairs about the 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)

    2008-02-01

    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. The document presents first the principle, challenge and stakes of the ITER project and the long negotiations that have led to chose France for the setting up of the research facility. Then, it presents the agreement of November 7, 2007, which fixes the general framework of the conditions necessary to the achievement of the project. The agreement comprises the classical rules, privileges and immunities of any international organization sited on the French ground, plus some more specific dispositions in particular in the domain of management of the wastes generated by the facility. (J.S.)

  15. Test results of the first 50 kA NbTi full size sample for ITER

    International Nuclear Information System (INIS)

    Ciazynski, D.; Zani, L.; Huber, S.; Stepanov, B.; Karlemo, B.

    2003-01-01

    Within the framework of the research studies for the International Thermonuclear Experimental Reactor (ITER) project, the first full size NbTi conductor sample was fabricated in industry and tested in the SULTAN facility (Villigen, Switzerland). This sample (PF-FSJS), which is relevant to the Poloidal Field coils of ITER, is composed of two parallel straight bars of conductor, connected at bottom through a joint designed according to the Cea twin-box concept. The two conductor legs are identical except for the use of different strands: a nickel plated NbTi strand with a pure copper matrix in one leg, and a bare NbTi strand with copper matrix and internal CuNi barrier in the other leg. The two conductors and the joint were extensively tested regarding DC (direct current) and AC (alternative current) properties. This paper reports on the tests results and analysis, stressing the differences between the two conductor legs and discussing the impact of the test results on the ITER design criteria for conductor and joint. While joint DC resistance, conductors and joint AC losses, fulfilled the ITER requirements, neither conductor could reach its current sharing temperature at relevant ITER currents, due to instabilities. Although the drop in temperature is slight for the CuNi strand cable, it is more significant for the Ni plated strand cable. (authors)

  16. Revised design for the Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1977-03-01

    A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost

  17. Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Sinev, V. V., E-mail: vsinev@pcbai10.inr.ruhep.ru [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

    2013-05-15

    Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.

  18. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri

    2008-01-01

    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  19. ITER TASK T26/28 (1995): Solubility, diffusion and absorption of hydrogen isotopes in potential fusion reactor ceramics

    International Nuclear Information System (INIS)

    Thompson, D.A.; Macauley-Newcombe, R.G.

    1996-04-01

    Ceramic insulators are integral parts of numerous components essential for the heating control and diagnostic measurement of fusion plasmas. For safe and reliable reactor operations it is necessary to be able to predict the resultant tritium inventories and permeation fluxes through these materials. Some materials being considered are Al 2 O 3 (both as single crystal sapphire and polycrystalline alumina) and BeO. This report contains results of ion-implantation, thermal absorption (diffusion loading) and ion-beam analysis experiments performed in 1994 and 1995 for ITER task T26/28. The combination of implantation and thermal absorption capabilities enable us to load samples with hydrogen isotopes under differing conditions. 13 figs., 1 tab., 11 refs

  20. Requirements for nuclear data from ITER/FER nuclear design

    International Nuclear Information System (INIS)

    Maki, Koichi

    1991-01-01

    Considering ITER and FER activities and future programme of fusion reactor developments, the present situations in fusion neutronics were explained up to now from the previous fusion nuclear data specialist meeting. Vicissitude of development of FUSION-J3 was also explained. From these discussions we clarified the required accuracies in nuclear data, their time limits and their priorities of nuclides to be improved for fusion reactor developments. (author)