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Sample records for iter candidate structural

  1. Compatibility of ITER candidate structural materials with static gallium

    International Nuclear Information System (INIS)

    Luebbers, P.R.; Michaud, W.F.; Chopra, O.K.

    1993-12-01

    Tests were conducted on the compatibility of gallium with candidate structural materials for the International Thermonuclear Experimental Reactor, e.g., Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, as well as Armco iron, Nickel 270, and pure chromium. Type 316 stainless steel is least resistant to corrosion in static gallium and Nb-5 Mo-1 Zr alloy is most resistant. At 400 degrees C, corrosion rates are ∼4.0, 0.5, and 0.03 mm/yr for type 316 SS, Inconel 625, and Nb-5 Mo- 1 Zr alloy, respectively. The pure metals react rapidly with gallium. In contrast to findings in earlier studies, pure iron shows greater corrosion than nickel. The corrosion rates at 400 degrees C are ≥88 and 18 mm/yr, respectively, for Armco iron and Nickel 270. The results indicate that at temperatures up to 400 degrees C, corrosion occurs primarily by dissolution and is accompanied by formation of metal/gallium intermetallic compounds. The solubility data for pure metals and oxygen in gallium are reviewed. The physical, chemical, and radioactive properties of gallium are also presented. The supply and availability of gallium, as well as price predictions through the year 2020, are summarized

  2. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Preliminary results

    International Nuclear Information System (INIS)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1993-01-01

    Candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at temperatures of either 60 or 250 degrees C. Preliminary results have been obtained for several of these materials irradiated at 60 degrees C. The results show that irradiation at this temperature reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The unloading compliance technique developed for the subsize disk compact specimens works quite well, particularly for materials with lower toughness. Specimens of materials with very high toughness deform excessively, and this results in experimental difficulties

  3. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

    International Nuclear Information System (INIS)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.; Shiba, Kiyoyuki

    1994-01-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

  4. ITER containment structures

    International Nuclear Information System (INIS)

    Sadakov, S.; Fauser, F.; Nelson, B.

    1991-01-01

    This document describes the results and recommendations of the Containment Structures Design Unit (CSDU) on the containment structures for ITER, made in the context of the Conceptual Design Phase. The document describes the following subsystems: (1) the primary vacuum vessel (VV), (2) the attaching locks (AL) of the invessel components, (3) the plasma passive and active stabilizers, (4) the cryostat vessel, and (5) the machine gravity supports. Although for most components reference designs were selected, for some of these alternative design options were described, because unresolved problems necessitate further research and development. Conclusions and future needs are summarized for each of the above subsystems: (1) a reference VV design was selected, while most critical VV future needs are the feasibility studies of manufacturing, assembly, and the repair/disassembly/reassembly by remote handling. Alternative, thin-wall options appear attractive and should be studied further during the Engineering Design Activities; (2) no reference design solution was selected for the AL system, as AL design requirements are extremely difficult and internally contradictory, while there is no existing tokamak precedent, but instead, five different approaches will be further researched early in the Engineering Design Phase; (3) significant progress is reported on passive loops, for which the ''twin-loops'' concept is ready to be advanced into the Engineering Design Phase, and on active coils, where a new coil positioning prevents interference with the blanket removal paths, and the current joints are located in a secondary vacuum or in the atmosphere of the reactor hall, repairable by remote handling; (4) a full metallic welded cryostat design with increased toroidal resistance was chosen, but with a design based on concrete with a thin inner metallic liner as a back-up in case detailed nuclear shielding requirements would force the cryostat to act as biological shield; (5) out

  5. Neutron irradiation behavior of ITER candidate beryllium grades

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

    1998-01-01

    Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

  6. 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.)

  7. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  8. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    International Nuclear Information System (INIS)

    Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-01-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  9. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

  10. Structural analysis of the ITER Divertor toroidal rails

    Energy Technology Data Exchange (ETDEWEB)

    Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

    2013-10-15

    The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

  11. Automatic Control of ITER-like Structures

    International Nuclear Information System (INIS)

    Bosia, G.; Bremond, S.

    2005-01-01

    In ITER Ion Cyclotron System requires a power transfer efficiency in excess of 90% from power source to plasma in quasi continuous operation. This implies the availability of a control system capable of optimizing the array radiation spectrum, automatically acquiring impedance match between the power source and the plasma loaded array at the beginning of the power pulse and maintaining it against load variations due to plasma position and plasma edge parameters fluctuations, rapidly detecting voltage breakdowns in the array and/or in the transmission system and reliably discriminating them from fast load variations. In this paper a proposal for a practical ITER control system, including power, phase, frequency and impedance matching is described. (authors)

  12. 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

  13. ITER vacuum vessel structural analysis completion during manufacturing phase

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F. [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Caixas, J.; Fernandez, E.; Zarzalejos, J.M. [F4E, c/Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019, Barcelona (Spain); Kim, H.-S.; Kim, Y.G. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Privalova, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); and others

    2016-11-01

    Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

  14. ITER vacuum vessel structural analysis completion during manufacturing phase

    International Nuclear Information System (INIS)

    Martinez, J.-M.; Alekseev, A.; Sborchia, C.; Choi, C.H.; Utin, Y.; Jun, C.H.; Terasawa, A.; Popova, E.; Xiang, B.; Sannazaro, G.; Lee, A.; Martin, A.; Teissier, P.; Sabourin, F.; Caixas, J.; Fernandez, E.; Zarzalejos, J.M.; Kim, H.-S.; Kim, Y.G.; Privalova, E.

    2016-01-01

    Highlights: • ITER Vacuum Vessel (VV) is a part of the first barrier to confine the plasma. • A Nuclear Pressure Equipment necessitates Agreed Notified Body to assure design, fabrication, and conformance testing and quality assurance. • Some supplementary RCC-MR margin targets have been considered to guarantee considerable structural margins in areas not inspected in operation. • Many manufacturing deviation requests (MDR) and project change requests (PCR) impose to re-evaluate the structural margin. • Several structural analyses were performed with global and local models to guarantee the structural integrity of the whole ITER Vacuum Vessel. - Abstract: Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

  15. Structural analysis of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G.; Ioki, K.; Johnson, G.; Onozuka, M.; Utin, Y. [ITER Joint Work Site, Garching (Germany); Nelson, B. [Oak Ridge National Lab., TN (United States); Swanson, J. [USHT, Raytheon, Princeton (United States)

    1998-07-01

    The ITER Vacuum Vessel (VV) must withstand a large number of loading conditions including electromagnetic, seismic, operational and upset pressure, thermal and test loads. All of the loading conditions and load combinations have been categorized and classified to permit the allowable stress to be defined in accordance with the recommendations of the ASME code. The most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and the load combination of seismic and electromagnetic loads due to a plasma vertical instability. The areas of high stress are the regions around the VV and the blanket supports, and the attachment of the ports to the main shell. In all of the loading conditions and load combinations the calculated stresses are below the allowable values. (authors)

  16. Structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka

    2004-12-01

    ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

  17. Asymptotic convergence for iterative optimization in electronic structure

    International Nuclear Information System (INIS)

    Lippert, Ross A.; Sears, Mark P.

    2000-01-01

    There have recently been a number of proposals for solving large electronic structure problems (local-density approximation, Hartree-Fock, and tight-binding methods) iteratively with a computational effort proportional to the size of the system. The effort needed to perform a single iteration in these schemes is well understood but the convergence rate has been an empirical matter. This paper will show that many of the proposed methods have a single underlying geometrical structure, which has a specific asymptotic convergence behavior, and that behavior can be understood in terms of some simple condition numbers based on the spectrum of the Hamiltonian. (c) 2000 The American Physical Society

  18. Reactor structure and superconducting magnet system of ITER

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  19. Structural analysis of ITER multi-purpose deployer

    International Nuclear Information System (INIS)

    Manuelraj, Manoah Stephen; Dutta, Pramit; Gotewal, Krishan Kumar; Rastogi, Naveen; Tesini, Alessandro; Choi, Chang-Hwan

    2016-01-01

    Highlights: • System modelling for structural analysis of the Multi-Purpose Deployer (MPD). • Finite element modeling of the Multi-Purpose Deployer (MPD). • Static, modal and seismic response analysis of the Multi-Purpose Deployer (MPD). • Iterative structural analysis and design update to satisfy the structural criteria. • Modal analysis for various kinematic configurations. • Reaction force calculations on the interfacing systems. - Abstract: The Multi-Purpose Deployer (MPD) is a general purpose ITER in-vessel remote handling (RH) system. The main handling equipment, known as the MPD Transporter, consists of a series of linked bodies, which provide anchoring to the vacuum vessel port and an articulated multi-degree of freedom motion to perform various in-vessel maintenance tasks. During the in-vessel operations, the structural integrity of the system should be guaranteed against various operational and seismic loads. This paper presents the structural analysis results of the concept design of the MPD Transporter considering the seismic events. Static structural, modal and frequency response spectrum analyses have been performed to verify the structural integrity of the system, and to provide reaction forces to the interfacing systems such as vacuum vessel and cask. Iterative analyses and design updates are carried out based on the reference design of the system to improve the structural behavior of the system. The frequency responses of the system in various kinematics and payloads are assessed.

  20. Structural analysis of ITER TBM Frame and Dummy TBM

    International Nuclear Information System (INIS)

    Marin, Anna; Kim, Byoung Yoon; Bertolini, Claudio; Lucca, Flavio; Komarov, Victor; Merola, Mario; Giancarli, Luciano; Gicquel, Stefan

    2013-01-01

    One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets. To accomplish these goals, three ITER equatorial ports are dedicated to the test of Test Blanket Modules (TBMs) that are mock-ups of tritium breeding blankets. These TBMs, associated with appropriate shield blocks, will also provide the same thermal and nuclear shielding as the main blanket. The main function of TBM Port Plug (PP) is to accommodate TBMs and provide a standardized interface with the vacuum vessel (VV)/port structure. The feasibility of the design concept of the Frame including two Dummy TBMs has been investigated by proposing design improvements of the reference design through an extensive set of thermal, electromagnetic (EM) and stress analyses. As well, the related static strength was evaluated in accordance with the structural design criteria for ITER in-vessel components (SDC-IC). This paper outlines the engineering aspects of the ITER TBM Frame and Dummy TBM and focuses on the feasibility of the present design by structural assessment

  1. Overview of ITER project. Objective, history, status of development, structure and its future

    International Nuclear Information System (INIS)

    Matsuda, S.

    2003-01-01

    The objective of ITER Project is to demonstrate the scientific and technological feasibility of fusion energy for peaceful use. To achieve this goal, the Conceptual Design Activities (CDA 1988-1990) and the Engineering Design Activities (EDA 1992-2001) were conducted by the joint effort of EU, Japan, RF and US under the auspices of IAEA. Engineering design has been developed consistently with plasma physics databases obtained by the world-wide magnetic fusion devices up to and during the EDA. Technology R and D including development, manufacturing and testing of scalable models of major components have been also conducted and ensured manufacturing and technical feasibility of ITER. Such technical readiness to start ITER construction urged formal government negotiation initially by Canada, EU, Japan and RF, and intense meetings have been held since November 2001. The US and China have joined since Feb. 2003. Four candidate sites have been proposed by Canada, EU and Japan for construction and have been evaluated. Once the agreement will be made, the construction takes about 10 years to complete, followed by the operation and exploitation phase lasting about 20 years. This paper gives an overview of ITER Project including its role in the long range fusion program, history, status of development, collaboration structure and its future program. (author)

  2. Vacuum vessel port structures for ITER-FEAT

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th.

    2001-01-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed

  3. Vacuum vessel port structures for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Utin, Yu.; Ioki, K.; Komarov, V.; Krylov, V.; Kuzmin, E.; Labusov, I.; Miki, N.; Onozuka, M.; Rozov, V.; Sannazzaro, G.; Tesini, A.; Yamada, M.; Barthel, Th

    2001-11-01

    The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

  4. Piping structural design for the ITER thermal shield manifold

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

  5. Structural analysis of ITER sub-assembly tools

    International Nuclear Information System (INIS)

    Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Lee, J.H.; Kim, K.K.; Im, K.; Shaw, R.

    2011-01-01

    The ITER Tokamak assembly tools are purpose-built assembly tools to complete the ITER Tokamak machine which includes the cryostat and the components contained therein. The sector sub-assembly tools descried in this paper are main assembly tools to assemble vacuum vessel, thermal shield and toroidal filed coils into a complete 40 o sector. The 40 o sector sub-assembly tools are composed of sector sub-assembly tool, including radial beam, vacuum vessel supports and mid-plane brace tools. These tools shall have sufficient strength to transport and handle heavy weight of the ITER Tokamak machine reached several hundred tons. Therefore these tools should be designed and analyzed to confirm both the strength and structural stability even in the case of conservative assumptions. To verify structural stabilities of the sector sub-assembly tools in terms of strength and deflection, ANSYS code was used for linear static analysis. The results of the analysis show that these tools are designed with sufficient strength and stiffness. The conceptual designs of these tools are briefly described in this paper also.

  6. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Alexandrov, Evgeny [Institution “Project Center ITER”, 1, Akademika Kurchatova sq., 123182 Moscow (Russian Federation); Batyunin, Alexander; Kashchuk, Yuri [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Korban, Svetlana; Lyublin, Boris [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Obudovsky, Sergey [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Senik, Konstantin [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity.

  7. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    International Nuclear Information System (INIS)

    Tanchuk, Victor; Alexandrov, Evgeny; Batyunin, Alexander; Kashchuk, Yuri; Korban, Svetlana; Lyublin, Boris; Obudovsky, Sergey; Senik, Konstantin

    2013-01-01

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity

  8. 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)

  9. 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

  10. Structural analysis for the joint of the ITER ELM coil

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Shanwen, E-mail: zhangshanwen123@163.com [College of Mechanical Engineering Yangzhou University, Yangzhou 225127 (China); Song, Yuntao; Wang, Zhongwei; Ji, Xiang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 200031 (China); Zhang, Jianfeng [College of Mechanical Engineering Yangzhou University, Yangzhou 225127 (China)

    2017-01-15

    Highlights: • The FE sub-model method is feasible and rapid for the joint design. • The components meet the static and fatigue criteria. • Nuclear heat is the key factor for the joint design. - Abstract: The joint is an important component of the Edge Localized Modes (ELM) coils in fusion reactor, which is used to connect the ELM coils. Like the ELM coils, the joints work in an environment with high radiation levels, high temperature and high magnetic field. These joints are mainly subject to nuclear heat from the plasma and cyclic electromagnetic (EM) loads induced by the interaction of ELM coil current with magnetic fields. Take the joint of ITER ELM coil for example. In order to assure the structural integrity of joints under these loads, it is necessary to estimate the strength and fatigue of the joints. As a local model, the joint without ELM coil is investigated by the sub-model method. Results show that the finite element sub-model method is feasible and rapid for the joint design. The maximum magnetic flux intensity occurs in the axial direction for the joint local reference, which parallels with the current and corresponds to the toroidal direction of the ITER. The two areas at the top of the Inconel sleeve appear high temperature. For the joint, the conductor, jacket and sleeve can meet the static and fatigue criteria and the joint design is valid and feasible. The thermal load from the nuclear heat is the key factor for the joint design.

  11. Development of manufacturing technology for ITER TF Coil Structure

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Takeru, E-mail: sakurai.takeru@jaea.go.jp; Iguchi, Masahide; Nakahira, Masataka; Inagaki, Takashi; Matsui, Kunihiro; Koizumi, Norikiyo

    2016-11-01

    Highlights: • Heavy thick welding (Max. 287 mm) was performed by balance welding. • Figured out Attachment welding deformation including heavy thick welding. • The deformation of Segments welding was suppressed to 1/3 of previous method. • Based on this study, JAEA started actual ITER TF coil structure manufacturing. - Abstract: Japan Atomic Energy Agency (JAEA) performed a trial of A1 Segment manufacturing of Toroidal Field (TF) coil structure, which is a piece with a radius of curvature 3 m with square channel for coil. Even though both-side welding (balance welding) was preferred to one-side welding considering the welding deformation, it could not be applied to the previous trial due to the difficulty of overhead or horizontal welding by machine. Hence, one-side welding with strong restriction jig was applied in the previous trials. In the latest trial, JAEA adopted a manual balance welding with a development of manufacturing technology. As the result of A1 Segment Mainbody welding trial, welding deformation of the Outer Plate and the Side Plate could have been controlled closer to the target value. JAEA also tried Attachments welding, in which Pre-Compression Flange (PCF) and Extension are welded to A1 Segment Mainbody, and a Segments welding trial, which is a weld between A1 Segment and a part of A2 Segment. A2 Segment is a 3 m straight part with square channel for coil. The inclination of A1 Segment and A2 Segment due to the welding was 2.7 mm. By applying balance welding, the deformation by Segments welding was suppressed to about 1/3 of the one-side welding. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

  12. Automatic Detection and Quantification of WBCs and RBCs Using Iterative Structured Circle Detection Algorithm

    Directory of Open Access Journals (Sweden)

    Yazan M. Alomari

    2014-01-01

    Full Text Available Segmentation and counting of blood cells are considered as an important step that helps to extract features to diagnose some specific diseases like malaria or leukemia. The manual counting of white blood cells (WBCs and red blood cells (RBCs in microscopic images is an extremely tedious, time consuming, and inaccurate process. Automatic analysis will allow hematologist experts to perform faster and more accurately. The proposed method uses an iterative structured circle detection algorithm for the segmentation and counting of WBCs and RBCs. The separation of WBCs from RBCs was achieved by thresholding, and specific preprocessing steps were developed for each cell type. Counting was performed for each image using the proposed method based on modified circle detection, which automatically counted the cells. Several modifications were made to the basic (RCD algorithm to solve the initialization problem, detecting irregular circles (cells, selecting the optimal circle from the candidate circles, determining the number of iterations in a fully dynamic way to enhance algorithm detection, and running time. The validation method used to determine segmentation accuracy was a quantitative analysis that included Precision, Recall, and F-measurement tests. The average accuracy of the proposed method was 95.3% for RBCs and 98.4% for WBCs.

  13. Comparing direct and iterative equation solvers in a large structural analysis software system

    Science.gov (United States)

    Poole, E. L.

    1991-01-01

    Two direct Choleski equation solvers and two iterative preconditioned conjugate gradient (PCG) equation solvers used in a large structural analysis software system are described. The two direct solvers are implementations of the Choleski method for variable-band matrix storage and sparse matrix storage. The two iterative PCG solvers include the Jacobi conjugate gradient method and an incomplete Choleski conjugate gradient method. The performance of the direct and iterative solvers is compared by solving several representative structural analysis problems. Some key factors affecting the performance of the iterative solvers relative to the direct solvers are identified.

  14. On Algebraic Structure of Improved Gauss-Seidel Iteration

    OpenAIRE

    O. M. Bamigbola; A. A. Ibrahim

    2014-01-01

    Analysis of real life problems often results in linear systems of equations for which solutions are sought. The method to employ depends, to some extent, on the properties of the coefficient matrix. It is not always feasible to solve linear systems of equations by direct methods, as such the need to use an iterative method becomes imperative. Before an iterative method can be employed to solve a linear system of equations there must be a guaranty that the process of solut...

  15. Low activation structural material candidates for fusion power plants

    International Nuclear Information System (INIS)

    Forty, C.B.A.; Cook, I.

    1997-06-01

    Under the SEAL Programme of the European Long-Term Fusion Safety Programme, an assessment was performed of a number of possible blanket structural materials. These included the steels then under consideration in the European Blanket Programme, as well as materials being considered for investigation in the Advanced Materials Programme. Calculations were performed, using SEAFP methods, of the activation properties of the materials, and these were related, based on the SEAFP experience, to assessments of S and E performance. The materials investigated were the SEAFP low-activation martensitic steel (LA12TaLC); a Japanese low-activation martensitic steel (F-82H), a range of compositional variants about this steel; the vanadium-titanium-chromium alloy which was the original proposal of the ITER JCT for the ITER in-vessel components; a titanium-aluminium intermetallic (Ti-Al) which is under investigation in Japan; and silicon carbide composite (SiC). Assessed impurities were included in the compositions of these materials, and they have very important impacts on the activation properties. Lack of sufficiently detailed data on the composition of chromium alloys precluded their inclusion in the study. (UK)

  16. Iterative structure within the five-particle two-loop amplitude

    International Nuclear Information System (INIS)

    Cachazo, Freddy; Spradlin, Marcus; Volovich, Anastasia

    2006-01-01

    We find an unexpected iterative structure within the two-loop five-gluon amplitude in N=4 supersymmetric Yang-Mills theory. Specifically, we show that a subset of diagrams contributing to the full amplitude, including a two-loop pentagon-box integral with nontrivial dependence on five kinematical variables, satisfies an iterative relation in terms of one-loop scalar box diagrams. The implications of this result for the possible iterative structure of the full two-loop amplitude are discussed

  17. Development of cutting and welding methods for thick-walled stainless steel support and containment structures for ITER

    International Nuclear Information System (INIS)

    Jones, L.; Maisonnier, D.; Goussain, J.; Johnson, G.; Petring, D.; Wernwag, L.

    1998-01-01

    In ITER the containment and support structures are made from 316L(N)-IG (ITER Grade) stainless steel plate, 40 to 70 mm thick. The structures are divided into twenty sectors which have to be welded together in situ. The three areas of work described in this paper are, CO 2 laser welding, plasma cutting and CO 2 laser cutting. CO 2 laser welding offers significant advantages due to its high speed and low distortion and the most powerful commercial laser in Europe has been used to investigate single pass welding of thick plates, with strong welds up to 35 mm thick being achieved in one pass. For cutting, the space available on the back-side to collect debris and protect fragile components from damage is limited to 30 mm. A static, water-cooled backside protection plate proved unable to contain the debris from plasma cutting so a reciprocating backside protection system with dry ceramic heat shield demonstrated a solution. A 10 kW CO 2 laser system for nitrogen-assisted laser cutting, provided successful results at 40 mm thickness. This technique shows considerable promise as significant reductions in through power and rate of debris production are expected compared with plasma cutting and thicker cuts appear feasible. The results presented herein represent significant technical advances and will be strong candidates for the mix of methods which will have to be used for the assembly and maintenance of the ITER machine. (authors)

  18. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  19. Strong coupling strategy for fluid-structure interaction problems in supersonic regime via fixed point iteration

    Science.gov (United States)

    Storti, Mario A.; Nigro, Norberto M.; Paz, Rodrigo R.; Dalcín, Lisandro D.

    2009-03-01

    In this paper some results on the convergence of the Gauss-Seidel iteration when solving fluid/structure interaction problems with strong coupling via fixed point iteration are presented. The flow-induced vibration of a flat plate aligned with the flow direction at supersonic Mach number is studied. The precision of different predictor schemes and the influence of the partitioned strong coupling on stability is discussed.

  20. Compatibility of candidate structural materials with static gallium

    International Nuclear Information System (INIS)

    Luebbers, P.R.; Michaud, W.F.; Chopra, O.K.

    1993-01-01

    Scoping tests were conducted on compatibility of gallium with candidate structural materials, e.g., Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, as well as Armco iron, Nickel 270, and pure chronimum. Type 316 stainless steel is least resistant and Nb-5 Mo-1 Zr alloy is most resistant to corrosion in static gallium. At 400 degrees C, corrosion rates are ∼4.0, 0.5, and 0.03 mm/y for Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, respectively. The pure metals react rapidly with gallium. In contrast to findings in earlier studies, pure iron shows greater corrosion than does nickel. The corrosion rates at 400 degrees C are ≥90 and 17 mm/y, respectively, for Armco iron and Nickel 270. The results indicate that at temperatures up to 400 degrees C, corrosion occurs primarily by dissolution accompanied by formation of metal/gallium intermetallic compounds

  1. A fast and robust iterative algorithm for prediction of RNA pseudoknotted secondary structures

    Science.gov (United States)

    2014-01-01

    Background Improving accuracy and efficiency of computational methods that predict pseudoknotted RNA secondary structures is an ongoing challenge. Existing methods based on free energy minimization tend to be very slow and are limited in the types of pseudoknots that they can predict. Incorporating known structural information can improve prediction accuracy; however, there are not many methods for prediction of pseudoknotted structures that can incorporate structural information as input. There is even less understanding of the relative robustness of these methods with respect to partial information. Results We present a new method, Iterative HFold, for pseudoknotted RNA secondary structure prediction. Iterative HFold takes as input a pseudoknot-free structure, and produces a possibly pseudoknotted structure whose energy is at least as low as that of any (density-2) pseudoknotted structure containing the input structure. Iterative HFold leverages strengths of earlier methods, namely the fast running time of HFold, a method that is based on the hierarchical folding hypothesis, and the energy parameters of HotKnots V2.0. Our experimental evaluation on a large data set shows that Iterative HFold is robust with respect to partial information, with average accuracy on pseudoknotted structures steadily increasing from roughly 54% to 79% as the user provides up to 40% of the input structure. Iterative HFold is much faster than HotKnots V2.0, while having comparable accuracy. Iterative HFold also has significantly better accuracy than IPknot on our HK-PK and IP-pk168 data sets. Conclusions Iterative HFold is a robust method for prediction of pseudoknotted RNA secondary structures, whose accuracy with more than 5% information about true pseudoknot-free structures is better than that of IPknot, and with about 35% information about true pseudoknot-free structures compares well with that of HotKnots V2.0 while being significantly faster. Iterative HFold and all data used in

  2. Structure and evolution of barley powdery mildew effector candidates

    Directory of Open Access Journals (Sweden)

    Pedersen Carsten

    2012-12-01

    Full Text Available Abstract Background Protein effectors of pathogenicity are instrumental in modulating host immunity and disease resistance. The powdery mildew pathogen of grasses Blumeria graminis causes one of the most important diseases of cereal crops. B. graminis is an obligate biotrophic pathogen and as such has an absolute requirement to suppress or avoid host immunity if it is to survive and cause disease. Results Here we characterise a superfamily predicted to be the full complement of Candidates for Secreted Effector Proteins (CSEPs in the fungal barley powdery mildew parasite B. graminis f.sp. hordei. The 491 genes encoding these proteins constitute over 7% of this pathogen’s annotated genes and most were grouped into 72 families of up to 59 members. They were predominantly expressed in the intracellular feeding structures called haustoria, and proteins specifically associated with the haustoria were identified by large-scale mass spectrometry-based proteomics. There are two major types of effector families: one comprises shorter proteins (100–150 amino acids, with a high relative expression level in the haustoria and evidence of extensive diversifying selection between paralogs; the second type consists of longer proteins (300–400 amino acids, with lower levels of differential expression and evidence of purifying selection between paralogs. An analysis of the predicted protein structures underscores their overall similarity to known fungal effectors, but also highlights unexpected structural affinities to ribonucleases throughout the entire effector super-family. Candidate effector genes belonging to the same family are loosely clustered in the genome and are associated with repetitive DNA derived from retro-transposons. Conclusions We employed the full complement of genomic, transcriptomic and proteomic analyses as well as structural prediction methods to identify and characterize the members of the CSEPs superfamily in B. graminis f

  3. Effects of coupling and asymmetries on load resilience of IC ITER-like structures

    International Nuclear Information System (INIS)

    Bosia, G.; Bremond, S.; Colas, L.

    2005-01-01

    ITER-like structures feature an intrinsic resilience to load variations, which is related to the symmetry of the currents in the two branches of the structure. It has been suggested that the effects of coupling between the array elements would significantly impair the load resilience of the structure. In this paper the effect of inter strap coupling and of however induced electrical array asymmetries on the structure load resilience are quantitatively examined

  4. Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

    International Nuclear Information System (INIS)

    Sannazzaro, G.; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H.

    2001-01-01

    In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses

  5. Critical issues of the structural integrity of the ITER-FEAT vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sannazzaro, G. E-mail: sannazg@itereu.de; Barabaschi, P.; Elio, F.; Ioki, K.; Miki, N.; Onozuka, M.; Utin, Y.; Verrecchia, M.; Yoshimura, H

    2001-11-01

    In the ITER-FEAT, the most severe loading conditions for the VV are the toroidal field coil fast discharge (TFCFD) and its load combination with electromagnetic loads due to a plasma vertical instability, which cause high compressive stresses in the VV inboard wall and increase the risk of buckling. Detailed analyses need to be performed to assess the stress level at the geometrical discontinuities and where concentrated loads are applied. The nuclear heating and the presence of gaps between the blanket modules cause concentrated nuclear heat loads. This paper describes the major structural issues of the ITER vacuum vessel (VV), and summarises the preliminary results of structural analyses.

  6. Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

    Energy Technology Data Exchange (ETDEWEB)

    Morimoto, M. [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)], E-mail: masaaki_morimoto@maia.eonet.ne.jp; Ioki, K.; Terasawa, A.; Utin, Yu.; Barabash, V.; Gribov, Y. [ITER Organization, 13108 St. Paul lez Durance (France)

    2009-12-15

    Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M and JT-60U . In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called 'ferromagnetic inserts'. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.

  7. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

    a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...

  8. Comparing two iteration algorithms of Broyden electron density mixing through an atomic electronic structure computation

    International Nuclear Information System (INIS)

    Zhang Man-Hong

    2016-01-01

    By performing the electronic structure computation of a Si atom, we compare two iteration algorithms of Broyden electron density mixing in the literature. One was proposed by Johnson and implemented in the well-known VASP code. The other was given by Eyert. We solve the Kohn-Sham equation by using a conventional outward/inward integration of the differential equation and then connect two parts of solutions at the classical turning points, which is different from the method of the matrix eigenvalue solution as used in the VASP code. Compared to Johnson’s algorithm, the one proposed by Eyert needs fewer total iteration numbers. (paper)

  9. Design and structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka; Shibanuma, Kiyoshi

    2004-01-01

    The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

  10. Basic performance tests on vibration of support structure with flexible plates for ITER tokamak device

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Shibanuma, Kiyoshi

    2005-01-01

    The vibration experiments of the support structures with flexible plates for the ITER major components such as toroidal field coil (TF coil) and vacuum vessel (VV) were performed using small-sized flexible plates aiming to obtain its basic mechanical characteristics such as dependence of the stiffness on the loading angle. The experimental results obtained by the hammering and frequency sweep tests were agreed each other, so that the experimental method is found to be reliable. In addition, the experimental results were compared with the analytical ones in order to estimate an adequate analytical model for ITER support structure with flexible plates. As a result, the bolt connection of the flexible plates on the base plate strongly affected on the stiffness of the flexible plates. After studies of modeling the bolts, it is found that the analytical results modeling the bolts with finite stiffness only in the axial direction and infinite stiffness in the other directions agree well with the experimental ones. Using this adequate model, the stiffness of the support structure with flexible plates for the ITER major components can be calculated precisely in order to estimate the dynamic behaviors such as eigen modes and amplitude of deformation of the major components of the ITER tokamak device. (author)

  11. Iterative Tensor Voting for Perceptual Grouping of Ill-Defined Curvilinear Structures: Application to Adherens Junctions

    Science.gov (United States)

    Loss, Leandro A.; Bebis, George; Parvin, Bahram

    2012-01-01

    In this paper, a novel approach is proposed for perceptual grouping and localization of ill-defined curvilinear structures. Our approach builds upon the tensor voting and the iterative voting frameworks. Its efficacy lies on iterative refinements of curvilinear structures by gradually shifting from an exploratory to an exploitative mode. Such a mode shifting is achieved by reducing the aperture of the tensor voting fields, which is shown to improve curve grouping and inference by enhancing the concentration of the votes over promising, salient structures. The proposed technique is applied to delineation of adherens junctions imaged through fluorescence microscopy. This class of membrane-bound macromolecules maintains tissue structural integrity and cell-cell interactions. Visually, it exhibits fibrous patterns that may be diffused, punctate and frequently perceptual. Besides the application to real data, the proposed method is compared to prior methods on synthetic and annotated real data, showing high precision rates. PMID:21421432

  12. Support structure concept for integration of ITER diagnostics in the port cell

    Energy Technology Data Exchange (ETDEWEB)

    Udintsev, V.S., E-mail: victor.udintsev@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Portalès, M.; Giacomin, T.; Darcourt, O.; Direz, M.-F.; Martins, J.P.; Penot, C.; Arumugam, A.P.; Drevon, J.-M.; Friconneau, J.P.; Levesy, B.; Maquet, P.; Patel, K.M.; Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Popova, E. [Russian Federation Domestic Agency, Moscow (Russian Federation); Proust, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Ronden, D.M.S. [DIFFER, Nieuwegein (Netherlands); Walker, C.I.; Walsh, M.J.; Watts, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France)

    2013-10-15

    Highlights: ► An interspace support structure to support the diagnostic systems from the back of the upper and equatorial port plugs to the biological shield plug. ► Port cell support structures are foreseen to handle the equipment in the port cell. ► Both ISS and PCSS will be supported by means of RH rail system. ► The structures will be positioned with a certain tolerance. ► The proposed concepts are found to fulfil the needs for support of the diagnostics in ITER. -- Abstract: Development of the diagnostics for ITER tokamak, which is presently under construction by several international partners at Cadarache in France, is a major challenge because of severe environment, strict engineering requirements, and the need for high reliability in the measurements. The diagnostic systems in the upper, equatorial and lower port cells on ITER are designed to be integrated within the interspace and port cell support structures. These structures are interfacing with remote handling rail system for the cask operations, thus facilitating the removal and installation of the diagnostics in the port and hence minimizing time for working close to the tokamak. In this paper, the challenges associated with the integration of the diagnostics in the port interspace and port cell, as well as their solutions will be addressed and presented. The interspace and the port cell support structures, as well as their interfaces with the biological shield, will be discussed.

  13. Aqueous stress corrosion of candidate austenitic steels for ITER structural applications

    International Nuclear Information System (INIS)

    Soppet, W.K.; French, D.M.; Kassner, T.F.

    1993-01-01

    Susceptibility of crevice-weldment specimens of Types 316L and 316NG stainless steel (SS) to SCC was investigated in slow-strain-rate-tensile (SSRT) tests in water that simulates important parameters anticipated in first-wall/blanket systems. The SSRT tests were performed in oxygenated water containing 0.06-10 ppm chloride at temperatures of 95 to 225 degrees C to establish the effects of water purity and temperature on SC resistance. These steels, including weldments, exhibit good resistance to SCC under crevice conditions at temperatures of 150 degrees C in oxygenated water containing 0.1-10 ppm chloride. Most specimens fractured in the base metal, and several others fractured in the heat-affected zone (HAZ) of the weld, but none failed in the weld metal

  14. 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

  15. Design progress of the ITER vacuum vessel sectors and port structures

    International Nuclear Information System (INIS)

    Utin, Yu.; Ioki, K.; Alekseev, A.; Bachmann, Ch.; Cho, S.; Chuyanov, V.; Jones, L.; Kuzmin, E.; Morimoto, M.; Nakahira, M.; Sannazzaro, G.

    2007-01-01

    Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (no. 1) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV Sectors no. 2 and no. 3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements

  16. Discrete Self-Similarity in Interfacial Hydrodynamics and the Formation of Iterated Structures.

    Science.gov (United States)

    Dallaston, Michael C; Fontelos, Marco A; Tseluiko, Dmitri; Kalliadasis, Serafim

    2018-01-19

    The formation of iterated structures, such as satellite and subsatellite drops, filaments, and bubbles, is a common feature in interfacial hydrodynamics. Here we undertake a computational and theoretical study of their origin in the case of thin films of viscous fluids that are destabilized by long-range molecular or other forces. We demonstrate that iterated structures appear as a consequence of discrete self-similarity, where certain patterns repeat themselves, subject to rescaling, periodically in a logarithmic time scale. The result is an infinite sequence of ridges and filaments with similarity properties. The character of these discretely self-similar solutions as the result of a Hopf bifurcation from ordinarily self-similar solutions is also described.

  17. Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

    Science.gov (United States)

    H, L. SWAMI; C, DANANI; A, K. SHAW

    2018-06-01

    Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a high-energy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic‑Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well

  18. A conceptual design of the ITER upper port plug structure and its cooling channels

    International Nuclear Information System (INIS)

    Pak, S.I.; Lee, H.G.; Jung, K.J.; Walker, C.I.; Kim, D.G.; Choi, K.S.

    2008-01-01

    A study is conducted on the conceptual design of the structure and cooling channels of the upper port plug of International Thermonuclear Experimental Reactor (ITER). Modification of the earlier port plug design is made and a simple fabrication method is proposed. It is shown that the newly designed port plug can accommodate the installation of both diagnostic and electron cyclotron heating (ECH) devices. Design assessment is carried out through structural and thermo-hydraulic analyses. Results of the analyses show that the port plug structure is stable against one of the most severe plasma events and the total pressure drop of the coolant is within the allowable level

  19. European TBM for ITER: Structural material assessment and breeding capability - Comparative analysis

    International Nuclear Information System (INIS)

    Herreras, Y.; Perlado, J.M.; Ibarra, A.

    2007-01-01

    Full text of publication follows: The ITER European Party is currently developing for DEMO reactor specifications two breeding blanket concepts: the Helium-Cooled Lithium-Lead blanket (HCLL), using a liquid breeder; and the Helium-Cooled Pebble-Bed blanket (HCPB), using a lithiated solid breeder. These two research lines are expected to be tested in ITER as Test Blanket Modules (TBM), in order to demonstrate their safety, economical and environmental suitability. In this sense, structural material activation and breeding blanket capability represent two major challenges. This paper presents new calculations regarding neutronic irradiation inside the ITER Vacuum Vessel. In particular, results are focused on the irradiation affecting the equatorial ports, where the TBM will be located for testing. The methodology employed mainly consists in calculating the neutronic irradiation levels at the required locations with the transport code MCNP, where the input geometry has been previously designed with the program CATIA V5. The main structural materials proposed for the European Test blanket Modules are selected in order to carry out a comparative analysis in safety terms: material activation and basic parameters for damage analysis are evaluated with the code ACAB, based on the neutronic irradiation results mentioned above. Finally, the breeding blanket capability is assessed for both breeding blanket concepts; the results are compared considering the choice of the structural material. (authors)

  20. Fast ion power loads on ITER first wall structures in the presence of NTMs and microturbulence

    International Nuclear Information System (INIS)

    Kurki-Suonio, T.; Asunta, O.; Hirvijoki, E.; Koskela, T.; Snicker, A.; Sipilae, S.; Hauff, T.; Jenko, F.; Poli, E.

    2011-01-01

    The level and distribution of the wall power flux of energetic ions in ITER have to be known accurately in order to ensure the integrity of the first wall. Until now, most quantitative estimates have been based on the assumption that fast ion transport is dictated by neoclassical effects only. However, in ITER, the fast ion distribution is likely to be affected by various MHD effects and probably also by microturbulence. We have now upgraded our orbit-following Monte Carlo code ASCOT so that it has simple, theory-based models for neoclassical tearing mode (NTM)-type islands as well as for turbulent diffusion. ASCOT also allows for full-orbit following, which is important close to the material surfaces and, possibly, also when strong toroidal inhomogeneities are present in the magnetic field. Here we introduce the new models, preliminary results obtained with them, and how these models could be made more realistic in the future. The simulations are carried out for thermonuclear alpha particles in ITER scenario 2 plasma, because we consider this combination to be most critical for the successful operation of ITER. Neither the turbulent transport nor NTM-type islands are found to introduce alarming changes in the wall loads. However, at this stage it was not possible to combine the island structures with the non-axisymmetric magnetic field of ITER, and it remains to be seen what the combined effect of drift islands together with the toroidal ripple and local field aberrations, such as those due to test blanket modules and resonant magnetic perturbations will be.

  1. Efficient approach to simulate EM loads on massive structures in ITER machine

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Andreeva, Z.; Belov, A.; Belyakov, V.; Filatov, O. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Gribov, Yu.; Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Kukhtin, V.; Labusov, A.; Lamzin, E.; Lyublin, B.; Malkov, A.; Mazul, I. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Rozov, V.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul-Lez-Durance (France); Sychevsky, S., E-mail: sytch@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    Highlights: ► A modelling technique to predict EM loads in ITER conducting structures is presented. ► The technique provides low computational cost and parallel computations. ► Detailed models were built for the system “vacuum vessel, cryostat, thermal shields”. ► EM loads on massive in-vessel structures were simulated with the use of local models. ► A flexible combination of models enables desired accuracy of load distributions. -- Abstract: Operation of the ITER machine is associated with high electromagnetic (EM) loads. An essential contributor to EM loads is eddy currents induced in passive conductive structures. Reasoning from the ITER construction, a modelling technique has been developed and applied in computations to efficiently predict anticipated loads. The technique allows us to avoid building a global 3D finite-element (FE) model that requires meshing of the conducting structures and their vacuum environment into 3D solid elements that leads to high computational cost. The key features of the proposed technique are: (i) the use of an existing shell model for the system “vacuum vessel (VV), cryostat, and thermal shields (TS)” implementing the magnetic shell approach. A solution is obtained in terms of a single-component, in this case, vector electric potential taken within the conducting shells of the “VV + cryostat + TS” system. (ii) EM loads on in-vessel conducting structures are simulated with the use of local FE models. The local models use either the 3D solid body or shell approximations. Reasoning from the simulation efficiency, the local boundary conditions are put with respect to the total field or an external field. The use of an integral-differential formulation and special procedures ensures smooth and accurate simulated distributions of fields from current sources of any geometry. The local FE models have been developed and applied for EM analyses of a variety of the ITER components including the diagnostic systems

  2. The structure analysis of ITER cryostat based on the finite element method

    International Nuclear Information System (INIS)

    Liang Chao; Ye, M.Y.; Yao, D.M.; Cao, Lei; Zhou, Z.B.; Xu, Teijun; Wang Jian

    2013-01-01

    In the ITER project the cryostat is one of the most important components. Cryostat shall transfer all the loads that derive from the TOKAMAK inner basic machine, and from the cryostat itself, to the floor of the TOKAMAK pit (during the normal and off-normal operational regimes, and at specified accidental conditions). This paper researches the dynamic structure strength of the ITER cryostat during the operation of TOKAMAK. Firstly the paper introduces the types of loads and the importance of every type load to the research. Then it gives out the method of building model and principle of simplified model, boundary conditions and the way of applying loads on the cryostat. Finally the author discussed the analysis result and the strength questions of cryostat, also, the author pointed out the opinions according to the analysis results.

  3. TurboFold: Iterative probabilistic estimation of secondary structures for multiple RNA sequences

    Directory of Open Access Journals (Sweden)

    Sharma Gaurav

    2011-04-01

    Full Text Available Abstract Background The prediction of secondary structure, i.e. the set of canonical base pairs between nucleotides, is a first step in developing an understanding of the function of an RNA sequence. The most accurate computational methods predict conserved structures for a set of homologous RNA sequences. These methods usually suffer from high computational complexity. In this paper, TurboFold, a novel and efficient method for secondary structure prediction for multiple RNA sequences, is presented. Results TurboFold takes, as input, a set of homologous RNA sequences and outputs estimates of the base pairing probabilities for each sequence. The base pairing probabilities for a sequence are estimated by combining intrinsic information, derived from the sequence itself via the nearest neighbor thermodynamic model, with extrinsic information, derived from the other sequences in the input set. For a given sequence, the extrinsic information is computed by using pairwise-sequence-alignment-based probabilities for co-incidence with each of the other sequences, along with estimated base pairing probabilities, from the previous iteration, for the other sequences. The extrinsic information is introduced as free energy modifications for base pairing in a partition function computation based on the nearest neighbor thermodynamic model. This process yields updated estimates of base pairing probability. The updated base pairing probabilities in turn are used to recompute extrinsic information, resulting in the overall iterative estimation procedure that defines TurboFold. TurboFold is benchmarked on a number of ncRNA datasets and compared against alternative secondary structure prediction methods. The iterative procedure in TurboFold is shown to improve estimates of base pairing probability with each iteration, though only small gains are obtained beyond three iterations. Secondary structures composed of base pairs with estimated probabilities higher than a

  4. FEM analyses and prototype tests of the UPP structure for the ECRH in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Strauss, D. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe (Germany)], E-mail: dirk.strauss@imf.fzk.de; Gantenbein, G. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Pulsed Power and Microwave Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Heidinger, R.; Kleefeldt, K. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe (Germany); Leonhardt, W. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Pulsed Power and Microwave Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Meier, A. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe (Germany); Mellein, D. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Pulsed Power and Microwave Technology, P.O. Box 3640, D-76021 Karlsruhe (Germany); Scherer, T. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe (Germany); Serikov, A. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Reactor Safety, P.O. Box 3640, D-76021 Karlsruhe (Germany); Spaeh, P.; Vaccaro, A. [Forschungszentrum Karlsruhe, Association FZK-Euratom, Karlsruhe (Germany); Institute for Materials Research I, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2009-06-15

    The ITER electron cyclotron upper port launching system has to provide local current drive in order to stabilize neoclassical tearing modes (NTMs) and the sawtooth instability. The mm-wave system is embedded in a rigid upper port plug structure optimized for the rough operating conditions. During regular operation the structure is heated by neutrons and photons which are simulated by the Monte-Carlo method. The structural system has to provide sufficient cooling especially in the front part where high heat loads occur. If the active plasma stabilization fails the worst case scenario for the upper port plugs is the upward vertical displacement event (VDE) followed by a fast current quench. The fast plasma disruption occurs close to the plugs; eddy currents are induced and interact with the strong static magnetic field. The resulting mechanical forces and torque moments stress the structural system to its physical limits; a detailed numerical analysis, prototype crosschecks and design optimization is required to obtain a working system for ITER. Numerical structural, thermal and fluid dynamic analyses are presented as well as prototype tests and finally the effect of plasma disruptions on the structural design is discussed.

  5. Structural analysis of the ITER vacuum vessel from disruption loading with halo asymmetry

    International Nuclear Information System (INIS)

    Riemer, B.W.; Sayer, R.O.

    1996-01-01

    Static structural analyses of the ITER vacuum vessel were performed with toroidally asymmetric disruption loads. Asymmetric halo current conditions were assumed to modify symmetric disruption loads which resulted in net lateral loading on the vacuum vessel torus. Structural analyses with the asymmetric loading indicated significantly higher vessel stress and blanket support forces than with symmetric disruption loads. A recent change in the vessel support design which provided toroidal constraints at each mid port was found to be effective in reducing torus lateral movement and vessel stress

  6. H-/H∞ structural damage detection filter design using an iterative linear matrix inequality approach

    International Nuclear Information System (INIS)

    Chen, B; Nagarajaiah, S

    2008-01-01

    The existence of damage in different members of a structure can be posed as a fault detection problem. It is also necessary to isolate structural members in which damage exists, which can be posed as a fault isolation problem. It is also important to detect the time instants of occurrence of the faults/damage. The structural damage detection filter developed in this paper is a model-based fault detection and isolation (FDI) observer suitable for detecting and isolating structural damage. In systems, possible faults, disturbances and noise are coupled together. When system disturbances and sensor noise cannot be decoupled from faults/damage, the detection filter needs to be designed to be robust to disturbances as well as sensitive to faults/damage. In this paper, a new H - /H ∞ and iterative linear matrix inequality (LMI) technique is developed and a new stabilizing FDI filter is proposed, which bounds the H ∞ norm of the transfer function from disturbances to the output residual and simultaneously does not degrade the component of the output residual due to damage. The reduced-order error dynamic system is adopted to form bilinear matrix inequalities (BMIs), then an iterative LMI algorithm is developed to solve the BMIs. The numerical example and experimental verification demonstrate that the proposed algorithm can successfully detect and isolate structural damage in the presence of measurement noise

  7. Iterative tensor voting for perceptual grouping of ill-defined curvilinear structures.

    Science.gov (United States)

    Loss, Leandro A; Bebis, George; Parvin, Bahram

    2011-08-01

    In this paper, a novel approach is proposed for perceptual grouping and localization of ill-defined curvilinear structures. Our approach builds upon the tensor voting and the iterative voting frameworks. Its efficacy lies on iterative refinements of curvilinear structures by gradually shifting from an exploratory to an exploitative mode. Such a mode shifting is achieved by reducing the aperture of the tensor voting fields, which is shown to improve curve grouping and inference by enhancing the concentration of the votes over promising, salient structures. The proposed technique is validated on delineating adherens junctions that are imaged through fluorescence microscopy. However, the method is also applicable for screening other organisms based on characteristics of their cell wall structures. Adherens junctions maintain tissue structural integrity and cell-cell interactions. Visually, they exhibit fibrous patterns that may be diffused, heterogeneous in fluorescence intensity, or punctate and frequently perceptual. Besides the application to real data, the proposed method is compared to prior methods on synthetic and annotated real data, showing high precision rates.

  8. Final design of the generic equatorial port plug structure for ITER diagnostic systems

    International Nuclear Information System (INIS)

    Udintsev, V.S.; Maquet, P.; Alexandrov, E.; Casal, N.; Cuenca, D.; Drevon, J.-M.; Feder, R.; Friconneau, J.P.; Giacomin, T.; Guirao, J.; Iglesias, S.; Josseaume, F.; Levesy, B.; Loesser, D.; Ordieres, J.; Quinn, E.; Pak, S.; Penot, C.; Pitcher, C.S.; Portalès, M.

    2015-01-01

    The Diagnostic Generic Equatorial Port Plug (GEPP) is designed to be common to all equatorial port-based diagnostic systems. It is designed to survive throughout the lifetime of ITER for 20 years, 30,000 discharges, and 3000 disruptions. The EPP structure dimensions (without Diagnostic First Walls and Diagnostic Shield Modules) are L2.9 × W1.9 × H2.4 m"3. The length of the fully integrated EPP is 3174 mm. The weight of the EPP structure is about 15 t, whereas the total weight of the integrated EPP may be up to 45 t. The EPP structure provides a flexible platform for a variety of diagnostics. The Diagnostic Shield Module assemblies, or drawers, allow a modular approach with respect to diagnostic integration and maintenance. In the nuclear phase of ITER operations, they will be remotely inserted into the EPP structure in the Hot Cell Facility. The port plug structure must also contribute to the nuclear shielding, or plugging, of the port and further contain circulated water to allow cooling during operation and heating during bake-out. The Final Design of the GEPP has been successfully passed in late 2013 and is now heading toward manufacturing. The final design of the GEPP includes interfaces, manufacturing, R&D, operation and maintenance, load cases and analysis of failure modes.

  9. 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

  10. 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.

  11. Mechanical design and construction qualification program on ITER correction coils structures

    Energy Technology Data Exchange (ETDEWEB)

    Foussat, A., E-mail: arnaud.foussat@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Weiyue, Wu; Jing, Wei; Shuangsong, Du [Academy of Science Institute of Plasma Physics, PO 1126, Hefei, Anhui 230031 (China); Sgobba, S. [European Center for Nuclear Research, CH-1211 Geneva 23 (Switzerland); Hongwei, Li [China International Nuclear Fusion Energy Program Execution Center, Ministry of Science and Technology, 15B Fuxing Rd., Beijing 100862 (China); Libeyre, Paul; Jong, Cornelis; Klofac, Kamil; Mitchell, Neil [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-04-01

    The ITER Magnet system consists of 4 main coils sub-systems, i.e. 18 toroidal field coils (TFC), a central solenoid (CS), 6 poloidal field coils (PF) and 3 sets of correction coils (CC). The ITER fusion project has selected the stainless steel 316LN as main material for the magnet structure. The CC contribute to reducing the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. During plasma operation, a large number of loading condition scenarios have been considered and structural analysis performed on key items like Cable-In-Conduit Conductor and the coil case. The results obtained are used for both static and fatigue structural assessment defining the present baseline design. For the construction of the structural cases, welding techniques such as GTAW (Gas Tungsten Arc Welding) and techniques resulting in low distortion and shrinkage like EBW (Electron Beam Welding) or Laser Beam Welding (LBW) with filler metal wire have been selected. Those methods are considered for future qualifications to guarantee proper weld parameters and specified weld properties. In order to determine the strength and fracture toughness of 316LN stainless steel welds with respect to design criteria, some mechanical tests have been carried out at 7 K (or 77 K), and room temperature.

  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. Structural Analysis for an Upper Port of the ITER Vacuum Vessel

    International Nuclear Information System (INIS)

    Yun-Seok Hong; Kwon, T. K.; Ahn, H. J.; Kim, Y.K.; Lee, C.D.

    2006-01-01

    The ITER vacuum vessel (VV) has numerous openings for the port structures including upper, equatorial, and lower ports used for equipment installation, utility feed through, vacuum pumping, and access into the vessel for maintenance. Every upper port, slanted upward slightly, has a trapezoidal/rectangular cross-section and consists of a port stub, a stub extension and a port extension with a connecting duct. To investigate the structural integrity and to increase the structural reliability of the VV and ports, the structural analyses of the upper port structure have been performed. The global structural analysis of the upper port with the in-port components has been carried out. The local analyses of a tangential key, an upper port flange, a connecting duct and a sealing unit have been performed. The design loads are dead weight, normal and abnormal pressure load, electromagnetic load, and seismic load in consideration of the dynamic amplification factors. The stress analyses were performed in a nonlinear elastic approach taking into account the contact surface between port extension flange and port plug flange. Two advanced designs from the ITER international team have been reviewed. To verify the strength of the reinforcing ribs for the connecting duct and of the fastening/sealing units, the local analyses utilizing the sub-modeling technique have been performed. The ASME code and the ITER design criteria were applied for the evaluation of the structural analysis results from the global and local analyses. The clearance between a port and a plug to accommodate the plug deformation has been assessed. The upper port flange based on the original design could withstand design loads, but there could be a gap on the flange surface under the design condition. The modified flange design, which is under the bolt friction only without tangential key was proposed. The deflection of the plug for an advanced design with a removable flange is higher than that for the original

  14. Structural analysis of the Passive Magnetic Shield for the ITER Heating Neutral Beam Injector system

    Energy Technology Data Exchange (ETDEWEB)

    Cabrera, Santiago, E-mail: santiago.cabrera@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); Rincón, Esther; Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Ramos, Francisco; Ríos, Luis [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); El-Ouazzani, Anass; García, Pablo [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

    2015-10-15

    The ITER Passive Magnetic Shield (PMS) main function is to protect the Neutral Beam Injector (NBI) from the external magnetic field coming from the tokamak, and to shield the NB cell from the radiation coming from all activated components. The shielding from the external magnetic field is performed in association with the Active Compensation Cooled Correction Coils (ACCC). The Bushing and Transmission Line (TL) PMS also provides structural support for HV bushing, allowing its maintenance and providing air sealing function between NBI cell and High Voltage deck room. The paper summarizes the structural analyses performed in order to evaluate the mechanical behaviour of the HNB PMS under operation combined with seismic event. The RCC-MR Code is used to validate the design, assuming creep is negligible, since the structure is expected to be at room temperature. P-type damage is assessed.

  15. Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs

    Energy Technology Data Exchange (ETDEWEB)

    Khaleel, Mohammad A.; Simonen, Fredric A.

    2009-05-01

    The work described in this report applies probabilistic structural mechanics models to predict the reliability of nuclear pressure boundary components. These same models are then applied to evaluate the effectiveness of alternative programs for inservice inspection to reduce these failure probabilities. Results of the calculations support the development and implementation of risk-informed inservice inspection of piping and vessels. Studies have specifically addressed the potential benefits of ultrasonic inspections to reduce failure probabilities associated with fatigue crack growth and stress-corrosion cracking. Parametric calculations were performed with the computer code pc-PRAISE to generate an extensive set of plots to cover a wide range of pipe wall thicknesses, cyclic operating stresses, and inspection strategies. The studies have also addressed critical inputs to fracture mechanics calculations such as the parameters that characterize the number and sizes of fabrication flaws in piping welds. Other calculations quantify uncertainties associated with the inputs calculations, the uncertainties in the fracture mechanics models, and the uncertainties in the resulting calculated failure probabilities. A final set of calculations address the effects of flaw sizing errors on the effectiveness of inservice inspection programs.

  16. Thermal-structural analysis for ITER in-wall shielding block

    International Nuclear Information System (INIS)

    Hao Junchuan; Song Yuntao; Wu Weiyue; Du Shuangsong; Wang, X.; Ioki, K.

    2012-01-01

    Highlights: ► IWS blocks shall withstand various types of mechanical loads including EM loads, inertial loads and thermal loads. ► Due to the complicated geometry, the finite element method is the suitable tool to solve the problem. ► Contact element has been selected to simulate the friction between the different components. ► At baking phase, secondary stresses due to preloading and temperature difference predominate in the total stress. ► At plasma operation phase, secondary stresses due to preloading and thermal loads were deducted from the total stresses. - Abstract: In order to verify the design strength of the in-wall shielding (IWS) blocks of the ITER, thermal-structural analyses of one IWS block under vacuum vessel (VV) baking and plasma operation conditions have been respectively performed with finite element (FE) method. Among the complicated operation scenarios of the ITER, two critical types of combined loads required by the load specification of IWS were applied on the shielding block. The stress of the block is judged by American Society of Mechanical Engineers (ASME) criterion. Results show that the structure of this block has enough safety margin, and it also supplies detailed information of the stress distribution in concerned region under certain loads.

  17. Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

    International Nuclear Information System (INIS)

    Kim, Byoung-Yoon; Ahn, Hee-Jae

    2011-01-01

    The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

  18. A simple iterative independent component analysis algorithm for vibration source signal identification of complex structures

    Directory of Open Access Journals (Sweden)

    Dong-Sup Lee

    2015-01-01

    Full Text Available Independent Component Analysis (ICA, one of the blind source separation methods, can be applied for extracting unknown source signals only from received signals. This is accomplished by finding statistical independence of signal mixtures and has been successfully applied to myriad fields such as medical science, image processing, and numerous others. Nevertheless, there are inherent problems that have been reported when using this technique: insta- bility and invalid ordering of separated signals, particularly when using a conventional ICA technique in vibratory source signal identification of complex structures. In this study, a simple iterative algorithm of the conventional ICA has been proposed to mitigate these problems. The proposed method to extract more stable source signals having valid order includes an iterative and reordering process of extracted mixing matrix to reconstruct finally converged source signals, referring to the magnitudes of correlation coefficients between the intermediately separated signals and the signals measured on or nearby sources. In order to review the problems of the conventional ICA technique and to vali- date the proposed method, numerical analyses have been carried out for a virtual response model and a 30 m class submarine model. Moreover, in order to investigate applicability of the proposed method to real problem of complex structure, an experiment has been carried out for a scaled submarine mockup. The results show that the proposed method could resolve the inherent problems of a conventional ICA technique.

  19. Numerical modeling of 3D halo current path in ITER structures

    Energy Technology Data Exchange (ETDEWEB)

    Bettini, Paolo; Marconato, Nicolò; Furno Palumbo, Maurizio; Peruzzo, Simone [Consorzio RFX, EURATOM-ENEA Association, C.so Stati Uniti 4, 35127 Padova (Italy); Specogna, Ruben, E-mail: ruben.specogna@uniud.it [DIEGM, Università di Udine, Via delle Scienze, 208, 33100 Udine (Italy); Albanese, Raffaele; Rubinacci, Guglielmo; Ventre, Salvatore; Villone, Fabio [Consorzio CREATE, EURATOM-ENEA Association, Via Claudio 21, 80125 Napoli (Italy)

    2013-10-15

    Highlights: ► Two numerical codes for the evaluation of halo currents in 3D structures are presented. ► A simplified plasma model is adopted to provide the input (halo current injected into the FW). ► Two representative test cases of ITER symmetric and asymmetric VDEs have been analyzed. ► The proposed approaches provide results in excellent agreement for both cases. -- Abstract: Disruptions represent one of the main concerns for Tokamak operation, especially in view of fusion reactors, or experimental test reactors, due to the electro-mechanical loads induced by halo and eddy currents. The development of a predictive tool which allows to estimate the magnitude and spatial distribution of the halo current forces is of paramount importance in order to ensure robust vessel and in-vessel component design. With this aim, two numerical codes (CARIDDI, CAFE) have been developed, which allow to calculate the halo current path (resistive distribution) in the passive structures surrounding the plasma. The former is based on an integral formulation for the eddy currents problem particularized to the static case; the latter implements a pair of 3D FEM complementary formulations for the solution of the steady-state current conduction problem. A simplified plasma model is adopted to provide the inputs (halo current injected into the first wall). Two representative test cases (ITER symmetric and asymmetric VDEs) have been selected to cross check the results of the proposed approaches.

  20. Improved understanding of physics processes in pedestal structure, leading to improved predictive capability for ITER

    International Nuclear Information System (INIS)

    Groebner, R.J.; Snyder, P.B.; Leonard, A.W.; Chang, C.S.; Maingi, R.; Boyle, D.P.; Diallo, A.; Hughes, J.W.; Davis, E.M.; Ernst, D.R.; Landreman, M.; Xu, X.Q.; Boedo, J.A.; Cziegler, I.; Diamond, P.H.; Eldon, D.P.; Callen, J.D.; Canik, J.M.; Elder, J.D.; Fulton, D.P.

    2013-01-01

    Joint experiment/theory/modelling research has led to increased confidence in predictions of the pedestal height in ITER. This work was performed as part of a US Department of Energy Joint Research Target in FY11 to identify physics processes that control the H-mode pedestal structure. The study included experiments on C-Mod, DIII-D and NSTX as well as interpretation of experimental data with theory-based modelling codes. This work provides increased confidence in the ability of models for peeling–ballooning stability, bootstrap current, pedestal width and pedestal height scaling to make correct predictions, with some areas needing further work also being identified. A model for pedestal pressure height has made good predictions in existing machines for a range in pressure of a factor of 20. This provides a solid basis for predicting the maximum pedestal pressure height in ITER, which is found to be an extrapolation of a factor of 3 beyond the existing data set. Models were studied for a number of processes that are proposed to play a role in the pedestal n e and T e profiles. These processes include neoclassical transport, paleoclassical transport, electron temperature gradient turbulence and neutral fuelling. All of these processes may be important, with the importance being dependent on the plasma regime. Studies with several electromagnetic gyrokinetic codes show that the gradients in and on top of the pedestal can drive a number of instabilities. (paper)

  1. The insulation structure of the 1 MV transmission line for the ITER neutral beam injector

    International Nuclear Information System (INIS)

    De Lorenzi, A.; Grando, L.; Gobbo, R.; Pesavento, G.; Bettini, P.; Specogna, R.; Trevisan, F.

    2007-01-01

    The paper describes the studies and the tests for the development of the insulation structure of the 1 MV-50 A gas insulated (SF 6 ) line of the ITER NBI in the SinGap configuration characterized by two kinds of spacers: at least a couple of disk-shaped spacers, designed to be gas tight, and a larger number (several tens) of inner conductor post spacers. To this aim a test campaign has been carried out to assess the capability of standard epoxy spacers to withstand a high dc voltage with frequent short circuits, simulating the operational condition for the ITER NBI. Two computational tools, the first for the epoxy spacer shape optimization under electrostatic distribution and the other for the nonlinear time variant evolution of the electric field and surface charge, have been developed specifically for designing epoxy spacer under dc voltage stress. The results on the optimization of the disk spacer and on the electric field-surface charge time evolution of the post spacer are reported and discussed. The effects of the SF 6 radiation induced conductivity on the post spacer are also reported

  2. The insulation structure of the 1 MV transmission line for the ITER neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    De Lorenzi, A. [Consorzio RFX, Associazione Euratom-Enea sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy)], E-mail: antonio.delorenzi@igi.cnr.it; Grando, L. [Consorzio RFX, Associazione Euratom-Enea sulla Fusione, Corso Stati Uniti 4, I-35127 Padova (Italy); Gobbo, R.; Pesavento, G. [DIE, Universita di Padova, Via Gradenigo 6A, I-35100 Padova (Italy); Bettini, P.; Specogna, R.; Trevisan, F. [DIEGM, Universita di Udine, Via delle Scienze 208, I-33100 Udine (Italy)

    2007-10-15

    The paper describes the studies and the tests for the development of the insulation structure of the 1 MV-50 A gas insulated (SF{sub 6}) line of the ITER NBI in the SinGap configuration characterized by two kinds of spacers: at least a couple of disk-shaped spacers, designed to be gas tight, and a larger number (several tens) of inner conductor post spacers. To this aim a test campaign has been carried out to assess the capability of standard epoxy spacers to withstand a high dc voltage with frequent short circuits, simulating the operational condition for the ITER NBI. Two computational tools, the first for the epoxy spacer shape optimization under electrostatic distribution and the other for the nonlinear time variant evolution of the electric field and surface charge, have been developed specifically for designing epoxy spacer under dc voltage stress. The results on the optimization of the disk spacer and on the electric field-surface charge time evolution of the post spacer are reported and discussed. The effects of the SF{sub 6} radiation induced conductivity on the post spacer are also reported.

  3. Conceptual design on structure and cooling channel of ITER upper port plug

    International Nuclear Information System (INIS)

    Pak, Sunil; Lee, Hyeon Gon; Jung, Ki Jung; Walker, C.; Kim, Doo Gi; Choi, Kwang Suk; Eo, Sang Gon

    2007-01-01

    This study has performed conceptual design on structure and cooling channel for the upper port plug of the International Thermonuclear Experimental Reactor (ITER), in which electron cyclotron heating (ECH) launcher and various diagnostic modules will be installed with the same structure. There are twelve diagnostic plugs and four ECH plugs at the upper port in ITER Tokamak. The use of the same port plug structure is beneficial for installation of diagnostic modules and ECH launcher from the viewpoint of cost reduction and simple RH maintenance. The diagnostic modules have rectangular cross-section and ECH modules have trapezoidal crosssection with the lower part wider. Here was suggested the bolt-jointed common structure of inverted-U shape beam and bottom plate, where the diagnostic and ECH modules are installed onto the bottom plate and then the assembly is bolted to the inverted-U beam from the bottom. The common structure of Inverted-U type was evaluated by considering several aspects, such as installation, remote handling (RH) maintenance, cooling line connection, manufacturing, and structural stiffness. For the inverted-U port plug structure developed here, this paper proposed a network of water channel for cooling and baking. Pressurized water as working fluid has to be supplied into the whole port plug. It consists of the structure, diagnostic/shielding modules fixed onto the bottom plate, and the blanket shield module (BSM) attached to the front. The internal water ways for these three components were designed in the direction that would not only minimize the RH connections, flow restrictors, and the length of water-vacuum welding, but also make the welding reliable. Independent coolant loops were composed for three parts of the structure, BSM, and diagnostic/shielding modules with bottom plate. These loops, therefore, make it possible to perform the leakage test for each one separately. Finally hydraulic analysis has been performed with ANSYS in order to

  4. ITER structural design criteria and their extension to advanced reactor blankets

    International Nuclear Information System (INIS)

    Majumdar, S.; Kalinin, G.

    2000-01-01

    Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method

  5. Newton`s iteration for inversion of Cauchy-like and other structured matrices

    Energy Technology Data Exchange (ETDEWEB)

    Pan, V.Y. [Lehman College, Bronx, NY (United States); Zheng, Ailong; Huang, Xiaohan; Dias, O. [CUNY, New York, NY (United States)

    1996-12-31

    We specify some initial assumptions that guarantee rapid refinement of a rough initial approximation to the inverse of a Cauchy-like matrix, by mean of our new modification of Newton`s iteration, where the input, output, and all the auxiliary matrices are represented with their short generators defined by the associated scaling operators. The computations are performed fast since they are confined to operations with short generators of the given and computed matrices. Because of the known correlations among various structured matrices, the algorithm is immediately extended to rapid refinement of rough initial approximations to the inverses of Vandermonde-like, Chebyshev-Vandermonde-like and Toeplitz-like matrices, where again, the computations are confined to operations with short generators of the involved matrices.

  6. Final design of the generic upper port plug structure for ITER diagnostic systems

    Energy Technology Data Exchange (ETDEWEB)

    Pak, Sunil, E-mail: paksunil@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Feder, Russell [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Giacomin, Thibaud; Guirao, Julio; Iglesias, Silvia; Josseaume, Fabien [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kalish, Michael; Loesser, Douglas [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Maquet, Philippe [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Ordieres, Javier; Panizo, Marcos [NATEC, Ingenieros, Gijón (Spain); Pitcher, Spencer; Portalès, Mickael [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Proust, Maxime [CEA, Cadarache, St. Paul-lez-Durance (France); Ronden, Dennis [FOM Institute DIFFER, Nieuwegein (Netherlands); Serikov, Arkady [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany); Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Tanchuk, Victor [NIIEFA, St.-Petersburg (Russian Federation); Udintsev, Victor; Vacas, Christian [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

    2016-01-15

    The generic upper port plug (GUPP) structure in ITER is a 6 m long metal box which deploys diagnostic components into the vacuum vessel. This structure is commonly used for all the diagnostic upper ports. The final design of the GUPP structure, which has successfully passed the final design review in 2013, is described here. The diagnostic port plug is cantilevered to the vacuum vessel with a heavy payload at the front, so called the diagnostic first wall (DFW) and the diagnostic shield module (DSM). Most of electromagnetic (EM) load (∼80%) occurs in DFW/DSM. Therefore, the mounting design to transfer the EM load from DFW/DSM to the GUPP structure is challenging, which should also comply with thermal expansion and tolerance for assembly and manufacturing. Another key design parameter to be considered is the gap between the port plug and the vacuum vessel port. The gap should be large enough to accommodate the remote handling of the heavy port plug (max. 25 t), the structural deflection due to external loads and machine assembly tolerance. At the same time, the gap should be minimized to stop the neutron streaming according to the ALARA (as low as reasonably achievable) principle. With these design constraints, the GUPP structure should also provide space for diagnostic integration as much as possible. This requirement has led to the single wall structure having the gun-drilled water channels inside the structure. Furthermore, intensive efforts have been made on the manufacturing study including material selection, manufacturing codes and French regulation related to nuclear equipment and safety. All these main design and manufacturing aspects are discussed in this paper, including requirements, interfaces, loads and structural assessment and maintenance.

  7. 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.)

  8. Iterative model building, structure refinement and density modification with the PHENIX AutoBuild wizard

    International Nuclear Information System (INIS)

    Terwilliger, Thomas C.; Grosse-Kunstleve, Ralf W.; Afonine, Pavel V.; Moriarty, Nigel W.; Zwart, Peter H.; Hung, Li-Wei; Read, Randy J.; Adams, Paul D.

    2008-01-01

    The highly automated PHENIX AutoBuild wizard is described. The procedure can be applied equally well to phases derived from isomorphous/anomalous and molecular-replacement methods. The PHENIX AutoBuild wizard is a highly automated tool for iterative model building, structure refinement and density modification using RESOLVE model building, RESOLVE statistical density modification and phenix.refine structure refinement. Recent advances in the AutoBuild wizard and phenix.refine include automated detection and application of NCS from models as they are built, extensive model-completion algorithms and automated solvent-molecule picking. Model-completion algorithms in the AutoBuild wizard include loop building, crossovers between chains in different models of a structure and side-chain optimization. The AutoBuild wizard has been applied to a set of 48 structures at resolutions ranging from 1.1 to 3.2 Å, resulting in a mean R factor of 0.24 and a mean free R factor of 0.29. The R factor of the final model is dependent on the quality of the starting electron density and is relatively independent of resolution

  9. The feasibility of beryllium as structural material for the ITER plasma-facing components (PFC)

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Gorenflo, H.

    1993-01-01

    Be as plasma-facing armour has attractive features including excellent plasma compatibility, no T-retention via co-deposition and the potential for in-situ repair via plasma spraying. In order to avoid the bonding of the Be-armour to a heatsink structure in e.g., Cu-alloys, the ITER Joint Central Team (JCT) proposed for the divertor tubular elements with monolithic Be, both as plasma-facing and structural material. The analysis of these Be-tubes with 5 mm wall thickness at a heat load of 5 MW/m 2 showed that even for the most favourable assumptions thermal stresses exceed by far the allowed values according to design codes. Damage by neutrons and disruptions would worsen further the case for Be as monolithic plasma-facing and structural material. For PFC at heat flux significantly above 1 MW/m 2 it appears evident that Be should be used merely as armour bonded to a suitable structural material as heatsink. (orig.)

  10. Iterative model-building, structure refinement, and density modification with the PHENIX AutoBuild Wizard

    Energy Technology Data Exchange (ETDEWEB)

    Los Alamos National Laboratory, Mailstop M888, Los Alamos, NM 87545, USA; Lawrence Berkeley National Laboratory, One Cyclotron Road, Building 64R0121, Berkeley, CA 94720, USA; Department of Haematology, University of Cambridge, Cambridge CB2 0XY, England; Terwilliger, Thomas; Terwilliger, T.C.; Grosse-Kunstleve, Ralf Wilhelm; Afonine, P.V.; Moriarty, N.W.; Zwart, P.H.; Hung, L.-W.; Read, R.J.; Adams, P.D.

    2007-04-29

    The PHENIX AutoBuild Wizard is a highly automated tool for iterative model-building, structure refinement and density modification using RESOLVE or TEXTAL model-building, RESOLVE statistical density modification, and phenix.refine structure refinement. Recent advances in the AutoBuild Wizard and phenix.refine include automated detection and application of NCS from models as they are built, extensive model completion algorithms, and automated solvent molecule picking. Model completion algorithms in the AutoBuild Wizard include loop-building, crossovers between chains in different models of a structure, and side-chain optimization. The AutoBuild Wizard has been applied to a set of 48 structures at resolutions ranging from 1.1 {angstrom} to 3.2 {angstrom}, resulting in a mean R-factor of 0.24 and a mean free R factor of 0.29. The R-factor of the final model is dependent on the quality of the starting electron density, and relatively independent of resolution.

  11. Thermal–structural analysis of ITER triangular support for dominant load verification

    International Nuclear Information System (INIS)

    Kim, Yu-Gyeong; Hwang, Jong-Hyun; Jung, Yung-Jin; Kim, Hyun-Soo; Ahn, Hee-Jae

    2014-01-01

    Highlights: • The load combination method is introduced to thermal–structural analysis for contradictive loads occurred simultaneously. • The one-way coupling analysis also conducted for thermal–structural analysis and its validity is checked by comparing with the load combination. • The dominant load for triangular support bracket is determined as the baking condition. - Abstract: The triangular support is located on the lower inner shell of vacuum vessel of ITER, which should be designed to withstand various loads such as nuclear heat, coolant pressure and so on. The appropriateness of its design is evaluated under the dominant load that could represent the most conservative condition among the design loads. In order to decide the dominant load, a valid method for thermal–structural analysis is firstly verified considering contradictory behaviors between heat and structural loads. In this paper, two approaches; one-way coupling and load combination, are introduced for thermal–structural analysis. The one-way coupling is a method generally used but has a limit to apply on contradictory conditions. The load combination could give a proper solution since it evaluates each load independently and then adds up each result linearly. Based on the results of each case, structural analysis for another load case, baking condition with incident, is conducted to find out which load is dominant for triangular support. Consequently, it is found that the baking condition is the dominant load for triangular support bracket. The proposed load combination method gives a physically reasonable solution which can be used as a reference for checking the validity of other thermal–structural analysis. It is expected that these results could be applied for manufacturing design of the triangular support under various load conditions

  12. Thermal–structural analysis of ITER triangular support for dominant load verification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yu-Gyeong, E-mail: aspirany@hhi.co.kr [Hyundai Heavy Industries Co., Ltd., 1000, Bangeojinsunhwando-ro, Dong-gu, Ulsan (Korea, Republic of); Hwang, Jong-Hyun; Jung, Yung-Jin [Hyundai Heavy Industries Co., Ltd., 1000, Bangeojinsunhwando-ro, Dong-gu, Ulsan (Korea, Republic of); Kim, Hyun-Soo; Ahn, Hee-Jae [National Fusion Research Institute, 113 Gwahangno, Yuseong-gu, Daejeon-si (Korea, Republic of)

    2014-12-15

    Highlights: • The load combination method is introduced to thermal–structural analysis for contradictive loads occurred simultaneously. • The one-way coupling analysis also conducted for thermal–structural analysis and its validity is checked by comparing with the load combination. • The dominant load for triangular support bracket is determined as the baking condition. - Abstract: The triangular support is located on the lower inner shell of vacuum vessel of ITER, which should be designed to withstand various loads such as nuclear heat, coolant pressure and so on. The appropriateness of its design is evaluated under the dominant load that could represent the most conservative condition among the design loads. In order to decide the dominant load, a valid method for thermal–structural analysis is firstly verified considering contradictory behaviors between heat and structural loads. In this paper, two approaches; one-way coupling and load combination, are introduced for thermal–structural analysis. The one-way coupling is a method generally used but has a limit to apply on contradictory conditions. The load combination could give a proper solution since it evaluates each load independently and then adds up each result linearly. Based on the results of each case, structural analysis for another load case, baking condition with incident, is conducted to find out which load is dominant for triangular support. Consequently, it is found that the baking condition is the dominant load for triangular support bracket. The proposed load combination method gives a physically reasonable solution which can be used as a reference for checking the validity of other thermal–structural analysis. It is expected that these results could be applied for manufacturing design of the triangular support under various load conditions.

  13. Basic study on weldability and machinability of structural materials for ITER toroidal field coils

    International Nuclear Information System (INIS)

    Onozuka, M.; Shimizu, K.; Urata, K.; Kimura, M.; Kadowaki, H.; Okamoto, M.; Nakajima, H.; Hamada, K.; Okuno, K.

    2006-01-01

    The toroidal field (TF) coils for ITER are very large components. The main structural component of the coil is the coil case, which requires a massive complex geometry with high fabrication accuracy to attain the required magnetic performance for plasma operations. To provide high mechanical strength and toughness at cryogenic temperature, the structural components employ high-strength austenite stainless steels that have been specially developed for ITER. However, one of the main drawbacks of using those materials is the difficulty of manufacturing capabilities. A manufacturing study has been conducted to examine welding and machining capabilities for JJ1 and ST-SS316LN, to be employed for TF coil structural components. Both materials include a high nitrogen content up to around 0.2%, which makes welding and machining difficult compared with conventional stainless steels. Electron beam welding conditions were studied for the JJ1 material. The applicable welding condition was found for a bead length of up to about 300 mm in the case of 40 mm thick plates. No optimal condition was found for plates thicker than 40 mm. An additional experimental study was also conducted to explore suitable welding conditions for different welding positions and directions. It was found that the appearance of defects depends on the welding positions and directions. A wider range of welding conditions was found for cases in the vertical upward direction, as opposed to those in the vertical downward and horizontal directions. Based on those results, a verification test up to 900 mm in length was conducted. The test results showed that vertical upward EB welding should be used for the coil case wherever possible. With respect to TIG welding, an average deposition rate as high as 26 g/min (i.e. the filler wire supplying speed of 3,000 mm/min) was achieved. A series of tests have been conducted to examine machinability of JJ1 and ST-SS316LN. Various types of milling tools, including face

  14. Theoretical predictor for candidate structure assignment from IMS data of biomolecule-related conformational space.

    Science.gov (United States)

    Schenk, Emily R; Nau, Frederic; Fernandez-Lima, Francisco

    2015-06-01

    The ability to correlate experimental ion mobility data with candidate structures from theoretical modeling provides a powerful analytical and structural tool for the characterization of biomolecules. In the present paper, a theoretical workflow is described to generate and assign candidate structures for experimental trapped ion mobility and H/D exchange (HDX-TIMS-MS) data following molecular dynamics simulations and statistical filtering. The applicability of the theoretical predictor is illustrated for a peptide and protein example with multiple conformations and kinetic intermediates. The described methodology yields a low computational cost and a simple workflow by incorporating statistical filtering and molecular dynamics simulations. The workflow can be adapted to different IMS scenarios and CCS calculators for a more accurate description of the IMS experimental conditions. For the case of the HDX-TIMS-MS experiments, molecular dynamics in the "TIMS box" accounts for a better sampling of the molecular intermediates and local energy minima.

  15. Upper-Lower Bounds Candidate Sets Searching Algorithm for Bayesian Network Structure Learning

    Directory of Open Access Journals (Sweden)

    Guangyi Liu

    2014-01-01

    Full Text Available Bayesian network is an important theoretical model in artificial intelligence field and also a powerful tool for processing uncertainty issues. Considering the slow convergence speed of current Bayesian network structure learning algorithms, a fast hybrid learning method is proposed in this paper. We start with further analysis of information provided by low-order conditional independence testing, and then two methods are given for constructing graph model of network, which is theoretically proved to be upper and lower bounds of the structure space of target network, so that candidate sets are given as a result; after that a search and scoring algorithm is operated based on the candidate sets to find the final structure of the network. Simulation results show that the algorithm proposed in this paper is more efficient than similar algorithms with the same learning precision.

  16. 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

  17. Basic analysis of weldability and machinability of structural materials for ITER Toroidal Field coils

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori [Mitsubishi Heavy Industries Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)], E-mail: masanori_onozuka@mnes-us.com; Shimizu, Katsusuke; Urata, Kazuhiro; Kimura, Masahiro; Kadowaki, Hirokazu; Okamoto, Mamoru [Mitsubishi Heavy Industries Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Nakajima, Hideo; Hamada, Kazuya; Okuno, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-shi, Ibaraki 311-0193 (Japan)

    2007-10-15

    A manufacturing study has been conducted to examine the welding and machining capabilities for strengthened austenitic stainless steels with a high nitrogen content, JJ1 and ST-SS316LN, to be employed for ITER Toroidal Field (TF) coil structural components. It was found that the applicable EB welding condition for JJ1 was limited to up to 40 mm thick plates. A wider range of welding conditions was found in the vertical upward direction. Based on those results, a verification test up to 900 mm in length was successfully conducted. With respect to TIG welding, an average deposition rate of 26 g/min (i.e. the filler wire supplying speed of 3000 mm/min) was achieved. In addition to the welding tests, a series of machining tests has been conducted to examine the machinability of JJ1 and ST-SS316LN. Various types of machining tools were examined. In practical application, the cutting speed should be low to extend the tool life. At a cutting speed of 40 m/min, a tool life of more than 2 h (at a traveling distance of up to 9 m) was attained. The degree of cutter wear after 30 min of operation, at a cutting speed of 40 m/min, was found to be around 0.1 mm, which is within an acceptable range.

  18. MAP: an iterative experimental design methodology for the optimization of catalytic search space structure modeling.

    Science.gov (United States)

    Baumes, Laurent A

    2006-01-01

    One of the main problems in high-throughput research for materials is still the design of experiments. At early stages of discovery programs, purely exploratory methodologies coupled with fast screening tools should be employed. This should lead to opportunities to find unexpected catalytic results and identify the "groups" of catalyst outputs, providing well-defined boundaries for future optimizations. However, very few new papers deal with strategies that guide exploratory studies. Mostly, traditional designs, homogeneous covering, or simple random samplings are exploited. Typical catalytic output distributions exhibit unbalanced datasets for which an efficient learning is hardly carried out, and interesting but rare classes are usually unrecognized. Here is suggested a new iterative algorithm for the characterization of the search space structure, working independently of learning processes. It enhances recognition rates by transferring catalysts to be screened from "performance-stable" space zones to "unsteady" ones which necessitate more experiments to be well-modeled. The evaluation of new algorithm attempts through benchmarks is compulsory due to the lack of past proofs about their efficiency. The method is detailed and thoroughly tested with mathematical functions exhibiting different levels of complexity. The strategy is not only empirically evaluated, the effect or efficiency of sampling on future Machine Learning performances is also quantified. The minimum sample size required by the algorithm for being statistically discriminated from simple random sampling is investigated.

  19. Validation and implementation of sandwich structure bottom plate to rib weld joint in the base section of ITER Cryostat

    Energy Technology Data Exchange (ETDEWEB)

    Prajapati, Rajnikant, E-mail: rajnikant@iter-india.org [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Bhardwaj, Anil K.; Gupta, Girish; Joshi, Vaibhav; Patel, Mitul; Bhavsar, Jagrut; More, Vipul; Jindal, Mukesh; Bhattacharya, Avik; Jogi, Gaurav; Palaliya, Amit; Jha, Saroj; Pandey, Manish [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Jadhav, Pandurang; Desai, Hemal [Larsen & Toubro Limited, Heavy Engineering, Hazira Manufacturing Complex, Gujarat (India)

    2016-11-01

    Highlights: • ITER Cryostat base section sandwich structure bottom plate to rib weld joint is qualified through mock-up. • Established welding sequence was successfully implemented on all six sectors of cryostat base section. • Each layer liquid penetrant examination has been carried out for these weld joints and found satisfactory. - Abstract: Cryostat is a large stainless steel vacuum vessel providing vacuum environment to ITER machine components. The cryostat is ∼30 m in diameter and ∼30 m in height having variable thickness from 25 mm to 180 mm. Sandwich structure of cryostat base section withstands vacuum loading and limits the deformation under service conditions. Sandwich structure consists of top and bottom plates internally strengthened with radial and circular ribs. In current work, sandwich structure bottom plate to rib weld joint has been designed with full penetration joint as per ITER Vacuum Handbook requirement considering nondestructive examinations and welding feasibility. Since this joint was outside the scope of ASME Section VIII Div. 2, it was decided to validate through mock-up of bottom plate to rib joint. Welding sequence was established to control the distortion. Tensile test, macro-structural examination and layer by layer LPE were carried out for validation of this weld joint. However possibility of ultrasonic examination method was also investigated. The test results from the welded joint mock-up were found to confirm all code and specification requirements. The same was implemented in first sector (0–60°) of base section sandwich structure.

  20. Perl Modules for Constructing Iterators

    Science.gov (United States)

    Tilmes, Curt

    2009-01-01

    The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

  1. The Insulation Structure of the 1 Megavolt Transmission Line for the ITER Neutral Beam Injector

    International Nuclear Information System (INIS)

    Lorenzi, A. de; Grando, L.; Gobbo, R; Pesavento, G.; Bettini, P.; Specogna, R.; Trevisan, F.

    2006-01-01

    The paper describes the studies and the tests for the development of the insulation structure of the 1 MV-50 A Gas Insulated (SF 6 ) Line of the ITER NBI in the SINGAP configuration. Basically, the Transmission Line for the SinGap configuration consists of a coaxial conductor with the inner electrode polarized at the negative value of 1 MV dc. Despite Transmission Line belongs to the Gas Insulated Switchgear (GIS) family, whose technology is nowadays mature, the design of spacers (usually epoxy post and conical type) for High Voltage DC with the same degree of reliability of spacer for High Voltage AC is far away to be attained, due to the substantial difference between AC and DC voltage electric field configuration and insulation behavior at Epoxy-SF 6 interface. Furthermore, the occurrence of very frequent short circuits during the NBI operation introduces further elements of uncertainty in the design of the insulation structure. The first step was then to assess the capability of standard epoxy spacers to withstand DC voltage with frequent short circuits, in order to determine if the choice of the material is suitable for such an application; a test circuit was set up, with the possibility to generate 200 kV DC and to produce a short-circuit every 15'. The system with the spacers was pressurized with SF 6 at 0.3 MPa. For the interpretation of the experimental results, the spacer has been modeled with commercial (ANSYS TM FEM) and research (Cell Method) numerical codes in order to evaluate both the capacitive and resistive electric field distributions. Once assessed the possibility of using epoxy compounds for the spacer construction, the post and cone spacers have been designed taking into consideration various cases, like electrostatic field configuration, resistive distribution for homogeneous and skin conductivity of the spacer, and for high SF 6 conductivity. In these cases the optimization of the triple point screening has been evaluated, leading to

  2. Discovery of Novel Leptospirosis Vaccine Candidates Using Reverse and Structural Vaccinology

    Directory of Open Access Journals (Sweden)

    Alan John Alexander McBride

    2017-04-01

    Full Text Available Leptospira spp. are diderm (two membranes bacteria that infect mammals causing leptospirosis, a public health problem with global implications. Thousands of people die every year due to leptospirosis, especially in developing countries with tropical climates. Prophylaxis is difficult due to multiple factors, including the large number of asymptomatic hosts that transmit the bacteria, poor sanitation, increasing numbers of slum dwellers, and the lack of an effective vaccine. Several leptospiral recombinant antigens were evaluated as a replacement for the inactivated (bacterin vaccine; however, success has been limited. A prospective vaccine candidate is likely to be a surface-related protein that can stimulate the host immune response to clear leptospires from blood and organs. In this study, a comprehensive bioinformatics approach based on reverse and structural vaccinology was applied toward the discovery of novel leptospiral vaccine candidates. The Leptospira interrogans serovar Copenhageni strain L1-130 genome was mined in silico for the enhanced identification of conserved β-barrel (βb transmembrane proteins and outer membrane (OM lipoproteins. Orthologs of the prospective vaccine candidates were screened in the genomes of 20 additional Leptospira spp. Three-dimensional structural models, with a high degree of confidence, were created for each of the surface-exposed proteins. Major histocompatibility complex II (MHC-II epitopes were identified, and their locations were mapped on the structural models. A total of 18 βb transmembrane proteins and 8 OM lipoproteins were identified. These proteins were conserved among the pathogenic Leptospira spp. and were predicted to have epitopes for several variants of MHC-II receptors. A structural and functional analysis of the sequence of these surface proteins demonstrated that most βb transmembrane proteins seem to be TonB-dependent receptors associated with transportation. Other proteins

  3. Investigation of Mathematics Teacher Candidates' Conceptual Structures about "Measurement" through Word Association Test: The Example of Turkey

    Science.gov (United States)

    Erdogan, Ahmet

    2017-01-01

    The purpose of this research is to determine mathematics teacher candidates' conceptual structures about the concept of "measurement" that is the one of the important learning fields of mathematics. Qualitative research method was used in this study. Participants of this study were 58 mathematics teacher candidates studying in one of the…

  4. Investigation of Physics Teacher Candidates' Cognitive Structures about "Electric Field": A Free Word Association Test Study

    Science.gov (United States)

    Türkkan, Ercan

    2017-01-01

    The aim of this study is to investigate the cognitive structures of physics teacher candidates about "electric field." Phenomenographic research method, one of the qualitative research patterns, was used in the study. The data of the study was collected from 91 physics teacher candidates who had taken General Physics II course at…

  5. Corrosion of candidate iron-base waste package structural barrier materials in moist salt environments

    International Nuclear Information System (INIS)

    Westerman, R.E.; Pitman, S.G.

    1984-11-01

    Mild steels are considered to be strong candidates for waste package structural barrier (e.g., overpack) applications in salt repositories. Corrosion rates of these materials determined in autoclave tests utilizing a simulated intrusion brine based on Permian Basin core samples are low, generally <25 μm (1 mil) per year. When the steels are exposed to moist salts containing simulated inclusion brines, the corrosion rates are found to increase significantly. The magnesium in the inclusion brine component of the environment is believed to be responsible for the increased corrosion rates. 1 reference, 4 figures, 2 tables

  6. Identification of candidate structured RNAs in the marine organism 'Candidatus Pelagibacter ubique'

    Directory of Open Access Journals (Sweden)

    Schwalbach Michael S

    2009-06-01

    Full Text Available Abstract Background Metagenomic sequence data are proving to be a vast resource for the discovery of biological components. Yet analysis of this data to identify functional RNAs lags behind efforts to characterize protein diversity. The genome of 'Candidatus Pelagibacter ubique' HTCC 1062 is the closest match for approximately 20% of marine metagenomic sequence reads. It is also small, contains little non-coding DNA, and has strikingly low GC content. Results To aid the discovery of RNA motifs within the marine metagenome we exploited the genomic properties of 'Cand. P. ubique' by targeting our search to long intergenic regions (IGRs with relatively high GC content. Analysis of known RNAs (rRNA, tRNA, riboswitches etc. shows that structured RNAs are significantly enriched in such IGRs. To identify additional candidate structured RNAs, we examined other IGRs with similar characteristics from 'Cand. P. ubique' using comparative genomics approaches in conjunction with marine metagenomic data. Employing this strategy, we discovered four candidate structured RNAs including a new riboswitch class as well as three additional likely cis-regulatory elements that precede genes encoding ribosomal proteins S2 and S12, and the cytoplasmic protein component of the signal recognition particle. We also describe four additional potential RNA motifs with few or no examples occurring outside the metagenomic data. Conclusion This work begins the process of identifying functional RNA motifs present in the metagenomic data and illustrates how existing completed genomes may be used to aid in this task.

  7. 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

  8. Rapid discovery of peptide capture candidates with demonstrated specificity for structurally similar toxins

    Science.gov (United States)

    Sarkes, Deborah A.; Hurley, Margaret M.; Coppock, Matthew B.; Farrell, Mikella E.; Pellegrino, Paul M.; Stratis-Cullum, Dimitra N.

    2016-05-01

    Peptides have emerged as viable alternatives to antibodies for molecular-based sensing due to their similarity in recognition ability despite their relative structural simplicity. Various methods for peptide capture reagent discovery exist, including phage display, yeast display, and bacterial display. One of the primary advantages of peptide discovery by bacterial display technology is the speed to candidate peptide capture agent, due to both rapid growth of bacteria and direct utilization of the sorted cells displaying each individual peptide for the subsequent round of biopanning. We have previously isolated peptide affinity reagents towards protective antigen of Bacillus anthracis using a commercially available automated magnetic sorting platform with improved enrichment as compared to manual magnetic sorting. In this work, we focus on adapting our automated biopanning method to a more challenging sort, to demonstrate the specificity possible with peptide capture agents. This was achieved using non-toxic, recombinant variants of ricin and abrin, RiVax and abrax, respectively, which are structurally similar Type II ribosomal inactivating proteins with significant sequence homology. After only two rounds of biopanning, enrichment of peptide capture candidates binding abrax but not RiVax was achieved as demonstrated by Fluorescence Activated Cell Sorting (FACS) studies. Further sorting optimization included negative sorting against RiVax, proper selection of autoMACS programs for specific sorting rounds, and using freshly made buffer and freshly thawed protein target for each round of biopanning for continued enrichment over all four rounds. Most of the resulting candidates from biopanning for abrax binding peptides were able to bind abrax but not RiVax, demonstrating that short peptide sequences can be highly specific even at this early discovery stage.

  9. Structure and interactions of a malarial vaccine candidate, AMA1, form the parasite plasmodium falciparum

    International Nuclear Information System (INIS)

    Miles, L.A.; Keizer, D.W.; Hodder, A.N.; Nair, M.; Hinds, M.G.; Norton, R.S.; Li, F.; Foley, M.; Coley, A.; Anders, R.F.

    2001-01-01

    Full text: Apical membrane antigen 1 (AMA1), a merozoite surface protein found in all species of Plasmodium and other apicomplexan parasites, is a strong candidate for inclusion in a malarial vaccine. Recombinant AMA1 protected against P. fragile in monkeys and P. chabaudi adami in mice. P. falciparum AMA1 which has a 62-kDa ectodomain consisting of three disulphide-stabilised domains, is a target of antibodies that inhibit merozoite invasion in vitro. Here we describe the solution structure of domain III (14 kDa), determined by NMR on 15 N- and 13 C/ 15 N-labelled samples. It has a well-defined disulphide-stabilised core interrupted by a disordered loop, and both the N- and C-terminal regions of the molecule are unstructured. The structured region includes all three disulphide bonds. Naturally-occurring mutations across 11 different P falciparum strains that are located far apart in the sequence cluster around the disulphide core in the 3D structure of domain III, suggesting that this region contains the major epitopes recognised by neutralising antibodies. Consistent with this, the disulphide-bond stabilised conformation of the ectodomain was essential for protection, as the antigen was not an effective vaccine after reduction and alkylation. Peptides have been found by phage display that bind to AMA1 and block merozoite invasion of erythrocytes. We have investigated their solution structures and interaction with full-length AMA1 ectodomain in an effort to understand the structure-function relationships of this important vaccine candidate

  10. Structural identifiability analyses of candidate models for in vitro Pitavastatin hepatic uptake.

    Science.gov (United States)

    Grandjean, Thomas R B; Chappell, Michael J; Yates, James W T; Evans, Neil D

    2014-05-01

    In this paper a review of the application of four different techniques (a version of the similarity transformation approach for autonomous uncontrolled systems, a non-differential input/output observable normal form approach, the characteristic set differential algebra and a recent algebraic input/output relationship approach) to determine the structural identifiability of certain in vitro nonlinear pharmacokinetic models is provided. The Organic Anion Transporting Polypeptide (OATP) substrate, Pitavastatin, is used as a probe on freshly isolated animal and human hepatocytes. Candidate pharmacokinetic non-linear compartmental models have been derived to characterise the uptake process of Pitavastatin. As a prerequisite to parameter estimation, structural identifiability analyses are performed to establish that all unknown parameters can be identified from the experimental observations available. Copyright © 2013. Published by Elsevier Ireland Ltd.

  11. Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR

    International Nuclear Information System (INIS)

    Litnovsky, A.; Philipps, V.; Wienhold, P.; Kreter, A.; Kirschner, A.; Matveev, D.; Brezinsek, S.; Sergienko, G.; Pospieszczyk, A.; Schweer, B.; Schulz, C.; Schmitz, O.; Coenen, J.W.; Samm, U.; Krieger, K.; Hirai, T.; Emmoth, B.; Rubel, M.; Bazylev, B.; Breuer, U.

    2011-01-01

    Plasma-facing components (PFCs) in ITER will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are concerns in particular on retention of codeposited radioactive fuel in the gaps. An R and D program is underway in TEXTOR addressing this acute issue of castellation. Material migration and fuel inventory are investigated using long- and short-term discharge-resolved experiments with castellated structures in TEXTOR. Significant impurity transport to the gaps was detected and results were in part quantitatively reproduced with 3D-GAPS code. Deposits containing up to 70 at.% of tungsten on the gap areas closest to the plasma were detected in recent experiments. Deposition in the gaps accompanied by metal mixing demand for development of effective cleaning techniques. In experiments with ITER-like castellation, the gaps were cleaned from carbonaceous deposits using oxygen plasmas at 350 deg. C. This contribution contains an overview of experimental and modeling results along with recommendations for PFCs in ITER.

  12. Evidence of novel fine-scale structural variation at autism spectrum disorder candidate loci

    Directory of Open Access Journals (Sweden)

    Hedges Dale J

    2012-04-01

    Full Text Available Abstract Background Autism spectrum disorders (ASD represent a group of neurodevelopmental disorders characterized by a core set of social-communicative and behavioral impairments. Gamma-aminobutyric acid (GABA is the major inhibitory neurotransmitter in the brain, acting primarily via the GABA receptors (GABR. Multiple lines of evidence, including altered GABA and GABA receptor expression in autistic patients, indicate that the GABAergic system may be involved in the etiology of autism. Methods As copy number variations (CNVs, particularly rare and de novo CNVs, have now been implicated in ASD risk, we examined the GABA receptors and genes in related pathways for structural variation that may be associated with autism. We further extended our candidate gene set to include 19 genes and regions that had either been directly implicated in the autism literature or were directly related (via function or ancestry to these primary candidates. For the high resolution CNV screen we employed custom-designed 244 k comparative genomic hybridization (CGH arrays. Collectively, our probes spanned a total of 11 Mb of GABA-related and additional candidate regions with a density of approximately one probe every 200 nucleotides, allowing a theoretical resolution for detection of CNVs of approximately 1 kb or greater on average. One hundred and sixty-eight autism cases and 149 control individuals were screened for structural variants. Prioritized CNV events were confirmed using quantitative PCR, and confirmed loci were evaluated on an additional set of 170 cases and 170 control individuals that were not included in the original discovery set. Loci that remained interesting were subsequently screened via quantitative PCR on an additional set of 755 cases and 1,809 unaffected family members. Results Results include rare deletions in autistic individuals at JAKMIP1, NRXN1, Neuroligin4Y, OXTR, and ABAT. Common insertion/deletion polymorphisms were detected at several

  13. The remote handling compatibility analysis of the ITER generic upper port plug structure

    Energy Technology Data Exchange (ETDEWEB)

    Ronden, D.M.S., E-mail: d.m.s.ronden@differ.nl [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Dammann, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Elzendoorn, B. [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Giacomin, T. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Heemskerk, C. [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Loesser, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Maquet, P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Oosterhout, J. van [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Pak, S.; Pitcher, C.S.; Portales, M.; Proust, M.; Udintsev, V.S.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France)

    2014-10-15

    Highlights: • We describe the remote handling compatibility of the ITER generic upper port plug. • Concepts are presented of specific design solutions to improve RH compatibility. • Simulation in VR of the GUPP DSM replacement indicates possible collisions. • Specific tooling concepts are proposed for GUPP handling equipment for the hot cell. - Abstract: The ITER diagnostics generic upper port plug (GUPP) is developed as a standardized design for all diagnostic upper port plugs, in which a variety of payloads can be mounted. Here, the remote handling compatibility analysis (RHCA) of the GUPP design is presented that was performed for the GUPP final design review. The analysis focuses mainly on the insertion and extraction procedure of the diagnostic shield module (DSM), a removable cassette that contains the diagnostic in-vessel components. It is foreseen that the DSM is a replaceable component – the procedure of which is to be performed inside the ITER hot cell facility (HCF), where the GUPP can be oriented in a vertical position. The DSM removal procedure in the HCF consists of removing locking pins, an M30 sized shoulder bolt and two electrical straps through the use of a dexterous manipulator, after which the DSM is lifted out of the GUPP by an overhead crane. For optimum access to its internals, the DSM is mounted in a handling device. The insertion of a new or refurbished DSM follows the reverse procedure. The RHCA shows that the GUPP design requires a moderate amount of changes to become fully compatible with RH maintenance requirements.

  14. The remote handling compatibility analysis of the ITER generic upper port plug structure

    International Nuclear Information System (INIS)

    Ronden, D.M.S.; Dammann, A.; Elzendoorn, B.; Giacomin, T.; Heemskerk, C.; Loesser, D.; Maquet, P.; Oosterhout, J. van; Pak, S.; Pitcher, C.S.; Portales, M.; Proust, M.; Udintsev, V.S.; Walsh, M.J.

    2014-01-01

    Highlights: • We describe the remote handling compatibility of the ITER generic upper port plug. • Concepts are presented of specific design solutions to improve RH compatibility. • Simulation in VR of the GUPP DSM replacement indicates possible collisions. • Specific tooling concepts are proposed for GUPP handling equipment for the hot cell. - Abstract: The ITER diagnostics generic upper port plug (GUPP) is developed as a standardized design for all diagnostic upper port plugs, in which a variety of payloads can be mounted. Here, the remote handling compatibility analysis (RHCA) of the GUPP design is presented that was performed for the GUPP final design review. The analysis focuses mainly on the insertion and extraction procedure of the diagnostic shield module (DSM), a removable cassette that contains the diagnostic in-vessel components. It is foreseen that the DSM is a replaceable component – the procedure of which is to be performed inside the ITER hot cell facility (HCF), where the GUPP can be oriented in a vertical position. The DSM removal procedure in the HCF consists of removing locking pins, an M30 sized shoulder bolt and two electrical straps through the use of a dexterous manipulator, after which the DSM is lifted out of the GUPP by an overhead crane. For optimum access to its internals, the DSM is mounted in a handling device. The insertion of a new or refurbished DSM follows the reverse procedure. The RHCA shows that the GUPP design requires a moderate amount of changes to become fully compatible with RH maintenance requirements

  15. Vibrations versus collisions and the iterative structure of two-body dynamics

    International Nuclear Information System (INIS)

    Pfitzner, A.; Cassing, W.; Peter, A.

    1993-11-01

    The two-body correlation function is decomposed into two channel correlation functions for the pp- and the ph-channel. The associated coupled equations describe the evolution in the respective channels as well as their mixing. Integration of the ph-channel in terms of vibrational RPA-states yields a closed equation for the correlations in the pp-channel comprising phonon-particle coupling and a memory term. In the stationary limit the equation for a generalised effective interaction is derived which iterates both the G-matrix (ladders) and the polarisation matrix (loops), thus accounting nonperturbatively for the mixing of ladders and loops. (orig.)

  16. Physics Structure Analysis of Parallel Waves Concept of Physics Teacher Candidate

    International Nuclear Information System (INIS)

    Sarwi, S; Linuwih, S; Supardi, K I

    2017-01-01

    The aim of this research was to find a parallel structure concept of wave physics and the factors that influence on the formation of parallel conceptions of physics teacher candidates. The method used qualitative research which types of cross-sectional design. These subjects were five of the third semester of basic physics and six of the fifth semester of wave course students. Data collection techniques used think aloud and written tests. Quantitative data were analysed with descriptive technique-percentage. The data analysis technique for belief and be aware of answers uses an explanatory analysis. Results of the research include: 1) the structure of the concept can be displayed through the illustration of a map containing the theoretical core, supplements the theory and phenomena that occur daily; 2) the trend of parallel conception of wave physics have been identified on the stationary waves, resonance of the sound and the propagation of transverse electromagnetic waves; 3) the influence on the parallel conception that reading textbooks less comprehensive and knowledge is partial understanding as forming the structure of the theory. (paper)

  17. Development of conductively cooled first wall armor and actively cooled divertor structure for ITER/FER

    International Nuclear Information System (INIS)

    Ioki, K.; Yamada, M.; Sakata, S.; Okada, K.; Toyoda, M.; Shimizu, K.; Tsujimura, S.; Iimura, M.; Akiba, M.; Araki, M.; Seki, M.

    1991-01-01

    Based on the design requirements for the plasma facing components in ITER/FER, we have performed design studies on the conductively cooled first wall armor and the divertor plate with sliding supports. The full-scale armor tiles were fabricated for heat load tests, and good thermal performances were obtained in heat load tests of 0.2-0.4 MW/m 2 . It is shown by the thermomechanical analysis on the divertor plate that thermal stresses and bending deformation are reduced significantly by using the sliding supports. The divertor test module with the sliding supports has been fabricated to investigate its fabricability and to verify the functions of the sliding supports during a high heat load of about 10 MW/m 2 . (orig.)

  18. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  19. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Ha, Min-Su, E-mail: msha12@nfri.re.kr [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Sa-Woong; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Duck-Hoi [ITER Organization, Route de Vinon sur Verdon - CS 90046, 13067 Sant Paul Lez Durance (France)

    2016-11-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K{sub e} factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  20. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    International Nuclear Information System (INIS)

    Shim, Hee-Jin; Ha, Min-Su; Kim, Sa-Woong; Jung, Hun-Chea; Kim, Duck-Hoi

    2016-01-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K_e factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  1. Structure-activity relationship analysis of cytotoxic cyanoguanidines: selection of CHS 828 as candidate drug

    Directory of Open Access Journals (Sweden)

    Gullbo Joachim

    2009-06-01

    Full Text Available Abstract Background N-(6-(4-chlorophenoxyhexyl-N'-cyano-N''-4-pyridyl guanidine (CHS 828 is the first candidate drug from a novel group of anti-tumour agents – the pyridyl cyanoguanidines, shown to be potent compounds interfering with cellular metabolism (inhibition of nicotinamide phosphoribosyl transferase and NF-κB signalling. Substituted cyanoguanidines are also found in anti-hypertensive agents such as the potassium channel opener pinacidil (N-cyano-N'-(4-pyridyl-N''-(1,2,2-trimethylpropylguanidine and histamine-II receptor antagonists (e.g. cimetidine, N-cyano-N'-methyl-N''-[2-[[(5-methylimidazol-4-yl]methyl]thio]ethylguanidine. In animal studies, CHS 828 has shown very promising activity, and phase I and II studies resulted in further development of a with a water soluble prodrug. Findings To study the structural requirements for cyanoguanidine cytotoxicity a set of 19 analogues were synthesized. The cytotoxic effects were then studied in ten cell lines selected for different origins and mechanisms of resistance, using the fluorometric microculture cytotoxicity assay (FMCA. The compounds showed varying cytotoxic activity even though the dose-response curves for some analogues were very shallow. Pinacidil and cimetidine were found to be non-toxic in all ten cell lines. Starting with cyanoguanidine as the crucial core it was shown that 4-pyridyl substitution was more efficient than was 3-pyridyl substitution. The 4-pyridyl cyanoguanidine moiety should be linked by an alkyl chain, optimally a hexyl, heptyl or octyl chain, to a bulky end group. The exact composition of this end group did not seem to be of crucial importance; when the end group was a mono-substituted phenyl ring it was shown that the preferred position was 4-substitution, followed by 3- and, finally, 2-substitution as the least active. Whether the substituent was a chloro, nitro or methoxy substituent seemed to be of minor importance. Finally, the activity patterns in the

  2. submitter Physical Properties of a High-Strength Austenitic Stainless Steel for the Precompression Structure of the ITER Central Solenoid

    CERN Document Server

    Sgobba, Stefano; Arauzo, Ana; Roussel, Pascal; Libeyre, Paul

    2016-01-01

    The ITER central solenoid (CS) consists of six independent coils kept together by a precompression support structure that must react vertical tensile loads and provide sufficient preload to maintain coil-to-coil contact when the solenoid is energized. The CS precompression system includes tie plates, lower and upper key blocks, load distribution and isolation plates and other attachment, support and insulating hardware. The tie plates operating at 4 K are manufactured starting from forgings in a high-strength austenitic stainless steel (FXM-19) with a stringent specification. Moreover, forged components for the lower and upper key blocks have to be provided in the same FXM-19 grade with comparably strict requirements. FXM-19 is a high-nitrogen austenitic stainless steel, featuring high strength and toughness, ready weldability, and forgeability. It features as well higher integral thermal contraction down to 4 K compared with the very high Mn steel grade selected for the CS coil jackets, hence providing an ad...

  3. ITER overview

    International Nuclear Information System (INIS)

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

    2001-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  4. ITER Overview

    International Nuclear Information System (INIS)

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

    1999-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  5. Improving search for low energy protein structures with an iterative niche genetic algorithm

    DEFF Research Database (Denmark)

    Helles, Glennie

    2010-01-01

    In attempts to predict the tertiary structure of proteins we use almost exclusively metaheuristics. However, despite known differences in performance of metaheuristics for different problems, the effect of the choice of metaheuristic has received precious little attention in this field. Particula......In attempts to predict the tertiary structure of proteins we use almost exclusively metaheuristics. However, despite known differences in performance of metaheuristics for different problems, the effect of the choice of metaheuristic has received precious little attention in this field...

  6. ITER Port Plug Engineering Trainee Program: Design, manufacturing and integration of structural components (analysis of the attachment)

    International Nuclear Information System (INIS)

    Zeile, Christian; Neuberger, Heiko; Dolensky, Bernhard

    2011-01-01

    The structural connection of helium cooled plasma facing components in ITER to the water cooled structural port plug shield requires an attachment system, which is able to cope with two main contradicting requirements: The attachment system has to be rigid in order to withstand mechanical loads, e.g. due to the deadweight or static and transient electro-magnetic loads. On the other hand, the attachment system has to be flexible in order to compensate the different thermal strains in between the plasma facing test devices (300-550 deg. C) and the port plug structure (∼120 deg. C). The paper presents the latest developments of an attachment system consisting of flexible attachment blocks with lamellae. The optimization steps as well as the connection to the shield are described. The results of the thermo-mechanical analyses under a defined worst-case scenario confirm the feasibility of the lamella design. This work has been performed in the frame of the EFDA goal orientated trainee program on port plug engineering.

  7. Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Tokitani, M., E-mail: tokitani.masayuki@LHD.nifs.ac.jp [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Miyamoto, M. [Shimane University, Matsue, Shimane 690-8504 (Japan); Masuzaki, S. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Fujii, Y. [Shimane University, Matsue, Shimane 690-8504 (Japan); Sakamoto, R. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Oya, Y. [Shizuoka University, Shizuoka 422-8529 (Japan); Hatano, Y. [University of Toyama, Toyama 930-8555 (Japan); Otsuka, T. [Kindai University, Higashi-Osaka, Osaka, 577-8502 (Japan); Oyaidzu, M.; Kurotaki, H.; Suzuki, T.; Hamaguchi, D.; Isobe, K.; Asakura, N. [National Institute for Quantum and Radiological Science and Technology (QST), Rokkasho Aomori 039-3212 (Japan); Widdowson, A. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Rubel, M. [Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden)

    2017-03-15

    Highlights: • Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall were studied. • The stratified mixed-material deposition layer composed by W, C, O, Mo and Be with the thickness of ∼1.5 μm was formed on the apron of Tile 1. • The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. - Abstract: Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was ∼1.5 μm. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation.

  8. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  9. IRMA iterative relaxation matrix approach for NMR structure determination application to DNA fragments

    International Nuclear Information System (INIS)

    Koning, M.M.G.

    1990-01-01

    The subject of this thesis is the structure determination of DNA molecules in solution with the use of NMR. For this purpose a new relaxation matrix approach is introduced. The emphasis is on the interpretation of nuclear Overhauser effects (NOEs) in terms of proton-proton distances and related three dimensional structures. The DNA molecules studied are obligonucleotides, unmodifief as well as modified molecules bu UV radiation. From comparison with unmodified molecules it turned out that UV irradiation scarcely influences the helical structure of the DNA string. At one place of the string a nucleotide is rotated towards the high-ANTI conformation which results in a slight unwinding of the DNA string but sufficient for blocking of the normal reading of genetic information. (H.W.). 456 refs.; 50 figs.; 30 tabs

  10. ITER licensing

    International Nuclear Information System (INIS)

    Gordon, C.W.

    2005-01-01

    ITER was fortunate to have four countries interested in ITER siting to the point where licensing discussions were initiated. This experience uncovered the challenges of licensing a first of a kind, fusion machine under different licensing regimes and helped prepare the way for the site specific licensing process. These initial steps in licensing ITER have allowed for refining the safety case and provide confidence that the design and safety approach will be licensable. With site-specific licensing underway, the necessary regulatory submissions have been defined and are well on the way to being completed. Of course, there is still work to be done and details to be sorted out. However, the informal international discussions to bring both the proponent and regulatory authority up to a common level of understanding have laid the foundation for a licensing process that should proceed smoothly. This paper provides observations from the perspective of the International Team. (author)

  11. Exploring Writing Circles as Innovative, Collaborative Writing Structures with Teacher Candidates

    Science.gov (United States)

    Roberts, Sherron Killingsworth; Blanch, Norine; Gurjar, Nandita

    2017-01-01

    Writing circles are "small groups... meeting regularly to share drafts, choose common writing topics, practice positive response, and in general, help each other become better writers" (Vopat, 2009, p. 6). In this exploratory study, writing circles were employed with elementary teacher candidates in hopes of enhancing their perceptions…

  12. Explosive welding method for manufacturing ITER-grade 316L(N)/CuCrZr hollow structural member

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Rui, E-mail: mr9980@163.com [PLA University of Science and Technology, Nanjing 210007 (China); Wang, Yaohua [PLA University of Science and Technology, Nanjing 210007 (China); Wu, Jihong [Southwestern Institute of Physics, Chengdu 610041 (China); Duan, Mianjun [PLA University of Science and Technology, Nanjing 210007 (China)

    2014-12-15

    Highlights: • Develop a new explosive welding method to fabricate the cooling channel of FW. • Utilize effective energy model to accurately calculate optimal welding parameters. • Provide an efficient way for manufacturing high-ductility hollow structural member. - Abstract: In this study, a new explosive welding method provided an effective way for manufacturing ITER-grade 316L(N)/CuCrZr hollow structural member. The welding parameters (stand-off distance and explosion rate) were calculated respectively using equivalent frontal collision wave model and effective energy model. The welded samples were subject to two step heat treatment cycles (solution annealing and aging). Optical microscopy (OM) and scanning electron microscopy (SEM) were utilized to analyze the microstructure of bonding interface. The mechanical properties of the welded samples were evaluated through microhardness test and tensile test. Moreover, the sealing property of the welded specimens was measured through helium leak test. Microstructural analysis showed that the welded sample using effective energy model had an ideal wavy interface. The results of microhardness test revealed an increase in hardness for both sides near to the bonding interface. And the hardening phenomenon of interface region disappeared after the solution annealing. SEM observation indicated that the samples with the post heat treatments exhibited a ductile fracture with dimple features after tensile test. After the specimens undergo aging strengthening, there was an obvious increase in the strength for all specimens. The helium leak test results have proven that the welded specimens are soundness.

  13. 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

  14. Development of an efficient iterative solver for linear systems in FE structural analysis

    International Nuclear Information System (INIS)

    Saint-Georges, P.; Warzee, G.; Beauwens, R.; Notay, Y.

    1993-01-01

    The preconditioned conjugate gradient is a well-known and powerful method to solve sparse symmetric positive definite systems of linear equations. Such systems are generated by the finite element discretization in structural analysis but users of finite element in this context generally still rely on direct methods. It is our purpose in the present paper to highlight the improvement brought forward by some new preconditioning techniques and show that the preconditioned conjugate gradient method is more performant than any direct method. (author)

  15. Iterative solution of multiple radiation and scattering problems in structural acoustics using the BL-QMR algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Malhotra, M. [Stanford Univ., CA (United States)

    1996-12-31

    Finite-element discretizations of time-harmonic acoustic wave problems in exterior domains result in large sparse systems of linear equations with complex symmetric coefficient matrices. In many situations, these matrix problems need to be solved repeatedly for different right-hand sides, but with the same coefficient matrix. For instance, multiple right-hand sides arise in radiation problems due to multiple load cases, and also in scattering problems when multiple angles of incidence of an incoming plane wave need to be considered. In this talk, we discuss the iterative solution of multiple linear systems arising in radiation and scattering problems in structural acoustics by means of a complex symmetric variant of the BL-QMR method. First, we summarize the governing partial differential equations for time-harmonic structural acoustics, the finite-element discretization of these equations, and the resulting complex symmetric matrix problem. Next, we sketch the special version of BL-QMR method that exploits complex symmetry, and we describe the preconditioners we have used in conjunction with BL-QMR. Finally, we report some typical results of our extensive numerical tests to illustrate the typical convergence behavior of BL-QMR method for multiple radiation and scattering problems in structural acoustics, to identify appropriate preconditioners for these problems, and to demonstrate the importance of deflation in block Krylov-subspace methods. Our numerical results show that the multiple systems arising in structural acoustics can be solved very efficiently with the preconditioned BL-QMR method. In fact, for multiple systems with up to 40 and more different right-hand sides we get consistent and significant speed-ups over solving the systems individually.

  16. Structural model for the first wall W-based material in ITER project

    Institute of Scientific and Technical Information of China (English)

    Dehua Xu; Xinkui He; Shuiquan Deng; Yong Zhao

    2014-01-01

    The preparation, characterization, and test of the first wall materials designed to be used in the fusion reactor have remained challenging problems in the material science. This work uses the first-principles method as implemented in the CASTEP package to study the influ-ences of the doped titanium carbide on the structural sta-bility of the W–TiC material. The calculated total energy and enthalpy have been used as criteria to judge the structural models built with consideration of symmetry. Our simulation indicates that the doped TiC tends to form its own domain up to the investigated nano-scale, which implies a possible phase separation. This result reveals the intrinsic reason for the composite nature of the W–TiC material and provides an explanation for the experimen-tally observed phase separation at the nano-scale. Our approach also sheds a light on explaining the enhancing effects of doped components on the durability, reliability, corrosion resistance, etc., in many special steels.

  17. Interaction of a tin-based capillary porous structure with ITER/DEMO relevant plasma conditions

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, T.W., E-mail: t.w.morgan@differ.nl; Bekerom, D.C.M. van den; De Temmerman, G.

    2015-08-15

    Sn filled capillary porous structures were exposed to high flux low temperature plasma conditions at the Pilot-PSI linear device. Enhanced erosion above that expected classically was investigated via spectroscopic observation of Sn{sup 0} emission from the plasma in front of the target surface while the surface temperature was monitored by both thermography and pyrometry. An anomalous erosion flux was observed as temperature increases, with onset for this occurrence varying strongly between different ion species. The results appear incompatible with existing ‘adatom’ models for the anomalous erosion flux. Further targets were exposed in turn to increasing heat fluxes and the heat removed determined from cooling water calorimetry, which was then compared to a solid Mo reference target. At high powers the total energy of the cooling water is reduced, indicating a shielding of the surface from the plasma heat flux by the vapour cloud in front.

  18. New concurrent iterative methods with monotonic convergence

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Qingchuan [Michigan State Univ., East Lansing, MI (United States)

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

  19. Statistical iterative reconstruction for streak artefact reduction when using multidetector CT to image the dento-alveolar structures.

    Science.gov (United States)

    Dong, J; Hayakawa, Y; Kober, C

    2014-01-01

    When metallic prosthetic appliances and dental fillings exist in the oral cavity, the appearance of metal-induced streak artefacts is not avoidable in CT images. The aim of this study was to develop a method for artefact reduction using the statistical reconstruction on multidetector row CT images. Adjacent CT images often depict similar anatomical structures. Therefore, reconstructed images with weak artefacts were attempted using projection data of an artefact-free image in a neighbouring thin slice. Images with moderate and strong artefacts were continuously processed in sequence by successive iterative restoration where the projection data was generated from the adjacent reconstructed slice. First, the basic maximum likelihood-expectation maximization algorithm was applied. Next, the ordered subset-expectation maximization algorithm was examined. Alternatively, a small region of interest setting was designated. Finally, the general purpose graphic processing unit machine was applied in both situations. The algorithms reduced the metal-induced streak artefacts on multidetector row CT images when the sequential processing method was applied. The ordered subset-expectation maximization and small region of interest reduced the processing duration without apparent detriments. A general-purpose graphic processing unit realized the high performance. A statistical reconstruction method was applied for the streak artefact reduction. The alternative algorithms applied were effective. Both software and hardware tools, such as ordered subset-expectation maximization, small region of interest and general-purpose graphic processing unit achieved fast artefact correction.

  20. Thermo-structural optimization of the ITER ICRH Four Port Junction and Straps against in-vessel design criteria

    International Nuclear Information System (INIS)

    Lafuente, Antonio; Fursdon, Mike; Shannon, Mark

    2014-01-01

    Design optimization work has been conducted on the ITER Ion Cyclotron Resonance Heating (ICRH) Four Port Junction (4PJ) and Straps – a sub-assembly of the antenna. The aim of the analysis is to evaluate ways of making the component compliant with SDC-IC rules while balancing the competing demands of different performance requirements. Of particular interest are the bends that connect the 316L(N) strap pipes to its housing, where previous work had shown that primary plus secondary stresses would result in a low predicted fatigue life. The aim of the study was to explore the possibility of reducing stresses in these bends. Coupled ANSYS CFX and structural models are used to calculate coolant and metal temperatures and resulting stresses due to incident and self-generated heat. Although all of the modifications explored resulted in primary plus secondary stresses exceeding the cyclic damage design criteria, some avenues are identified for future studies and a reduction in stress toward the target is obtained

  1. Structural damages prevention of the ITER vacuum vessel and ports by elasto-plastic analysis with regards to RCC-MR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Jean-Marc, E-mail: jean-marc.martinez@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Jun, Chang Hoon; Portafaix, Christophe; Alekseev, Alexander; Sborchia, Carlo; Choi, Chang-Ho [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Albin, Vincent [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Borrelly, Stephane [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Cambazar, Magali [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France); Gaucher, Thomas [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Sfarni, Samir; Tailhardat, Olivier [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France)

    2015-10-15

    Highlights: • ITER vacuum vessel (VV) is a part of the first barrier to confine the plasma. • ITER VV as NPE necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, and conformance testing and quality assurance, i.e. ANB. • Several types of damages have to be prevented in order to guarantee the structural integrity with regards to RCC-MR. • It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. • Several structural analyses were performed with many different global and local models of the whole ITER VV. - Abstract: Several types of damages have to be prevented in order to guarantee the structural integrity of a structure with regards to RCC-MR; the P-type damages which can result from the application to a structure of a steadily and regularly increasing loading or a constant loading and the S-type damages during operational loading conditions which can only result from repeated application of loadings associated to the progressive deformations and fatigue. Following RCC-MR, the S-type damages prevention has to be started only when the structural integrity is guaranteed against P-type damages. The verification of the last one on the ITER vacuum vessel and ports has been performed by limit analysis with elasto-(perfectly)plastic material behavior. It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. Some elasto-plastic analyses have been performed considering several cyclic loadings to evaluate also more realistic structural margins of the against S-type damages.

  2. Catalytic Metal Free Production of Large Cage Structure Carbon Particles: A Candidate for Hydrogen Storage

    Science.gov (United States)

    Kimura, Yuki; Nuth, Joseph A., III; Ferguson, Frank T.

    2005-01-01

    We will demonstrate that carbon particles consisting of large cages can be produced without catalytic metal. The carbon particles were produced in CO gas as well as by introduction of 5% methane gas into the CO gas. The gas-produced carbon particles were able to absorb approximately 16.2 wt% of hydrogen. This value is 2.5 times higher than the 6.5 wt% goal for the vehicular hydrogen storage proposed by the Department of Energy in the USA. Therefore, we believe that this carbon particle is an excellent candidate for hydrogen storage for fuel cells.

  3. Iterated multidimensional wave conversion

    International Nuclear Information System (INIS)

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-01-01

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

  4. 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

  5. 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.)

  6. Does Changing Examiner Stations During UK Postgraduate Surgery Objective Structured Clinical Examinations Influence Examination Reliability and Candidates' Scores?

    Science.gov (United States)

    Brennan, Peter A; Croke, David T; Reed, Malcolm; Smith, Lee; Munro, Euan; Foulkes, John; Arnett, Richard

    2016-01-01

    Objective structured clinical examinations (OSCE) are widely used for summative assessment in surgery. Despite standardizing these as much as possible, variation, including examiner scoring, can occur which may affect reliability. In study of a high-stakes UK postgraduate surgical OSCE, we investigated whether examiners changing stations once during a long examining day affected marking, reliability, and overall candidates' scores compared with examiners who examined the same scenario all day. An observational study of 18,262 examiner-candidate interactions from the UK Membership of the Royal College of Surgeons examination was carried at 3 Surgical Colleges across the United Kingdom. Scores between examiners were compared using analysis of variance. Examination reliability was assessed with Cronbach's alpha, and the comparative distribution of total candidates' scores for each day was evaluated using t-tests of unit-weighted z scores. A significant difference was found in absolute scores differences awarded in the morning and afternoon sessions between examiners who changed stations at lunchtime and those who did not (p design and examiner experience in surgical OSCEs and beyond. Copyright © 2016 Association of Program Directors in Surgery. Published by Elsevier Inc. All rights reserved.

  7. Structural origination of charge transfer complex nanostructures: Excellent candidate for field emission

    International Nuclear Information System (INIS)

    Pal, Shreyasi; Chattopadhyay, Kalyan Kumar

    2016-01-01

    Worldwide strategies for amalgamating rationally controlled one-dimensional organic nanowires are of fundamental importance for their applications in flexible, cheaper and lighter electronics. In this work we have fabricated large-area, ordered CuTCNQ (copper-7,7,8,8-tetracyanoquinodimethane) nano architecture arrays over flexible conducting substrate and discussed the rational growth and integration of nanostructures. Here we adopted the organic solid phase reaction (VLS) technique for the growth of organic hierarchies and investigated how field emission properties changes by tuning the nanostructures morphology i.e., by varying length, diameter, alignment and orientation over flexible substrate. The CuTCNQ nanowires with optimized geometry exhibit excellent high field emission performance with low turn-on and threshold field values. The result strongly indicate that CuTCNQ nanowires on flexible carbon cloth substrate are promising candidates for constructing cold cathode based emission display devices, vacuum nanoelectronics, and etc.

  8. Structural origination of charge transfer complex nanostructures: Excellent candidate for field emission

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Shreyasi; Chattopadhyay, Kalyan Kumar [Thin Films and Nanoscience Laboratory, Department of Physics, Jadavpur University, Kolkata 700032 (India)

    2016-05-23

    Worldwide strategies for amalgamating rationally controlled one-dimensional organic nanowires are of fundamental importance for their applications in flexible, cheaper and lighter electronics. In this work we have fabricated large-area, ordered CuTCNQ (copper-7,7,8,8-tetracyanoquinodimethane) nano architecture arrays over flexible conducting substrate and discussed the rational growth and integration of nanostructures. Here we adopted the organic solid phase reaction (VLS) technique for the growth of organic hierarchies and investigated how field emission properties changes by tuning the nanostructures morphology i.e., by varying length, diameter, alignment and orientation over flexible substrate. The CuTCNQ nanowires with optimized geometry exhibit excellent high field emission performance with low turn-on and threshold field values. The result strongly indicate that CuTCNQ nanowires on flexible carbon cloth substrate are promising candidates for constructing cold cathode based emission display devices, vacuum nanoelectronics, and etc.

  9. Structural impact of armor monoblock dimensions on the failure behavior of ITER-type divertor target components: Size matters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; You, Jeong-Ha, E-mail: you@ipp.mpg.de

    2016-12-15

    Highlights: • Quantitative assessment of size effects was conducted numerically for W monoblock. • Decreasing the width of W monoblock leads to a lower risk of failure. • The Cu interlayer was not affected significantly by varying armor thickness. • The predicted trends were in line with the experimental observations. - Abstract: Plenty of high-heat-flux tests conducted on tungsten monoblock type divertor target mock-ups showed that the threshold heat flux density for cracking and fracture of tungsten armor seems to be related to the dimension of the monoblocks. Thus, quantitative assessment of such size effects is of practical importance for divertor target design. In this paper, a computational study about the thermal and structural impact of monoblock size on the plastic fatigue and fracture behavior of an ITER-type tungsten divertor target is reported. As dimensional parameters, the width and thickness of monoblock, the thickness of sacrificial armor, and the inner diameter of cooling tube were varied. Plastic fatigue lifetime was estimated for the loading surface of tungsten armor and the copper interlayer by use of a cyclic-plastic constitutive model. The driving force of brittle crack growth through the tungsten armor was assessed in terms of J-integral at the crack tip. Decrease of the monoblock width effectively reduced accumulation of plastic strain at the armor surface and the driving force of brittle cracking. Decrease of sacrificial armor thickness led to decrease of plastic deformation at the loading surface due to lower surface temperature, but the thermal and mechanical response of the copper interlayer was not affected by the variation of armor thickness. Monoblock with a smaller tube diameter but with the same armor thickness and shoulder thickness experienced lower fatigue load. The predicted trends were in line with the experimental observations.

  10. Fracture mechanics analysis approach to assess structural integrity of the first confinement boundaries in ITER Generic Upper Port Plug structure

    Energy Technology Data Exchange (ETDEWEB)

    Guirao, Julio, E-mail: julio@natec-ingenieros.com [Numerical Analysis Technologies S.L. (NATEC), Gijon (Spain); Iglesias, Silvia; Vacas, Christian; Udintsev, Victor [CHD, Diagnostic Division, ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Pak, Sunil [Diagnostic and Control Team, National Fusion Research Institute, Daejeon (Korea, Republic of); Maquet, Philippe [CHD, Diagnostic Division, ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Rodriguez, Eduardo; Roces, Jorge [Department of Construction and Manufacturing Engineering, University of Oviedo, Gijon (Spain)

    2015-10-15

    Highlights: • A parametric submodel of the spot under study is developed. • The associated macro has the capability to successively re-build the submodel implementing the crack with the geometry of the updated crack front as a function of the predicted increments of length in the apexes of the crack from the calculated stress intensity factor at the crack front. • The analysis incorporates the crack behavior model to predict the evolution of the postulated defect under the application of the different transients. • The analysis is based on the Elasto-Plastic Fracture Mechanics (EPFM) theory to account for the ductility of the materials (316LN type stainless steel). - Abstract: This paper demonstrates structural integrity of the first confinement boundary in Generic Upper Port Plug structures against cracking during service. This constitutes part of the justification to demonstrate that the non-aggression to the confinement barrier requirement may be compatible with the absent of a specific in-service inspections (ISI) program in the trapezoidal section. Since the component will be subjected to 100% volumetric inspections it can be assumed that no defects below the threshold of applied Nondestructive Evaluation techniques will be present before its commissioning. Cracks during service would be associated to defects under Code acceptance limit. This limit can be reasonably taken as 2 mm. Using elastic–plastic fracture mechanics an initial defect is postulated at the worst location in terms of probability and impact on the confinement boundary. Its evolution is simulated through finite element analysis and final dimension at the end of service is estimated. Applying the procedures in RCC-MR 2007 (App-16) the stability of the crack is assessed. As relative high safety margin was achieved, a complementary assessment postulating an initial defect of 6 mm was also conducted. New margin calculated provides a more robust design.

  11. On One-Point Iterations and DIIS

    DEFF Research Database (Denmark)

    Østerby, Ole; Sørensen, Hans Henrik Brandenborg

    2009-01-01

    We analyze various iteration procedures in many dimensions inspired by the SCF iteration used in first principles electronic structure calculations. We show that the simple mixing of densities can turn a divergent (or slowly convergent) iteration into a (faster) convergent process provided all...

  12. ITER council proceedings: 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Continuing the ITER EDA, two further ITER Council Meetings were held since the publication of ITER EDA documentation series no, 20, namely the ITER Council Meeting on 27-28 February 2001 in Toronto, and the ITER Council Meeting on 18-19 July in Vienna. That Meeting was the last one during the ITER EDA. This volume contains records of these Meetings, including: Records of decisions; List of attendees; ITER EDA status report; ITER EDA technical activities report; MAC report and advice; Final report of ITER EDA; and Press release

  13. Neutron diffraction study of internal stresses in brazed CFC/Mo divertor structures for NET/ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ceretti, M [Laboratoire Leon-Brillouin, CEA/CE Saclay, F-91191, Gif-sur-Yvette (France); Coppola, R [ENEA/Casaccia, INN-FIS, C.P. 2400, I-00100 Rome (Italy); Di Pietro, E [ENEA/Frascati, Dip. FUS, C.P. 2400, I-00100 Rome (Italy); Lodini, A [Laboratoire Leon-Brillouin, CEA/CE Saclay, F-91191, Gif-sur-Yvette (France) Universite de Champagne-Ardennes, Reims (France); Perrin, M [Laboratoire Leon-Brillouin, CEA/CE Saclay, F-91191, Gif-sur-Yvette (France); Piant, A [Laboratoire Leon-Brillouin, CEA/CE Saclay, F-91191, Gif-sur-Yvette (France); Rustichelli, F [Istituto di Scienze Fisiche, Universita di Ancona (Italy)

    1994-09-01

    This contribution presents the first results of a study, performed by neutron diffraction, on the internal stresses remaining after brazing at 860 C in graphite/molybdenum samples developed for NET/ITER. Samples of polycrystalline graphite and a carbon-fiber composite are considered. The deformation field is characterized close to the brazing interface, within a linear spatial resolution of approximately 2 mm. The results are discussed with reference to those obtainable by other methods and to theoretical considerations. ((orig.))

  14. ITER EDA Newsletter. Vol. 1, No. 1

    International Nuclear Information System (INIS)

    1992-11-01

    After the ITER Engineering Design Activities (EDA) Agreement and Protocol 1 had been signed by the four ITER parties on July 21, 1992 and had entered into force, the ITER Council suggested at its first meeting (Vienna, September 10-11, 1992) that the publication of the ITER Newsletter be continued during the EDA with assistance of the International Atomic Energy Agency. This suggestion was supported by the Agency and subsequently the ITER office in Vienna assumed its responsibilities for planning and executing activities related to the publication of the Newsletter. The ITER EDA Newsletter is planned to be a monthly publication aimed at disseminating broad information and understanding, including the description of the personal and institutional involvements in the ITER project in addition to technical facts about it. The responsibility for the Newsletter rests with the ITER council. In this first issue the signing of the ITER EDA Activities and Protocol 1 is reported. The EDA organizational structure is described. This issue also reports on the first ITER EDA council meeting, the opening of the ITER EDA NAKA Co-Centre, the first meeting of the ITER Technical Advisory Committee, activities of special working groups, an ITER Technical Meeting, as well as ''News in Brief'' and ''Coming Events''

  15. Design of the ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

    1995-01-01

    The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

  16. Event-horizon-scale structure in the supermassive black hole candidate at the Galactic Centre.

    Science.gov (United States)

    Doeleman, Sheperd S; Weintroub, Jonathan; Rogers, Alan E E; Plambeck, Richard; Freund, Robert; Tilanus, Remo P J; Friberg, Per; Ziurys, Lucy M; Moran, James M; Corey, Brian; Young, Ken H; Smythe, Daniel L; Titus, Michael; Marrone, Daniel P; Cappallo, Roger J; Bock, Douglas C-J; Bower, Geoffrey C; Chamberlin, Richard; Davis, Gary R; Krichbaum, Thomas P; Lamb, James; Maness, Holly; Niell, Arthur E; Roy, Alan; Strittmatter, Peter; Werthimer, Daniel; Whitney, Alan R; Woody, David

    2008-09-04

    The cores of most galaxies are thought to harbour supermassive black holes, which power galactic nuclei by converting the gravitational energy of accreting matter into radiation. Sagittarius A* (Sgr A*), the compact source of radio, infrared and X-ray emission at the centre of the Milky Way, is the closest example of this phenomenon, with an estimated black hole mass that is 4,000,000 times that of the Sun. A long-standing astronomical goal is to resolve structures in the innermost accretion flow surrounding Sgr A*, where strong gravitational fields will distort the appearance of radiation emitted near the black hole. Radio observations at wavelengths of 3.5 mm and 7 mm have detected intrinsic structure in Sgr A*, but the spatial resolution of observations at these wavelengths is limited by interstellar scattering. Here we report observations at a wavelength of 1.3 mm that set a size of 37(+16)(-10) microarcseconds on the intrinsic diameter of Sgr A*. This is less than the expected apparent size of the event horizon of the presumed black hole, suggesting that the bulk of Sgr A* emission may not be centred on the black hole, but arises in the surrounding accretion flow.

  17. 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)

  18. 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)

  19. ICP (ITER Collaborative Platform)

    Energy Technology Data Exchange (ETDEWEB)

    Capuano, C.; Carayon, F.; Patel, V. [ITER, 13 - St. Paul-Lez Durance (France)

    2009-07-01

    The ITER organization has the necessity to manage a massive amount of data and processes. Each team requires different process and databases often interconnected with those of others teams. ICP is the current central ITER repository of structured and unstructured data. All data in ICP is served and managed via a web interface that provides global accessibility with a common user friendly interface. This paper will explain the model used by ICP and how it serves the ITER project by providing a robust and agile platform. ICP is developed in ASP.NET using MSSQL Server for data storage. It currently houses 15 data driven applications, 150 different types of record, 500 k objects and 2.5 M references. During European working hours the system averages 150 concurrent users and 20 requests per second. ICP connects to external database applications to provide a single entry point to ITER data and a safe shared storage place to maintain this data long-term. The Core model provides an easy to extend framework to meet the future needs of the Organization. ICP follows a multi-tier architecture, providing logical separation of process. The standard three-tier architecture is expanded, with the data layer separated into data storage, data structure, and data access components. The business or applications logic layer is broken up into a common business functionality layer, a type specific logic layer, and a detached work-flow layer. Finally the presentation tier comprises a presentation adapter layer and an interface layer. Each layer is built up from small blocks which can be combined to create a wide range of more complex functionality. Each new object type developed gains access to a wealth of existing code functionality, while also free to adapt and extend this. The hardware structure is designed to provide complete redundancy, high availability and to handle high load. This document is composed of an abstract followed by the presentation transparencies. (authors)

  20. ITER EDA status

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    The Project has focused on drafting the Plant Description Document (PDD), which will be published as the Technical Basis for the ITER Final Design Report (FDR), and its related documentation in time for the ITER review process. The preparations have involved continued intensive detailed design work, analyses and assessments by the Home Teams and the Joint Central Team, who have co-operated closely and efficiently. The main technical document has been completed in time for circulation, as planned, to TAC members for their review at TAC-17 (19-22 February 2001). Some of the supporting documents, such as the Plant Design Specification (PDS), Design Requirements and Guidelines (DRG1 and DRG2), and the Plant Safety Requirement (PSR) are also available for reference in draft form. A summary paper of the PDD for the Council's information is available as a separate document. A new documentation structure for the Project has been established. This hierarchical structure for documentation facilitates the entire organization in a way that allows better change control and avoids duplications. The initiative was intended to make this documentation system valid for the construction and operation phases of ITER. As requested, the Director and the JCT have been assisting the Explorations to plan for future joint technical activities during the Negotiations, and to consider technical issues important for ITER construction and operation for their introduction in the draft of a future joint implementation agreement. As charged by the Explorers, the Director has held discussions with the Home Team Leaders in order to prepare for the staffing of the International Team and Participants Teams during the Negotiations (Co-ordinated Technical Activities, CTA) and also in view of informing all ITER staff about their future directions in a timely fashion. One important element of the work was the completion by the Parties' industries of costing studies of about 83 ''procurement packages

  1. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  2. Alternative modelling of brittle structures in a sub-area of the SKB candidate area at Forsmark, eastern Sweden

    International Nuclear Information System (INIS)

    Askling, Per; Tiren, Sven A.; Beckholmen, Monica; Straeng, Thomas

    2008-11-01

    One way to test the confidence of a presented model is to construct an alternative model. Such work is cognitive process of skill acquisition and also a process of understanding data in the sense of sorting and classifying data. This is of particular interest for the Swedish Radiation Safety Authority (SSM) in their technical review of SKB's on-going site investigation programme for potential repository sites. In this study, an alternative brittle deformation model of a selected part of the SKB candidate area in eastern Sweden was constructed. The input data set was obtained from SKB's database SICADA and is a selected set of data from five cored boreholes drilled from two drill-sites and comprises geophysical borehole logs, geological core-logs, hydrological logs (PFL; Posiva Flow Log) and borehole deviation measurements. Statistical cluster analysis applied on the geophysical borehole data were used to obtain the locations of bedrock with contrasting physical characteristics similar to those of brittle deformation zones. The cluster analysis is an objective procedure, contrasting with SKB's more subjective approach to the single-hole interpretation. Thus some differences are expected which could illustrate the effect of methodology that includes subjective 'expert judgement.' and indicate the possibility of alternative interpretations. The information about brittle structures in the geological boreholes logs was sorted and classification was made according to character of the structures (all fractures, open fractures, partly open fractures, frequency, orientate on/identification of fracture sets, sections of crush rock, and alteration). A separate study was performed to relate rock alteration with structures. The resolution applied in the fracture statistics is one metre, i.e. all studied entities were expressed per metre borehole length. All clusters were structurally characterized by the fractures inside the clusters (orientation and density of fractures) and

  3. Alternative modelling of brittle structures in a sub-area of the SKB candidate area at Forsmark, eastern Sweden.

    Energy Technology Data Exchange (ETDEWEB)

    Askling, Per; Tiren, Sven A.; Beckholmen, Monica; Straeng, Thomas (Geosigma AB, Uppsala (Sweden))

    2008-11-15

    One way to test the confidence of a presented model is to construct an alternative model. Such work is cognitive process of skill acquisition and also a process of understanding data in the sense of sorting and classifying data. This is of particular interest for the Swedish Radiation Safety Authority (SSM) in their technical review of SKB's on-going site investigation programme for potential repository sites. In this study, an alternative brittle deformation model of a selected part of the SKB candidate area in eastern Sweden was constructed. The input data set was obtained from SKB's database SICADA and is a selected set of data from five cored boreholes drilled from two drill-sites and comprises geophysical borehole logs, geological core-logs, hydrological logs (PFL; Posiva Flow Log) and borehole deviation measurements. Statistical cluster analysis applied on the geophysical borehole data were used to obtain the locations of bedrock with contrasting physical characteristics similar to those of brittle deformation zones. The cluster analysis is an objective procedure, contrasting with SKB's more subjective approach to the single-hole interpretation. Thus some differences are expected which could illustrate the effect of methodology that includes subjective 'expert judgement.' and indicate the possibility of alternative interpretations. The information about brittle structures in the geological boreholes logs was sorted and classification was made according to character of the structures (all fractures, open fractures, partly open fractures, frequency, orientate on/identification of fracture sets, sections of crush rock, and alteration). A separate study was performed to relate rock alteration with structures. The resolution applied in the fracture statistics is one metre, i.e. all studied entities were expressed per metre borehole length. All clusters were structurally characterized by the fractures inside the clusters (orientation and

  4. The effect of adaptive statistical iterative reconstruction on the assessment of diagnostic image quality and visualisation of anatomical structures in paediatric cerebral CT examinations

    International Nuclear Information System (INIS)

    Larsson, Joel; Baath, Magnus; Thilander-Klang, Anne; Ledenius, Kerstin

    2016-01-01

    The purpose of this study was to investigate the effect of adaptive statistical iterative reconstruction (ASiR) on the visualisation of anatomical structures and diagnostic image quality in paediatric cerebral computed tomography (CT) examinations. Forty paediatric patients undergoing routine cerebral CT were included in the study. The raw data from CT scans were reconstructed into stacks of 5 mm thick axial images at various levels of ASiR. Three paediatric radiologists rated six questions related to the visualisation of anatomical structures and one question on diagnostic image quality, in a blinded randomised visual grading study. The evaluated anatomical structures demonstrated enhanced visibility with increasing level of ASiR, apart from the cerebrospinal fluid space around the brain. In this study, 60 % ASiR was found to be the optimal level of ASiR for paediatric cerebral CT examinations. This shows that the commonly used 30 % ASiR may not always be the optimal level. (authors)

  5. HA03 as an Iranian Candidate Concealed Antigen for Vaccination against Hyalomma anatolicum anatolicum: Comparative Structural and In silico Studies

    Directory of Open Access Journals (Sweden)

    Mohammadi, A.

    2013-12-01

    Full Text Available In the last decades researchers had focused on developing a vaccine against tick based on protective antigen. Recombinant vaccines based on concealed antigen from Boophilus microplus have been developed in Australia and Cuba by the name of TICKGARD and GAVAC (De La Fuente and Kocan, 2006. Further studies on this antigen have shown some extent of protection against other species (De Vos et al., 2001. In Iran most important species is Hyalomma anatolicum and limited information about its control are available. This paper reports structural and polymorphic analysis of HA03 as an Iranian candidate concealed antigen of H. a. anatolicum deposited in Gen-Bank .(Aghaeipour et al. GQ228820. The comparison between this antigen and other mid gut concealed antigen that their characteristics are available in GenBank showed there are high rate of similarity between them. The HA03 amino acid sequence had a homology of around 89%, 64%, 56% with HA98, BM86, BM95 respectively. Potential of MHC class I and II binding region indicated a considerable variation between BM86 antigen and its efficiency against Iranian H. a. anatolicum. In addition, predicted major of hydrophobisity and similarity in N-glycosylation besides large amount of cystein and seven EGF like regions presented in protein structure revealed that value of HA03 as a new protective antigen and the necessity of the development, BM86 homolog of H. a. anatolicum HA03 based recombinant vaccine.

  6. ITER council proceedings: 1998

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains documents of the 13th and the 14th ITER council meeting as well as of the 1st extraordinary ITER council meeting. Documents of the ITER meetings held in Vienna and Yokohama during 1998 are also included. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics are given

  7. Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Belyakov, V.; Kukhtin, V.; Lamzin, E.; Mazul, I.; Sytchevsky, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2014-11-15

    Highlights: • We present the way of modeling transient electro-magnetic loads on local conductive domains in the large magnetic system. • Simplification is achieved by decomposing of the problem, multi-scale integral-differential modeling and use of integral parameters. • The intrinsic scale of loads on a localized conductor with eddy is quantified through the load susceptibility tensor. • Solution is searched as response of a simple equivalent dynamic simulator, using control theory methods. • The concept is exemplified with multi-scenario assessment of EM eddy loads on ITER blanket modules. - Abstract: The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail.

  8. Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

    International Nuclear Information System (INIS)

    Rozov, Vladimir; Belyakov, V.; Kukhtin, V.; Lamzin, E.; Mazul, I.; Sytchevsky, S.

    2014-01-01

    Highlights: • We present the way of modeling transient electro-magnetic loads on local conductive domains in the large magnetic system. • Simplification is achieved by decomposing of the problem, multi-scale integral-differential modeling and use of integral parameters. • The intrinsic scale of loads on a localized conductor with eddy is quantified through the load susceptibility tensor. • Solution is searched as response of a simple equivalent dynamic simulator, using control theory methods. • The concept is exemplified with multi-scenario assessment of EM eddy loads on ITER blanket modules. - Abstract: The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail

  9. ITER EDA newsletter. V. 10, no. 1

    International Nuclear Information System (INIS)

    2001-01-01

    This article provides a summary of results of the ITER Physics Committee Meeting, which was held on 14 October 2000 at the ITER Garching Joint Work Site, Germany. The ITER Physics Committee is the body responsible for overseeing, through the seven specialized Expert Groups, the R and D activities contributed voluntarily by the ITER Parties. The Parties' Physics Designated Persons, the Chairs and Co-Chairs of ITER Physics Expert Groups and the JCT members involved attended the Meeting. As usual, the meeting was chaired by the ITER Director, Dr. R. Aymar, who reported on the status of the ITER EDA. Dr. Aymar described the steps being taken in preparing the ITER-FEAT Final Design Report (FDR), and further stated that the Report would be available in time to be of benefit to the Negotiations on the ITER Joint Implementation, expected to start around May 2001. All Parties recognize that the ITER Physics Expert Group structure has been useful in focusing the tokamak physics activity on the ITER-relevant issues and provides an efficient worldwide collaboration on confirming innovative solutions. The concept of an international workshop to be organized as a pre-meeting of each Expert Group meeting, in order to involve U.S. scientists in the discussion of generic tokamak physics issues, was introduced in 2000, with some success, and its goal should be pursued

  10. Nb-base FS-85 Alloy as a Candidate Structural Material for Space Reactor Applications: Effects of Thermal Aging

    International Nuclear Information System (INIS)

    Leonard, Keith J.; Busby, Jeremy T.; Hoelzer, David T.; Zinkle, Steven J.

    2009-01-01

    The proposed use of fission reactors for manned or deep space missions have typically relied on the potential use of refractory metal alloys as structural materials. Throughout the history of these programs, the lead candidate has been Nb-1Zr due to its good fabrication and welding characteristics. However, the less than optimal creep resistance of this alloy has encouraged interest in the more complex FS-85 (Nb-28Ta-10W-1Zr) alloy. Despite this interest, a relatively small database exists for the properties of FS-85. These gaps include potential microstructural instabilities that can lead to mechanical property degradation. In this work, changes in microstructure and mechanical properties of FS-85 were investigated following 1100 h of thermal aging at 1098, 1248 and 1398 K. The changes in electrical resistivity, hardness and tensile properties between the as-annealed and aged materials are compared. Evaluation of the microstructural changes was performed through optical, scanning and transmission electron microscopy. The development of intragranular and grain boundary precipitation of Zr-rich compounds as a function of aging temperature was followed. Brittle tensile behavior was measured in the 1248 K aged material, while ductile behavior occurred in material aged above and below this temperature. The effect of temperature on the under and overaging of the grain boundary particles are believed to have contributed to the mechanical property behavior of the aged material

  11. Nb-Base FS-85 Alloy as a Candidate Structural Material for Space Reactor Applications: Effects of Thermal Aging

    Science.gov (United States)

    Leonard, Keith J.; Busby, Jeremy T.; Hoelzer, David T.; Zinkle, Steven J.

    2009-04-01

    The proposed uses of fission reactors for manned or deep space missions have typically relied on the potential use of refractory metal alloys as structural materials. Throughout the history of these programs, a leading candidate has been Nb-1Zr, due to its good fabrication and welding characteristics. However, the less-than-optimal creep resistance of this alloy has encouraged interest in the more complex FS-85 (Nb-28Ta-10W-1Zr) alloy. Despite this interest, only a relatively small database exists for the properties of FS-85. Database gaps include the potential microstructural instabilities that can lead to mechanical property degradation. In this work, changes in the microstructure and mechanical properties of FS-85 were investigated following 1100 hours of thermal aging at 1098, 1248, and 1398 K. The changes in electrical resistivity, hardness, and tensile properties between the as-annealed and aged materials are compared. Evaluation of the microstructural changes was performed through optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). The development of intragranular and grain-boundary precipitation of Zr-rich compounds as a function of aging temperature was followed. Brittle tensile behavior was measured in the material aged at 1248 K, while ductile behavior occurred in samples aged above and below this temperature. The effect of temperature on the under- and overaging of the grain-boundary particles is believed to have contributed to the mechanical property behavior of the aged materials.

  12. The radio structure of the peculiar narrow-line Seyfert 1 galaxy candidate J1100+4421

    Science.gov (United States)

    Gabányi, K. É.; Frey, S.; Paragi, Z.; Järvelä, E.; Morokuma, T.; An, T.; Tanaka, M.; Tar, I.

    2018-01-01

    Narrow-line Seyfert 1 galaxies (NLS1) are an intriguing subclass of active galactic nuclei. Their observed properties indicate low central black hole mass and high accretion rate. The extremely radio-loud NLS1 sources often show relativistic beaming and are usually regarded as younger counterparts of blazars. Recently, the object SDSS J110006.07+442144.3 was reported as a candidate NLS1 source. The characteristics of its dramatic optical flare indicated its jet-related origin. The spectral energy distribution of the object was similar to that of the γ-ray detected radio-loud NLS1, PMN J0948+0022. Our high-resolution European very long baseline interferometry network observations at 1.7 and 5 GHz revealed a compact core feature with a brightness temperature of ≳1010 K. Using the lowest brightness temperature value and assuming a moderate Lorentz factor of ∼9, the jet viewing angle is ≲26°. Archival Very Large Array data show a large-scale radio structure with a projected linear size of ∼150 kpc reminiscent of double-sided morphology.

  13. 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)

  14. THE CANDIDATE CLUSTER AND PROTOCLUSTER CATALOG (CCPC). II. SPECTROSCOPICALLY IDENTIFIED STRUCTURES SPANNING 2 <  z  < 6.6

    Energy Technology Data Exchange (ETDEWEB)

    Franck, J. R.; McGaugh, S. S. [Case Western Reserve University, 10900 Euclid Ave., Cleveland, OH 44106 (United States)

    2016-12-10

    The Candidate Cluster and Protocluster Catalog (CCPC) is a list of objects at redshifts z  > 2 composed of galaxies with spectroscopically confirmed redshifts that are coincident on the sky and in redshift. These protoclusters are identified by searching for groups in volumes corresponding to the expected size of the most massive protoclusters at these redshifts. In CCPC1 we identified 43 candidate protoclusters among 14,000 galaxies between 2.74 <  z  < 3.71. Here we expand our search to more than 40,000 galaxies with spectroscopic redshifts z  > 2.00, resulting in an additional 173 candidate structures. The most significant of these are 36 protoclusters with overdensities δ {sub gal} > 7. We also identify three large proto-supercluster candidates containing multiple protoclusters at z  = 2.3, 3.5 and z  = 6.56. Eight candidates with N  ≥ 10 galaxies are found at redshifts z  > 4.0. The last system in the catalog is the most distant spectroscopic protocluster candidate known to date at z  = 6.56.

  15. Water-cooled Pb-17Li test blanket module for ITER: impact of the structural material grade on the neutronic responses

    Energy Technology Data Exchange (ETDEWEB)

    Vella, G.; Aiello, G.; Oliveri, E. [Palermo Univ. (Italy). Dipt. di Ingegneria Nucl.; Fuetterer, M.A.; Giancarli, L. [CEA - Saclay, DRN/DMT/SERMA, Gif-sur-Yvette (France); Tavassoli, F. [CEA - Saclay, CEREM, Gif-sur-Yvette (France)

    1998-10-01

    The water-cooled lithium lead (WCLL) DEMO blanket is one of the two EU lines to be further developed with the aim of manufacturing by 2010 a test blanket module for ITER (TBM). In this paper results of a 3D-Monte Carlo neutronic analysis of the TBM design are reported. A fully 3D heterogeneous model of the WCLL-TBM has been inserted into an existing ITER model accounting for a proper D-T neutron source. The structural material assumed for the calculations was martensitic 9% Cr steel code named Z 10 CDV Nb 9-1. Results have been compared with those obtained using MANET. The main nuclear responses of the TBM have been determined, such as detailed power deposition density, material damage through DPA and He and H gas production rate, radial distribution of tritium production rate and total tritium production in the module. The impact of using natural lithium on the TBM system operation has also been evaluated. (orig.) 13 refs.

  16. ITER Council proceedings: 1993

    International Nuclear Information System (INIS)

    1994-01-01

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ''ITER EDA Agreement and Protocol 2'' (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tasks; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings

  17. ITER council proceedings: 2000

    International Nuclear Information System (INIS)

    2001-01-01

    No ITER Council Meetings were held during 2000. However, two ITER EDA Meetings were held, one in Tokyo, January 19-20, and one in Moscow, June 29-30. The parties participating in these meetings were those that partake in the extended ITER EDA, namely the EU, the Russian Federation, and Japan. This document contains, a/o, the records of these meetings, the list of attendees, the agenda, the ITER EDA Status Reports issued during these meetings, the TAC (Technical Advisory Committee) reports and recommendations, the MAC Reports and Advice (also for the July 1999 Meeting), the ITER-FEAT Outline Design Report, the TAC Reports and Recommendations both meetings), Site requirements and Site Design Assumptions, the Tentative Sequence of technical Activities 2000-2001, Report of the ITER SWG-P2 on Joint Implementation of ITER, EU/ITER Canada Proposal for New ITER Identification

  18. ITER council proceedings: 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ``ITER EDA Agreement and Protocol 2`` (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tastes; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings.

  19. ITER EDA Newsletter. V.2, no.5

    International Nuclear Information System (INIS)

    1993-05-01

    This ITER EDA (Engineering Design Activities), Newsletter issue includes reports on the third ITER council meeting in Tokyo on the involvement of other countries, on an outline of the report by the Management Advisory Committee (MAC), on such involvement, and on the improvement by the MAC and the ITER Council to proceed with Task Agreements on the Research and Development programme of the Superconductor Coils and Structures Division

  20. Effect of high flux plasma exposure on the micro-structural and -mechanical properties of ITER specification tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Dubinko, A., E-mail: adubinko@sckcen.be [Institute for Nuclear Material Sciences, SCK-CEN, 2400 Mol (Belgium); Department of Applied Physics, Ghent University, 9000 Ghent (Belgium); Terentyev, D. [Institute for Nuclear Material Sciences, SCK-CEN, 2400 Mol (Belgium); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoe shosse 31, 115409 Moscow (Russian Federation); Bakaeva, A. [Institute for Nuclear Material Sciences, SCK-CEN, 2400 Mol (Belgium); Department of Applied Physics, Ghent University, 9000 Ghent (Belgium); Pardoen, T. [Institute of Mechanics, Materials and Civil Engineering, Université catholique de Louvain, Place Sainte Barbe 2 L5.02.02, 1348 Louvain‐la‐Neuve (Belgium); Zibrov, M. [Department of Applied Physics, Ghent University, 9000 Ghent (Belgium); Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany); FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, De Zaale 20, 5612 AJ Eindhoven (Netherlands); Physik-Department E28, Technische Universität München, James-Franck-Straße 1, D-85748 Garching (Germany); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoe shosse 31, 115409 Moscow (Russian Federation); Morgan, T.W. [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, De Zaale 20, 5612 AJ Eindhoven (Netherlands)

    2017-02-15

    Highlights: • Plasma exposure induces dislocation-dominated microstructure. • The exposure-induced changes in microstructure vanish beyond a depth of 12–15 μm. • Surface hardness after the plasma exposure increases significantly in the sub-surface region of 1.5–3 μm. - Abstract: We have performed a combined study using transmission electron microscopy (TEM), nuclear reaction analysis (NRA) and nano-indentation (NI) techniques to reveal the impact of high flux plasma exposure on the properties of a sub-surface region of the commercially available pure tungsten fabricated following the ITER specification. TEM examination revealed the formation of a dense dislocation network and dislocation tangles, resulting in a strong increase in the dislocation density by at least one order of magnitude as compared to the bulk density. The plasma-induced dislocation microstructure vanishes within a depth of about 10–15 μm from the top of the exposed surface. Surface hardness after the plasma exposure was characterized by NI and was found to increase significantly in the sub-surface region of 1.5–3 μm. That was attributed to the resistance of the plasma-induced dislocation networks and deuterium-induced defects, whose presence within a depth of ∼1 μm was unambiguously detected by the NRA measurements as well.

  1. 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)

  2. ITER council proceedings: 1995

    International Nuclear Information System (INIS)

    1996-01-01

    Records of the 8. ITER Council Meeting (IC-8), held on 26-27 July 1995, in San Diego, USA, and the 9. ITER Council Meeting (IC-9) held on 12-13 December 1995, in Garching, Germany, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the ITER Interim Design Report Package and Relevant Documents. Figs, tabs

  3. ITER council proceedings: 1999

    International Nuclear Information System (INIS)

    1999-01-01

    In 1999 the ITER meeting in Cadarache (10-11 March 1999) and the Programme Directors Meeting in Grenoble (28-29 July 1999) took place. Both meetings were exclusively devoted to ITER engineering design activities and their agendas covered all issues important for the development of ITER. This volume presents the documents of these two important meetings

  4. ITER council proceedings: 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Records of the 10. ITER Council Meeting (IC-10), held on 26-27 July 1996, in St. Petersburg, Russia, and the 11. ITER Council Meeting (IC-11) held on 17-18 December 1996, in Tokyo, Japan, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the cost review and safety analysis. Figs, tabs

  5. 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

  6. ITER radio frequency systems

    International Nuclear Information System (INIS)

    Bosia, G.

    1998-01-01

    Neutral Beam Injection and RF heating are two of the methods for heating and current drive in ITER. The three ITER RF systems, which have been developed during the EDA, offer several complementary services and are able to fulfil ITER operational requirements

  7. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim

    2011-11-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  8. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim; El Sayed, Tamer S.; Ramadan, Hany E.

    2011-01-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop's task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application's latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  9. People's Republic of China joins ITER

    International Nuclear Information System (INIS)

    Huo Yuping

    2003-01-01

    The People's Republic of China is the largest developing country with a projected population of 1.6 - 2 billion people and an energy consumption growing from the current 1.3 Billion Tons Coal Equivalent (TCE) to more than 4 Billion TCE by 2050. This large demand needs to be accommodated in a sustainable way, requiring energy generation in an environmentally friendly way. Fusion is one of the most promising candidates to solve this important issue. This explains why in the second half of 2002, the ITER Participants' delegations to the ITER Negotiations received expression of interest from the People's Republic of China in the possibility of Chinese participation in ITER, including joining the ongoing Negotiations. The speed with which the Chinese authorities had made their decision to participate in the ITER Negotiations was impressive. The Prime Minister and the State Council had already confirmed their decision to apply to join ITER as soon as possible, and Mr. Xu Guanhua, Chinese Minister of Science and Technology, wrote on behalf of his government, on 10 January 2003, to the four heads of delegation in the ITER Negotiations, requesting that China participate in the present ITER Negotiations, pointing out that China intends to provide a substantial contribution to the Project, comparable to what is currently envisaged by some of the participants in the present Negotiations

  10. 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)

  11. ITER council proceedings: 1997

    International Nuclear Information System (INIS)

    1997-01-01

    This volume of the ITER EDA Documentation Series presents records of the 12th ITER Council Meeting, IC-12, which took place on 23-24 July, 1997 in Tampere, Finland. The Council received from the Parties (EU, Japan, Russia, US) positive responses on the Detailed Design Report. The Parties stated their willingness to contribute to fulfil their obligations in contributing to the ITER EDA. The summary discussions among the Parties led to the consensus that in July 1998 the ITER activities should proceed for additional three years with a general intent to enable an efficient start of possible, future ITER construction

  12. Fixed structure feedforward controller design exploiting iterative trials: application to a wafer stage and a desktop printer

    NARCIS (Netherlands)

    Meulen, van der S.H.; Tousain, R.L.; Bosgra, O.H.

    2008-01-01

    In this paper, the feedforward controller design problem for high-precision electromechanical servo systems that execute finite time tasks is addressed. The presented procedure combines the selection of the fixed structure of the feedforward controller and the optimization of the controller

  13. 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

  14. 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

  15. 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

  16. Technical basis for the ITER-FEAT outline design

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities.

  17. Technical basis for the ITER-FEAT outline design

    International Nuclear Information System (INIS)

    2000-01-01

    This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities

  18. Iteration and accelerator dynamics

    International Nuclear Information System (INIS)

    Peggs, S.

    1987-10-01

    Four examples of iteration in accelerator dynamics are studied in this paper. The first three show how iterations of the simplest maps reproduce most of the significant nonlinear behavior in real accelerators. Each of these examples can be easily reproduced by the reader, at the minimal cost of writing only 20 or 40 lines of code. The fourth example outlines a general way to iteratively solve nonlinear difference equations, analytically or numerically

  19. Structural stability at high pressure, electronic, and magnetic properties of BaFZnAs: A new candidate of host material of diluted magnetic semiconductors

    International Nuclear Information System (INIS)

    Chen Bi-Juan; Deng Zheng; Wang Xian-Cheng; Feng Shao-Min; Yuan Zhen; Zhang Si-Jia; Liu Qing-Qing; Jin Chang-Qing

    2016-01-01

    The layered semiconductor BaFZnAs with the tetragonal ZrCuSiAs-type structure has been successfully synthesized. Both the in-situ high-pressure synchrotron x-ray diffraction and the high-pressure Raman scattering measurements demonstrate that the structure of BaFZnAs is stable under pressure up to 17.5 GPa at room temperature. The resistivity and the magnetic susceptibility data show that BaFZnAs is a non-magnetic semiconductor. BaFZnAs is recommended as a candidate of the host material of diluted magnetic semiconductor. (special topic)

  20. Future plan of ITER

    International Nuclear Information System (INIS)

    Kitsunezaki, Akio

    1998-01-01

    In cooperation of four countries, Japan, USA, EU and Russia, ITER plan has been proceeding as ''the conceptual design activities'' from 1988 to 1990 and ''the industrial design activities'' since 1992. To construct ITER, the legal and work side of ITER operation has been investigated by four countries. However, their economic conditions have been changed to be wrong. So that, construction of ITER can not begin after end of industrial design activities in 1998. Accordingly, they determined to continue the industrial design activities more three years in order to study low cost options and to test the superconductive model·coil. (S.Y.)

  1. ITER test programme

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Casini, G.

    1991-01-01

    ITER has been designed to operate in two phases. The first phase which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER. 15 refs, figs and tabs

  2. United States rejoin ITER

    International Nuclear Information System (INIS)

    Roberts, M.

    2003-01-01

    Upon pressure from the United States Congress, the US Department of Energy had to withdraw from further American participation in the ITER Engineering Design Activities after the end of its commitment to the EDA in July 1998. In the years since that time, changes have taken place in both the ITER activity and the US fusion community's position on burning plasma physics. Reflecting the interest in the United States in pursuing burning plasma physics, the DOE's Office of Science commissioned three studies as part of its examination of the option of entering the Negotiations on the Agreement on the Establishment of the International Fusion Energy Organization for the Joint Implementation of the ITER Project. These were a National Academy Review Panel Report supporting the burning plasma mission; a Fusion Energy Sciences Advisory Committee (FESAC) report confirming the role of ITER in achieving fusion power production, and The Lehman Review of the ITER project costing and project management processes (for the latter one, see ITER CTA Newsletter, no. 15, December 2002). All three studies have endorsed the US return to the ITER activities. This historical decision was announced by DOE Secretary Abraham during his remarks to employees of the Department's Princeton Plasma Physics Laboratory. The United States will be working with the other Participants in the ITER Negotiations on the Agreement and is preparing to participate in the ITA

  3. ITER CTA newsletter. No. 3

    International Nuclear Information System (INIS)

    2001-11-01

    This ITER CTA newsletter comprises reports of Dr. P. Barnard, Iter Canada Chairman and CEO, about the progress of the first formal ITER negotiations and about the demonstration of details of Canada's bid on ITER workshops, and Dr. V. Vlasenkov, Project Board Secretary, about the meeting of the ITER CTA project board

  4. 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.)

  5. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    International Nuclear Information System (INIS)

    Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

    2008-01-01

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

  6. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

    2008-12-15

    The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

  7. ITER council proceedings: 1992

    International Nuclear Information System (INIS)

    1994-01-01

    At the signing of the ITER EDA Agreement on July, 1992, each of the Parties presented to the Director General the names of their designated members of the ITER Council. Upon receiving those names, the Director General stated that the ITER Engineering Design Activities were ''ready to begin''. The next step in this process was the convening of the first meeting of the ITER Council. The first meeting of the Council, held in Vienna, was opened by Director General Hans Blix. The second meeting was held in Moscow, the formal seat of the Council. This volume presents records of these first two Council meetings and, together with the previous volumes on the text of the Agreement and Protocol 1 and the preparations for their signing respectively, represents essential information on the evolution of the ITER EDA

  8. Full Dynamic Analysis of Mooring Solution Candidates - First Iteration

    DEFF Research Database (Denmark)

    Thomsen, Jonas Bjerg; Ferri, Francesco

    This report covers an initial full dynamic analysis of the mooring solutions for the four wave energy converters in the project “Mooring Solutions for Large Wave Energy Converters”. The analysis tends to provide the first understanding of the layouts and provide discussion on what parameters that...

  9. Testing of the OMERACT 8 draft validation criteria for a soluble biomarker reflecting structural damage in rheumatoid arthritis: a systematic literature search on 5 candidate biomarkers

    DEFF Research Database (Denmark)

    Syversen, Silje W; Landewe, Robert; van der Heijde, Désirée

    2009-01-01

    OBJECTIVE: To test the OMERACT 8 draft validation criteria for soluble biomarkers by assessing the strength of literature evidence in support of 5 candidate biomarkers. METHODS: A systematic literature search was conducted on the 5 soluble biomarkers RANKL, osteoprotegerin (OPG), matrix...... metalloprotease (MMP-3), urine C-telopeptide of types I and II collagen (U-CTX-I and U CTX-II), focusing on the 14 OMERACT 8 criteria. Two electronic voting exercises were conducted to address: (1) strength of evidence for each biomarker as reflecting structural damage according to each individual criterion...

  10. ITER towards the construction

    International Nuclear Information System (INIS)

    Shimomura, Y.

    2005-01-01

    The ITER Project has been significantly developed in the last few years in preparation for its construction. The ITER Participant's Negotiators have developed the Joint Implementation Agreement (JIA), ready for finalisation following selection of the construction site and nomination of the project's Director General. The ITER International Team and Participant Teams have continued technical and organisational preparations. Construction will be able to start immediately after the international ITER organisation is established, following signature of the JIA. The Project is strongly supported by the governments of the Participants as well as by the scientific community. The real negotiations, including siting and the final details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. Their wish to host ITER construction is too strong to allow convergence to a single site considering the ITER device in isolation. A broader collaboration among the Parties is therefore being contemplated, covering complementary activities to help accelerate fusion development towards a viable power source, and allow the Participants to reach a conclusion on ITER siting. This report reviews these preparations, and the status of negotiations

  11. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  12. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  13. Iterative Prototyping of Strategy Implementation Workshop Design

    DEFF Research Database (Denmark)

    Kryger, Anders

    2018-01-01

    Purpose: The purpose of this paper is to demonstrate how a strategy implementation workshop design can be developed and tested while minimizing the time spent on developing the design. Design/methodology/approach: This multiple case study at a diesel engine company shows how iterative prototyping...... can be used to structure the design process of a strategy implementation workshop. Findings: Strategy implementation workshop design can be developed in resource-constrained environments through iterative prototyping of the workshop design. Each workshop iteration can generate value in its own right...... draw on his/her experience as well as add to his/her knowledge base. Originality/value: Introducing iterative prototyping in an organizational context can facilitate fast yet structured development of a rigorous workshop design. Strategy consultants are provided with empirical examples of how...

  14. Structural materials for high-heat flux applications

    International Nuclear Information System (INIS)

    Rybin, V.V.; Smith, D.L.

    1991-01-01

    The structural materials for the ITER, (International Thermonuclear Experimental Reactor) divertor must perform reliably under complex and diverse operating requirements. Only a limited number of materials offer a potential for meeting these requirements for the wide temperature range of interest. The candidate materials considered in the ITER design activity include copper, molybdenum, niobium alloys. Molybdenum alloys being considered include dilute alloys of the TZM type and the Mo-Re system. Niobium alloys under consideration include Nb-V-Zr and Nb-Zr systems. Copper alloys being considered include precipitation strengthened alloys of the Glidcop and MAGT type, alloys of Cu-Mo system and dispersion hardened bronzes. The projected operating conditions for the ITER divertor and the criteria for evaluating the candidate materials are reviewed. This paper summarizes the data base and presents recent experimental results on these candidate divertor structural alloys

  15. Quantitative Structure-Use Relationship Model thresholds for Model Validation, Domain of Applicability, and Candidate Alternative Selection

    Data.gov (United States)

    U.S. Environmental Protection Agency — This file contains value of the model training set confusion matrix, domain of applicability evaluation based on training set to predicted chemicals structural...

  16. 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

  17. 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.

  18. Design iteration in construction projects – Review and directions

    Directory of Open Access Journals (Sweden)

    Purva Mujumdar

    2018-03-01

    Full Text Available Design phase of any construction project involves several designers who exchange information with each other most often in an unstructured manner throughout the design phase. When these information exchanges happen to occur in cycles/loops, it is termed as design iteration. Iteration is an inherent and unavoidable aspect of any design phase which requires proper planning. Till date, very few researchers have explored the design iteration (“complexity” in construction sector. Hence, the objective of this paper was to document and review the complexities of iteration during design phase of construction projects for efficient design planning. To achieve this objective, exhaustive literature review on design iteration was done for four sectors – construction, manufacturing, aerospace, and software development. In addition, semi-structured interviews and discussions were done with a few design experts to verify the different dimensions of iteration. Finally, a design iteration framework was presented in this study that facilitates successful planning. Keywords: Design iteration, Types of iteration, Causes and impact of iteration, Models of iteration, Execution strategies of iteration

  19. 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

  20. Toward construction of ITER

    International Nuclear Information System (INIS)

    Shimomura, Yasuo

    2005-01-01

    The ITER Project has been significantly developed in the past years in preparation for its construction. The ITER Negotiators have developed a draft Joint Implementation Agreement (JIA), ready for completion following the nomination of the Project's Director General (DG). The ITER International Team and Participant Teams have continued technical and organizational preparations. The actual construction will be able to start immediately after the international ITER organization will be established, following signature of the JIA. The Project is now strongly supported by all the participants as well as by the scientific community with the final high-level negotiations, focused on siting and the concluding details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. The extent to which they both wish to host the ITER facility is such that large contributions to a broader collaboration among the Parties are also proposed by them. This covers complementary activities to help accelerate fusion development towards a viable power source, as well as may allow the Participants to reach a conclusion on ITER siting. (author)

  1. 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.

  2. Extensive characterisation of advanced manufacturing solutions for the ITER Central Solenoid pre-compression system

    CERN Document Server

    Langeslag, S.A.E.; Libeyre, P.; Marcinek, D.J.; Zhang, Z.

    2015-01-01

    The ITER Central Solenoid (CS), positioned in the center of the ITER tokamak, will provide a magnetic field, contributing to the confinement of the plasma. The 13 m high CS consists of a vertical stack of 6 independently driven modules, dynamically activated. Resulting opposing currents can lead to high separation forces. A pre-compression structure is implemented to counteract these opposing forces, by realising a continuous 180 MN coil-to-coil contact loading. Preload is applied by mechanical fastening via 9 subunits, positioned along the coil stack, each consisting of 2 outer and 1 inner tie plate. The tie plates therefore need to feature outstanding mechanical behaviour in a large temperature range. High strength, Nitronic®-50 type F XM-19 austenitic stainless steel is selected as candidate material. The linearised stress distribution reaches approximately 250 MPa, leading to a required yield strength of 380 MPa at room temperature. Two different manufacturing methods are being studied for the procuremen...

  3. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.N. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [EFDA-CSU, Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Pestchanyi, S.E. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2007-06-15

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  4. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Science.gov (United States)

    Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Loarte, A.; Pestchanyi, S. E.

    2007-06-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  5. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    International Nuclear Information System (INIS)

    Bazylev, B.N.; Janeschitz, G.; Landman, I.S.; Loarte, A.; Pestchanyi, S.E.

    2007-01-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated

  6. The ITER Remote Maintenance Management System

    International Nuclear Information System (INIS)

    Tesini, Alessandro; Rolfe, A.C.

    2009-01-01

    A major challenge for the ITER project is to develop and implement a Remote Maintenance System, which can deliver high Tokamak availability within the constraints of the overall ITER programme objectives. Much of the maintenance of ITER will be performed using remote handling methods and some with combined manual and remote activities working together. The organization and management of the ITER remote handling facilities will be of a scale unlike any other remote handling application in the world. The ITER remote handling design and procurement activities will require co-ordination and management across many different sites throughout the world. It will be a major challenge for the ITER project to ensure a consistent quality and technical approach in all of the contributing parties. To address this issue the IO remote handling team are implementing the ITER Maintenance Management Plan (IMMP) comprising an overarching document defining the policies and methodologies (ITER Remote Maintenance Management System or IMMS) and an associated ITER remote handling code of practise (IRHCOP). The IMMS will be in document form available as a pdf file or similar. The IRHCOP will be implemented as a web based application and will provide access to the central resource of the entire code of practise from any location in the world. The IRHCOP data library will be centrally controlled in order that users can be assured of the data relevance and authenticity. This paper will describe the overall approach being taken to deal with this challenge and go on to detail the structure and content of both the IMMS and the IRHCOP.

  7. Eighth ITER negotiations meeting (N-8)

    International Nuclear Information System (INIS)

    Korzhavin, V.

    2003-01-01

    The eighth ITER Negotiations meeting was held on 18-19 February, 2003 in St. Petersburg, Russia. The delegations of People's Republic of China and the USA joined those from Canada, The European Union, Japan and the Russian federation in their efforts to reach agreement on the implementation of the ITER project. The delegations took note of the progress of discussions on procurement allocations, ITER decommissioning issues, management structure and intellectual property rights. The Negotiators approved the report on the Joint Assessment of Specific Sites and noted the report on the start of the ITER Transitional Arrangements (ITA). The delegations also noted that China has stated its willingness to participate in ITA and that the USA is considering participation

  8. ITER CTA newsletter. No. 6

    International Nuclear Information System (INIS)

    2002-01-01

    This ITER CTA Newsletter issue comprises information about the following ITER Meetings: The second negotiation meeting on the joint implementation of ITER, held in Tokyo(Japan) on 22-23 January 2002, and an international ITER symposium on burning plasma science and technology, held the day later after the second negotiation meeting at the same place

  9. ITER CTA newsletter. No. 2

    International Nuclear Information System (INIS)

    2001-10-01

    This ITER CTA newsletter contains results of the ITER toroidal field model coil project presented by ITER EU Home Team (Garching) and an article in commemoration of the late Dr. Charles Maisonnier, one of the former leaders of ITER who made significant contributions to its development

  10. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

    The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

  11. 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)

  12. On the theoretical–numerical study of the ITER Upper Port Plug structure hydraulic behaviour under steady state and draining and drying transient conditions

    International Nuclear Information System (INIS)

    Di Maio, P.A.; Paradiso, D.; Dell’Orco, G.; Pitcher, C.S.; Kalish, M.

    2011-01-01

    Highlights: ► UPP TS hydraulic behaviour has been investigated under steady state and D and D transient conditions. ► A thermal–hydraulic system code has been adopted and a UPP TS model has been set-up and validated against results of steady state CFD analyses. ► The TS steady state hydraulic characteristic functions have been derived for two coolant flow paths showing that right plate inlet one is the most promising. ► Draining simulations indicate that the 4 MPa injection pressure is high enough to drain almost completely the circuit in a reasonable time (∼6 s). ► Results indicate that right plate inlet flow path allows the TS complete draining, eliminating the need for the drying procedure. - Abstract: The ITER diagnostic Upper Port Plug (UPP) is a water-cooled stainless steel structure aimed to integrate within vacuum vessel the plasma diagnostic systems, shielding them from neutron and photon irradiation. Due to the very intense heat loads expected, a proper cooling circuit has been designed to ensure an adequate UPP cooling with an acceptable thermal rise and an unduly high pumping power and to perform its draining and drying procedure by injection of pressurized nitrogen. A theoretical research activity has been launched at the Department of Nuclear Engineering of the University of Palermo aiming to investigate the hydraulic behaviour of the UPP Trapezoid Section cooling circuit under steady state conditions and during its draining and drying transient procedure. The research activity has been performed following a theoretical–computational approach and adopting the RELAP5 thermal–hydraulic system code. The Trapezoid Section cooling circuit characteristic functions have been derived under steady state conditions at various coolant temperatures for both the coolant flow paths at the present under consideration for this circuit. The distributions of coolant mass flow rates along the channels of the cooling circuit have been calculated too

  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. The Use of Chemical-Chemical Interaction and Chemical Structure to Identify New Candidate Chemicals Related to Lung Cancer.

    Directory of Open Access Journals (Sweden)

    Lei Chen

    Full Text Available Lung cancer causes over one million deaths every year worldwide. However, prevention and treatment methods for this serious disease are limited. The identification of new chemicals related to lung cancer may aid in disease prevention and the design of more effective treatments. This study employed a weighted network, constructed using chemical-chemical interaction information, to identify new chemicals related to two types of lung cancer: non-small lung cancer and small-cell lung cancer. Then, a randomization test as well as chemical-chemical interaction and chemical structure information were utilized to make further selections. A final analysis of these new chemicals in the context of the current literature indicates that several chemicals are strongly linked to lung cancer.

  15. 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

  16. 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.)

  17. 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)

  18. ITER plant systems

    International Nuclear Information System (INIS)

    Kolbasov, B.; Barnes, C.; Blevins, J.

    1991-01-01

    As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

  19. 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

  20. It was the demonstration of industrial steel production capacity ferritic-martensitic Spanish ASTURFER scale demand ITER; Hacia la demostracion de capacidad de produccion industrial del acero ferritico-martensitico espanol ASTURFER a escala de demanda ITER

    Energy Technology Data Exchange (ETDEWEB)

    Coto, R.; Serrano, M.; Moran, A.; Rodriguez, D.; Artimez, J. A.; Belzunce, J.; Sedano, L.

    2013-07-01

    Reduced Activation Ferritic-Martensitic (RAFM) structural steels are considered as candidate materials with notable possibilities to be incorporated to fusion reactor ITER, nowadays under construction, and future fusion reactor DEMO, involving a notable forecasting of supply materials, with a considerable limitation due to the few number of furnishes currently on the market. The manufacture at an industrial scale of the ASTURFER steel, developed at laboratory scale by ITMA Materials Technology and the Structural Materials Division of the Technology Division of CIEMAT would be a significant business opportunity for steelwork companies.

  1. Dark matter candidates

    International Nuclear Information System (INIS)

    Turner, M.S.

    1989-01-01

    One of the simplest, yet most profound, questions we can ask about the Universe is, how much stuff is in it, and further what is that stuff composed of? Needless to say, the answer to this question has very important implications for the evolution of the Universe, determining both the ultimate fate and the course of structure formation. Remarkably, at this late date in the history of the Universe we still do not have a definitive answer to this simplest of questions---although we have some very intriguing clues. It is known with certainty that most of the material in the Universe is dark, and we have the strong suspicion that the dominant component of material in the Cosmos is not baryons, but rather is exotic relic elementary particles left over from the earliest, very hot epoch of the Universe. If true, the Dark Matter question is a most fundamental one facing both particle physics and cosmology. The leading particle dark matter candidates are: the axion, the neutralino, and a light neutrino species. All three candidates are accessible to experimental tests, and experiments are now in progress. In addition, there are several dark horse, long shot, candidates, including the superheavy magnetic monopole and soliton stars. 13 refs

  2. Physics research needs for ITER

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1995-01-01

    Design of ITER entails the application of physics design tools that have been validated against the world-wide data base of fusion research. In many cases, these tools do not yet exist and must be developed as part of the ITER physics program. ITER's considerable increases in power and size demand significant extrapolations from the current data base; in several cases, new physical effects are projected to dominate the behavior of the ITER plasma. This paper focuses on those design tools and data that have been identified by the ITER team and are not yet available; these needs serve as the basis for the ITER Physics Research Needs, which have been developed jointly by the ITER Physics Expert Groups and the ITER design team. Development of the tools and the supporting data base is an on-going activity that constitutes a significant opportunity for contributions to the ITER program by fusion research programs world-wide

  3. It was the demonstration of industrial steel production capacity ferritic-martensitic Spanish ASTURFER scale demand ITER

    International Nuclear Information System (INIS)

    Coto, R.; Serrano, M.; Moran, A.; Rodriguez, D.; Artimez, J. A.; Belzunce, J.; Sedano, L.

    2013-01-01

    Reduced Activation Ferritic-Martensitic (RAFM) structural steels are considered as candidate materials with notable possibilities to be incorporated to fusion reactor ITER, nowadays under construction, and future fusion reactor DEMO, involving a notable forecasting of supply materials, with a considerable limitation due to the few number of furnishes currently on the market. The manufacture at an industrial scale of the ASTURFER steel, developed at laboratory scale by ITMA Materials Technology and the Structural Materials Division of the Technology Division of CIEMAT would be a significant business opportunity for steelwork companies.

  4. Assessment of vanadium alloys for ITER application

    International Nuclear Information System (INIS)

    Borgstedt, H.U.; Clemens, H.; Ehrlich, K.; Fromm, E.; Kelzenberg, S.; Moeslang, A.; Pick, M.; Ruehle, M.; Schaaf, B. van der; Schaefer, L.; Schiller, P.; Schirra, M.; Witwer, M.; Witzenburg, W. van; Zolti, E.; Zucchetti, M.

    1993-09-01

    The assessment effort concerned required evaluation of various relevant properties of vanadium alloys. The outcome predictably shows that these properties, as well as timing, funding, manufacturing and licensing aspects, each set their own specific boundary conditions for application of these alloys in ITER. Some of these are not really felt as constraints. Their capacity to accommodate high heat loads, for example, is better than other candidate materials and appears to be the main reason for the present interest in these alloys. Other favourable properties include neutronic properties (low nuclear heating rates, good tritium breeding performance and low helium generation rates), intrinsically low activation, excellent tensile and creep properties up to high temperatures and high strength-to-density ratio. Not all of these properties necessarily are relevant for ITER, but they would be important for longer term application. (orig.)

  5. Studying uniform thickness II: Transversely nonsimple iterated torus knots

    DEFF Research Database (Denmark)

    LaFountain, Douglas

    2011-01-01

    We prove that an iterated torus knot type in the standard contact 3-sphere fails the uniform thickness property (UTP) if and only if it is formed from repeated positive cablings, which is precisely when an iterated torus knot supports the standard contact structure. This is the first complete UTP...... classification for a large class of knots. We also show that all iterated torus knots that fail the UTP support cabling knot types that are transversely non-simple....

  6. Iterative List Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom; Hjaltason, Johan

    2005-01-01

    We analyze the relation between iterative decoding and the extended parity check matrix. By considering a modified version of bit flipping, which produces a list of decoded words, we derive several relations between decodable error patterns and the parameters of the code. By developing a tree...... of codewords at minimal distance from the received vector, we also obtain new information about the code....

  7. 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.

  8. 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

  9. 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)

  10. 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.)

  11. 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

  12. 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)

  13. US ITER Management Plan

    International Nuclear Information System (INIS)

    1991-12-01

    This US ITER Management Plan is the plan for conducting the Engineering Design Activities within the US. The plan applies to all design, analyses, and associated physics and technology research and development (R ampersand D) required to support the program. The plan defines the management considerations associated with these activities. The plan also defines the management controls that the project participants will follow to establish, implement, monitor, and report these activities. The activities are to be conducted by the project in accordance with this plan. The plan will be updated to reflect the then-current management approach required to meet the project objectives. The plan will be reviewed at least annually for possible revision. Section 2 presents the ITER objectives, a brief description of the ITER concept as developed during the Conceptual Design Activities, and comments on the Engineering Design Activities. Section 3 discusses the planned international organization for the Engineering Design Activities, from which the tasks will flow to the US Home Team. Section 4 describes the US ITER management organization and responsibilities during the Engineering Design Activities. Section 5 describes the project management and control to be used to perform the assigned tasks during the Engineering Design Activities. Section 6 presents the references. Several appendices are provided that contain detailed information related to the front material

  14. 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

  15. 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

  16. 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)

  17. Iterative software kernels

    Energy Technology Data Exchange (ETDEWEB)

    Duff, I.

    1994-12-31

    This workshop focuses on kernels for iterative software packages. Specifically, the three speakers discuss various aspects of sparse BLAS kernels. Their topics are: `Current status of user lever sparse BLAS`; Current status of the sparse BLAS toolkit`; and `Adding matrix-matrix and matrix-matrix-matrix multiply to the sparse BLAS toolkit`.

  18. 11 CFR 110.13 - Candidate debates.

    Science.gov (United States)

    2010-01-01

    ... debates include at least two candidates; and (2) The staging organization(s) does not structure the... PROHIBITIONS § 110.13 Candidate debates. (a) Staging organizations. (1) Nonprofit organizations described in 26..., subparts D and E. (b) Debate structure. The structure of debates staged in accordance with this section and...

  19. ITER Safety and Licensing

    International Nuclear Information System (INIS)

    Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

    2006-01-01

    The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

  20. Behaviour of the ASDEX pressure gauge at high neutral gas pressure and applications for ITER

    International Nuclear Information System (INIS)

    Scarabosio, A.; Haas, G.

    2008-01-01

    The ASDEX Pressure Gauge is, at present, the main candidate for in-vessel neutral pressure measurement in ITER. Although the APG output is found to saturate at around 15 Pa, below the ITER requirement of 20 Pa. We show, here, that with small modifications of the gauge geometry and potentials settings we can achieve satisfactory behaviour up to 30 Pa at 6 T

  1. Status of ITER

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

    At the end of engineering design activities (EDA) in July 2001, all the essential elements became available to make a decision on construction of ITER. A sufficiently detailed and integrated engineering design now exists for a generic site, has been assessed for feasibility, and costed, and essential physics and technology R and D has been carried out to underpin the design choices. Formal negotiations have now begun between the current participants--Canada, Euratom, Japan, and the Russian Federation--on a Joint Implementation Agreement for ITER which also establishes the legal entity to run ITER. These negotiations are supported on technical aspects by Coordinated Technical Activities (CTA), which maintain the integrity of the project, for the good of all participants, and concentrate on preparing for procurement by industry of the longest lead items, and for formal application for a construction license with the host country. This paper highlights the main features of the ITER design. With cryogenically-cooled magnets close to neutron-generating plasma, the design of shielding with adequate access via port plugs for auxiliaries such as heating and diagnostics, and of remote replacement and refurbishing systems for in-vessel components, are particularly interesting nuclear technology challenges. Making a safety case for ITER to satisfy potential regulators and to demonstrate, as far as possible at this stage, the environmental attractiveness of fusion as an energy source, is also important. The paper gives illustrative details on this work, and an update on the progress of technical preparations for construction, as well as the status of the above negotiations

  2. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  3. A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

    2007-07-01

    The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

  4. Electromagnetic scattering using the iterative multi-region technique

    CERN Document Server

    Al Sharkawy, Mohamed H

    2007-01-01

    In this work, an iterative approach using the finite difference frequency domain method is presented to solve the problem of scattering from large-scale electromagnetic structures. The idea of the proposed iterative approach is to divide one computational domain into smaller subregions and solve each subregion separately. Then the subregion solutions are combined iteratively to obtain a solution for the complete domain. As a result, a considerable reduction in the computation time and memory is achieved. This procedure is referred to as the iterative multiregion (IMR) technique.Different enhan

  5. 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)

  6. ITER CTA newsletter. No. 4

    International Nuclear Information System (INIS)

    2001-12-01

    This ITER CTA Newsletter contains information about the organization of the ITER Co-ordinated Technical Activities (CTA) International Team as the follow-up of the ITER CTA project board meeting in Toronto on 7 November 2001. It also includes a summary on the start of the international tokamak physics activity by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee

  7. ITER CTA newsletter. No. 9

    International Nuclear Information System (INIS)

    2002-06-01

    This ITER CTA newsletter contains information about the Fourth Negotiations Meeting on the Joint Implementation of ITER held in Cadarache, France on 4-6 June 2002 and about the meeting of the ITER CTA Project Board which took place on the occasion of the N4 Meeting at Cadarache on 3-4 June 2002

  8. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 23 February in Garching, Germany. The main topics were: the consideration of the report by the Director on the ITER EDA Status, the review of the Work Programme, the review of the Joint Fund, the review of a schedule of ITER meetings, and the arrangements for termination and wind-up of the EDA

  9. ITER CTA newsletter. No. 1

    International Nuclear Information System (INIS)

    2001-01-01

    This ITER CTA newsletter comprises reports on ITER co-ordinated technical activities, information about the Meeting of the ITER CTA project board which took place in Vienna on 16 July 2001, and the Meeting of the expert group on MHD, disruptions and plasma control which was held on 25-26 June 2001 in Funchal, Madeira

  10. Status of the ITER EDA

    International Nuclear Information System (INIS)

    Aymar, R.

    2000-01-01

    This article summarizes progress made in the ITER Engineering Design Activities in the period between the ITER Meeting in Tokyo (January 2000) and June 2000. Topics: Termination of EDA, Joint Central Team and Support, Task Assignments, ITER Physics, Urgent and High Priority Physics Research Areas

  11. iHadoop: Asynchronous Iterations Support for MapReduce

    KAUST Repository

    Elnikety, Eslam

    2011-08-01

    MapReduce is a distributed programming framework designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter- iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This thesis also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  12. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola

    2014-05-02

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  13. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola; Hanafy, Sherif M.; Schuster, Gerard T.

    2014-01-01

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  14. ITER technical basis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-01-01

    Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties.

  15. ITER technical basis

    International Nuclear Information System (INIS)

    2002-01-01

    Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties

  16. Developments in ITER physics R and D

    International Nuclear Information System (INIS)

    Aymar, R.

    1999-01-01

    The monograph details how by restructuring the ITER Physics Experts Groups the physics basis for the design effort can be improved while aiming for options for cost reduction. Tables with the new expert group structure are given together with tables of urgent physics research areas and personnel listings of the physics expert groups

  17. Cryptanalysis of an Iterated Halving-based hash function: CRUSH

    DEFF Research Database (Denmark)

    Bagheri, Nasour; Henricksen, Matt; Knudsen, Lars Ramkilde

    2009-01-01

    Iterated Halving has been suggested as a replacement to the Merkle–Damgård (MD) construction in 2004 anticipating the attacks on the MDx family of hash functions. The CRUSH hash function provides a specific instantiation of the block cipher for Iterated Halving. The authors identify structural pr...

  18. Magnet design technical report---ITER definition phase

    International Nuclear Information System (INIS)

    Henning, C.

    1989-01-01

    This report contains papers on the following topics: conceptual design; radiation damage of ITER magnet systems; insulation system of the magnets; critical current density and strain sensitivity; toroidal field coil structural analysis; stress analysis for the ITER central solenoid; and volt-second capabilities and PF magnet configurations

  19. Towards the procurement of the ITER divertor

    International Nuclear Information System (INIS)

    Merola, M.; Tivey, R.; Martin, A.; Pick, M.

    2006-01-01

    The procurement of the ITER divertor is planned to start in 2009. On the basis of the present common understanding of the sharing of the ITER components, the Japanese Participating Team (JAPT) will supply the outer vertical target, the Russian Federation (RF) PT the dome liner and will perform the high heat flux testing, the EU PT will supply the inner vertical targets and the cassette bodies, including final assembly of the divertor plasma-facing components (PFCs). The manufacturing of the PFCs of the ITER divertor represents a challenging endeavor due to the high technologies which are involved, and due to the unprecedented series production. To mitigate the associated risks, special arrangements need to be put in place prior to and during procurement to ensure quality and to keep to the time schedule. Before procurement can start, an ITER review of the qualification and production capability of each candidate PT is planned. Well in advance of the assumed start of the procurement, each PT which would like to contribute to the divertor PFC procurement, should first demonstrate its technical qualification to carry out the procurement with the required quality, and in an efficient and timely manner. Appropriate precautions, like subdivision of the procurement into stages, are also to be adopted during the procurement phase to mitigate the consequences of possible unexpected manufacturing problems. In preparation for writing the procurement specification for the vertical targets, the topic of setting acceptance criteria is also being addressed. This activity has the objective of defining workable acceptance criteria for the PFC armour joints. A complete set of analyses is also in progress to assess the latest design modifications against the design requirements. This task includes neutronic, shielding, thermo-mechanical and electromagnetic analyses. More than half of the ITER plasma parameters that must be measured and the related diagnostics are located in the

  20. Conformable variational iteration method

    Directory of Open Access Journals (Sweden)

    Omer Acan

    2017-02-01

    Full Text Available In this study, we introduce the conformable variational iteration method based on new defined fractional derivative called conformable fractional derivative. This new method is applied two fractional order ordinary differential equations. To see how the solutions of this method, linear homogeneous and non-linear non-homogeneous fractional ordinary differential equations are selected. Obtained results are compared the exact solutions and their graphics are plotted to demonstrate efficiency and accuracy of the method.

  1. Iterated Leavitt Path Algebras

    International Nuclear Information System (INIS)

    Hazrat, R.

    2009-11-01

    Leavitt path algebras associate to directed graphs a Z-graded algebra and in their simplest form recover the Leavitt algebras L(1,k). In this note, we introduce iterated Leavitt path algebras associated to directed weighted graphs which have natural ± Z grading and in their simplest form recover the Leavitt algebras L(n,k). We also characterize Leavitt path algebras which are strongly graded. (author)

  2. Metrology for ITER Assembly

    International Nuclear Information System (INIS)

    Bogusch, E.

    2006-01-01

    The overall dimensions of the ITER Tokamak and the particular assembly sequence preclude the use of conventional optical metrology, mechanical jigs and traditional dimensional control equipment, as used for the assembly of smaller, previous generation, fusion devices. This paper describes the state of the art of the capabilities of available metrology systems, with reference to the previous experience in Fusion engineering and in other industries. Two complementary procedures of transferring datum from the primary datum network on the bioshield to the secondary datum s inside the VV with the desired accuracy of about 0.1 mm is described, one method using the access directly through the ports and the other using transfer techniques, developed during the co-operation with ITER/EFDA. Another important task described is the development of a method for the rapid and easy measurement of the gaps between sectors, required for the production of the customised splice plates between them. The scope of the paper includes the evaluation of the composition and cost of the systems and team of technical staff required to meet the requirements of the assembly procedure. The results from a practical, full-scale demonstration of the methodologies used, using the proposed equipment, is described. This work has demonstrated the feasibility of achieving the necessary accuracies for the successful building of ITER. (author)

  3. The ITER tritium systems

    International Nuclear Information System (INIS)

    Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

    2007-01-01

    ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

  4. Neutron cameras for ITER

    International Nuclear Information System (INIS)

    Johnson, L.C.; Barnes, C.W.; Batistoni, P.

    1998-01-01

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from 16 N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with 16 N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins

  5. 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

  6. ITER CTA newsletter. No. 10

    International Nuclear Information System (INIS)

    2002-07-01

    This ITER CTA newsletter issue comprises the ITER backgrounder, which was approved as an official document by the participants in the Negotiations on the ITER Implementation agreement at their fourth meeting, held in Cadarache from 4-6 June 2002, and information about two ITER meetings: one is the third meeting of the ITER parties' designated Safety Representatives, which took place in Cadarache, France from 6-7 June 2002, and the other is the second meeting of the International Tokamak Physics Activity (ITPA) topical group on diagnostics, which was held at General Atomics, San Diego, USA, from 4-8 March 2002

  7. 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

  8. Selection of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-01-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

  9. Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials

    International Nuclear Information System (INIS)

    Thompson, L.D.

    1981-02-01

    This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H 2 /5 Pa CO/5 Pa CH 4 / 2 O (500 μatm H 2 /50 μatm CO/50 μatm CH 4 / 2 O) at 900 0 C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables

  10. ITER EDA newsletter. V. 7, no. 7

    International Nuclear Information System (INIS)

    1998-07-01

    This newsletter contains the articles: 'Extraordinary ITER council meeting', 'ITER EDA final safety meeting' and 'Summary report of the 3rd combined workshop of the ITER confinement and transport and ITER confinement database and modeling expert groups'

  11. 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)

  12. 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)

  13. Iterative-build OMIT maps: map improvement by iterative model building and refinement without model bias

    International Nuclear Information System (INIS)

    Terwilliger, Thomas C.; Grosse-Kunstleve, Ralf W.; Afonine, Pavel V.; Moriarty, Nigel W.; Adams, Paul D.; Read, Randy J.; Zwart, Peter H.; Hung, Li-Wei

    2008-01-01

    An OMIT procedure is presented that has the benefits of iterative model building density modification and refinement yet is essentially unbiased by the atomic model that is built. A procedure for carrying out iterative model building, density modification and refinement is presented in which the density in an OMIT region is essentially unbiased by an atomic model. Density from a set of overlapping OMIT regions can be combined to create a composite ‘iterative-build’ OMIT map that is everywhere unbiased by an atomic model but also everywhere benefiting from the model-based information present elsewhere in the unit cell. The procedure may have applications in the validation of specific features in atomic models as well as in overall model validation. The procedure is demonstrated with a molecular-replacement structure and with an experimentally phased structure and a variation on the method is demonstrated by removing model bias from a structure from the Protein Data Bank

  14. Citizen Candidates Under Uncertainty

    OpenAIRE

    Eguia, Jon X.

    2005-01-01

    In this paper we make two contributions to the growing literature on "citizen-candidate" models of representative democracy. First, we add uncertainty about the total vote count. We show that in a society with a large electorate, where the outcome of the election is uncertain and where winning candidates receive a large reward from holding office, there will be a two-candidate equilibrium and no equilibria with a single candidate. Second, we introduce a new concept of equilibrium, which we te...

  15. ITER EDA newsletter. V. 10, special issue

    International Nuclear Information System (INIS)

    2001-07-01

    This ITER EDA Newsletter includes summaries of the reports of ITER EDA JCT Physics unit about ITER physics R and D during the Engineering Design Activities (EDA), ITER EDA JCT Naka JWC ITER technology R and D during the EDA, and Safety, Environment and Health group of ITER EDA JCT, Garching JWS on EDA activities related to safety

  16. Iterative and iterative-noniterative integral solutions in 3-loop massive QCD calculations

    International Nuclear Information System (INIS)

    Ablinger, J.; Radu, C.S.; Schneider, C.; Behring, A.; Imamoglu, E.; Van Hoeij, M.; Von Manteuffel, A.; Raab, C.G.

    2017-11-01

    Various of the single scale quantities in massless and massive QCD up to 3-loop order can be expressed by iterative integrals over certain classes of alphabets, from the harmonic polylogarithms to root-valued alphabets. Examples are the anomalous dimensions to 3-loop order, the massless Wilson coefficients and also different massive operator matrix elements. Starting at 3-loop order, however, also other letters appear in the case of massive operator matrix elements, the so called iterative non-iterative integrals, which are related to solutions based on complete elliptic integrals or any other special function with an integral representation that is definite but not a Volterra-type integral. After outlining the formalism leading to iterative non-iterative integrals,we present examples for both of these cases with the 3-loop anomalous dimension γ (2) qg and the structure of the principle solution in the iterative non-interative case of the 3-loop QCD corrections to the ρ-parameter.

  17. Iterative and iterative-noniterative integral solutions in 3-loop massive QCD calculations

    Energy Technology Data Exchange (ETDEWEB)

    Ablinger, J.; Radu, C.S.; Schneider, C. [Johannes Kepler Univ., Linz (Austria). Research Inst. for Symbolic Computation (RISC); Behring, A. [RWTH Aachen Univ. (Germany). Inst. fuer Theoretische Teilchenphysik und Kosmologie; Bluemlein, J.; Freitas, A. de [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Imamoglu, E.; Van Hoeij, M. [Florida State Univ., Tallahassee, FL (United States). Dept. of Mathematics; Von Manteuffel, A. [Michigan State Univ., East Lansing, MI (United States). Dept. of Physics and Astronomy; Raab, C.G. [Johannes Kepler Univ., Linz (Austria). Inst. for Algebra

    2017-11-15

    Various of the single scale quantities in massless and massive QCD up to 3-loop order can be expressed by iterative integrals over certain classes of alphabets, from the harmonic polylogarithms to root-valued alphabets. Examples are the anomalous dimensions to 3-loop order, the massless Wilson coefficients and also different massive operator matrix elements. Starting at 3-loop order, however, also other letters appear in the case of massive operator matrix elements, the so called iterative non-iterative integrals, which are related to solutions based on complete elliptic integrals or any other special function with an integral representation that is definite but not a Volterra-type integral. After outlining the formalism leading to iterative non-iterative integrals,we present examples for both of these cases with the 3-loop anomalous dimension γ{sup (2)}{sub qg} and the structure of the principle solution in the iterative non-interative case of the 3-loop QCD corrections to the ρ-parameter.

  18. The ITER divertor concept

    International Nuclear Information System (INIS)

    Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

    1995-01-01

    The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

  19. Iteration of adjoint equations

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1994-01-01

    Adjoint functions are the basis of variational methods and now widely used for perturbation theory and its extension to higher order theory as used, for example, in modelling fuel burnup and optimization. In such models, the adjoint equation is to be solved in a critical system with an adjoint source distribution that is not zero but has special properties related to ratios of interest in critical systems. Consequently the methods of solving equations by iteration and accumulation are reviewed to show how conventional methods may be utilized in these circumstances with adequate accuracy. (author). 3 refs., 6 figs., 3 tabs

  20. Feedback and rotational stabilization of resistive wall modes in ITER

    International Nuclear Information System (INIS)

    Liu Yueqiang; Bondeson, A.; Chu, M.S.; La Haye, R.J.; Favez, J.-Y.; Lister, J.B.; Gribov, Y.; Gryaznevich, M.; Hender, T.C.; Howell, D.F.

    2005-01-01

    Different models have been introduced in the stability code MARS-F in order to study the damping effect of resistive wall modes (RWM) in rotating plasmas. Benchmark of MARS-F calculations with RWM experiments on JET and D3D indicates that the semi-kinetic damping model is a good candidate for explaining the damping mechanisms. Based on these results, the critical rotation speeds required for RWM stabilization in an advanced ITER scenario are predicted. Active feedback control of the n = 1 RWM in ITER is also studied using the MARS-F code. (author)

  1. iterClust: a statistical framework for iterative clustering analysis.

    Science.gov (United States)

    Ding, Hongxu; Wang, Wanxin; Califano, Andrea

    2018-03-22

    In a scenario where populations A, B1 and B2 (subpopulations of B) exist, pronounced differences between A and B may mask subtle differences between B1 and B2. Here we present iterClust, an iterative clustering framework, which can separate more pronounced differences (e.g. A and B) in starting iterations, followed by relatively subtle differences (e.g. B1 and B2), providing a comprehensive clustering trajectory. iterClust is implemented as a Bioconductor R package. andrea.califano@columbia.edu, hd2326@columbia.edu. Supplementary information is available at Bioinformatics online.

  2. Extending ITER materials design to welded joints

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F. [DMN/Dir, CEA/Saclay, Commissariat a l' Energie Atomique, 91191 Gif sur Yvette cedex (France)]. E-mail: tavassoli@cea.fr

    2007-08-01

    This paper extends the ITER materials properties documentation to weld metals and incorporates the needs of Test Blanket Modules for higher temperature materials properties. Since the main structural material selected for ITER is type 316L(N)-IG, the paper is focused on weld metals and joining techniques for this steel. Materials properties data are analysed according to the French design and construction rules for nuclear components (RCC-MR) and design allowables are equally derived using the same rules. Particular attention is paid to the type of weld metal, to the type and position of welding and their influence on the materials properties data and design allowables. The primary goal of this work, starting with 19-12-2 weld metal, is to produce comprehensive materials properties documentations that when combined with codification and inspection documents would satisfy ITER licensing needs. As a result, structural stability and capability of welded joints during manufacturing of ITER components and their subsequent service, including the effects of irradiation and eventual incidental or accidental situations, are also covered.

  3. Cogon grass (Imperata cylindrica), a potential biomass candidate for bioethanol: cell wall structural changes enhancing hydrolysis in a mild alkali pretreatment regime.

    Science.gov (United States)

    Haque, Md Azizul; Barman, Dhirendra Nath; Kim, Min Keun; Yun, Han Dae; Cho, Kye Man

    2016-03-30

    Imperata cylindrica is being considered as a biomass candidate for bioethanol. This work aimed to evaluate a mild alkali pretreatment effect on the Imperata recalcitrant structure. Therefore, varied concentrations of NaOH (0, 7.5, 15, 20, and 25 g L(-1) ) were applied as treatments to Imperata at 105 °C for 10 min. Scanning electron microscopy, atomic force microscopy, and Fourier transform infrared spectroscopy studies revealed that 20 to 25 g L(-1) NaOH-treated Imperata exposed amorphous cellulose on surface granules composed of lignin, waxes, and partly hemicelluloses were abolished due to the comprehensive disruption of the linkages between lignin and carbohydrates. The cellulose crystalline index was increased with 7.5 to 20 g L(-1) NaOH treatments and reduced with a 25 g L(-1) NaOH treatment. In fact, the cellulose content in solids increased with the increasing NaOH concentration and was estimated to be 720 and 740 g kg(-1) for the 20 and 25 g L(-1) NaOH treatments, respectively. The yield of the reducing sugar was obtained 805 and 813 mg g(-1) from 20 and 25 g L(-1) NaOH-treated Imperata, respectively. Considering the cost of pretreatment, the 20 g L(-1) NaOH treatment is judged to be effective for disrupting Imperata recalcitrance in this pretreatment regime. © 2015 Society of Chemical Industry.

  4. ITER assembly and maintenance

    International Nuclear Information System (INIS)

    Honda, T.; Davis, F.; Lousteau, D.

    1991-01-01

    This document is intended to describe the work conducted by the ITER Assembly and Maintenance (A and M) Design Unit and the supporting home teams during the ITER Conceptual Design Activities, carried out from 1988 through 1990. Its content consists of two main sections, i.e., Chapter III, which describes the identified tasks to be performed by the A and M system and a general description of the required equipment; and Chapter IV, which provides a more detailed description of the equipment proposed to perform the assigned tasks. A two-stage R and D program is now planned, i.e., (1) a prototype equipment functional tests using full scale mock-ups and (2) a full scale integration demonstration test facility with real components (vacuum vessel with ports, blanket modules, divertor modules, armor tiles, etc.). Crucial in-vessel and ex-vessel operations and the associated remote handling equipment, including handling of divertor plates and blanket modules will be demonstrated in the first phase, whereby the database needed to proceed with the engineering phase will be acquired. The second phase will demonstrate the ability of the overall system to execute the required maintenance procedures and evaluate the performance of the prototype equipment

  5. Rokkasho: Japanese site for ITER

    International Nuclear Information System (INIS)

    Ohtake, S.; Yamaguchi, V.; Matsuda, S.; Kishimoto, H.

    2003-01-01

    The Atomic Energy Commission of Japan authorized ITER as the core machine of the Third Phase Basic Program of Fusion Energy Development. After a series of discussions in the Atomic Energy Commission and the Council of Science and Technology Policy, Japanese Government concluded formally with the Cabinet Agreement on 31 May 2002 that Japan should participate in the ITER Project and offer the Rokkasho-Mura site for construction of ITER to the Negotiations among Canada (CA), the European Union (EU), Japan (JA), and the Russian Federation (RF). The JA site proposal is now under the international assessment in the framework of the ITER Negotiations. (author)

  6. IAEA activities related to ITER

    International Nuclear Information System (INIS)

    Dolan, T.J.; Schneider, U.

    2001-01-01

    As agreed between the IAEA and the ITER Parties, special sessions are dedicated to ITER at the IAEA Fusion Energy Conferences. At the 18th IAEA Fusion Energy Conference, held on 4-10 October 2000 in Sorrento, Italy, in the Artsimovich-Kadomtsev Memorial opening session there were special lectures by Carlo Rubbia (President, ENEA, Italy), A. Arima (Japan), and E.P. Velikhov (Russia); an overview talk on ITER by R. Aymar (ITER Director); and a talk on the FTU experiment by F. Romanelli. In total, 573 participants from 34 countries presented 389 papers (including 11 post-deadline papers and the 4 summaries)

  7. ITER CTA newsletter. No. 13, October 2002

    International Nuclear Information System (INIS)

    2002-11-01

    This ITER CTA newsletter issue comprises concise information about an ITER related meeting concerning the joint implementation of ITER - the fifth ITER Negotiations Meeting - which was held in Toronto, Canada, 19-20 September, 2002, and information about assessment of the possible ITER site in Clarington, Ontario, Canada, which was the subject of the first official stage of the Joint Assessment of Specific Sites (JASS) for the ITER Project. This assessment was completed just before the Fifth ITER Negotiations Meeting

  8. 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

  9. ITER waste management

    International Nuclear Information System (INIS)

    Rosanvallon, S.; Na, B.C.; Benchikhoune, M.; Uzan, J. Elbez; Gastaldi, O.; Taylor, N.; Rodriguez, L.

    2010-01-01

    ITER will produce solid radioactive waste during its operation (arising from the replacement of components and from process and housekeeping waste) and during decommissioning (de-activation phase and dismantling). The waste will be activated by neutrons of energies up to 14 MeV and potentially contaminated by activated corrosion products, activated dust and tritium. This paper describes the waste origin, the waste classification as a function of the French national agency for radioactive waste management (ANDRA), the optimization process put in place to reduce the waste radiotoxicity and volumes, the estimated waste amount based on the current design and maintenance procedure, and the overall strategy from component removal to final disposal anticipated at this stage of the project.

  10. Iterated crowdsourcing dilemma game

    Science.gov (United States)

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-02-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space.

  11. ITER cooling systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

    1992-06-01

    The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

  12. Divertor development for ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

    1998-01-01

    The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

  13. Designing a prototype of the ITER pulse scheduling system

    International Nuclear Information System (INIS)

    Yamamoto, T.; Yonekawa, I.; Ohta, K.; Hosoyama, H.; Hashimoto, Y.; Wallander, A.; Winter, A.; Sugie, T.; Kusama, Y.; Kawano, Y.; Yoshino, R.

    2012-01-01

    Highlights: ► We designed a prototype of the ITER pulse scheduling system. ► Structure of ITER pulse schedules was designed. ► Validation and automatic value assignment functions were adopted. ► A prototype will be implemented in 2011. - Abstract: A prototype of the ITER pulse scheduling system that prepares and manages parameters for ITER plasma operations has been designed. Based on the analyzed requirements on the system, structure of the parameters and necessary functions were determined. Segment and module structures were tuned to the ITER requirements. Three types of validations assure sanity of the parameters. The design limits check and the operation window check verify whether the values of the parameters do not exceed the limits. The consistency check calculates dependency among parameters in accordance with logics described in a scripting language. The ITER pulse scheduling system provides interface with a physics model and simulator. Some abstract physics parameters are converted to engineering parameters with the physics simulation. The results of simulation such as plasma characteristics of specified parameters are also shown to the researchers. The tool to specify the parameters is data-driven. Therefore, it is flexible for changes of number of the parameters. A prototype is being implemented in 2011. Using the prototype, this design will be verified and refined. The evaluation of the prototype will be a basis of the final production of the ITER pulse scheduling system.

  14. A functional approach for managing ITER operations

    International Nuclear Information System (INIS)

    Houtte, Didier van; Sagot, François; Okayama, Katsumi; Blackler, Kenneth

    2012-01-01

    Highlights: ► A function-oriented approach for defining and organizing all the functions required to perform the mission has been developed. ► A Functional Breakdown Structure providing a complete hierarchy of functions on multiple levels is presented. ► The FBS is used for giving a good visibility of ITER project needs and requirements. ► Reliability (R) and Inherent Availability (A I ) of basic functions are calculated from data on the structures, systems and components (failure rate and time to repair) for obtaining the Availability objectives of the ITER project. - Abstract: ITER is currently the most ambitious project on nuclear fusion research. Its objective is to demonstrate the feasibility of fusion as an energy source for the future. The complexity of the systems required to meet this challenge present many opportunities for omissions or incorrect assumptions. System engineering allows the engineer to deal with such a complexity by developing a Functional Breakdown Structure (FBS). Unlike a Plant Breakdown Structure (PBS), the FBS is a function-oriented tree, not a product-oriented tree. It details operations or activities that have to be performed as needed functions of the architecture, allowing identification of any missing elements, defining the personnel skills required to operate the architecture and managing the machine availability.

  15. 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

  16. ITER-FEAT - outline design report. Report by the ITER Director. ITER meeting, Tokyo, January 2000

    International Nuclear Information System (INIS)

    2001-01-01

    It is now possible to define the key elements of ITER-FEAT. This report provides the results, to date, of the joint work of the Special Working Group in the form of an Outline Design Report on the ITER-FEAT design which, subject to the views of ITER Council and of the Parties, will be the focus of further detailed design work and analysis in order to provide to the Parties a complete and fully integrated engineering design within the framework of the ITER EDA extension

  17. Development of sampling techniques for ITER Type B radwaste

    International Nuclear Information System (INIS)

    Hong, Kwon Pyo; Kim, Sung Geun; Jung, Sang Hee; Oh, Wan Ho; Park, Myung Chul; Kim, Hee Moon; Ahn, Sang Bok

    2016-01-01

    There are several difficulties and limitation in sampling activities. As the Type B radwaste components are mostly metallic(mostly stainless steel) and bulk(∼ 1 m in size and ∼ 100 mm in thickness), it is difficult in taking samples from the surface of Type B radwaste by remote operation. But also, sampling should be performed without use of any liquid coolant to avoid the spread of contamination. And all sampling procedures are carried in the hot cell red zone with remote operation. Three kinds of sampling techniques are being developed. They are core sampling, chip sampling, and wedge sampling, which are the candidates of sampling techniques to be applied to ITER hot cell. Object materials for sampling are stainless steel or Cu alloy block in order to simulate ITER Type B radwaste. The best sampling technique for ITER Type B radwaste among the three sampling techniques will be suggested in several months after finishing the related experiment

  18. Development of sampling techniques for ITER Type B radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Kim, Sung Geun; Jung, Sang Hee; Oh, Wan Ho; Park, Myung Chul; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are several difficulties and limitation in sampling activities. As the Type B radwaste components are mostly metallic(mostly stainless steel) and bulk(∼ 1 m in size and ∼ 100 mm in thickness), it is difficult in taking samples from the surface of Type B radwaste by remote operation. But also, sampling should be performed without use of any liquid coolant to avoid the spread of contamination. And all sampling procedures are carried in the hot cell red zone with remote operation. Three kinds of sampling techniques are being developed. They are core sampling, chip sampling, and wedge sampling, which are the candidates of sampling techniques to be applied to ITER hot cell. Object materials for sampling are stainless steel or Cu alloy block in order to simulate ITER Type B radwaste. The best sampling technique for ITER Type B radwaste among the three sampling techniques will be suggested in several months after finishing the related experiment.

  19. 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.

  20. 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.)

  1. 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

  2. ITER CTA newsletter. No. 8

    International Nuclear Information System (INIS)

    2002-05-01

    This ITER CTA newsletter contains information about the Third Negotiations Meeting on the Joint Implementation of ITER held in Moscow on 23-24 April 2002 and about the visit of Canadian officials and members of the Canadian delegation to RF research center 'Kurchatov Institute'

  3. ITER physics design guidelines: 1989

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1990-01-01

    The physics basis for ITER has been developed from an assessment of the results of the last twenty-five years of tokamak research and from detailed analysis of important physics issues specifically for the ITER design. This assessment has been carried out with direct participation of members of the experimental teams of each of the major tokamaks in the world fusion program through participation in ITER workshops, contributions to the ITER Physics R and D Program, and by direct contacts between the ITER team and the cognizant experimentalists. Extrapolations to the present data base, where needed, are made in the most cautious way consistent with engineering constraints and performance goals of the ITER. In cases where a working assumptions had to be introduced, which is insufficiently supported by the present data base, is explicitly stated. While a strong emphasis has been placed on the physics credibility of the design, the guidelines also take into account that ITER should be designed to be able to take advantage of potential improvements in tokamak physics that may occur before and during the operation of ITER. (author). 33 refs

  4. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held in Vienna on 16 July 2001. It was the last MAC Meeting and the main topics were consideration of the report by the Director on the ITER EDA status, review of the Work Programme, review of the Joint Fund and arrangements for termination and wind-up of the EDA

  5. ITER CTA newsletter. No. 7

    International Nuclear Information System (INIS)

    2002-04-01

    This issue of ITER CTA newsletter contains information about the meeting of the ITER CTA project board, which took place in Moscow, Russian Federation on 22 April 2002 on the occasion of the Third Negotiators Meeting (N3), and about the meeting 'EU divertor celebration day' organized on 16 January 2002 at Plansee AG, Reutte, Austria

  6. ITER EDA Newsletter. V. 3, no. 8

    International Nuclear Information System (INIS)

    1994-08-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on the sixth ITER council meeting; introduces the newly appointed ITER director and reports on his address to the ITER council. The vacuum tank for the ITER model coil testing, installed at JAERI, Naka, Japan is also briefly described

  7. ITER ITA newsletter. No. 6, July 2003

    International Nuclear Information System (INIS)

    2003-09-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities. One of them was the farewell party for for Annick Lyraud and Robert Aymar, who will take up his position as Director-General of CERN in January 2004, another is information about Dr. Yasuo Shimomura, ITER interim project leader, and ITER technical work during the transitional arrangements

  8. ITER ITA newsletter. No. 8, September 2003

    International Nuclear Information System (INIS)

    2003-10-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including Robert Aymar's leaving ITER for CERN, ITER related issues at the IAEA General Conference and status and prospects of thermonuclear power and activity during the ITA on materials foe vessel and in-vessel components

  9. ITER interim design report package documents

    International Nuclear Information System (INIS)

    1996-01-01

    This publication contains the Excerpt from the ITER Council (IC-8), the ITER Interim Design Report, Cost Review and Safety Analysis, ITER Site Requirements and ITER Site Design Assumptions and the Excerpt from the ITER Council (IC-9). 8 figs, 2 tabs

  10. Plasma control concepts for ITER

    International Nuclear Information System (INIS)

    Lister, J.B.; Nieswand, C.

    1997-01-01

    This overview paper skims over a wide range of issues related to the control of ITER plasmas. Although operation of the ITER project will require extensive developmental work to achieve the degree of control required, there is no indication that any of the identified problems will present overwhelming difficulties compared with the operation of present tokamaks. However, the precision of control required and the degree of automation of the final ITER plasma control system will present a challenge which is somewhat greater than for present tokamaks. In order to operate ITER optimally, integrated use of a large amount of diagnostic information will be necessary, evaluated and interpreted automatically. This will challenge both the diagnostics themselves and their supporting interpretation codes. The intervening years will provide us with the opportunity to implement and evaluate most of the new features required for ITER on existing tokamaks, with the exception of the control of an ignited plasma. (author) 7 figs., 7 refs

  11. ITER technical advisory committee meeting

    International Nuclear Information System (INIS)

    Fujiwara, M.

    2001-01-01

    The 17th Meeting of the ITER Technical Advisory Committee (TAC-17) was held on February 19-22, the ITER Garching Work Site in Germany. The objective of the meeting was to review the Draft Final Design Report of ITER-FEAT and assess the ability of the self-consistent overall design both to satisfy the technical objectives previously defined and to meet the cost limitations. TAC-17 was also organized to confirm that the design and critical elements, with emphasis on the key recommendations made at previous TAC meetings, are such as to extend the confidence in starting ITER construction. It was also intended to provide the ITER Council, scheduled to meet on 27 and 28 February in Toronto, with a technical assessment and key recommendations of the above mentioned report

  12. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    International Nuclear Information System (INIS)

    Hong, Bong Guen; In, S. R.; Bae, Y. D.

    2006-02-01

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  13. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; In, S. R.; Bae, Y. D. (and others)

    2006-02-15

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  14. Irreducible complexity of iterated symmetric bimodal maps

    Directory of Open Access Journals (Sweden)

    J. P. Lampreia

    2005-01-01

    Full Text Available We introduce a tree structure for the iterates of symmetric bimodal maps and identify a subset which we prove to be isomorphic to the family of unimodal maps. This subset is used as a second factor for a ∗-product that we define in the space of bimodal kneading sequences. Finally, we give some properties for this product and study the ∗-product induced on the associated Markov shifts.

  15. 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

  16. Industry participation in the ITER engineering designing

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2006-01-01

    Involvement of the European industry promoted elaboration of the ITER engineering design. The EFDA is responsible for coordination of the industry involvement under the signed contracts the total amount of which is about 70 MEURO. Diversified remote handling equipment is available to replace internal structures and to transfer them to and back from hot cell. The contribution of the European industry consists mainly of divertor equipment, of air cushion transfer system and transfer casks [ru

  17. Structural determination of individual chemical species in a mixed system by iterative transformation factor analysis-based X-ray absorption spectroscopy combined with UV-visible absorption and quantum chemical calculation.

    Science.gov (United States)

    Ikeda, Atsushi; Hennig, Christoph; Rossberg, André; Tsushima, Satoru; Scheinost, Andreas C; Bernhard, Gert

    2008-02-15

    A multitechnique approach using extended X-ray absorption fine structure (EXAFS) spectroscopy based on iterative transformation factor analysis (ITFA), UV-visible absorption spectroscopy, and density functional theory (DFT) calculations has been performed in order to investigate the speciation of uranium(VI) nitrate species in acetonitrile and to identify the complex structure of individual species in the system. UV-visible spectral titration suggests that there are four different species in the system, that is, pure solvated species, mono-, di-, and trinitrate species. The pure EXAFS spectra of these individual species are extracted by ITFA from the measured spectral mixtures on the basis of the speciation distribution profile calculated from the UV-visible data. Data analysis of the extracted EXAFS spectra, with the help of DFT calculations, reveals the most probable complex structures of the individual species. The pure solvated species corresponds to a uranyl hydrate complex with an equatorial coordination number (CNeq) of 5, [UO2(H2O)5]2+. Nitrate ions tend to coordinate to the uranyl(VI) ion in a bidentate fashion rather than a unidentate one in acetonitrile for all the nitrate species. The mononitrate species forms the complex of [UO2(H2O)3NO3]+ with a CNeq value of 5, while the di- and trinitrate species have a CNeq value of 6, corresponding to [UO2(H2O)2(NO3)2]0 (D2h) and [UO2(NO3)3]- (D3h), respectively.

  18. Second meeting of the ITER Preparatory Committee

    International Nuclear Information System (INIS)

    Drew, M.

    2003-01-01

    The committee charged to oversee the ITER ITA (ITER transitional arrangements) the ITER preparatory committee, held its second meeting on 24 September at the JET facilities at Culham, UK. Dr. Umberto Finzi of the European Commission was chairman. This meeting was also the first since the succession by Dr. Yasuo Shimomura to Dr. Robert Aymar as Interim Project Leader (IPL). Welcoming Dr. Shimomura in his new capacity, the Committee paid tribute to the outstanding contributions of his predecessor to the definition, design and promotion of ITER, and expressed the gratitude of all Participants to Dr. Aymar and its best wishes for future success in his new appointment.The technical activities of the ITA were the main focus of the Committee's discussions. The Committee took note of the IPL's Status Report on ITA Technical Activities and endorsed the IPL's proposals for the top level structure of the International Team, including the designation of Dr. Pietro Barabaschi as Deputy to the IPL. The Committee took note of the IPL's proposals on Participants' contributions to the ITA and of the Participants' stated intentions and expectations in this regard. Several Delegations pointed out that access to necessary resources would depend strongly on progress made towards the Agreement. All Participants were invited, in the shared interests of the project, to respond constructively to the specific technical areas where the IPL reported a lack of resources. Following a presentation from the IT on Project Management Tools, the Committee expressed support, in general, for the proposed strategy designed to provide the current team with the CAD and Data Management elements necessary to prepare for an efficient start of ITER construction, and asked the IT Leader to report on an estimate and time profile of expenditure during the period to mid-2004. The Committee supported the proposals to re-establish the ITER Test Blanket Working. The Committee agreed that the phasing of planned

  19. Achievements in the development of the Water Cooled Solid Breeder Test Blanket Module of Japan to the milestones for installation in ITER

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Tanigawa, Hisashi; Hirose, Takanori; Mohri, Kensuke; Seki, Yohji; Enoeda, Mikio; Ezato, Koichiro; Suzuki, Satoshi; Nishi, Hiroshi; Akiba, Masato

    2009-01-01

    As the primary candidate of ITER Test Blanket Module (TBM) to be tested under the leadership of Japan, a water cooled solid breeder (WCSB) TBM is being developed. This paper shows the recent achievements towards the milestones of ITER TBMs prior to the installation, which consist of design integration in ITER, module qualification and safety assessment. With respect to the design integration, targeting the detailed design final report in 2012, structure designs of the WCSB TBM and the interfacing components (common frame and backside shielding) that are placed in a test port of ITER and the layout of the cooling system are presented. As for the module qualification, a real-scale first wall mock-up fabricated by using the hot isostatic pressing method by structural material of reduced activation martensitic ferritic steel, F82H, and flow and irradiation test of the mock-up are presented. As for safety milestones, the contents of the preliminary safety report in 2008 consisting of source term identification, failure mode and effect analysis (FMEA) and identification of postulated initiating events (PIEs) and safety analyses are presented.

  20. ITER management advisory committee meeting in NAKA

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    1999-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 17 December 1999 in Naka, Japan. The main topics were the ITER EDA Status, Task Status Summary and Work Program and a schedule of ITER meetings

  1. ITER EDA newsletter. V. 7, no. 6

    International Nuclear Information System (INIS)

    1998-06-01

    This newsletter contains the articles: 'ITER representation at the 11th Pacific Basin Nuclear Conference', 'Summary of discussion points and further deliberations in the special committee on the ITER project in the Atomic Energy Commission', and 'ITER radio frequency systems'

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

    International Nuclear Information System (INIS)

    2000-02-01

    This ITER EDA Newsletter reports on the seventh ITER technical meeting on safety and environment and contains the executive summary of the eleventh ITER scrape-off layer and divertor physics expert group meeting. Individual abstracts have been prepared

  3. SPARSE ELECTROMAGNETIC IMAGING USING NONLINEAR LANDWEBER ITERATIONS

    KAUST Repository

    Desmal, Abdulla; Bagci, Hakan

    2015-01-01

    minimization problem is solved using nonlinear Landweber iterations, where at each iteration a thresholding function is applied to enforce the sparseness-promoting L0/L1-norm constraint. The thresholded nonlinear Landweber iterations are applied to several two

  4. ITER cooling system

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-11-01

    The present specification of the ITER cooling system does not permit its operation with water above 150 C. However, the first wall needs to be heated to higher temperatures during conditioning at 250 C and bake-out at 350 C. In order to use the cooling water for these operations the cooling system would have to operate during conditioning at 37 Bar and during bake-out at 164 Bar. This is undesirable from the safety analysis point of view, and alternative heating methods are to be found. This review suggests that superheated steam or gas heating can be used for both baking and conditioning. The blanket design must consider the use of dual heat transfer media, allowing for change from one to another in both directions. Transfer from water to gas or steam is the most intricate and risky part of the entire heating process. Superheated steam conditioning appears unfavorable. The use of inert gas is recommended, although alternative heating fluids such as organic coolant should be investigated

  5. 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

  6. 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

  7. 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

  8. 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.

  9. Nuclear energy and sustainability: Understanding ITER

    International Nuclear Information System (INIS)

    Fiore, Karine

    2006-01-01

    Deregulation and new environmental requirements combined with the growing scarcity of fossil resources and the increasing world energy demand lead to a renewal of the debate on tomorrow's energies. Specifically, nuclear energy, which has undeniable assets, faces new constraints. On the one hand, nuclear energy is very competitive and harmless to greenhouse effect. From this point, it seems to be an ideal candidate to reach future objectives of sustainability, availability and acceptability. On the other hand, its technology of production - based on fission - remains imperfect and generates risks for environment and health. In this respect, it is less desirable. Therefore, world researchers turn today towards another type of nuclear technique, fusion, on which the project ITER is founded. This worldwide project is interesting for our analysis because, as a technological revolution, it takes into consideration all the global challenges of nuclear energy for the future, and particularly its capacity to meet the increasing energy needs of developing countries. It is the example par excellence of a successful international scientific collaboration oriented towards very long-run energy ends that involve huge technological, economic and political stakes. Focusing on this project, we thus have to reconsider the future place of nuclear energy in a more and more demanding world. Considering the magnitude of the efforts undertaken to implement ITER, this paper aims at analysing, in a detailed way, its goals, its challenges and its matter

  10. ITER EDA Newsletter. V. 10, no. 7

    International Nuclear Information System (INIS)

    2001-07-01

    This ITER EDA Newsletter presents an overview of meetings held at IAEA Headquarters in Vienna during the week 16-20 July 2001 related to the successful completion of the ITER Engineering Design Activities (EDA). Among them were the final meeting of the ITER Council, the closing ceremony to commemorate the EDA completion, the final meeting of the ITER Management Advisory Committee, a briefing of issues related to ITER developments, and discussions on the possible joint implementation of ITER

  11. 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

  12. 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

  13. ITER technical advisory committee meeting

    International Nuclear Information System (INIS)

    Fujiwara, M.

    1999-01-01

    The ITER Technical Advisory Committee (TAC) meeting took place on December 20-22, 1999 at the Naka Joint Work Site. The objective of this meeting was to review the document 'Technical Basis for ITER-FEAT Outline Design (ODR)' issued by the Director on December 10. It was also aimed at providing the ITER Meeting scheduled for January 19-20, 2000 in Tokyo with a technical assessment of ODR and recommendations for the optimization of the anticipated plasma performance and engineering design, based on the guidelines approved by the Council in June 1998 and recommendations of the last TAC meeting

  14. Mechanical design features and challenges for the ITER ICRH antenna

    Energy Technology Data Exchange (ETDEWEB)

    Borthwick, A. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)], E-mail: andy.borthwick@yahoo.co.uk; Agarici, G. [Fusion for Energy, Barcelona (Spain); Davis, A. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Dumortier, P.; Durodie, F. [LPP-ERM-KMS, Association EURATOM-Belgian State, Brussels (Belgium); Fanthome, J.; Hamlyn-Harris, C.; Hancock, A.D.; Lockley, D. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Mitteau, R. [Euratom-CEA Association, DSM/IRFM, CEA-Cadarache, 13108 St Paul lez Durance (France); Nightingale, M. [UKAEA/Euratom Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sartori, R. [Fusion for Energy, Barcelona (Spain); Vulliez, K. [Euratom-CEA Association, DSM/IRFM, CEA-Cadarache, 13108 St Paul lez Durance (France)

    2009-06-15

    The ITER Ion Cyclotron Resonant Heating (ICRH) antenna provides plasma heating at a power of 20 MW. Operation in the ITER environment imposes significant thermal power handling capability, structural integrity, shielding and operations requirements. The design will require a step change over any predecessor in terms of power, scale and complexity. This paper reports the main mechanical design features that address the challenges and often conflicting requirements during the conceptual design phase.

  15. Fabrication of divertor cassette for ITER

    International Nuclear Information System (INIS)

    Sanguinetti, G.P.

    2008-01-01

    The Divertor is the component located on the bottom of the ITER vacuum vessel, whose main function is to adsorb the high thermal flux generated by the plasma whilst keeping the plasma impurity at a reasonable low level. The divertor consist of 54 units, each comprising outer components, facing the plasma and a component supporting the plasma facing components (PFC) and providing coolant distribution to them (divertor cassette). The divertor cassette is a box structure, butt welded and machined, made from plates and forgins of austenitic stainless steels. The cassette fabrication, which is in detail described, includes manufacturing of the attachments of the PFC to the cassette, the coolant distribution channels, and the cassette to vacuum vessel locking system. The divertor cassette is a pressure component (the cooling water runs at 40 bar) and therefore divertor cassette design, fabrication and service shall comply with the European PED and the applicable French law for the ITER. (orig.)

  16. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  17. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  18. TIG of Reduced Activation Ferrite/Martensitic Steel for the Korean ITER-TBM

    International Nuclear Information System (INIS)

    Ku, Duck Young; Ahn, Mu Young; Yu, In Keun; Cho, Seun Gyon; Oh, Seung Jin

    2010-01-01

    Test Blanket Modules (TBM) will be tested in ITER to verify the capability of tritium breeding and recovery and the extraction of thermal energy suitable for the production of electricity. A Helium Cooled Solid Breeder (HCSB) TBM has been developed in Korea to accomplish these goals. Reduced Activation Ferritic/Martensitic (RAFM) steel has been chosen as the primary candidate structural material for Korean TBM. Due to the complexity of the First wall (FW) and Side wall (SW), it is necessary to develop various joining technologies, such as Hot Isostatic Pressing (HIP), Electron Beam Welding (EBW) and Tungsten Inert Gas (TIG) welding, for the successful fabrication of TBM. In this study, the mechanical properties of TIG welded RAFM steel were investigated. Various mechanical tests of TIG-welded RAFM steel were performed to obtain the optimized TIG welding process for RAFM steel

  19. TIG of Reduced Activation Ferrite/Martensitic Steel for the Korean ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Ku, Duck Young; Ahn, Mu Young; Yu, In Keun; Cho, Seun Gyon [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Oh, Seung Jin [KHNP, Daejeon (Korea, Republic of)

    2010-10-15

    Test Blanket Modules (TBM) will be tested in ITER to verify the capability of tritium breeding and recovery and the extraction of thermal energy suitable for the production of electricity. A Helium Cooled Solid Breeder (HCSB) TBM has been developed in Korea to accomplish these goals. Reduced Activation Ferritic/Martensitic (RAFM) steel has been chosen as the primary candidate structural material for Korean TBM. Due to the complexity of the First wall (FW) and Side wall (SW), it is necessary to develop various joining technologies, such as Hot Isostatic Pressing (HIP), Electron Beam Welding (EBW) and Tungsten Inert Gas (TIG) welding, for the successful fabrication of TBM. In this study, the mechanical properties of TIG welded RAFM steel were investigated. Various mechanical tests of TIG-welded RAFM steel were performed to obtain the optimized TIG welding process for RAFM steel

  20. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    International Nuclear Information System (INIS)

    Tanigawa, H.; Hirose, T.; Shiba, K.; Kasada, R.; Wakai, E.; Serizawa, H.; Kawahito, Y.; Jitsukawa, S.; Kimura, A.; Kohno, Y.; Kohyama, A.; Katayama, S.; Mori, H.; Nishimoto, K.; Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J.

    2008-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed

  1. Iterants, Fermions and Majorana Operators

    Science.gov (United States)

    Kauffman, Louis H.

    Beginning with an elementary, oscillatory discrete dynamical system associated with the square root of minus one, we study both the foundations of mathematics and physics. Position and momentum do not commute in our discrete physics. Their commutator is related to the diffusion constant for a Brownian process and to the Heisenberg commutator in quantum mechanics. We take John Wheeler's idea of It from Bit as an essential clue and we rework the structure of that bit to a logical particle that is its own anti-particle, a logical Marjorana particle. This is our key example of the amphibian nature of mathematics and the external world. We show how the dynamical system for the square root of minus one is essentially the dynamics of a distinction whose self-reference leads to both the fusion algebra and the operator algebra for the Majorana Fermion. In the course of this, we develop an iterant algebra that supports all of matrix algebra and we end the essay with a discussion of the Dirac equation based on these principles.

  2. Updated safety analysis of ITER

    International Nuclear Information System (INIS)

    Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

    2011-01-01

    An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

  3. Rollout sampling approximate policy iteration

    NARCIS (Netherlands)

    Dimitrakakis, C.; Lagoudakis, M.G.

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions, which focus on policy representation using classifiers and address policy learning as a

  4. Updated safety analysis of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

  5. 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

  6. Cooperation between CERN and ITER

    CERN Document Server

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. Kaname Ikeda, the Director-General of the International Fusion Energy Organisation (ITER) (on the right) and Robert Aymar, Director-General of CERN, signing the agreement.The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March. One of the main purposes of this agreement is for CERN to give ITER the benefit of its experience in the field of technology as well as in administrative domains such as finance, procurement, human resources and informatics through the provision of consultancy services. Currently in its start-up phase at its Cadarache site, 70 km from Marseilles (France), ITER will focus its research on the scientific and technical feasibility of using fusion energy as a fu...

  7. 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)

  8. 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

  9. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

    Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

    2013-01-01

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

  10. ITER leader to head CERN

    CERN Document Server

    Feder, Toni

    2003-01-01

    After successfully chairing an external review committee for CERN last year, Robert Aymar will leave ITER to become director general of the European particle physics laboratory rom 2004. Before ITER he also successfully managed the startup or Tore Supra. He will attempt to ensure that the LHC begins operating in 2007 - two years late - and is paid for by 2010 and will also start the planning for life after the LHC (1 page)

  11. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  12. 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

  13. 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.)

  14. ITER primary cryopump test facility

    International Nuclear Information System (INIS)

    Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

    1998-01-01

    A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

  15. The ITER remote maintenance system

    International Nuclear Information System (INIS)

    Tesini, A.; Palmer, J.

    2007-01-01

    ITER is a joint international research and development project that aims to demonstrate the scientific and technological feasibility of fusion power. As soon as the plasma operation begins using tritium, the replacement of the vacuum vessel internal components will need to be done with remote handling techniques. To accomplish these operations ITER has equipped itself with a Remote Maintenance System; this includes the Remote Handling equipment set and the Hot Cell facility. Both need to work in a cooperative way, with the aim of minimizing the machine shutdown periods and to maximize the machine availability. The ITER Remote Handling equipment set is required to be available, robust, reliable and retrievable. The machine components, to be remotely handle-able, are required to be designed simply so as to ease their maintenance. The baseline ITER Remote Handling equipment is described. The ITER Hot Cell Facility is required to provide a controlled and shielded area for the execution of repair operations (carried out using dedicated remote handling equipment) on those activated components which need to be returned to service, inside the vacuum vessel. The Hot Cell provides also the equipment and space for the processing and temporary storage of the operational and decommissioning radwaste. A conceptual ITER Hot Cell Facility is described. (orig.)

  16. Overview of magnetic control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Zabeo, L., E-mail: luca.zabeo@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Ambrosino, G., E-mail: ambrosin@unina.it [CREATE/Universitá di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’informazione, Naples (Italy); Cavinato, M., E-mail: mario.cavinato@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Torres Diagonal Litoral - B3, 08019 Barcelona (Spain); Gribov, Y., E-mail: yuri.gribov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Kavin, A., E-mail: kavina@sintez.niiefa.spb.su [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Lukash, V., E-mail: lukash@nfi.kiae.ru [Kurchatov Institute, Moscow (Russian Federation); Mattei, M., E-mail: massimiliano.mattei@unina2.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Pironti, A., E-mail: pironti@unina.it [CREATE/Seconda Universitá di Napoli, Dip. Ingegneria Industriale e dell’informazione, Naples (Italy); Snipes, J.A., E-mail: joseph.snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Vayakis, G., E-mail: george.vayakis@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France); Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul Lez Durance (France)

    2014-05-15

    ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination. Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios. Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design. Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the

  17. Modeling of ELM Dynamics in ITER

    International Nuclear Information System (INIS)

    Pankin, A.Y.; Bateman, G.; Kritz, A.H.; Brennan, D.P.; Snyder, P.B.; Kruger, S.

    2007-01-01

    Edge localized modes (ELMs) are large scale instabilities that alter the H-mode pedestal, reduce the total plasma stored energy, and can result in heat pulses to the divertor plates. These modes can be triggered by pressure driven ballooning modes or by current driven peeling instabilities. In this study, stability analyses are carried out for a series of ITER equilibria that are generated with the TEQ and TOQ equilibrium codes. The H-mode pedestal pressure and parallel component of plasma current density are varied in a systematic way in order to include the relevant parameter space for a specific ITER discharge. Ideal MHD stability codes, DCON, ELITE, and BALOO code, are employed to determine whether or not each ITER equilibrium profile is unstable to peeling or ballooning modes in the pedestal region. Several equilibria that are close to the marginal stability boundary for peeling and ballooning modes are tested with the NIMROD non-ideal MHD code. The effects of finite resistivity are studied in a series of linear NIMROD computations. It is found that the peeling-ballooning stability threshold is very sensitive to the resistivity and viscosity profiles, which vary dramatically over a wide range near the separatrix. Due to the effects of finite resistivity and viscosity, the peeling-ballooning stability threshold is shifted compared to the ideal threshold. A fundamental question in the integrated modeling of ELMy H-mode discharges concerning how much plasma and current density is removed during each ELM crash can be addressed with nonlinear non-ideal MHD simulations. In this study, the NIMROD computer simulations are continued into the nonlinear stage for several ITER equilibria that are marginally unstable to peeling or ballooning modes. The role of two-fluid and finite Larmor radius effects on the ELM dynamics in ITER geometry is examined. The formation of ELM filament structures, which are observed in many existing tokamak experiments, is demonstrated for ITER

  18. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Smith, John [General Atomics, San Diego, CA (United States)

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  19. Divertor armour issues: lifetime, safety and influence on ITER performance

    International Nuclear Information System (INIS)

    Pestchanyi, S.

    2009-01-01

    Comprehensive simulations of the ITER divertor armour vaporization and brittle destruction under ELMs of different sizes have revealed that the erosion rate of CFC armour is intolerable for an industrial reactor, but it can be considerably reduced by the armour fibre structure optimization. The ITER core contamination with carbon is tolerable for medium size ELMs, but large type I ELM can run the confinement into the disruption. Erosion of tungsten, an alternative armour material, under ELMs influence is satisfactory, but the danger of the core plasma contamination with tungsten is still not enough understood and potentially it could be very dangerous. Vaporization of tungsten, its cracking and dust production during ELMs are rather urgent issues to be investigated for proper choice of the divertor armour material for ITER. However, the erosion rate under action of the disruptive heat loads is tolerable for both armour materials assuming few hundred disruptions falls out during ITER lifetime

  20. 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

  1. System engineering and configuration management in ITER

    International Nuclear Information System (INIS)

    Chiocchio, S.; Martin, E.; Barabaschi, P.; Bartels, Hans Werner; How, J.; Spears, W.

    2007-01-01

    The construction of ITER will represent a major challenge for the fusion community at large, because of the intrinsic complexity of the tokamak design, the large number of different systems which are all essential for its operation, the worldwide distribution of the design activities and the unusual procurement scheme based on a combination of in-kind and directly funded deliverables. A key requirement for the success of such a large project is that a systematic approach to ensure the consistency of the design with the required performance is adopted. Also, effective project management methods, tools and working practices must be deployed to facilitate the communication and collaboration among the institutions and industries involved in the project. The authors have been involved in the definition and practical implementation of the design integration and configuration control structure inside ITER and in the system engineering process during the selection and optimization of the machine configuration. In parallel, they have assessed design, drawing and documentation management software to be used for the construction phase. Here, they describe the experience gained in recent years, explain the drivers behind the selection of the documents and drawings management systems, and illustrate the scope and issues of the configuration management activities to ensure the congruence of the design, to control and track the design changes and to manage the interfaces among the ITER systems

  2. Toward Generalization of Iterative Small Molecule Synthesis.

    Science.gov (United States)

    Lehmann, Jonathan W; Blair, Daniel J; Burke, Martin D

    2018-02-01

    Small molecules have extensive untapped potential to benefit society, but access to this potential is too often restricted by limitations inherent to the customized approach currently used to synthesize this class of chemical matter. In contrast, the "building block approach", i.e., generalized iterative assembly of interchangeable parts, has now proven to be a highly efficient and flexible way to construct things ranging all the way from skyscrapers to macromolecules to artificial intelligence algorithms. The structural redundancy found in many small molecules suggests that they possess a similar capacity for generalized building block-based construction. It is also encouraging that many customized iterative synthesis methods have been developed that improve access to specific classes of small molecules. There has also been substantial recent progress toward the iterative assembly of many different types of small molecules, including complex natural products, pharmaceuticals, biological probes, and materials, using common building blocks and coupling chemistry. Collectively, these advances suggest that a generalized building block approach for small molecule synthesis may be within reach.

  3. Toward Generalization of Iterative Small Molecule Synthesis

    Science.gov (United States)

    Lehmann, Jonathan W.; Blair, Daniel J.; Burke, Martin D.

    2018-01-01

    Small molecules have extensive untapped potential to benefit society, but access to this potential is too often restricted by limitations inherent to the customized approach currently used to synthesize this class of chemical matter. In contrast, the “building block approach”, i.e., generalized iterative assembly of interchangeable parts, has now proven to be a highly efficient and flexible way to construct things ranging all the way from skyscrapers to macromolecules to artificial intelligence algorithms. The structural redundancy found in many small molecules suggests that they possess a similar capacity for generalized building block-based construction. It is also encouraging that many customized iterative synthesis methods have been developed that improve access to specific classes of small molecules. There has also been substantial recent progress toward the iterative assembly of many different types of small molecules, including complex natural products, pharmaceuticals, biological probes, and materials, using common building blocks and coupling chemistry. Collectively, these advances suggest that a generalized building block approach for small molecule synthesis may be within reach. PMID:29696152

  4. ITER EDA Newsletter. V. 4, no. 5

    International Nuclear Information System (INIS)

    1995-05-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains comments on the ITER project by the Permanent Representative of the Russian Federation to the International Organizations in Vienna; a report on the ITER Magnet Technical Meeting held at the Joint Work Site at Naka, Japan, April 19-21, 1995; and a contribution entitled ''ITER spouses cross the cultures''

  5. ITER EDA newsletter. V. 10, no. 6

    International Nuclear Information System (INIS)

    2001-06-01

    This ITER EDA Newsletter issue includes information about the ITER Management Advisory Committee Meeting held in Vienna on 16 July 2001 and also a summary of the ninth ITER Technical Meeting on safety and environment held at the ITER Garching Joint Work site, 8 to 10 May, 2001

  6. ITER ITA newsletter. No. 27, January 2006

    International Nuclear Information System (INIS)

    2006-02-01

    This issue of ITER ITA (ITER transitional arrangements) newsletter contains concise information about two ITER related meetings including the twelfth ITER Negotiations Meeting and The Ninth Meeting of the ITPA Topical Group (TG) on Diagnostics was held at the National Fusion Research Centre (NFRC), Daejeon, Korea, from 10-14 October 2005

  7. ITER EDA newsletter. V. 8, no. 9

    International Nuclear Information System (INIS)

    1999-09-01

    This edition of the ITER EDA Newsletter contains a contribution by the ITER Director, R. Aymar, on the subject of developments in ITER Physics R and D report on the completion of the ITER central solenoid model coils installation by H. Tsuji, Head fo the Superconducting Magnet Laboratory at JAERI in Naka, Japan. Individual abstracts are prepared for each of the two articles

  8. ITER EDA Newsletter. V.3, no.3

    International Nuclear Information System (INIS)

    1994-03-01

    This ITER EDA Newsletter issue contains reports on (i) the completion of the ITER EDA Protocol 1, (ii) the signing of ITER EDA Protocol 2, (iii) a technical meeting on pumping and fuelling and (iv) a technical meeting on the ITER Tritium Plant

  9. ITER EDA newsletter. V. 4, no. 9

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the first meeting of the ITER Test Blanket Working Group held 19-21 July 1995 at the ITER Garching Joint Work Site, and on the second workshop of the ITER Expert Group on Confinement and Transport.

  10. ITER EDA newsletter. V. 4, no. 9

    International Nuclear Information System (INIS)

    1995-09-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the first meeting of the ITER Test Blanket Working Group held 19-21 July 1995 at the ITER Garching Joint Work Site, and on the second workshop of the ITER Expert Group on Confinement and Transport

  11. ITER ITA newsletter. No. 10, November 2003

    International Nuclear Information System (INIS)

    2003-12-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about an ITER related meeting, namely, the Ninth ITER Negotiations Meeting (N-9), which was held on 9-10 November 2003 at the Fragrant Hill Golden Resources Commerce Hotel in Beijing and information about research on magnetic confinement fusion (MCF) in China

  12. ITER EDA newsletter. V. 8, no. 12

    International Nuclear Information System (INIS)

    1999-12-01

    This ITER EDA Newsletter reports about the ITER Management Advisory Committee Meeting in Naka, the ITER Technical Advisory Committee Meeting in Naka and the meeting of the ITER SWG-P2 in Vienna. A separate abstract is prepared for each meeting

  13. ITER EDA newsletter. V. 5, no. 9

    International Nuclear Information System (INIS)

    1996-09-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER project contains an overview of one of the seven large ITER Research and Development Projects identified by the ITER Director, namely the Vacuum Vessel Sector, as well as an account of computer animation created for ITER

  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. ITER ITA newsletter. No. 22, May 2005

    International Nuclear Information System (INIS)

    2005-06-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about Japanese Participant Team's recent activities in the ITER Transitional Arrangements(ITA) phase and ITER related meeting the Fourth IAEA Technical Meeting (IAEA-TM) on Negative Ion Based Neutral Beam Injectors which was held in Padova, Italy from 9-11 May 2005

  16. 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

  17. Rotation and neoclassical ripple transport in ITER

    Science.gov (United States)

    Paul, E. J.; Landreman, M.; Poli, F. M.; Spong, D. A.; Smith, H. M.; Dorland, W.

    2017-11-01

    Neoclassical transport in the presence of non-axisymmetric magnetic fields causes a toroidal torque known as neoclassical toroidal viscosity (NTV). The toroidal symmetry of ITER will be broken by the finite number of toroidal field coils and by test blanket modules (TBMs). The addition of ferritic inserts (FIs) will decrease the magnitude of the toroidal field ripple. 3D magnetic equilibria in the presence of toroidal field ripple and ferromagnetic structures are calculated for an ITER steady-state scenario using the variational moments equilibrium code (VMEC). Neoclassical transport quantities in the presence of these error fields are calculated using the stellarator Fokker-Planck iterative neoclassical conservative solver (SFINCS). These calculations fully account for E r , flux surface shaping, multiple species, magnitude of ripple, and collisionality rather than applying approximate analytic NTV formulae. As NTV is a complicated nonlinear function of E r , we study its behavior over a plausible range of E r . We estimate the toroidal flow, and hence E r , using a semi-analytic turbulent intrinsic rotation model and NUBEAM calculations of neutral beam torque. The NTV from the \\vert{n}\\vert = 18 ripple dominates that from lower n perturbations of the TBMs. With the inclusion of FIs, the magnitude of NTV torque is reduced by about 75% near the edge. We present comparisons of several models of tangential magnetic drifts, finding appreciable differences only for superbanana-plateau transport at small E r . We find the scaling of calculated NTV torque with ripple magnitude to indicate that ripple-trapping may be a significant mechanism for NTV in ITER. The computed NTV torque without ferritic components is comparable in magnitude to the NBI and intrinsic turbulent torques and will likely damp rotation, but the NTV torque is significantly reduced by the planned ferritic inserts.

  18. Repair of manufacturing defects in the armor of plasma facing units of the ITER Divertor Dome

    International Nuclear Information System (INIS)

    Litunovsky, Nikolay; Alekseenko, Evgeny; Kuznetsov, Vladimir; Lyanzberg, Dmitriy; Makhankov, Aleksey; Rulev, Roman

    2013-01-01

    Highlights: • Sporadic manufacturing defects in ITER Divertor Dome PFUs may be repaired. • We have developed a repair technique for ITER Divertor Dome PFUs. • Armor repair technique for ITER Divertor Dome PFUs is successfully tested. -- Abstract: The paper describes the repair procedure developed for removal of manufacturing defects occurring sporadically during armoring of plasma facing units (PFUs) of the ITER Divertor Dome. Availability of armor repair technique is prescribed by the procurement arrangement for the ITER Divertor Dome concluded in 2009 between the ITER Organization and the ITER Domestic Agency of Russia. The paper presents the detailed description of the procedure, data on its effect on the joints of the rest part of the armor and on the grain structure of the PFU heat sink. The results of thermocycling of large-scale Dome PFU mock-ups manufactured with demonstration of armor repair are also given

  19. Repair of manufacturing defects in the armor of plasma facing units of the ITER Divertor Dome

    Energy Technology Data Exchange (ETDEWEB)

    Litunovsky, Nikolay, E-mail: nlitunovsky@sintez.niiefa.spb.su; Alekseenko, Evgeny; Kuznetsov, Vladimir; Lyanzberg, Dmitriy; Makhankov, Aleksey; Rulev, Roman

    2013-10-15

    Highlights: • Sporadic manufacturing defects in ITER Divertor Dome PFUs may be repaired. • We have developed a repair technique for ITER Divertor Dome PFUs. • Armor repair technique for ITER Divertor Dome PFUs is successfully tested. -- Abstract: The paper describes the repair procedure developed for removal of manufacturing defects occurring sporadically during armoring of plasma facing units (PFUs) of the ITER Divertor Dome. Availability of armor repair technique is prescribed by the procurement arrangement for the ITER Divertor Dome concluded in 2009 between the ITER Organization and the ITER Domestic Agency of Russia. The paper presents the detailed description of the procedure, data on its effect on the joints of the rest part of the armor and on the grain structure of the PFU heat sink. The results of thermocycling of large-scale Dome PFU mock-ups manufactured with demonstration of armor repair are also given.

  20. ITER ITA newsletter No. 33, August-September-October 2006

    International Nuclear Information System (INIS)

    2006-11-01

    This issue of ITER ITA (ITER transitional arrangements) newsletter contains concise information about ITER related events such as public debate on ITER in Provence and fiftieth annual General Conference of the IAEA. Eight ITER related statements were made during Conference

  1. ITER Construction--Plant System Integration

    International Nuclear Information System (INIS)

    Tada, E.; Matsuda, S.

    2009-01-01

    This brief paper introduces how the ITER will be built in the international collaboration. The ITER Organization plays a central role in constructing ITER and leading it into operation. Since most of the ITER components are to be provided in-kind from the member countries, integral project management should be scoped in advance of real work. Those include design, procurement, system assembly, testing, licensing and commissioning of ITER.

  2. 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

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

    International Nuclear Information System (INIS)

    1995-12-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a report on the ninth ITER council meeting held December 12 - 13, 1995 in Garching near Munich, Germany (by Dr. E. Canobbio), a report on the status of the ITER EDA (by Dr. R. Aymar, ITER Director) and a report on the ninth meeting of the ITER Technical Advisory Committee (by Professor P. Rutherford, TAC Chair) held 27 - 29 November 1995, in Garching near Munich, Germany

  4. ITER ITA newsletter. No. 4, May 2003

    International Nuclear Information System (INIS)

    2003-07-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related meetings, one of them the eighth meeting of the ITER negotiators' standing sub-group (NSSG-8) and a number of related meetings from 14 to 22 May 2003 at Garching, Germany, another was bilateral blanket meeting between ITER International Team (IT) and the Research and Development Institute of Power Engineering (ENTEK), which was held in Moscow, Russian Federation on 22 and 23 May, 2003

  5. ITER ITA newsletter. Special issue - December 2006

    International Nuclear Information System (INIS)

    2006-12-01

    This issue of ITER ITA (ITER transitional arrangements) newsletter contains information about signing ITER Agreement, which took place on 21 November 2006 in Paris, France. It was great day for fusion research as Ministers from the seven ITER Parties in the presence of President Jacques Chirac and President of European Commission Jose Barroso and some 400 invited guests signed the Agreement setting up the ITER International Fusion Energy Organization. This issues contains the speeches, statements and remarks of Presidents and Ministers

  6. ITER EDA newsletter. V. 5, no. 7

    International Nuclear Information System (INIS)

    1996-07-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on the Tenth ITER Council Meeting, held July 24-25, 1996, in St. Petersburg, Russia; a description of the Status of the ITER EDA by the ITER Director, Dr. R. Aymar; and a report on the so-called Task Number One by the ITER Special Working Group (Basis for the Start of Explorations, presenting possible scenarios toward siting, licensing and host support)

  7. ITER ITA newsletter. No. 1, February 2003

    International Nuclear Information System (INIS)

    2003-04-01

    This first issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related meetings including eighth ITER Negotiations meeting, held on 18-19 February, 2003 in St. Petersburg, Russia, first meeting of the ITER preparatory committee, held on 17 February, 2003 in St. Petersburg, Russia and the third meeting of the ITPA (International Tokamak Physics Activity) coordinating committee, held on 24-25 October 2002 at the Max-Planck-Institut fuer Plasmaphysik, Garching

  8. 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)

  9. ITER project and fusion technology

    International Nuclear Information System (INIS)

    Takatsu, H.

    2011-01-01

    In the sessions of ITR, FTP and SEE of the 23rd IAEA Fusion Energy Conference, 159 papers were presented in total, highlighted by the remarkable progress of the ITER project: ITER baseline has been established and procurement activities have been started as planned with a target of realizing the first plasma in 2019; ITER physics basis is sound and operation scenarios and operational issues have been extensively studied in close collaboration with the worldwide physics community; the test blanket module programme has been incorporated into the ITER programme and extensive R and D works are ongoing in the member countries with a view to delivering their own modules in a timely manner according to the ITER master schedule. Good progress was also reported in the areas of a variety of complementary activities to DEMO, including Broader Approach activities and long-term technology. This paper summarizes the highlights of the papers presented in the ITR, FTP and SEE sessions with a minimum set of background information.

  10. Remote-LIBS characterization of ITER-like plasma facing materials

    International Nuclear Information System (INIS)

    Almaviva, S.; Caneve, L.; Colao, F.; Fantoni, R.; Maddaluno, G.

    2012-01-01

    Graphical abstract: Display Omitted Highlights: ► Description of a LIBS set-up as remote diagnostics in new generation fusion machines. ► Identification of the atomic composition of samples simulating plasma facing components. ► Submicrometric resolution in depth profiling the elemental composition of the samples. ► Identification of elements present in traces or as impurities on the sample surface. ► Discussion on the applicability of the Calibration Free method for quantitative analysis. - Abstract: The occurrence of several plasma-wall interaction processes, eventually affecting the overall system performances, is expected in a working fusion device chamber. Monitoring the changes in the composition of the plasma facing component (PFC) surface layer, as a result of erosion and redeposition mechanisms, can provide useful information on the possible plasma pollution and fuel retention. To this aim, suitable diagnostic techniques able to perform depth profiling analysis of the superficial layers on the PFCs have been developed. Due to the constraints commonly found in fusion devices, the measuring apparatus must be non invasive, remote and sensitive to light elements. These requirements make LIBS (Laser Induced Breakdown Spectroscopy) an ideal candidate for on-line monitoring the walls of current and of next generation (as ITER) fusion devices. LIBS is a well established tool for qualitative, semi-quantitative and quantitative analysis of surfaces, with micro-destructive characteristics and some capabilities for stratigraphy. In this work, LIBS depth profiling capability has been verified for the determination of the composition of multilayer structures simulating plasma facing components covered with deposited impurity layers. A new experimental setup has been designed and realized in order to optimize the characteristics of a LIBS system working in vacuum conditions and remotely, two noticeable properties for an ITER-relevant diagnostics. A quantitative

  11. Alternative dark matter candidates. Axions

    International Nuclear Information System (INIS)

    Ringwald, Andreas

    2017-01-01

    The axion is arguably one of the best motivated candidates for dark matter. For a decay constant >or similar 10 9 GeV, axions are dominantly produced non-thermally in the early universe and hence are ''cold'', their velocity dispersion being small enough to fit to large scale structure. Moreover, such a large decay constant ensures the stability at cosmological time scales and its behaviour as a collisionless fluid at cosmological length scales. Here, we review the state of the art of axion dark matter predictions and of experimental efforts to search for axion dark matter in laboratory experiments.

  12. Alternative dark matter candidates. Axions

    Energy Technology Data Exchange (ETDEWEB)

    Ringwald, Andreas

    2017-01-15

    The axion is arguably one of the best motivated candidates for dark matter. For a decay constant >or similar 10{sup 9} GeV, axions are dominantly produced non-thermally in the early universe and hence are ''cold'', their velocity dispersion being small enough to fit to large scale structure. Moreover, such a large decay constant ensures the stability at cosmological time scales and its behaviour as a collisionless fluid at cosmological length scales. Here, we review the state of the art of axion dark matter predictions and of experimental efforts to search for axion dark matter in laboratory experiments.

  13. The ITER project technological challenges

    CERN Multimedia

    CERN. Geneva; Lister, Joseph; Marquina, Miguel A; Todesco, Ezio

    2005-01-01

    The first lecture reminds us of the ITER challenges, presents hard engineering problems, typically due to mechanical forces and thermal loads and identifies where the physics uncertainties play a significant role in the engineering requirements. The second lecture presents soft engineering problems of measuring the plasma parameters, feedback control of the plasma and handling the physics data flow and slow controls data flow from a large experiment like ITER. The last three lectures focus on superconductors for fusion. The third lecture reviews the design criteria and manufacturing methods for 6 milestone-conductors of large fusion devices (T-7, T-15, Tore Supra, LHD, W-7X, ITER). The evolution of the designer approach and the available technologies are critically discussed. The fourth lecture is devoted to the issue of performance prediction, from a superconducting wire to a large size conductor. The role of scaling laws, self-field, current distribution, voltage-current characteristic and transposition are...

  14. 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.

  15. 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

  16. 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

  17. The danger of iteration methods

    International Nuclear Information System (INIS)

    Villain, J.; Semeria, B.

    1983-01-01

    When a Hamiltonian H depends on variables phisub(i), the values of these variables which minimize H satisfy the equations deltaH/deltaphisub(i) = O. If this set of equations is solved by iteration, there is no guarantee that the solution is the one which minimizes H. In the case of a harmonic system with a random potential periodic with respect to the phisub(i)'s, the fluctuations have been calculated by Efetov and Larkin by means of the iteration method. The result is wrong in the case of a strong disorder. Even in the weak disorder case, it is wrong for a one-dimensional system and for a finite system of 2 particles. It is argued that the results obtained by iteration are always wrong, and that between 2 and 4 dimensions, spin-pair correlation functions decay like powers of the distance, as found by Aharony and Pytte for another model

  18. ITER Operating Limits and Conditions

    International Nuclear Information System (INIS)

    Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.

    2006-01-01

    The Operating Limits and Conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which together define the domain of the safe operation of ITER in all foreseen ITER status (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components important for personnel or public safety (Safety Important Class, SIC) are identified from the overall plant safety analysis on functional importance to safety of the components. SIC classification has to be presented already inside the preliminary safety analysis report and approved by the licensing safety authority before the relevant construction. OLCs comprise the safety limits, i.e. that if exceeded could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems and the operational limits on equipment which warn from or stop a functional departure from a planned operational status that could challenge equipment and functions. The safety limits have to indicate clearly states that leave the nominal safety state of ITER; they are derived from the safety analysis of ITER. OLCs can represent in some cases few parameters grouping together. Some operational conditions, e.g. inventories, will be controlled through no real time measurements and procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants is considered at the maximum possible extent. This paper presents the guidelines to develop the ITER OLCs with particular reference to safety limits. A few examples are reported as well as open issues on some OLCs control and measurement and the relevant R-and-D planned to solve the issues. (author)

  19. Progress in standardization for ITER Remote Handling control system

    International Nuclear Information System (INIS)

    Hamilton, David Thomas; Tesini, Alessandro; Ranz, Roberto; Kozaka, Hiroshi

    2014-01-01

    Graphical abstract: - Highlights: • Standard parts specified for ITER Remote Handling (RH) control system. • Standard approach for VR modeling of structural deformations in real-time. • RH Core System produced as standard platform for RH controller applications. • Synthetic Viewing investigated and demonstrated. • Structured language defined for RH operation procedures and motion sequences. - Abstract: An integrated control system architecture has been defined for the ITER Remote Handling (RH) equipment systems, and work has been continuing to develop and validate standards for this architecture. Evaluations of standard parts and a standard control room work-cell have contributed to an update of the RH Control System Design Handbook, while R and D activities have been carried out to validate concepts for standard solutions to ITER RH problems: the use of a standard master arm with different slave arms, the achievement of high accuracy tracking of RH operations within virtual reality, and condition monitoring of RH equipment systems. The standardization efforts have been consolidated through the development of a freely distributable software platform to support the adoption of the ITER RH standards. The RH Core System installs on top of the CODAC Core System and provides the basic platform for the development of ITER RH equipment controller applications. The standardization work has continued in the areas of RH viewing, network communication protocols, and a structured language for programming ITER RH operations. Prototyping has been done on high-level control system applications, and R and D has been carried out in the area of synthetic viewing for ITER RH. These developments will be reflected in a new version of the RH Core System to be produced during 2013

  20. Progress in standardization for ITER Remote Handling control system

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, David Thomas, E-mail: david.hamilton@iter.org [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Tesini, Alessandro [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Ranz, Roberto [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Kozaka, Hiroshi [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan)

    2014-10-15

    Graphical abstract: - Highlights: • Standard parts specified for ITER Remote Handling (RH) control system. • Standard approach for VR modeling of structural deformations in real-time. • RH Core System produced as standard platform for RH controller applications. • Synthetic Viewing investigated and demonstrated. • Structured language defined for RH operation procedures and motion sequences. - Abstract: An integrated control system architecture has been defined for the ITER Remote Handling (RH) equipment systems, and work has been continuing to develop and validate standards for this architecture. Evaluations of standard parts and a standard control room work-cell have contributed to an update of the RH Control System Design Handbook, while R and D activities have been carried out to validate concepts for standard solutions to ITER RH problems: the use of a standard master arm with different slave arms, the achievement of high accuracy tracking of RH operations within virtual reality, and condition monitoring of RH equipment systems. The standardization efforts have been consolidated through the development of a freely distributable software platform to support the adoption of the ITER RH standards. The RH Core System installs on top of the CODAC Core System and provides the basic platform for the development of ITER RH equipment controller applications. The standardization work has continued in the areas of RH viewing, network communication protocols, and a structured language for programming ITER RH operations. Prototyping has been done on high-level control system applications, and R and D has been carried out in the area of synthetic viewing for ITER RH. These developments will be reflected in a new version of the RH Core System to be produced during 2013.

  1. Array architectures for iterative algorithms

    Science.gov (United States)

    Jagadish, Hosagrahar V.; Rao, Sailesh K.; Kailath, Thomas

    1987-01-01

    Regular mesh-connected arrays are shown to be isomorphic to a class of so-called regular iterative algorithms. For a wide variety of problems it is shown how to obtain appropriate iterative algorithms and then how to translate these algorithms into arrays in a systematic fashion. Several 'systolic' arrays presented in the literature are shown to be specific cases of the variety of architectures that can be derived by the techniques presented here. These include arrays for Fourier Transform, Matrix Multiplication, and Sorting.

  2. ITER oriented issues-2 (etc.)

    International Nuclear Information System (INIS)

    Goryayev, G.V.; Savchuk, V.V.; Shakhvorostov, Yu. V.

    2004-01-01

    The study analyzes the possibilities of utilization beryllium ingots produced at UMZ (Ulba Metallurgical Plant) for the purpose of ITER program. The results of comparative analysis of specification requirement to S-65 grade chemical compound and statistics data of UMZ beryllium ingots impurities content are presented. It has been demonstrated that beryllium industrial ingots produced at UMZ can be used for a production of powders and billets conforming the requirements of ITER specification. Beryllium ingots production flow chart, description of basic process equipment, the layout of metallurgical production upgrade, the results of such upgrade implementation are complimentary data to this study. The study illustrated with explanatory drawings. (author)

  3. Radwaste management aspects of the test blanket systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Canas, D. [CEA, DEN/DADN, centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Chaudhari, V. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Iseli, M. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Kawamura, Y. [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Petit, P. [European Commission, DG ENER, Brussels (Belgium); Pitcher, C.S.; Torcy, D. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Ugolini, D. [Fusion for Energy, Barcelona (Spain); Zhang, H. [China Nuclear Energy Industry Corporation, Beijing 100032 (China)

    2016-11-01

    Highlights: • Test Blanket Systems are operated in ITER to test tritium breeding technologies. • The in-vessel parts of TBS become radio-active during the ITER nuclear phase. • For each TBM campaign the TBM, its shield and the Pipe Forests are removed. • High tritium contents and novel materials are specific TBS radwaste features. • A preliminary assessment confirmed RW routing, provided its proper conditioning. - Abstract: Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

  4. 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

  5. Final report of the ITER EDA. Final report of the ITER Engineering Design Activities. Prepared by the ITER Council

    International Nuclear Information System (INIS)

    2001-01-01

    This is the Final Report by the ITER Council on work carried out by ITER participating countries on cooperation in the Engineering Design Activities (EDA) for the ITER. In this report the main ITER EDA technical objectives, the scope of ITER EDA, its organization and resources, engineering design of ITER tokamak and its main parameters are presented. This Report also includes safety and environmental assessments, site requirements and proposed schedule and estimates of manpower and cost as well as proposals on approaches to joint implementation of the project

  6. An Iterative Travel Time Inversion and Waveform Modeling Method to Determine the Crust Structure and Focal Mechanism: Case Study of 2015 Alxa Left Banner Ms5.8 Earthquake

    Science.gov (United States)

    Song, C.; Ge, Z.

    2017-12-01

    The boundary region between Alxa Block and Ordos Block is an area of stress concentration with strong seismicity and frequent small earthquakes. However, the knowledge of this area is limited since only a few seismic stations were deployed in this area. The 2015 Ms5.8 Alxa Left Banner Earthquake on April 15 is the largest one occurred in the surroundings since the 1976 Ms6.2 Bayinmuren Earthquake. Abundant stations built in the northern part of Chinese North-South Seismic Belt recorded this event sequence well within short distance, which provides us a great opportunity to carry out studies. We use these data to obtain a mean 1-D layered velocity structure via iterative inversion based on both travel time and waveform misfits. Then we use the travel time difference between data and synthetic seismograms to relocate the epicenter. Finally we invert the best double-couple focal mechanism and centroid depths of the source. As the result, the source is located at (39.7027° N, 106.4207° E) with a depth of 18 km and Mw 5.28. Nodal plane Ⅰ has strike 86°, dip angle 90° and slip angle -3°, while plane Ⅱ has strike 176°, dip angle 87° and slip angle 180°. Considering the dynamic structure of regional fault zone, we believe this earthquake is caused by a nearly pure left-lateral strike-slip fault with nodal plane Ⅰ being the fault plane. The seismogenic structure is likely to be an E-W striking buried fault nearby. There develops several groups of NE, NEE and E-W striking faults in Jilantai tectonic zone, parts of which have been verified by geophysical investigations. But we still know little about the dynamic nature of them. From our study, the corresponding fault of this event may indicate all groups of faults with same E-W strike has the common character of large-dip left-lateral strike-slip. Moreover, there may be some buried faults being newly born or not found yet. These results could be an important supplement to the future research of seismicity and

  7. ITER ITA newsletter No. 31, June 2006

    International Nuclear Information System (INIS)

    2006-07-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about initialling the ITER Agreement and its related instruments by seven ITER parties, which too place in Brussels on 24 May 2006. The initialling constituted the final act of the ITER negotiations. It confirmed the Parties' common acceptance of the negotiated texts, ad referendum, and signalled their intentions to move forward towards the entry into force of the ITER Agreement as soon as possible. 'ITER - Uniting science today, global energy tomorrow' was the theme of a number of media events timed to accompany a remarkable day in the history of the ITER international venture, May 24th 2006, initialling of the ITER international agreement

  8. Absorption of lower hybrid waves by alpha particles in ITER

    International Nuclear Information System (INIS)

    Imbeaux, F.; Peysson, Y.; Eriksson, L.G.

    2003-01-01

    Absorption of lower hybrid (LH) waves by alpha particles may reduce significantly the current drive efficiency of the waves in a reactor or burning plasma experiment. This absorption is quantified for ITER using the ray-tracing+2D relativistic Fokker-Planck code Delphine. The absorption is calculated as a function of the superthermal alpha particle density, which is constant in these simulations, for two candidate frequencies for the LH system of ITER. Negligible absorption by alpha particles at 3.7 GHz requires n(alpha,supra) = 7.5 10 16 m -3 , while no significant impact on the driven current is found at 5 GHz, even if n(alpha,supra) = 1.5 10 18 m -3 . (authors)

  9. Recovery act. Characterizing structural controls of EGS-candidate and conventional geothermal reservoirs in the Great Basin. Developing successful exploration strategies in extended terranes

    Energy Technology Data Exchange (ETDEWEB)

    Faulds, James [Univ. of Nevada, Reno, NV (United States)

    2015-06-25

    We conducted a comprehensive analysis of the structural controls of geothermal systems within the Great Basin and adjacent regions. Our main objectives were to: 1) Produce a catalogue of favorable structural environments and models for geothermal systems. 2) Improve site-specific targeting of geothermal resources through detailed studies of representative sites, which included innovative techniques of slip tendency analysis of faults and 3D modeling. 3) Compare and contrast the structural controls and models in different tectonic settings. 4) Synthesize data and develop methodologies for enhancement of exploration strategies for conventional and EGS systems, reduction in the risk of drilling non-productive wells, and selecting the best EGS sites.

  10. ITER EDA Newsletter. V. 4, no. 7

    International Nuclear Information System (INIS)

    1995-07-01

    This ITER EDA (Engineering Design Activities) Newsletter issue contains reports on (i) the 8th meeting of the ITER Technical Advisory Committee (TAC-8) held on June 29 - July 7, 1995 at the ITER San Diego Work Site, (ii) the 8th meeting of the ITER Management Advisory Committee (MAC-8) held at the ITER San Diego Work Site on July 9-10, 1995, (iii) the 33rd meeting of the International Fusion Research Council (FRC), held July 11, 1995 at the IAEA Headquarters in Vienna, Austria, and (iv) the ITER participation in the fifth topical meeting on Tritium Technology in Fission, Fusion and Isotopic Applications

  11. ITER ITA Newsletter. No. 29, March 2006

    International Nuclear Information System (INIS)

    2006-05-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities and meetings, namely, the ITER Director-General Nominee, Dr. Kaname Ikeda, took up his position as ITER Project Leader in Cadarache on 13 March, the consolidation of information technology infrastructure for ITER and about he Thirty-Fifth Meeting of the Fusion Power Co-ordinating Committee (FPCC), which was held on 28 February-1 March 2006 at the headquarters of the International Energy Agency (IEA) in Paris

  12. ITER Council tour of Clarington site

    International Nuclear Information System (INIS)

    Dautovich, D.

    2001-01-01

    The ITER Council meeting was recently held in Toronto on 27 and 28 February. ITER Canada provided local arrangements for the Council meeting on behalf of Europe as the Official host. Following the meeting, on 1 March, ITER Canada conducted a tour of the proposed ITER construction site at Charington, and the ITER Council members attended a luncheon followed by a speech by Dr. Peter Barnard, Chairman and CEO of ITER Canada, at the Empire Club of Canada. The official invitation to participate in these events came from Dr. Peter Harrison, Deputy Minister of Natural Resources Canada. This report provides a brief summary of the events on 1 March

  13. ITER operating limit definition criteria

    International Nuclear Information System (INIS)

    Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.; Chiocchio, S.; Hureau, D.; Girard, J.Ph.; Gordon, C.; Portone, A.; Rodrigo, L. Rodriguez; Roldan, C.; Saibene, G.; Uzan-Elbez, J.

    2009-01-01

    The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.

  14. 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)

  15. Status of the ITER EDA

    International Nuclear Information System (INIS)

    Aymar, R.

    1999-01-01

    This article summarises progress made in the ITER Design Activities between October 1998 and February 1999. The three main focusses of the activity were on design work, on R and D work and on the physics basis. The consequences of diminishing financial funds and personnel are discussed and the state of the individual R and D projects is given briefly

  16. Japanese site for ITER: Rokkasho

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi

    2003-01-01

    This paper describes the status of Japanese efforts for hosting ITER in Japan. In May 2002, Japanese Government decided to propose an ITER site, Rokkasho in Aomori Prefecture, a Northern part of the main island, based on the comprehensive/intensive assessments by the Site Selection Committee established by Japanese Government. ITER is designed basically with a potential flexibility beyond the detailed technical objectives to have more clear scope for developing technical key elements in a future power plant. Various flexibilities in the construction, operation and decommissioning of ITER are totally assessed. Consequently the Japanese site has been chosen and it satisfies sufficiently not only the Site Requirements and the Site Design Assumptions but also the further extension and flexibilities. In particular the potential for more flexible construction schedule and operations is technically described as well as the fulfillment of the site requirements and its assumptions as the minimum requirements in this paper. The socio-cultural environment is also described briefly because of a key aspect for the scientists and engineers who will participate in the project

  17. New developments in iterated rounding

    NARCIS (Netherlands)

    Bansal, N.; Raman, V.; Suresh, S.P.

    2014-01-01

    Iterated rounding is a relatively recent technique in algorithm design, that despite its simplicity has led to several remarkable new results and also simpler proofs of many previous results. We will briefly survey some applications of the method, including some recent developments and giving a high

  18. Iterative method for Amado's model

    International Nuclear Information System (INIS)

    Tomio, L.

    1980-01-01

    A recently proposed iterative method for solving scattering integral equations is applied to the spin doublet and spin quartet neutron-deuteron scattering in the Amado model. The method is tested numerically in the calculation of scattering lengths and phase-shifts and results are found better than those obtained by using the conventional Pade technique. (Author) [pt

  19. Informal meeting on ITER developments

    International Nuclear Information System (INIS)

    Canobbio, E.

    2000-01-01

    The International Fusion Research Council (IFRC), advisory body of the IAEA, organized an informal meeting on the general status and outlook for ITER, held October 9 at Sorrento, Italy, in conjunction with the 18th IAEA Fusion Energy Conference. This article describes the main events at the meeting

  20. Enhanced mucosal immune responses induced by a combined candidate mucosal vaccine based on Hepatitis A virus and Hepatitis E virus structural proteins linked to tuftsin.

    Science.gov (United States)

    Gao, Yan; Su, Qiudong; Yi, Yao; Jia, Zhiyuan; Wang, Hao; Lu, Xuexin; Qiu, Feng; Bi, Shengli

    2015-01-01

    Hepatitis A virus (HAV) and Hepatitis E virus (HEV) are the most common causes of infectious hepatitis. These viruses are spread largely by the fecal-oral route and lead to clinically important disease in developing countries. To evaluate the potential of targeting hepatitis A and E infection simultaneously, a combined mucosal candidate vaccine was developed with the partial open reading frame 2 (ORF2) sequence (aa 368-607) of HEV (HE-ORF2) and partial virus protein 1 (VP1) sequence (aa 1-198) of HAV (HA-VP1), which included the viral neutralization epitopes. Tuftsin is an immunostimulatory peptide which can enhance the immunogenicity of a protein by targeting it to macrophages and dendritic cells. Here, we developed a novel combined protein vaccine by conjugating tuftsin to HE-ORF2 and HA-VP1 and used synthetic CpG oligodeoxynucleotides (ODNs) as the adjuvant. Subsequent experiments in BALB/c mice demonstrated that tuftsin enhanced the serum-specific IgG and IgA antibodies against HEV and HAV at the intestinal, vaginal and pulmonary interface when delivered intranasally. Moreover, mice from the intranasally immunized tuftsin group (HE-ORF2-tuftsin + HA-VP1-tuftsin + CpG) showed higher levels of IFN-γ-secreting splenocytes (Th1 response) and ratio of CD4+/CD8+ T cells than those of the no-tuftsin group (HE-ORF2 + HA-VP1 + CpG). Thus, the tuftsin group generated stronger humoral and cellular immune responses compared with the no-tuftsin group. Moreover, enhanced responses to the combined protein vaccine were obtained by intranasal immunization compared with intramuscular injection. By integrating HE-ORF2, HA-VP1 and tuftsin in a vaccine, this study validated an important concept for further development of a combined mucosal vaccine against hepatitis A and E infection.

  1. Development of ITER relevant laser techniques for deposited layer characterisation and tritium inventory

    NARCIS (Netherlands)

    Malaquias, A.; Philipps, V.; Huber, A.; Hakola, A.; Likonen, J.; Kolehmainen, J.; Tervakangas, S.; Aints, M.; Paris, P.; Laan, M.; Lissovski, A.; Almaviva, S.; Caneve, L.; Colao, F.; Maddaluno, G.; Kubkowska, M.; Gasior, P.; van der Meiden, H. J.; Lof, A. R.; van Emmichoven, P. A. Zeijlma; Petersson, P.; Rubel, M.; Fortuna, E.; Xiao, Q.

    2013-01-01

    Laser Induced Breakdown Spectroscopy (LIBS) is a potential candidate to monitor the layer composition and fuel retention during and after plasma shots on specific locations of the main chamber and divertor of ITER. This method is being investigated in a cooperative research programme on plasma

  2. In situ synthesis of 3D CoS nanoflake/Ni(OH)_2 nanosheet nanocomposite structure as a candidate supercapacitor electrode

    International Nuclear Information System (INIS)

    Li, Songzhan; Wen, Jian; Chen, Tian; Xiong, Liangbin; Wang, Jianbo; Fang, Guojia

    2016-01-01

    A three-dimensional (3D) CoS/Ni(OH)_2 nanocomposite structure based on CoS nanoflakes and two-dimensional (2D) Ni(OH)_2 nanosheets were in situ synthesized on Ni foam by a whole hydrothermal reaction and electrodeposition process. The 3D CoS/Ni(OH)_2 nanocomposite structures demonstrate the combined advantages of a sustained cycle stability of CoS and high specific capacitance from Ni(OH)_2. The obtained CoS/Ni(OH)_2 nanocomposite structures on Ni foam can directly serve as a binder-free electrode for a supercapacitor. For the 3D CoS/Ni(OH)_2 nanocomposite electrode, the high specific capacitance is 1837 F g"−"1 at a scan rate of 1 mV s"−"1, which is obviously higher than both the bare CoS electrode and Ni(OH)_2 electrode. The galvanostatic charge and discharge measurements illustrate that the 3D CoS/Ni(OH)_2 nanocomposite electrode possesses excellent cycle stability, and it keeps a 95.8% retention of the initial capacity after 5000 cycles. Electrochemical impedance spectroscopy measurements also confirm that the 3D CoS/Ni(OH)_2 nanocomposite electrode has better electrochemical characteristics. These remarkable performances can be attributed to the unique 3D nanoporous structure of CoS/Ni(OH)_2 which leads to a large accessible surface area and a high stability during long-term operation. In addition, 2D Ni(OH)_2 nanosheets in 3D nanocomposite structures can afford rapid mass transport and a strong synergistic effect of CoS and Ni(OH)_2 as individual compositions contribute to the high performance of the nanocomposite structure electrode. These results may promote the design and implementation of nanocomposite structures in advanced supercapacitors. (paper)

  3. A linear iterative unfolding method

    International Nuclear Information System (INIS)

    László, András

    2012-01-01

    A frequently faced task in experimental physics is to measure the probability distribution of some quantity. Often this quantity to be measured is smeared by a non-ideal detector response or by some physical process. The procedure of removing this smearing effect from the measured distribution is called unfolding, and is a delicate problem in signal processing, due to the well-known numerical ill behavior of this task. Various methods were invented which, given some assumptions on the initial probability distribution, try to regularize the unfolding problem. Most of these methods definitely introduce bias into the estimate of the initial probability distribution. We propose a linear iterative method (motivated by the Neumann series / Landweber iteration known in functional analysis), which has the advantage that no assumptions on the initial probability distribution is needed, and the only regularization parameter is the stopping order of the iteration, which can be used to choose the best compromise between the introduced bias and the propagated statistical and systematic errors. The method is consistent: 'binwise' convergence to the initial probability distribution is proved in absence of measurement errors under a quite general condition on the response function. This condition holds for practical applications such as convolutions, calorimeter response functions, momentum reconstruction response functions based on tracking in magnetic field etc. In presence of measurement errors, explicit formulae for the propagation of the three important error terms is provided: bias error (distance from the unknown to-be-reconstructed initial distribution at a finite iteration order), statistical error, and systematic error. A trade-off between these three error terms can be used to define an optimal iteration stopping criterion, and the errors can be estimated there. We provide a numerical C library for the implementation of the method, which incorporates automatic

  4. A comprehensive iterative approach is highly effective in diagnosing individuals who are exome negative.

    Science.gov (United States)

    Shashi, Vandana; Schoch, Kelly; Spillmann, Rebecca; Cope, Heidi; Tan, Queenie K-G; Walley, Nicole; Pena, Loren; McConkie-Rosell, Allyn; Jiang, Yong-Hui; Stong, Nicholas; Need, Anna C; Goldstein, David B

    2018-06-15

    Sixty to seventy-five percent of individuals with rare and undiagnosed phenotypes remain undiagnosed after exome sequencing (ES). With standard ES reanalysis resolving 10-15% of the ES negatives, further approaches are necessary to maximize diagnoses in these individuals. In 38 ES negative patients an individualized genomic-phenotypic approach was employed utilizing (1) phenotyping; (2) reanalyses of FASTQ files, with innovative bioinformatics; (3) targeted molecular testing; (4) genome sequencing (GS); and (5) conferring of clinical diagnoses when pathognomonic clinical findings occurred. Certain and highly likely diagnoses were made in 18/38 (47%) individuals, including identifying two new developmental disorders. The majority of diagnoses (>70%) were due to our bioinformatics, phenotyping, and targeted testing identifying variants that were undetected or not prioritized on prior ES. GS diagnosed 3/18 individuals with structural variants not amenable to ES. Additionally, tentative diagnoses were made in 3 (8%), and in 5 individuals (13%) candidate genes were identified. Overall, diagnoses/potential leads were identified in 26/38 (68%). Our comprehensive approach to ES negatives maximizes the ES and clinical data for both diagnoses and candidate gene identification, without GS in the majority. This iterative approach is cost-effective and is pertinent to the current conundrum of ES negatives.

  5. Detailed design of the ITER central solenoid

    International Nuclear Information System (INIS)

    Libeyre, P.; Mitchell, N.; Bessette, D.; Gribov, Y.; Jong, C.; Lyraud, C.

    2009-01-01

    The central solenoid (CS) of the ITER tokamak contributes to the inductive flux to drive the plasma, to the shaping of the field lines in the divertor region and to vertical stability control. It is made of 6 independent coils, using a Nb3Sn cable-in-conduit superconducting conductor, held together by a vertical precompression structure. This design enables ITER to access a wide operating window of plasma parameters, up to 17 MA and covering inductive and non-inductive operation. Each coil is based on a stack of multiple pancake winding units to minimise joints. A glass-polyimide electrical insulation, impregnated with epoxy resin, is giving a high voltage operating capability, tested up to 29 kV. The CS performance is fatigue driven mainly by the stress levels in the conductor jacket and in the precompression structure needed to keep the modules in contact during the repulsive forces which can arise in operation. A rigid connection to the TF coils provided at one end and a centering support at the other end allow to resist net vertical forces as well as unbalanced radial forces while avoiding torsion transmission from the TF Coils to the CS assembly.

  6. ITER plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.A.

    2000-01-01

    The ITER site has not yet been determined. 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 electrical power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, demineralized water, steam and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radiation exposure of workers and the public. To prevent 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 geographical centre of the site. The locations of the buildings on the generic site were selected to meet all design requirements at minimum total project cost. A similar approach was used to determine the locations 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)

  7. Process and overview of diagnostics integration in ITER ports

    International Nuclear Information System (INIS)

    Drevon, J.M.; Walsh, M.; Andrew, P.; Barnsley, R.; Bertalot, L.; Bock, M. de; Bora, D.; Bouhamou, R.; Direz, M.F.; Encheva, A.; Fang, T.; Feder, R.; Giacomin, T.; Hellermann, M. von; Jakhar, S.; Johnson, D.; Kaschuk, Y.; Kusama, Y.; Lee, H.G.; Levesy, B.

    2013-01-01

    Highlights: ► An overview of the Port Integration hardware for tenant system hosting inside ITER diagnostics ports is given. ► The main challenges for diagnostic port integration engineering are presented. ► The actions taken for a common modular approach and a coordinated design are detailed. -- Abstract: ITER will have a set of 45 diagnostics to ensure controlled operation. Many of them are integrated in the ITER ports. This paper addresses the integration process of the diagnostic systems and the approach taken to enable coordinated progress. An overview of the Port Integration hardware introduces the various structures needed for hosting tenant systems inside ITER diagnostics ports. The responsibilities of the different parties involved (ITER Organization and the Domestic Agencies) are outlined. The main challenges for diagnostic port integration engineering are summarized. The plan for a common approach to design and manufacture of the supporting structures, in particular the Port Plug is detailed. A coordinated design including common components and a common approach for neutronic analyses is proposed. One particular port, the equatorial port 11, is used to illustrate the approach

  8. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

    2007-01-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  9. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

    2007-07-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  10. Influence of materials choice on occupational radiation exposure in ITER

    International Nuclear Information System (INIS)

    Forty, C.B.A.; Firth, J.D.; Butterworth, G.J.

    1998-01-01

    In fission reactor plant, the radiation doses associated with inspection and maintenance of the primary cooling circuit account for a substantial fraction of the collective occupational radiation exposure (ORE). Similarly, it is anticipated that much of the ORE occurring during normal operation of ITER will arise from active deposits in the cooling loop. Using a number of calculation steps ranging from neutron activation analysis, mobilisation and transport modelling and Monte Carlo simulation, estimates for the gamma photon flux and radiation dose fields around a typical 'hot-leg' cooling pipe have been made taking SS316, OPTSTAB, MANET-II and F-82H steels as alternative candidate loop materials. (orig.)

  11. ITER EDA newsletter. V. 5, no. 8

    International Nuclear Information System (INIS)

    1996-08-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on the divertor remote handling development (and of a summer party at the ITER Joint Work Site in Garching, Germany)

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

    International Nuclear Information System (INIS)

    1999-03-01

    This newsletter contains reports on: 'The ITER management advisory committee (MAC) meeting in Cadarache' and on 'ITER technical advisory committee meeting at Garching'. It contains photos of the participants of both meetings

  13. ITER EDA newsletter. V. 5, no. 8

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This issue of the Newsletter on the Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on the divertor remote handling development (and of a summer party at the ITER Joint Work Site in Garching, Germany).

  14. Iteration and Prototyping in Creating Technical Specifications.

    Science.gov (United States)

    Flynt, John P.

    1994-01-01

    Claims that the development process for computer software can be greatly aided by the writers of specifications if they employ basic iteration and prototyping techniques. Asserts that computer software configuration management practices provide ready models for iteration and prototyping. (HB)

  15. ITER EDA newsletter. V. 7, special issue

    International Nuclear Information System (INIS)

    1998-07-01

    In conjunction with the ITER Council Meeting, a ceremony was held at the IAEA Headquarters in Vienna on 22 July 1998 to celebrate the achievements of the ITER Engineering Design Activities during the period 1992-1998

  16. Iterative perceptual learning for social behavior synthesis

    NARCIS (Netherlands)

    de Kok, I.A.; Poppe, Ronald Walter; Heylen, Dirk K.J.

    We introduce Iterative Perceptual Learning (IPL), a novel approach to learn computational models for social behavior synthesis from corpora of human–human interactions. IPL combines perceptual evaluation with iterative model refinement. Human observers rate the appropriateness of synthesized

  17. Iterative Perceptual Learning for Social Behavior Synthesis

    NARCIS (Netherlands)

    de Kok, I.A.; Poppe, Ronald Walter; Heylen, Dirk K.J.

    We introduce Iterative Perceptual Learning (IPL), a novel approach for learning computational models for social behavior synthesis from corpora of human-human interactions. The IPL approach combines perceptual evaluation with iterative model refinement. Human observers rate the appropriateness of

  18. ITER EDA Newsletter. V.3, no.4

    International Nuclear Information System (INIS)

    1994-04-01

    This ITER EDA Newsletter issue contains a report on the fifth meeting of the ITER Management Advisory Committee and a summary of a magnet and safety technical meeting held at Naka, February 22-25, 1994

  19. ITER EDA newsletter. V. 8, no. 8

    International Nuclear Information System (INIS)

    1999-08-01

    This ITER EDA newsletter reports on the programme directors meeting of 28-29 July 1999, the Snowmass Fusion Summer Study Group workshop and the ITER Management Advisory Committee meeting in Garching. Individual abstracts are prepared for the 3 meetings

  20. ITER EDA newsletter. V. 8, no. 7

    International Nuclear Information System (INIS)

    1999-07-01

    This newsletter contains an article concerning the ITER divertor cassette project meeting in Bologna, Italy (May 26-28, 1999), and an emotional outburst, concerning the closure of the ITER site in San Diego, USA

  1. ITER EDA newsletter. V. 2, no. 11

    International Nuclear Information System (INIS)

    1993-11-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains an ITER EDA Status Report, and a report on the Fourth International Fusion Neutronics Workshop at the University of California, Los Angeles Campus, October 20-21, 1993

  2. ITER on display at Yokohama, Japan

    International Nuclear Information System (INIS)

    Green, B.

    1998-01-01

    During the 17th IAEA Fusion Energy Conference, held in Yokohama, Japan, from 19-24 October 1998, the ITER Project exhibited a display, primarily of the research and development which has taken place to support the design of ITER

  3. ITER EDA newsletter. V. 6, no. 12

    International Nuclear Information System (INIS)

    1997-12-01

    This issue of the ITER Newsletter contains summary reports (i) on the Sixth ITER Technical Meeting on Safety and Environment and (ii) on JAERI's Annual Public Seminar on Fusion Research and Development

  4. Thermomechanical analysis of the DFLL test blanket module for ITER

    International Nuclear Information System (INIS)

    Chen Hongli; Wu Yican; Bai Yunqing

    2006-01-01

    The finite element code is used to simulate two kinds of blanket design structure, which are SLL (Quasi-Static Lithium Lead) and DLL (Dual-cooled Lithium Lead) blanket concepts for the Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM) submitted to the ITER test blanket working group. The temperature and stress distributions have been presented for the two kinds of blanket structure on the basis of the structural design, thermal-hydraulic design and neutronics analysis. Also the mechanical performance is presented for the high temperature component of blanket structure according to the ITER Structural Design Criteria (ISDC). The rationality and feasibility of the two kinds of blanket structure design of DFLL-TBM have been analyzed based on the above results which also acted as the theoretical base for further optimized analysis. (authors)

  5. Meeting of the ITER Council

    International Nuclear Information System (INIS)

    Drew, M.

    2001-01-01

    Full text: A meeting of the ITER Council took place in Toronto, Canada, on 27-28 February 2001 (Canada participates in the ITER EDA as an associate of the EU Party). The delegations to the Council were led by Dr. U. Finzi, Principal Advisor in charge of Fusion R and D in the Directorate-General for Research of the European Commission, Mr. T. Sugawa, Deputy Director-General of the Research and Development Bureau of the Ministry of Education, Culture, Sport, Science and Technology of Japan, and Academician E. Velikhov, President of the RRC ''Kurchatov Institute''. The European delegation was joined by Canadian experts including a representative from the Canadian Department of Natural Resources. The Council heard presentations from Dr. H. Kishimoto on the successful completion of the Explorations concerning future joint implementation of ITER, and from Dr. J.-P. Rager on the ITER International Industry Liaison Meeting held in Toronto in November 2000. Having noted statements of Parties' status, in particular concerning the readiness to start negotiations and the progress toward site offers, the Council encouraged the Parties to pursue preparations toward future implementation of ITER along the general lines proposed in the Explorers' final report. The Council also noted the readiness the of the RF and EU Parties to instruct specified current JCT members to remain at their places of assignment after the end of the EDA, in preparation for a transition to the Co-ordinated Technical Activities foreseen as support to ITER negotiations. The Council was pleased to hear that meetings with the Director of the ITER Parties' Designated Safety Representatives had started, and commended the progress toward achieving timely licensing processes with a good common understanding. The Council noted with appreciation the Director's view that no difficulties of principle in the licensing approach had been identified during the informal discussions with the regulatory representatives and

  6. An iterative method for near-field Fresnel region polychromatic phase contrast imaging

    Science.gov (United States)

    Carroll, Aidan J.; van Riessen, Grant A.; Balaur, Eugeniu; Dolbnya, Igor P.; Tran, Giang N.; Peele, Andrew G.

    2017-07-01

    We present an iterative method for polychromatic phase contrast imaging that is suitable for broadband illumination and which allows for the quantitative determination of the thickness of an object given the refractive index of the sample material. Experimental and simulation results suggest the iterative method provides comparable image quality and quantitative object thickness determination when compared to the analytical polychromatic transport of intensity and contrast transfer function methods. The ability of the iterative method to work over a wider range of experimental conditions means the iterative method is a suitable candidate for use with polychromatic illumination and may deliver more utility for laboratory-based x-ray sources, which typically have a broad spectrum.

  7. Evaluating ITER remote handling middleware concepts

    Energy Technology Data Exchange (ETDEWEB)

    Koning, J.F., E-mail: j.f.koning@differ.nl [FOM Institute DIFFER, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M.; Schoen, P.; Smedinga, D. [Heemskerk Innovative Technology, Noordwijk (Netherlands); Boode, A.H. [University of Applied Sciences InHolland, Alkmaar (Netherlands); Hamilton, D.T. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Remote Handling Study Centre: middleware system setup and modules built. ► Aligning to ITER RH Control System Layout: prototype of database, VR and simulator. ► OpenSplice DDS, ZeroC ICE messaging and object oriented middlewares reviewed. ► Windows network latency found problematic for semi-realtime control over the network. -- Abstract: Remote maintenance activities in ITER will be performed by a unique set of hardware systems, supported by an extensive software kit. A layer of middleware will manage and control a complex set of interconnections between teams of operators, hardware devices in various operating theatres, and databases managing tool and task logistics. The middleware is driven by constraints on amounts and timing of data like real-time control loops, camera images, and database access. The Remote Handling Study Centre (RHSC), located at FOM institute DIFFER, has a 4-operator work cell in an ITER relevant RH Control Room setup which connects to a virtual hot cell back-end. The centre is developing and testing flexible integration of the Control Room components, resulting in proof-of-concept tests of this middleware layer. SW components studied include generic human-machine interface software, a prototype of a RH operations management system, and a distributed virtual reality system supporting multi-screen, multi-actor, and multiple independent views. Real-time rigid body dynamics and contact interaction simulation software supports simulation of structural deformation, “augmented reality” operations and operator training. The paper presents generic requirements and conceptual design of middleware components and Operations Management System in the context of a RH Control Room work cell. The simulation software is analyzed for real-time performance and it is argued that it is critical for middleware to have complete control over the physical network to be able to guarantee bandwidth and latency to the components.

  8. Evaluating ITER remote handling middleware concepts

    International Nuclear Information System (INIS)

    Koning, J.F.; Heemskerk, C.J.M.; Schoen, P.; Smedinga, D.; Boode, A.H.; Hamilton, D.T.

    2013-01-01

    Highlights: ► Remote Handling Study Centre: middleware system setup and modules built. ► Aligning to ITER RH Control System Layout: prototype of database, VR and simulator. ► OpenSplice DDS, ZeroC ICE messaging and object oriented middlewares reviewed. ► Windows network latency found problematic for semi-realtime control over the network. -- Abstract: Remote maintenance activities in ITER will be performed by a unique set of hardware systems, supported by an extensive software kit. A layer of middleware will manage and control a complex set of interconnections between teams of operators, hardware devices in various operating theatres, and databases managing tool and task logistics. The middleware is driven by constraints on amounts and timing of data like real-time control loops, camera images, and database access. The Remote Handling Study Centre (RHSC), located at FOM institute DIFFER, has a 4-operator work cell in an ITER relevant RH Control Room setup which connects to a virtual hot cell back-end. The centre is developing and testing flexible integration of the Control Room components, resulting in proof-of-concept tests of this middleware layer. SW components studied include generic human-machine interface software, a prototype of a RH operations management system, and a distributed virtual reality system supporting multi-screen, multi-actor, and multiple independent views. Real-time rigid body dynamics and contact interaction simulation software supports simulation of structural deformation, “augmented reality” operations and operator training. The paper presents generic requirements and conceptual design of middleware components and Operations Management System in the context of a RH Control Room work cell. The simulation software is analyzed for real-time performance and it is argued that it is critical for middleware to have complete control over the physical network to be able to guarantee bandwidth and latency to the components

  9. Baseline Architecture of ITER Control System

    Science.gov (United States)

    Wallander, A.; Di Maio, F.; Journeaux, J.-Y.; Klotz, W.-D.; Makijarvi, P.; Yonekawa, I.

    2011-08-01

    The control system of ITER consists of thousands of computers processing hundreds of thousands of signals. The control system, being the primary tool for operating the machine, shall integrate, control and coordinate all these computers and signals and allow a limited number of staff to operate the machine from a central location with minimum human intervention. The primary functions of the ITER control system are plant control, supervision and coordination, both during experimental pulses and 24/7 continuous operation. The former can be split in three phases; preparation of the experiment by defining all parameters; executing the experiment including distributed feed-back control and finally collecting, archiving, analyzing and presenting all data produced by the experiment. We define the control system as a set of hardware and software components with well defined characteristics. The architecture addresses the organization of these components and their relationship to each other. We distinguish between physical and functional architecture, where the former defines the physical connections and the latter the data flow between components. In this paper, we identify the ITER control system based on the plant breakdown structure. Then, the control system is partitioned into a workable set of bounded subsystems. This partition considers at the same time the completeness and the integration of the subsystems. The components making up subsystems are identified and defined, a naming convention is introduced and the physical networks defined. Special attention is given to timing and real-time communication for distributed control. Finally we discuss baseline technologies for implementing the proposed architecture based on analysis, market surveys, prototyping and benchmarking carried out during the last year.

  10. Neutral transport and helium pumping of ITER

    International Nuclear Information System (INIS)

    Ruzic, D.N.

    1990-08-01

    A 2-D Monte-Carlo simulation of the neutral atom densities in the divertor, divertor throat and pump duct of ITER was made using the DEGAS code. Plasma conditions in the scrape-off layer and region near the separatrix were modeled using the B2 plasma transport code. Wall reflection coefficients including the effect of realistic surface roughness were determined by using the fractal TRIM code. The DEGAS and B2 coupling was iterated until a consistent recycling was predicted. Results were obtained for a helium and a deuterium/tritium mixture on 7 different ITER divertor throat geometries for both the physics phase reference base case and a technology phase case. The geometry with a larger structure on the midplane-side of the throat opening closing the divertor throat and a divertor plate which maintains a steep slope well into the throat removed helium 1.5 times better than the reference geometry for the physics phase case and 2.2 times better for the technology phase case. At the same time the helium to hydrogen pumping ratio shows a factor of 2.34 ± .41 enhancement over the ratio of helium to hydrogen incident on the divertor plate in the physics phase and an improvement of 1.61 ± .31 in the technology phase. If the helium flux profile on the divertor plate is moved outward by 20 cm with respect to the D/T flux profile for this particular geometry, the enhancement increases to 4.36 ± .90 in the physics phase and 5.10 ± .92 in the technology phase

  11. ITER CTA newsletter. No. 15, December 2002

    International Nuclear Information System (INIS)

    2003-03-01

    This ITER CTA newsletter issue contains brief information about several meetings related to ITER. One of them is the seventh ITER Negotiations Meetings that took place in Barcelona, Spain on 9-10 December 2002, another is the final ITER CTA Project Board Meeting, which took place in Barcelona, Spain on 8 December 2002 and the last one is the Third Meeting of the International Tokamak Physics Activity (ITPA) Topical Group on diagnostics held in Toki, Japan on 18-21 September 2002

  12. ITER EDA newsletter. V. 5, no. 10

    International Nuclear Information System (INIS)

    1996-10-01

    This issue of the newsletter on the Engineering Design Activities (EDA) for the ITER Tokamak project contains a report on the Fifth ITER Technical Meeting on Safety, Environment, and Regulatory Approval, held September 29 - October 7, 1996 at the ITER San Diego Joint Work Site; and a report on the Fifth ITER Diagnostics Expert Group Workshop and Technical Meeting on Diagnostics held in Montreal, Canada, 12-13 October 1996

  13. ITER EDA newsletter. V. 5, no. 5

    International Nuclear Information System (INIS)

    1996-05-01

    This issues of the ITER Engineering Design Activities Newsletter contains a report on the Tenth Meeting of the ITER Management Advisory Committee held at JAERI Headquarters, Tokyo, June 5-6, 1996; on the Fourth ITER Divertor Physics and Divertor Modelling and Database Expert Group Workshop, held at the San Diego ITER Joint Worksite, March 11-15, 1996, and on the Agenda for the 16th IAEA Fusion Energy Conference (7-11 October 1996)

  14. ITER EDA newsletter. V. 9, no. 11

    International Nuclear Information System (INIS)

    2000-11-01

    This issue of the ITER EDA Newsletter contains discussions of three meetings, i.e., (1) the Third ITER International Industry Liaison Meeting held in Toronto, Canada (November 7-9, 2000), (2) an informal meeting on ITER developments held in Sorrento, Italy (October 9, 2000), and (3) the Thirteenth Meeting of the ITER Physics Expert Group on Diagnostics held in Naka, Japan (September 21-22, 2000)

  15. ITER EDA newsletter. V. 9, no. 8

    International Nuclear Information System (INIS)

    2000-08-01

    This ITER EDA Newsletter reports on the ITER meeting on 29-30 June 2000 in Moscow, summarizes the status report on the ITER EDA by R. Aymar, the ITER Director, and gives overviews of the expert group workshop on transport and internal barrier physics, confinement database and modelling and edge and pedestal physics, and the IEA workshop on transport barriers at edge and core. Individual abstracts have been prepared

  16. Colorado Conference on iterative methods. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The conference provided a forum on many aspects of iterative methods. Volume I topics were:Session: domain decomposition, nonlinear problems, integral equations and inverse problems, eigenvalue problems, iterative software kernels. Volume II presents nonsymmetric solvers, parallel computation, theory of iterative methods, software and programming environment, ODE solvers, multigrid and multilevel methods, applications, robust iterative methods, preconditioners, Toeplitz and circulation solvers, and saddle point problems. Individual papers are indexed separately on the EDB.

  17. Next-generation sequencing of 100 candidate genes in young victims of suspected sudden cardiac death with structural abnormalities of the heart

    DEFF Research Database (Denmark)

    Hertz, C L; Christiansen, S L; Ferrero-Miliani, Laura

    2016-01-01

    with non-diagnostic structural abnormalities of the heart. METHODS AND RESULTS: We screened 72 suspected SCD cases (HaloPlex Target Enrichment System (Agilent) and NGS (Illumina MiSeq) for 100 genes previously associated with inherited cardiomyopathies and channelopathies. Fifty......-two cases had non-diagnostic structural cardiac abnormalities and 20 cases, diagnosed with a cardiomyopathy post-mortem (ARVC = 14, HCM = 6), served as comparators. Fifteen (29 %) of the deceased individuals with non-diagnostic findings had variants with likely functional effects based on conservation......, computational prediction, allele-frequency and supportive literature. The corresponding frequency in deceased individuals with cardiomyopathies was 35 % (p = 0.8). CONCLUSION: The broad genetic screening revealed variants with likely functional effects at similar high rates, i.e. in 29 and 35 % of the suspected...

  18. ITER technical advisory committee meeting at Garching

    International Nuclear Information System (INIS)

    Fujiwara, M.

    1999-01-01

    The ITER Technical Advisory Committee meeting took place on 24-27 February at the Garching Joint Work Site. According to the discussions at the ITER meeting in Yokohama in October 1998, the Technical Advisory Committee was requested to conduct a thorough review of the document 'Options for the reduced technical objectives / reduced cost ITER'

  19. ITER EDA newsletter. V. 9, no. 9

    International Nuclear Information System (INIS)

    2000-09-01

    This ITER EDA Newsletter contains the following 5 contributions: CSMC and CSIC charging tests successfully completed; The ITER divertor cassette project meeting; Blanket R and D and design task meeting; IAEA technical committee meeting on fusion safety; ITER L-6 large project ''blanket remote handling and maintenance''

  20. 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

  1. ITER ITA newsletter. No. 21, April 2005

    International Nuclear Information System (INIS)

    2005-05-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about Russian federation Participant Team's activity in the area of preparation for ITER construction and information about International Fusion materials irradiation Facility(IRMIF) project and prospects for implementation

  2. ITER EDA newsletter. V. 7, no. 12

    International Nuclear Information System (INIS)

    1998-12-01

    This edition of the ITER EDA Newsletter is dedicated to celebrate the achievements of the ITER activities at the San Diego Joint Work Site. Articles by E. Velikhov, A. Davies and R. Aymar mark the final days of American participation in the ITER program

  3. ITER EDA newsletter. V. 7, no. 10

    International Nuclear Information System (INIS)

    1998-10-01

    This newsletter contains three articles, namely a report on an ITER meeting (October 20-21,1998) in Yokohama, Japan, a short note on the 17th IAEA Fusion Energy Conference (October 19-24, 1998) in Yokohama and a monograph by ITER Director R. Aymar on 'the Legacy of Artsimovitch and the lessons of ITER'

  4. ITER EDA newsletter. V. 6, no. 2

    International Nuclear Information System (INIS)

    1997-02-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter reports on the ITER divertor development project and its objectives; contains a report on the 16th Energy IAEA Fusion Conference (ITER and other Tokamak Issues) held in Montreal, Canada; 287 papers were selected by the Programme Committee for presentation and 178 posters were presented. 3 figs

  5. ITER EDA newsletter. V. 10, no. 3

    International Nuclear Information System (INIS)

    2001-03-01

    This issue contains a report on the meeting of the ITER Council (M. Drew), a report on the ITER EDA status (Dr. R. Aymar), a report on the ITER Council tour of the Clarington Site (Dr. D. Dautovich) . Abstracts of the indivdual reports have been included in the database

  6. ITER EDA newsletter. V. 7, No. 3

    International Nuclear Information System (INIS)

    1998-03-01

    This issue of the ITER Newsletter contains an article of the Status of the ITER EDA and the progress of the ITER activities and a report on the 5th Technical Meeting on Quality which was held in San Diego on 20-22 October 1997

  7. 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

  8. ITER EDA Newsletter. V. 2, no. 1

    International Nuclear Information System (INIS)

    1993-01-01

    This ITER EDA (Engineering Design Activities) Newsletter issue is dedicated to the description of the ITER EDA Home Teams (European Community, Japan, Russian Federation, USA), in particular their composition, tasks, responsibilities, national support and activities, aimed to design the ITER tokamak

  9. ITER ITA newsletter. No. 2, March 2003

    International Nuclear Information System (INIS)

    2003-05-01

    This ITER ITA newsletter contains concise information about the JASS (Joint Assessment of Specific Sites) final report. The main conclusion is that all four sites are sound and fully capable to respond to all ITER Site Requirements and Design assumptions, as approved by the ITER Council in its January 2000 Meeting

  10. ITER EDA newsletter. V. 2, no. 12

    International Nuclear Information System (INIS)

    1993-12-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a report of the Second ITER Technical Committee Meeting on Safety, Environment, and Regulatory Approval, San Diego, USA, November 3-12, 1993, and a summary report on an ITER Magnet Technical Meeting, Naka, Japan, October 5-8, 1993

  11. 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

  12. ITER-FEAT outline design report

    International Nuclear Information System (INIS)

    2001-01-01

    In July 1998 the ITER Parties were unable, for financial reasons, to proceed with construction of the ITER design proposed at that time, to meet the detailed technical objectives and target cost set in 1992. It was therefore decided to investigate options for the design of ITER with reduced technical objectives and with possibly decreased technical margins, whose target construction cost was one half that of the 1998 ITER design, while maintaining the overall programmatic objective. To identify designs that might meet the revised objectives, task forces involving the JCT and Home Teams met during 1998 and 1999 to analyse and compare a range of options for the design of such a device. This led at the end of 1999 to a single configuration for the ITER design with parameters considered to be the most credible consistent with technical limitations and the financial target, yet meeting fully the objectives with appropriate margins. This new design of ITER, called ''ITER-FEAT'', was submitted to the ITER Director to the ITER Parties as the ''ITER-FEAT Outline Design Report'' (ODR) in January 2000, at their meeting in Tokyo. The Parties subsequently conducted their domestic assessments of this report and fed the resulting comments back into the progressing design. The progress on the developing design was reported to the ITER Technical Advisory Committee (TAC) in June 2000 in the report ''Progress in Resolving Open Design Issues from the ODR'' alongside a report on Progress in Technology R and D for ITER. In addition, the progress in the ITER-FEAT Design and Validating R and D was reported to the ITER Parties. The ITER-FEAT design was subsequently approved by the governing body of ITER in Moscow in June 2000 as the basis for the preparation of the Final Design Report, recognising it as a single mature design for ITER consistent with its revised objectives. This volume contains the documents pertinent to the process described above. More detailed technical information

  13. 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

  14. The ITER poloidal field system

    Energy Technology Data Exchange (ETDEWEB)

    Wesley, J [General Atomics, San Diego, CA (USA); Beljakov, V; Kavin, A; Korshakov, V; Kostenko, A; Roshal, A; Zakharov, L [Kurchatov Inst. of Atomic Energy, Moscow (USSR); Bulmer, R; Kaiser, T; Miller, J R; Pearlstein, L D [Lawrence Livermore National Lab., CA (USA); Hogan, J [Oak Ridge National Lab., TN (USA); Kurihara, K; Shimomura, Y; Sugihara, M; Yoshino, R [Japan Atomic Energy Resea

    1990-12-15

    The ITER poloidal field (PF) system uses superconducting coils to provide the plasma equilibrium fields, slow equilibrium control and plasma flux linkage (V-s) needed for the ITER Operations and Research Program. Double-null (DN) divertor plasmas and operation scenarios for 22 MA Physics (high-Q/ignition) and 15 MA Technology (high-fluence testing) phases are provided. For 22 MA plasmas, total PF flux swing is 333 V-s. This provides inductive current drive (CD) for start-up with 66 V-s of resistive loss and 440-s (330-s minimum) sustained burn. The PF system also allows plasma start-up and shutdown scenarios, and can maintain the plasma configuration during burn over a range of current and pressure profiles. Other capabilities include increased plasma current (25 MA with inductive CD; 28 MA with non-inductive CD assist), divertor separatrix sweeping, and semi-DN and single-null plasmas.

  15. Iterative analysis of concrete gravity dam-nonlinear foundation ...

    African Journals Online (AJOL)

    The solution of the coupled system is accomplished by solving the two systems separately and then considering the interaction effects at the soil–structure interface enforced by a developed iterative scheme. Emphasis has been laid on the study of material nonlinearity of the foundation material in the interaction analysis.

  16. ITER central solenoid model coil heat treatment complete and assembly started

    International Nuclear Information System (INIS)

    Thome, R.J.; Okuno, K.

    1998-01-01

    A major R and D task in the ITER program is to fabricate a Superconducting Model Coil for the Central Solenoid to establish the design and fabrication methods for ITER size coils and to demonstrate conductor performance. Completion of its components is expected in 1998, to be followed by assembly with structural components and testing in a facility at JAERI

  17. Progress in LHCD: a tool for advanced regimes on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Tuccillo, A. A.

    2005-07-01

    Since the early 80s, Lower Hybrid (LH) waves have driven plasma current non-inductively in tokamak experiments with efficiency far higher than other auxiliary systems, especially at low plasma temperature. The latter aspect makes LH the natural candidate for off axis current drive (CD) in ITER where current profile control will be required to maintain burning performance on a long time scale. The difficulties in coupling LH waves in high performance H-mode plasmas led in the 90s to the decision, by the ITER JCT, of not supporting directly the development of a LH system, with subsequent exclusion of LH from ITER initial Hn and CD systems. Nevertheless, the LH activity continued in all areas: theory, experiments, modelling and R and D. Recently, in JET, the problem of coupling LH waves in severe edge conditions (ELMy plasmas and very low SOL densities) has been solved by localised gas injection. LH has then boosted the research on advanced scenarios allowing a variety of current profiles, from peaked to deeply hollow, to be used in generating Internal Transport Barriers (ITBs). In JET, long ITBs (11s) have been sustained for times close to the resistive time scale in full CD conditions, additionally pulses lasting minutes, relying on LHCD, are routinely obtained in Tore Supra. Advanced scenarios have also been obtained in conditions of dominant electron heating (FTU, Tore Supra) and efficient CD has been demonstrated at ITER relevant densities on FTU. In these experiments, an improvement of the Electron Cyclotron (EC) CD efficiency (up to 4?) due to synergy with LH is also observed both in normal operation and with EC resonating at a down shifted frequency. Modelling all these experiments has been a good test bench for LH codes and has increased the confidence in their use to predict future experiments. The use of LH as an actuator for Real Time Control of the plasma current profile has become a powerful tool to optimise ITB dynamics in JET and JT-60U. An LH system

  18. 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

  19. Truncated States Obtained by Iteration

    International Nuclear Information System (INIS)

    Cardoso, W. B.; Almeida, N. G. de

    2008-01-01

    We introduce the concept of truncated states obtained via iterative processes (TSI) and study its statistical features, making an analogy with dynamical systems theory (DST). As a specific example, we have studied TSI for the doubling and the logistic functions, which are standard functions in studying chaos. TSI for both the doubling and logistic functions exhibit certain similar patterns when their statistical features are compared from the point of view of DST

  20. Infrared laser diagnostics for ITER

    International Nuclear Information System (INIS)

    Hutchinson, D.P.; Richards, R.K.; Ma, C.H.

    1995-01-01

    Two infrared laser-based diagnostics are under development at ORNL for measurements on burning plasmas such as ITER. The primary effort is the development of a CO 2 laser Thomson scattering diagnostic for the measurement of the velocity distribution of confined fusion-product alpha particles. Key components of the system include a high-power, single-mode CO 2 pulsed laser, an efficient optics system for beam transport and a multichannel low-noise infrared heterodyne receiver. A successful proof-of-principle experiment has been performed on the Advanced Toroidal Facility (ATF) stellerator at ORNL utilizing scattering from electron plasma frequency satellites. The diagnostic system is currently being installed on Alcator C-Mod at MIT for measurements of the fast ion tail produced by ICRH heating. A second diagnostic under development at ORNL is an infrared polarimeter for Faraday rotation measurements in future fusion experiments. A preliminary feasibility study of a CO 2 laser tangential viewing polarimeter for measuring electron density profiles in ITER has been completed. For ITER plasma parameters and a polarimeter wavelength of 10.6 microm, a Faraday rotation of up to 26 degree is predicted. An electro-optic polarization modulation technique has been developed at ORNL. Laboratory tests of this polarimeter demonstrated a sensitivity of ≤ 0.01 degree. Because of the similarity in the expected Faraday rotation in ITER and Alcator C-Mod, a collaboration between ORNL and the MIT Plasma Fusion Center has been undertaken to test this polarimeter system on Alcator C-Mod. A 10.6 microm polarimeter for this measurement has been constructed and integrated into the existing C-Mod multichannel two-color interferometer. With present experimental parameters for C-Mod, the predicted Faraday rotation was on the order of 0.1 degree. Significant output signals were observed during preliminary tests. Further experiment and detailed analyses are under way