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Sample records for iterated functions systems

  1. Image based rendering of iterated function systems

    NARCIS (Netherlands)

    Wijk, van J.J.; Saupe, D.

    2004-01-01

    A fast method to generate fractal imagery is presented. Iterated function systems (IFS) are based on repeatedly copying transformed images. We show that this can be directly translated into standard graphics operations: Each image is generated by texture mapping and blending copies of the previous

  2. Iterated function systems for DNA replication

    Science.gov (United States)

    Gaspard, Pierre

    2017-10-01

    The kinetic equations of DNA replication are shown to be exactly solved in terms of iterated function systems, running along the template sequence and giving the statistical properties of the copy sequences, as well as the kinetic and thermodynamic properties of the replication process. With this method, different effects due to sequence heterogeneity can be studied, in particular, a transition between linear and sublinear growths in time of the copies, and a transition between continuous and fractal distributions of the local velocities of the DNA polymerase along the template. The method is applied to the human mitochondrial DNA polymerase γ without and with exonuclease proofreading.

  3. A note on iterated function systems with discontinuous probabilities

    International Nuclear Information System (INIS)

    Jaroszewska, Joanna

    2013-01-01

    Highlights: ► Certain iterated function system with discontinuous probabilities is discussed. ► Existence of an invariant measure via the Schauder–Tychonov theorem is established. ► Asymptotic stability of the system under examination is proved. -- Abstract: We consider an example of an iterated function system with discontinuous probabilities. We prove that it posses an invariant probability measure. We also prove that it is asymptotically stable provided probabilities are positive

  4. Normalization of the collage regions of iterated function systems

    Science.gov (United States)

    Zhang, Zhengbing; Zhang, Wei

    2012-11-01

    Fractal graphics, generated with iterated function systems (IFS), have been applied in broad areas. Since the collage regions of different IFS may be different, it is difficult to respectively show the attractors of iterated function systems in a same region on a computer screen using one program without modifying the display parameters. An algorithm is proposed in this paper to solve this problem. A set of transforms are repeatedly applied to modify the coefficients of the IFS so that the collage region of the resulted IFS changes toward the unit square. Experimental results demonstrate that the collage region of any IFS can be normalized to the unit square with the proposed method.

  5. Approximate convex hull of affine iterated function system attractors

    International Nuclear Information System (INIS)

    Mishkinis, Anton; Gentil, Christian; Lanquetin, Sandrine; Sokolov, Dmitry

    2012-01-01

    Highlights: ► We present an iterative algorithm to approximate affine IFS attractor convex hull. ► Elimination of the interior points significantly reduces the complexity. ► To optimize calculations, we merge the convex hull images at each iteration. ► Approximation by ellipses increases speed of convergence to the exact convex hull. ► We present a method of the output convex hull simplification. - Abstract: In this paper, we present an algorithm to construct an approximate convex hull of the attractors of an affine iterated function system (IFS). We construct a sequence of convex hull approximations for any required precision using the self-similarity property of the attractor in order to optimize calculations. Due to the affine properties of IFS transformations, the number of points considered in the construction is reduced. The time complexity of our algorithm is a linear function of the number of iterations and the number of points in the output approximate convex hull. The number of iterations and the execution time increases logarithmically with increasing accuracy. In addition, we introduce a method to simplify the approximate convex hull without loss of accuracy.

  6. ON RANDOM ITERATED FUNCTION SYSTEMS WITH GREYSCALE MAPS

    Directory of Open Access Journals (Sweden)

    Matthew Demers

    2012-05-01

    Full Text Available In the theory of Iterated Function Systems (IFSs it is known that one can find an IFS with greyscale maps (IFSM to approximate any target signal or image with arbitrary precision, and a systematic approach for doing so was described. In this paper, we extend these ideas to the framework of random IFSM operators. We consider the situation where one has many noisy observations of a particular target signal and show that the greyscale map parameters for each individual observation inherit the noise distribution of the observation. We provide illustrative examples.

  7. 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

  8. Iterative scheme for electronic systems: using one-electron Green's functions

    International Nuclear Information System (INIS)

    Hyslop, J.; Rees, D.

    1976-01-01

    An iterative generalization of the minimum principle proposed for electronic systems by Hall, Hyslop, and Rees is investigated. It is shown that this generalization still retains the advantage of using members of a larger class of trial wave functions, for example those with discontinuities, as initial approximations to the wave functions. This scheme has the advantage that, at each stage of iteration, an upper bound is obtained which is at least as good as that obtained previously. The theory is first applied to the hydrogen atom. It is then adapted to estimate the Hartree--Fock energy of the helium atom, the Hartree--Fock limit being obtained after a relatively small number of iterations

  9. On extension of solutions of a simultaneous system of iterative functional equations

    Directory of Open Access Journals (Sweden)

    Janusz Matkowski

    2009-01-01

    Full Text Available Some sufficient conditions which allow to extend every local solution of a simultaneous system of equations in a single variable of the form \\[ \\varphi(x = h (x, \\varphi[f_1(x],\\ldots,\\varphi[f_m(x],\\] \\[\\varphi(x = H (x, \\varphi[F_1(x],\\ldots,\\varphi[F_m(x],\\] to a global one are presented. Extensions of solutions of functional equations, both in single and in several variables, play important role (cf. for instance [M. Kuczma, Functional equations in a single variable, Monografie Mat. 46, Polish Scientific Publishers, Warsaw, 1968, M. Kuczma, B. Choczewski, R. Ger, Iterative functional equations, Encyclopedia of Mathematics and Its Applications v. 32, Cambridge, 1990, J. Matkowski, Iteration groups, commuting functions and simultaneous systems of linear functional equations, Opuscula Math. 28 (2008 4, 531-541].

  10. 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

  11. Adaptive iterated function systems filter for images highly corrupted with fixed - Value impulse noise

    Science.gov (United States)

    Shanmugavadivu, P.; Eliahim Jeevaraj, P. S.

    2014-06-01

    The Adaptive Iterated Functions Systems (AIFS) Filter presented in this paper has an outstanding potential to attenuate the fixed-value impulse noise in images. This filter has two distinct phases namely noise detection and noise correction which uses Measure of Statistics and Iterated Function Systems (IFS) respectively. The performance of AIFS filter is assessed by three metrics namely, Peak Signal-to-Noise Ratio (PSNR), Mean Structural Similarity Index Matrix (MSSIM) and Human Visual Perception (HVP). The quantitative measures PSNR and MSSIM endorse the merit of this filter in terms of degree of noise suppression and details/edge preservation respectively, in comparison with the high performing filters reported in the recent literature. The qualitative measure HVP confirms the noise suppression ability of the devised filter. This computationally simple noise filter broadly finds application wherein the images are highly degraded by fixed-value impulse noise.

  12. Efficient fractal-based mutation in evolutionary algorithms from iterated function systems

    Science.gov (United States)

    Salcedo-Sanz, S.; Aybar-Ruíz, A.; Camacho-Gómez, C.; Pereira, E.

    2018-03-01

    In this paper we present a new mutation procedure for Evolutionary Programming (EP) approaches, based on Iterated Function Systems (IFSs). The new mutation procedure proposed consists of considering a set of IFS which are able to generate fractal structures in a two-dimensional phase space, and use them to modify a current individual of the EP algorithm, instead of using random numbers from different probability density functions. We test this new proposal in a set of benchmark functions for continuous optimization problems. In this case, we compare the proposed mutation against classical Evolutionary Programming approaches, with mutations based on Gaussian, Cauchy and chaotic maps. We also include a discussion on the IFS-based mutation in a real application of Tuned Mass Dumper (TMD) location and optimization for vibration cancellation in buildings. In both practical cases, the proposed EP with the IFS-based mutation obtained extremely competitive results compared to alternative classical mutation operators.

  13. 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

  14. 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)

  15. 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

  16. Rigorous approximation of stationary measures and convergence to equilibrium for iterated function systems

    International Nuclear Information System (INIS)

    Galatolo, Stefano; Monge, Maurizio; Nisoli, Isaia

    2016-01-01

    We study the problem of the rigorous computation of the stationary measure and of the rate of convergence to equilibrium of an iterated function system described by a stochastic mixture of two or more dynamical systems that are either all uniformly expanding on the interval, either all contracting. In the expanding case, the associated transfer operators satisfy a Lasota–Yorke inequality, we show how to compute a rigorous approximations of the stationary measure in the L "1 norm and an estimate for the rate of convergence. The rigorous computation requires a computer-aided proof of the contraction of the transfer operators for the maps, and we show that this property propagates to the transfer operators of the IFS. In the contracting case we perform a rigorous approximation of the stationary measure in the Wasserstein–Kantorovich distance and rate of convergence, using the same functional analytic approach. We show that a finite computation can produce a realistic computation of all contraction rates for the whole parameter space. We conclude with a description of the implementation and numerical experiments. (paper)

  17. 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

  18. Entropy of Iterated Function Systems and Their Relations with Black Holes and Bohr-Like Black Holes Entropies

    Directory of Open Access Journals (Sweden)

    Christian Corda

    2018-01-01

    Full Text Available In this paper we consider the metric entropies of the maps of an iterated function system deduced from a black hole which are known the Bekenstein–Hawking entropies and its subleading corrections. More precisely, we consider the recent model of a Bohr-like black hole that has been recently analysed in some papers in the literature, obtaining the intriguing result that the metric entropies of a black hole are created by the metric entropies of the functions, created by the black hole principal quantum numbers, i.e., by the black hole quantum levels. We present a new type of topological entropy for general iterated function systems based on a new kind of the inverse of covers. Then the notion of metric entropy for an Iterated Function System ( I F S is considered, and we prove that these definitions for topological entropy of IFS’s are equivalent. It is shown that this kind of topological entropy keeps some properties which are hold by the classic definition of topological entropy for a continuous map. We also consider average entropy as another type of topological entropy for an I F S which is based on the topological entropies of its elements and it is also an invariant object under topological conjugacy. The relation between Axiom A and the average entropy is investigated.

  19. 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

  20. 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.

  1. 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

  2. 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

  3. ITMETH, Iterative Routines for Linear System

    International Nuclear Information System (INIS)

    Greenbaum, A.

    1989-01-01

    1 - Description of program or function: ITMETH is a collection of iterative routines for solving large, sparse linear systems. 2 - Method of solution: ITMETH solves general linear systems of the form AX=B using a variety of methods: Jacobi iteration; Gauss-Seidel iteration; incomplete LU decomposition or matrix splitting with iterative refinement; diagonal scaling, matrix splitting, or incomplete LU decomposition with the conjugate gradient method for the problem AA'Y=B, X=A'Y; bi-conjugate gradient method with diagonal scaling, matrix splitting, or incomplete LU decomposition; and ortho-min method with diagonal scaling, matrix splitting, or incomplete LU decomposition. ITMETH also solves symmetric positive definite linear systems AX=B using the conjugate gradient method with diagonal scaling or matrix splitting, or the incomplete Cholesky conjugate gradient method

  4. Iterative photoinduced chain functionalization as a generic platform for advanced polymeric drug delivery systems

    Czech Academy of Sciences Publication Activity Database

    Al Samad, A.; Bethry, A.; Janoušková, Olga; Ciccione, J.; Wenk, C.; Coll, J.-L.; Subra, G.; Etrych, Tomáš; El Omar, F.; Bakkour, Y.; Coudane, J.; Nottelet, B.

    2018-01-01

    Roč. 39, č. 3 (2018), s. 1-5, č. článku 1700502. ISSN 1022-1336 R&D Projects: GA MŠk(CZ) LO1507; GA MŠk(CZ) LQ1604 Institutional support: RVO:61389013 Keywords : drug delivery systems * functionalization of polymers * photochemistry Subject RIV: CD - Macromolecular Chemistry OBOR OECD: Polymer science Impact factor: 4.265, year: 2016

  5. 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.)

  6. 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)

  7. ITER isotope separation system

    International Nuclear Information System (INIS)

    Busigin, A.; Sood, S.K.; Kveton, O.K.; Sherman, R.H.; Anderson, J.L.

    1990-09-01

    This document presents the results of a study that examined the technical operating and economic viability of an alternative Isotope Separation System (ISS) design based on the distributed design concept. In the distributed design, the ISS is broken up into local independently operable subsystems matched to local processing requirements. The distributed design accepts the same feeds and produces essentially the same products as the reference design. The distributed design consists of two separate, independent subsystems. The first, called ISS-H, receives only protium-dominated streams and waste water from tritium extraction. It has two cryogenic distillation columns and can produce a 50 percent D, 50 percent T product since it lacks D/T separation capability. A final 80 percent T 2 concentration product can be obtained by blending the 50 percent T 2 stream from ISS-H with the more than 99 percent T 2 stream from the second subsystem, ISS-D. The second subsystem receives only deuterium-dominated feeds, which also contain some protium. ISS-D is as complex as the reference design, but smaller. Although each subsystem has some advantages, such as only two cryogenic distillation columns in ISS-H and better than 99 percent steady state T 2 product in ISS-D, the combined subsystems do not offer any real advantage compared to the reference IISS. The entire distributed ISS design has been simulated using Ontario Hydro's FLOSHEET steady state process simulator. Dynamic analysis has not been done for the distributed design. (10 refs., 3 figs., 8 tabs.)

  8. Fractals via iterated functions and multifunctions

    International Nuclear Information System (INIS)

    Singh, S.L.; Prasad, Bhagwati; Kumar, Ashish

    2009-01-01

    Fractals have wide applications in biology, computer graphics, quantum physics and several other areas of applied sciences (see, for instance [Daya Sagar BS, Rangarajan Govindan, Veneziano Daniele. Preface - fractals in geophysics. Chaos, Solitons and Fractals 2004;19:237-39; El Naschie MS. Young double-split experiment Heisenberg uncertainty principles and cantorian space-time. Chaos, Solitons and Fractals 1994;4(3):403-09; El Naschie MS. Quantum measurement, information, diffusion and cantorian geodesics. In: El Naschie MS, Rossler OE, Prigogine I, editors. Quantum mechanics, diffusion and Chaotic fractals. Oxford: Elsevier Science Ltd; 1995. p. 191-205; El Naschie MS. Iterated function systems, information and the two-slit experiment of quantum mechanics. In: El Naschie MS, Rossler OE, Prigogine I, editors. Quantum mechanics, diffusion and Chaotic fractals. Oxford: Elsevier Science Ltd; 1995. p. 185-9; El Naschie MS, Rossler OE, Prigogine I. Forward. In: El Naschie MS, Rossler OE, Prigogine I, editors. Quantum mechanics, diffusion and Chaotic fractals. Oxford: Elsevier Science Ltd; 1995; El Naschie MS. A review of E-infinity theory and the mass spectrum of high energy particle physics. Chaos, Solitons and Fractals 2004;19:209-36; El Naschie MS. Fractal black holes and information. Chaos, Solitons and Fractals 2006;29:23-35; El Naschie MS. Superstring theory: what it cannot do but E-infinity could. Chaos, Solitons and Fractals 2006;29:65-8). Especially, the study of iterated functions has been found very useful in the theory of black holes, two-slit experiment in quantum mechanics (cf. El Naschie, as mentioned above). The intent of this paper is to give a brief account of recent developments of fractals arising from IFS. We also discuss iterated multifunctions.

  9. 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.

  10. 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.

  11. ITER fuel cycle systems layout

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-10-01

    The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

  12. Proposals for Iterated Hash Functions

    DEFF Research Database (Denmark)

    Knudsen, Lars Ramkilde; Thomsen, Søren Steffen

    2008-01-01

    The past few years have seen an increase in the number of attacks on cryptographic hash functions. These include attacks directed at specific hash functions, and generic attacks on the typical method of constructing hash functions. In this paper we discuss possible methods for protecting against...... some generic attacks. We also give a concrete proposal for a new hash function construction, given a secure compression function which, unlike in typical existing constructions, is not required to be resistant to all types of collisions. Finally, we show how members of the SHA-family can be turned...

  13. Proposals for iterated hash functions

    DEFF Research Database (Denmark)

    Knudsen, Lars Ramkilde; Thomsen, Søren Steffen

    2006-01-01

    The past few years have seen an increase in the number of attacks on cryptographic hash functions. These include attacks directed at specific hash functions, and generic attacks on the typical method of constructing hash functions. In this paper we discuss possible methods for protecting against...... some generic attacks. We also give a concrete proposal for a new hash function construction, given a secure compression function which, unlike in typical existing constructions, is not required to be resistant to all types of collisions. Finally, we show how members of the SHA-family can be turned...

  14. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-04-01

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

  15. ETR/ITER systems code

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  16. Advanced fuelling system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Raman, Roger [University of Washington, Seattle, WA (United States)], E-mail: raman@aa.washington.edu

    2008-12-15

    Steady-state high-performance discharges in reactors, such as the Advanced Tokamak (AT) scenarios would rely on optimized density and pressure profiles that must be maintained. This maximizes the bootstrap current fraction, reduces reactor recycling power and reduces thermal stresses. Other than a system for the balance of current drive not provided by bootstrap current drive, no other sources of input power, such as from neutral beams, are allowed. For these systems, a precision fuelling system would be the ideal way to control the fusion burn by controlling and maintaining the required pressure profile. This requires a fuelling system that is capable of depositing fuel at any radial location within the plasma while at the same time not altering the density profile to a level that degrades the required pressure profile. Present fuelling systems are incapable of meeting these requirements. An advanced fuelling system based on Compact Toroid injection has the potential to meet these needs while simultaneously providing a source of toroidal momentum input. Description of a conceptual Compact Toroid fueller for ITER is presented in conjunction with a plan for developing this much needed technology.

  17. Modified ITER In-Vessel Viewing System

    International Nuclear Information System (INIS)

    Ahola, H.; Heikkinen, V.; Keraenen, K.; Suomela, J.

    2001-01-01

    The original ITER In-Vessel Viewing System (IVVS) prototype (Proc. of the 20th SOFT, vol. 2 (1998) 1051), which demonstrates the feasibility of linear fibre arrays for ITER in-vessel viewing, has been modified. In order to reduce the viewing time and to improve the image quality the beam dispersing mirrors was replaced by a diffractive optics element (DOE), which enhanced the laser illumination considerably. The performance of the system was tested using various target surfaces: the results obtained clearly indicate its adequacy for in-vessel viewing. Mechanical damage on smooth metal surfaces (scratches etc.) can be easily distinguished and the viewing resolution at a distance of 2 m is better than 1 mm. The IVVS has been re-designed to be compatible with the new ITER-FEAT. A conceptual study which covers all the functions and subsystems required for viewing has been completed. These results will be used to further modify the prototype: items to be tested include horizontal probe operation and laser illumination with an optical fibre

  18. 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.

  19. ITER ISS system alternative specification study

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-08-01

    Recent comments suggested that the fuel systems, in particular the ISS, could be simplified if the ITER specifications were relaxed from the data specified for ITER. This interim report addresses the first part of the analysis, which considers the impact of design specifications on fuel systems design

  20. RAMI analysis of the ITER Central Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [ITER Project Unit, Japan Atomic Energy Agency (JAEA), Naka, 311-0193 Ibaraki (Japan); Okayama, Katsumi [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neyatani, Yuzuru [ITER Project Unit, Japan Atomic Energy Agency (JAEA), Naka, 311-0193 Ibaraki (Japan); Sagot, Francois; Houtte, Didier van [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-06-15

    Highlights: • We performed the functional analysis of the ITER CSS. • We performed a failure mode analysis of the ITER CSS. • We estimated the reliability and availability of the ITER CSS. • The ITER RAMI approach was applied to the ITER CSS for technical risk control in the design phase. - Abstract: ITER is the first worldwide international project aiming to design a facility to produce nuclear fusion energy. The technical requirements of its plant systems have been established in the ITER Project Baseline. In the project, the Reliability, Availability, Maintainability and Inspectability (RAMI) approach has been adopted for technical risk control to help aid the design of the components in preparation for operation and maintenance. A RAMI analysis was performed on the conceptual design of the ITER Central Safety System (CSS). A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 2 main functions and 20 sub-functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CSS expected after implementation of mitigation actions was calculated to be 99.80% over 2 years, which is the typical interval of the scheduled maintenance cycles. This is consistent with the project required value of 99.9 ± 0.1%. A Failure Modes, Effects and Criticality Analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability of the plasma operation. This analysis defined when risk mitigation actions were required in terms of design, testing, operation procedures and/or maintenance to reduce the risk levels and increase the availability of the

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

    Science.gov (United States)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  3. 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.

  4. The remote handling systems for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Isabel, E-mail: mir@isr.ist.utl.pt [Institute for Systems and Robotics/Instituto Superior Tecnico, Lisboa (Portugal); Damiani, Carlo [Fusion for Energy, Barcelona (Spain); Tesini, Alessandro [ITER Organization, Cadarache (France); Kakudate, Satoshi [ITER Tokamak Device Group, Japan Atomic Energy Agency, Ibaraki (Japan); Siuko, Mikko [VTT Systems Engineering, Tampere (Finland); Neri, Carlo [Associazione EURATOM ENEA, Frascati (Italy)

    2011-10-15

    The ITER remote handling (RH) maintenance system is a key component in ITER operation both for scheduled maintenance and for unexpected situations. It is a complex collection and integration of numerous systems, each one at its turn being the integration of diverse technologies into a coherent, space constrained, nuclearised design. This paper presents an integrated view and recent results related to the Blanket RH System, the Divertor RH System, the Transfer Cask System (TCS), the In-Vessel Viewing System, the Neutral Beam Cell RH System, the Hot Cell RH and the Multi-Purpose Deployment System.

  5. 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.

  6. ITER Remote Maintenance System (IRMS) lifecycle management

    Energy Technology Data Exchange (ETDEWEB)

    Tesini, Alessandro, E-mail: alessandro.tesini@iter.org [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Otto' , Bede [Oxford Technologies Ltd, 7, Nuffield Way, Abingdon, Oxon OX14 1RJ (United Kingdom); Blight, John [FAAST 31c Allee de la Granette, 13600 Ceyreste (France); Choi, Chang-Hwan; Friconneau, Jean-Pierre; Gotewal, Krishan Kumar; Hamilton, David [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Heckendorn, Frank [FD Technologies, PO Box 6686, Aiken, SC (United States); Martins, Jean-Pierre [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Marty, Thomas [Westinghouse, 122, avenue de Hambourg, 13008 Marseille (France); Nakahira, Masataka; Palmer, Jim; Subramanian, Rajendran [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2011-10-15

    The availability of the ITER machine to perform its scientific program is strongly dependent on the performance of the different Remote Handling (RH) systems constituting the ITER Remote Maintenance System (IRMS). The lifecycle of the IRMS will largely exceed 40 years from initial concept design and proof testing through to machine decommissioning. Such a long lifecycle requires that a rigorous approach is put in place to guarantee the technical capabilities of the highly innovative IRMS, its efficiency and its availability. For this purpose, an IRMS System Engineering and IRMS lifecycle management approach has been adopted by ITER. The approach aims at ensuring the IRMS full operability and availability at an acceptable cost of ownership over the full ITER machine assembly and operations period. The IRMS lifecycle management method described in this paper covers such subjects as specific requirements for IRMS design reviews, monitoring during manufacture, factory and site acceptance testing, integrated commissioning, decontamination, maintenance and re-qualification strategies, requirements for Integrated Logistical Support during operations. The updating and implementation of the IRMS lifecycle strategy and this procedure will be managed and monitored by the Remote Handling Integrated Product Team (RH-IPT). Although developed for the IRMS, the basic principles and procedures of lifecycle management could be applied to other ITER plant systems whose reliability and availability will be essential for the continued operation of the ITER machine.

  7. ITER Remote Maintenance System (IRMS) lifecycle management

    International Nuclear Information System (INIS)

    Tesini, Alessandro; Otto', Bede; Blight, John; Choi, Chang-Hwan; Friconneau, Jean-Pierre; Gotewal, Krishan Kumar; Hamilton, David; Heckendorn, Frank; Martins, Jean-Pierre; Marty, Thomas; Nakahira, Masataka; Palmer, Jim; Subramanian, Rajendran

    2011-01-01

    The availability of the ITER machine to perform its scientific program is strongly dependent on the performance of the different Remote Handling (RH) systems constituting the ITER Remote Maintenance System (IRMS). The lifecycle of the IRMS will largely exceed 40 years from initial concept design and proof testing through to machine decommissioning. Such a long lifecycle requires that a rigorous approach is put in place to guarantee the technical capabilities of the highly innovative IRMS, its efficiency and its availability. For this purpose, an IRMS System Engineering and IRMS lifecycle management approach has been adopted by ITER. The approach aims at ensuring the IRMS full operability and availability at an acceptable cost of ownership over the full ITER machine assembly and operations period. The IRMS lifecycle management method described in this paper covers such subjects as specific requirements for IRMS design reviews, monitoring during manufacture, factory and site acceptance testing, integrated commissioning, decontamination, maintenance and re-qualification strategies, requirements for Integrated Logistical Support during operations. The updating and implementation of the IRMS lifecycle strategy and this procedure will be managed and monitored by the Remote Handling Integrated Product Team (RH-IPT). Although developed for the IRMS, the basic principles and procedures of lifecycle management could be applied to other ITER plant systems whose reliability and availability will be essential for the continued operation of the ITER machine.

  8. 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

  9. RAMI analysis for ITER radial X-ray camera system

    Energy Technology Data Exchange (ETDEWEB)

    Qin, Shijun, E-mail: sjqin@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Liqun; Chen, Kaiyun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Barnsley, Robin; Sirinelli, Antoine [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Song, Yuntao; Lu, Kun; Yao, Damao; Chen, Yebin; Li, Shi; Cao, Hongrui; Yu, Hong; Sheng, Xiuli [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The functional analysis of the ITER RXC system was performed. • A failure modes, effects and criticality analysis of the ITER RXC system was performed. • The reliability and availability of the ITER RXC system and its main functions were calculated. • The ITER RAMI approach was applied to the ITER RXC system for technical risk control in the preliminary design phase. - Abstract: ITER is the first international experimental nuclear fusion device. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to mitigate all the possible failure of components in preparation for operation and maintenance. RAMI analysis of the ITER Radial X-ray Camera diagnostic (RXC) system during preliminary design phase was required, which insures the system with a very high performance to measure the X-ray emission and research the MHD of plasma with high accuracy on the ITER machine. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 6 intermediate functions and 28 basic functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to calculate the reliability and availability of each function under assumption of operating conditions and failure data. Initial and expected scenarios were analyzed to define risk-mitigation actions. The initial availability of RXC system was 92.93%, while after optimization the expected availability was 95.23% over 11,520 h (approx. 16 months) which corresponds to ITER typical operation cycle. A Failure Modes, Effects and Criticality Analysis (FMECA) was performed to the system initial risk. Criticality charts highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. There are 28 risks for the initial state, including 8 major risks. No major risk remains after taking into

  10. Development of pellet injection systems for ITER

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. 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...

  12. Central system of Interlock of ITER, high integrity architecture

    International Nuclear Information System (INIS)

    Prieto, I.; Martinez, G.; Lopez, C.

    2014-01-01

    The CIS (Central Interlock System), along with the CODAC system and CSS (Central Safety System), form the central I and C systems of ITER. The CIS is responsible for implementing the core functions of protection (Central Interlock Functions) through different systems of plant (Plant Systems) within the overall strategy of investment protection for ITER. IBERDROLA supports engineering to define and develop the control architecture of CIS according to the stringent requirements of integrity, availability and response time. For functions with response times of the order of half a second is selected PLC High availability of industrial range. However, due to the nature of the machine itself, certain functions must be able to act under the millisecond, so it has had to develop a solution based on FPGA (Field Programmable Gate Array) capable of meeting the requirements architecture. In this article CIS architecture is described, as well as the process for the development and validation of the selected platforms. (Author)

  13. Iterative solution of large linear systems

    CERN Document Server

    Young, David Matheson

    1971-01-01

    This self-contained treatment offers a systematic development of the theory of iterative methods. Its focal point resides in an analysis of the convergence properties of the successive overrelaxation (SOR) method, as applied to a linear system with a consistently ordered matrix. The text explores the convergence properties of the SOR method and related techniques in terms of the spectral radii of the associated matrices as well as in terms of certain matrix norms. Contents include a review of matrix theory and general properties of iterative methods; SOR method and stationary modified SOR meth

  14. ITER lower port systems integration

    Energy Technology Data Exchange (ETDEWEB)

    Levesy, B., E-mail: bruno.levesy@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, D.; Boussier, B.; Bryan, S.; Cordier, J.J.; Dremel, M.; Dell' Orco, G.; Daly, E.; Doshi, B.; Jeannoutot, T.; Friconneau, J.P.; Gliss, C.; Le Barbier, R.; Lachevre, F.; Loughlin, M.; Martin, A.; Martins, J.P.; Maruyama, S.; Palmer, J.; Reichle, R. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    The lower port systems are installed inside the vacuum vessel lower ports and in the adjacent port cells. The vacuum vessel ports and penetrations are allocated as follow: -4 ports dedicated to remote handling of the divertor cassettes, contain diagnostics racks and divertor cooling pipes. -5 ports connecting the main vessel to the torus cryopumps, contain divertor cooling pipes, pellet and gas injection pipes and vertical stabilization coil feeders. -3 penetrations connecting torus cryopumps are connected to the vacuum vessel by branch pipes. -Specific penetrations for divertor cooling lines, in-vessel viewing and glow discharge systems. The general layout of the port systems has been revised recently to improve the cryopump (8 t weight, 1.8 m diameter and 2.5 m long) maintenance scheme with remote handling tools and integrate the in-vessel vertical stabilization coil feeders. The port allocation, the pumping ports design, and interfaces in-between ports and cryostat and in-between cryopumps and cryostat have been up-dated. The integration inside the 18 port cells (11 m x 4 m each) has been reviewed to avoid clashes in between systems and to fix the openings in the port cell concrete walls. The new layout integrates safety and neutron-shielding requirements as well as remote handling and maintenance compatibility for the different systems. The paper presents the up-dated integration of the lower port systems inside the ports and the port cells. Interfaces of the port systems with the vacuum vessel, the cryostat and the port cells are described.

  15. Availability analysis of the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

    2015-01-01

    The ITER blanket remote handling system (BRHS) is required to replace 440 blanket first wall panels in a two-year maintenance period. To investigate this capability, an availability analysis of the system was carried out. Following the analysis procedure defined by the ITER organization, the availability analysis consists of a functional analysis and a reliability block diagram analysis. In addition, three measures to improve availability were implemented: procurement of spare parts, in-vessel replacement of cameras, and simultaneous replacement of umbilical cables. The availability analysis confirmed those measures improve the availability and capability of the BRHS to replace 440 blanket first wall panels in two years. (author)

  16. 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

  17. EU Developments of the ITER ECRH System

    International Nuclear Information System (INIS)

    Henderson, M.

    2006-01-01

    The electron cyclotron (EC) heating and current drive (H (and) CD) system of ITER will deliver 20 MW/CW in the plasma at 170 GHz for H (and) CD in addition to 2.5 MW/3 s at 120 GHz for plasma start-up. The EC system is composed of power supplies (PS), up to 24 H (and) CD gyrotrons (1 to 2 MW tubes), 3 start-up gyrotrons (1 MW tubes), 24 transmission lines and two sets of launching antennas: equatorial (EL) and upper (UL) launchers. Under the present ITER procurement package the EU is responsible for one third of the H (and) CD 170 GHz gyrotrons, all PSs associated with the H (and) CD system, and the whole set (4) of upper launchers. In all areas of participation, the EU EC partnership (coordinated by the European Fusion Development Association - EFDA) aims toward advancing the technology of each of these subsystems. For example, procurement of Pulse Step Modulator (PSM) HVPS is under consideration, which might have equivalent costs to the present ITER design (thyristor HVPS and HV series switch), but with an increased flexibility in operation and variation in the EC power waveform. The EU is at the forefront in gyrotron research and is developing a 2 MW CW 170 GHz coaxial cavity gyrotron offering an increase in output power while maintaining moderate power densities in the gyrotron cavity and collector. THALES R in collaboration with its EFDA partners (FZK, CRPP, TEKES) is manufacturing a series of prototype tubes in three phases of typically 1 s, 100 s and then CW pulse capacity (∼ 20 10 ). A 2 MW, CW gyrotron test facility is being built at CRPP that will be used to develop the 2 MW coaxial tube, in addition to testing various components required by the EC system. EFDA has undertaken a parallel development of two launcher options: front (FS) and remote (RS) steering, with the aim of providing an optimum launcher for ITER weighing EC physics aspects and operation reliability. The FS launcher (ITER reference design) offers a significant enhancement in physics

  18. 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

  19. Iterative solution of high order compact systems

    Energy Technology Data Exchange (ETDEWEB)

    Spotz, W.F.; Carey, G.F. [Univ. of Texas, Austin, TX (United States)

    1996-12-31

    We have recently developed a class of finite difference methods which provide higher accuracy and greater stability than standard central or upwind difference methods, but still reside on a compact patch of grid cells. In the present study we investigate the performance of several gradient-type iterative methods for solving the associated sparse systems. Both serial and parallel performance studies have been made. Representative examples are taken from elliptic PDE`s for diffusion, convection-diffusion, and viscous flow applications.

  20. Concatenated coding system with iterated sequential inner decoding

    DEFF Research Database (Denmark)

    Jensen, Ole Riis; Paaske, Erik

    1995-01-01

    We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...

  1. Architectural concept for the ITER Plasma Control System

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Humphreys, D., E-mail: humphreys@fusion.gat.com [General Atomics, San Diego, CA (United States); Raupp, G., E-mail: Gerhard.Raupp@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Schuster, E., E-mail: schuster@lehigh.edu [Lehigh University, Bethlehem, PA (United States); Snipes, J., E-mail: Joseph.Snipes@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France); De Tommasi, G., E-mail: detommas@unina.it [CREATE/Università di Napoli Federico II, Napoli (Italy); Walker, M., E-mail: walker@fusion.gat.com [General Atomics, San Diego, CA (United States); Winter, A., E-mail: Axel.Winter@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France)

    2014-05-15

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  2. Architectural concept for the ITER Plasma Control System

    International Nuclear Information System (INIS)

    Treutterer, W.; Humphreys, D.; Raupp, G.; Schuster, E.; Snipes, J.; De Tommasi, G.; Walker, M.; Winter, A.

    2014-01-01

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  3. An integrated architecture for the ITER RH control system

    International Nuclear Information System (INIS)

    Hamilton, David Thomas; Tesini, Alessandro

    2012-01-01

    Highlights: ► Control system architecture integrating ITER remote handling equipment systems. ► Standard control system architecture for remote handling equipment systems. ► Research and development activities to validate control system architecture. ► Standardization studies to select standard parts for control system architecture. - Abstract: The ITER remote handling (RH) system has been divided into 7 major equipment system procurements that deliver complete systems (operator interfaces, equipment controllers, and equipment) according to task oriented functional specifications. Each equipment system itself is an assembly of transporters, power manipulators, telemanipulators, vehicular systems, cameras, and tooling with a need for controllers and operator interfaces. From an operational perspective, the ITER RH systems are bound together by common control rooms, operations team, and maintenance team; and will need to achieve, to a varying degree, synchronization of operations, co-operation on tasks, hand-over of components, and sharing of data and resources. The separately procured RH systems must, therefore, be integrated to form a unified RH system for operation from the RH control rooms. The RH system will contain a heterogeneous mix of specially developed RH systems and off-the-shelf RH equipment and parts. The ITER Organization approach is to define a control system architecture that supports interoperable heterogeneous modules, and to specify a standard set of modules for each system to implement within this architecture. Compatibility with standard parts for selected modules is required to limit the complexity for operations and maintenance. A key requirement for integrating the control system modules is interoperability, and no module should have dependencies on the implementation details of other modules. The RH system is one of the ITER Plant systems that are integrated and coordinated through the hierarchical structure of the ITER CODAC system

  4. The ITER Plasma Control System Simulation Platform

    International Nuclear Information System (INIS)

    Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W.; Welander, A.S.; Winter, A.

    2015-01-01

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  5. The ITER Plasma Control System Simulation Platform

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.L., E-mail: walker@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Ambrosino, G.; De Tommasi, G. [CREATE/Università di Napoli Federico II, Napoli (Italy); Humphreys, D.A. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Napoli (Italy); Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Welander, A.S. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Winter, A. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  6. NITSOL: A Newton iterative solver for nonlinear systems

    Energy Technology Data Exchange (ETDEWEB)

    Pernice, M. [Univ. of Utah, Salt Lake City, UT (United States); Walker, H.F. [Utah State Univ., Logan, UT (United States)

    1996-12-31

    Newton iterative methods, also known as truncated Newton methods, are implementations of Newton`s method in which the linear systems that characterize Newton steps are solved approximately using iterative linear algebra methods. Here, we outline a well-developed Newton iterative algorithm together with a Fortran implementation called NITSOL. The basic algorithm is an inexact Newton method globalized by backtracking, in which each initial trial step is determined by applying an iterative linear solver until an inexact Newton criterion is satisfied. In the implementation, the user can specify inexact Newton criteria in several ways and select an iterative linear solver from among several popular {open_quotes}transpose-free{close_quotes} Krylov subspace methods. Jacobian-vector products used by the Krylov solver can be either evaluated analytically with a user-supplied routine or approximated using finite differences of function values. A flexible interface permits a wide variety of preconditioning strategies and allows the user to define a preconditioner and optionally update it periodically. We give details of these and other features and demonstrate the performance of the implementation on a representative set of test problems.

  7. New proposal on the development of machine protection functions for ITER diagnostics control

    International Nuclear Information System (INIS)

    Yamamoto, Tsuyoshi; Yatsuka, Eiichi; Hatae, Takaki; Takeuchi, Masaki; Kitazawa, Sin-iti; Ogawa, Hiroaki; Kawano, Yasunori; Itami, Kiyoshi; Ota, Kazuya; Hashimoto, Yasunori; Nakamura, Kitaru; Sugie, Tatsuo

    2016-01-01

    There is a need to develop ITER instrumentation and control (I and C) systems with high reliabilities. Interlock systems that activate machine protection functions are implemented on robust wired-logic systems such as programmable logic controllers (PLCs). We herein propose a software tool that generates program code templates for the control systems using PLC logic. This tool decreases careless mistakes by developers and increases reliability of the program codes. A large-scale engineering database has been implemented in the ITER project. To derive useful information from this database, we propose adding semantic data to it using the Resource Description Framework format. In our novel proposal for the ITER diagnostic control system, a guide words generator that analyzes the engineering data by inference is applied to the hazard and operability study. We validated the methods proposed in this paper by applying them to the preliminary design for the I and C system of the ITER edge Thomson scattering system. (author)

  8. Value Iteration Adaptive Dynamic Programming for Optimal Control of Discrete-Time Nonlinear Systems.

    Science.gov (United States)

    Wei, Qinglai; Liu, Derong; Lin, Hanquan

    2016-03-01

    In this paper, a value iteration adaptive dynamic programming (ADP) algorithm is developed to solve infinite horizon undiscounted optimal control problems for discrete-time nonlinear systems. The present value iteration ADP algorithm permits an arbitrary positive semi-definite function to initialize the algorithm. A novel convergence analysis is developed to guarantee that the iterative value function converges to the optimal performance index function. Initialized by different initial functions, it is proven that the iterative value function will be monotonically nonincreasing, monotonically nondecreasing, or nonmonotonic and will converge to the optimum. In this paper, for the first time, the admissibility properties of the iterative control laws are developed for value iteration algorithms. It is emphasized that new termination criteria are established to guarantee the effectiveness of the iterative control laws. Neural networks are used to approximate the iterative value function and compute the iterative control law, respectively, for facilitating the implementation of the iterative ADP algorithm. Finally, two simulation examples are given to illustrate the performance of the present method.

  9. Fatigue behavior of an insulation system for the ITER magnets

    International Nuclear Information System (INIS)

    Prokopec, R.; Humer, K.; Weber, H.W.

    2006-01-01

    The application of glass-fiber reinforced plastics as insulation materials for fusion magnet coils (e.g. the Toroidal Field Coils of ITER) requires the full characterization of their mechanical performance under ITER-relevant conditions. One of the methods of testing material's response under dynamic load is the tension-tension fatigue procedure. This test can be used to simulate the pulsed tokamak-operation of the ITER coils over a lifetime of more than 20 years. Furthermore, it provides information on the maximum tensile or shear stress in the ITER-relevant range of 10 4 -10 5 cycles. In order to simulate the operation conditions of ITER as closely as possible, several fatigue parameters can be set in the test programme, e.g., the minimum-to-peak stress ratio R and the frequency ν of the sinusoidal load function. Further, the fatigue process can be run under load or strain control. All of these parameters may influence the mechanical response of the insulation system under cyclic load. Therefore, it is highly desirable to investigate the influence of test parameter variations on the measured stress-lifetime diagrams. The investigations were performed at 77 K using an industrial glass-fiber reinforced composite impregnated with epoxy resin. For both the load and the strain controlled mode, R-values of 0.3 and 0.5 and a frequency of 10 Hz were chosen. The results showed almost no deviations in the lifetime behavior between the load and the strain controlled mode, up to the ITER specified number of pulses, i.e. 3 x 10 4 cycles. Beyond this point, the residual strength levels were lower by 5-30 % under strain control than under load control. This effect is more pronounced at higher cycle numbers and for lower R-ratios. (author)

  10. Implementation strategy for the ITER plasma control system

    International Nuclear Information System (INIS)

    Winter, A.; Ambrosino, G.; Bauvir, B.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neto, A.; Raupp, G.; Snipes, J.A.; Stephen, A.V.; Treutterer, W.; Walker, M.L.; Zabeo, L.

    2015-01-01

    This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

  11. Implementation strategy for the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Ambrosino, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Bauvir, B. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); De Tommasi, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Humphreys, D.A. [General Atomics, San Diego, CA (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Dip. Ingegneria Industriale e dell’Informazione (Italy); Neto, A. [Fusion for Energy, Barcelona (Spain); Raupp, G. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Snipes, J.A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Stephen, A.V. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon (United Kingdom); Treutterer, W. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Walker, M.L. [General Atomics, San Diego, CA (United States); Zabeo, L. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

  12. LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration and Root-finding

    International Nuclear Information System (INIS)

    Hindmarsh, A.D.; Brown, P.N.

    1996-01-01

    1 - Description of program or function: LSODKR is a new initial value ODE solver for stiff and non-stiff systems. It is a variant of the LSODPK and LSODE solvers, intended mainly for large stiff systems. The main differences between LSODKR and LSODE are the following: a) for stiff systems, LSODKR uses a corrector iteration composed of Newton iteration and one of four preconditioned Krylov subspace iteration methods. The user must supply routines for the preconditioning operations, b) within the corrector iteration, LSODKR does automatic switching between functional (fix point) iteration and modified Newton iteration, c) LSODKR includes the ability to find roots of given functions of the solution during the integration. 2 - Method of solution: Integration is by Adams or BDF (Backward Differentiation Formula) methods, at user option. Corrector iteration is by Newton or fix point iteration, determined dynamically. Linear system solution is by a preconditioned Krylov iteration, selected by user from Incomplete Orthogonalization Method, Generalized Minimum Residual Method, and two variants of Preconditioned Conjugate Gradient Method. Preconditioning is to be supplied by the user. 3 - Restrictions on the complexity of the problem: None

  13. IVVS actuating system compatibility test to ITER gamma radiation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F. [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Baccaro, S.; Piegari, A. [ENEA CR Casaccia, 301 Via Anguillarese, 00123 Santa Maria di Galeria, Rome (Italy); Damiani, C.; Dubus, G. [Fusion For Energy c/Josep Pla, n° 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

  14. IVVS actuating system compatibility test to ITER gamma radiation conditions

    International Nuclear Information System (INIS)

    Rossi, Paolo; Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F.; Baccaro, S.; Piegari, A.; Damiani, C.; Dubus, G.

    2013-01-01

    Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

  15. Shielding design of ITER pressure suppression system

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu

    2006-01-01

    The duct shield from streaming D-T neutrons has been designed for the ITER pressure suppression system. Streaming calculations are performed with the DUCT-III code for the region from the inlet of the pressure relief line to the rupture disk. Next, the neutron permeation through the shield is studied by Monte Carlo calculations with the MCNP code. It is found that 0.15 m thick iron shield is enough to suppress the permeating component from the outside. In addition, it is suggested that the volume of the shield can be reduced by about 30% if the optimized iron shield structure having localized thickness across intense permeation paths is employed to shield the pressure suppression line. (T.I.)

  16. Control systems for ITER diagnostics, heating and current drive

    Energy Technology Data Exchange (ETDEWEB)

    Simrock, Stefan, E-mail: stefan.simrock@iter.org

    2016-11-15

    The ITER Diagnostic, Heating and Current Drive systems might appear, on the face of it, to have very different control requirements. There are approximately 45 diagnostic systems, including magnetic sensors for plasma position and shape determination, imaging systems in the IR and visible, Thompson scattering for electron temperature and density, neutron detectors and collective scattering for alpha particle density and energy distribution. The H&CD systems encompass Electron Cyclotron Heating, using 24 1MW, 170 GHz gyrotrons and 5 steerable launchers to deliver 20 MW to the plasma, Ion Cyclotron Heating, using 8 3MW, 40–55 MHz sources and two multi-element launchers to deliver 20 MW to the plasma, and 2 Negative Ion Neutral Beam Injectors, each of which can deliver up to 16.5 MW of 1 MeV beams to the plasma. Although there are substantial differences in the needs for protection, when handling multi-MW heating systems, and in data throughput for many diagnostics, the formal processes needed to translate system requirements into Instrumentation and Control are identical. Due to the distributed procurement of ITER sub-systems and the need to integrate as painlessly as possible to CODAC, the formal processes, together with a substantial degree of standardization, are even more than usually essential. Starting from the technical, safety and protection, integration and operation requirements, a loop of functional analysis and signal listing is used to generate the controller configuration and the conceptual architecture. These elements in their turn lead to the physical and software design. The paper will describe the formal processes of control system design and the methods used by the ITER project to achieve the standardization of systems engineering practices. These have been applied to several use-cases covering all operation relevant phases such as plasma operation, maintenance, testing and conditioning. There are a number of running contracts that are developing

  17. Biometric identification using local iterated function

    Science.gov (United States)

    Al-Saidi, N. M. G.; Said, M. R. M.

    2014-06-01

    Biometric identification protocol has been received an increasing interest recently. It is a process that determines person identity by making use of their biometric features. A new biometric identification method is presented in this paper based on partial self-similarity that used to identify features within fingerprint images. This approach is already used in Fractal Image Compression (FIC) due to their ability to represent the images by a limited number of affine transformations, and its variation of scale, translation or rotation. These features give the recognition process high impact and good performance. To process data in a fingerprint image, it first converted into digital format using Optical Fingerprint Reader (OFR). The verification process is done by comparing these data with the server data. The system analysis shows that the proposed method is efficient in terms of memory and time complexity.

  18. ITER Fast Plant System Controller prototype based on PXIe platform

    International Nuclear Information System (INIS)

    Ruiz, M.; Vega, J.; Castro, R.; Sanz, D.; López, J.M.; Arcas, G. de; Barrera, E.; Nieto, J.; Gonçalves, B.; Sousa, J.; Carvalho, B.; Utzel, N.; Makijarvi, P.

    2012-01-01

    Highlights: ► Implementation of Fast Plant System Controller (FPSC) for ITER CODAC. ► Efficient data acquisition and data movement using EPICS. ► Performance of PCIe technologies in the implementation of FPSC. - Abstract: The ITER Fast Plant System Controller (FPSC) is based on embedded technologies. The FPSC will be devoted to both data acquisition tasks (sampling rates higher than 1 kHz) and control purposes (feedback loop actuators). Some of the essential requirements of these systems are: (a) data acquisition and data preprocessing; (b) interfacing with different networks and high speed links (Plant Operation Network, timing network based on IEEE1588, synchronous data transference and streaming/archiving networks); and (c) system setup and operation using EPICS (Experimental Physics and Industrial Control System) process variables. CIEMAT and UPM have implemented a prototype of FPSC using a PXIe (PCI eXtension for Instrumentation) form factor in a R and D project developed in two phases. The paper presents the main features of the two prototypes developed that have been named alpha and beta. The former was implemented using LabVIEW development tools as it was focused on modeling the FPSC software modules, using the graphical features of LabVIEW applications, and measuring the basic performance in the system. The alpha version prototype implements data acquisition with time-stamping, EPICS monitoring using waveform process variables (PVs), and archiving. The beta version prototype is a complete IOC implemented using EPICS with different software functional blocks. These functional blocks are integrated and managed using an ASYN driver solution and provide the basic functionalities required by ITER FPSC such as data acquisition, data archiving, data pre-processing (using both CPU and GPU) and streaming.

  19. Methodology for dimensional variation analysis of ITER integrated systems

    International Nuclear Information System (INIS)

    Fuentes, F. Javier; Trouvé, Vincent; Cordier, Jean-Jacques; Reich, Jens

    2016-01-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

  20. Methodology for dimensional variation analysis of ITER integrated systems

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes, F. Javier, E-mail: FranciscoJavier.Fuentes@iter.org [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France); Trouvé, Vincent [Assystem Engineering & Operation Services, rue J-M Jacquard CS 60117, 84120 Pertuis (France); Cordier, Jean-Jacques; Reich, Jens [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France)

    2016-11-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Yoshida, Kiyoshi; Sugimoto, Makoto; Tsuji, Hiroshi

    1991-08-01

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

  4. ITER neutral beam system US conceptual design

    International Nuclear Information System (INIS)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D - source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus

  5. Cryogenic distribution system for ITER proto-type cryoline test

    International Nuclear Information System (INIS)

    Bhattacharya, R.; Shah, N.; Badgujar, S.; Sarkar, B.

    2012-01-01

    Design validation for ITER cryoline will be carried out by proto-type test on cryoline. The major objectives of the test will be to ensure the mechanical integrity, reliability, thermal stress and heat load as well as checking of assembly and fabrication procedures. The cryogenics system has to satisfy the functional operating scenario of the cryoline. Cryoplant, distribution box (DB) including liquid helium (LHe) tank constitute the cryogenic system for the test. Conceptual system architecture is proposed with a commercially available refrigerator/liquefier and custom designed DB housing cold compressor, cold circulator as well as phase separator with sub-merged heat exchanger. System level optimization, mainly with DB and LHe tank with options, has been studied to minimize the cold power required for the system. Aspen HYSYS is used for the purpose of process simulation. The paper describes the system architecture and the optimized design as well as process simulation with associated results. (author)

  6. Implications of ITER requirements on R and D of RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.; Agarici, G.; Beaumont, B.

    2003-01-01

    Heating and Current Drive (H and CD) systems have an essential role in ITER-FEAT operation, as all phases of ITER operation are driven and controlled by the auxiliary power flow. The RF (Electron Cyclotron and Ion Cyclotron) systems, planned to contribute for ∼ 60% of the total auxiliary power (72 MW), with Lower Hybrid used for the specialised function of current drive in the extended performance phase (20 MW), are at different level of technology development. All systems, need a significant development in order to meet ITER operation requirements In this paper these requirements are reviewed and CEA proposals for the development of the Ion cyclotron system presented. (author)

  7. Engineering Design of ITER Prototype Fast Plant System Controller

    Science.gov (United States)

    Goncalves, B.; Sousa, J.; Carvalho, B.; Rodrigues, A. P.; Correia, M.; Batista, A.; Vega, J.; Ruiz, M.; Lopez, J. M.; Rojo, R. Castro; Wallander, A.; Utzel, N.; Neto, A.; Alves, D.; Valcarcel, D.

    2011-08-01

    The ITER control, data access and communication (CODAC) design team identified the need for two types of plant systems. A slow control plant system is based on industrial automation technology with maximum sampling rates below 100 Hz, and a fast control plant system is based on embedded technology with higher sampling rates and more stringent real-time requirements than that required for slow controllers. The latter is applicable to diagnostics and plant systems in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers, interface to actuators and sensors and, if necessary, high performance networks. Two prototypes of a fast plant system controller specialized for data acquisition and constrained by ITER technological choices are being built using two different form factors. This prototyping activity contributes to the Plant Control Design Handbook effort of standardization, specifically regarding fast controller characteristics. Envisaging a general purpose fast controller design, diagnostic use cases with specific requirements were analyzed and will be presented along with the interface with CODAC and sensors. The requirements and constraints that real-time plasma control imposes on the design were also taken into consideration. Functional specifications and technology neutral architecture, together with its implications on the engineering design, were considered. The detailed engineering design compliant with ITER standards was performed and will be discussed in detail. Emphasis will be given to the integration of the controller in the standard CODAC environment. Requirements for the EPICS IOC providing the interface to the outside world, the prototype decisions on form factor, real-time operating system, and high-performance networks will also be discussed, as well as the requirements for data streaming to CODAC for visualization and

  8. LSODKR, Stiff Ordinary Differential Equations (ODE) System Solver with Krylov Iteration with Root-finding

    International Nuclear Information System (INIS)

    Hindmarsh, A.C.; Petzold, L.R.

    2005-01-01

    1 - Description of program or function: LSODKR is a new initial value ODE solver for stiff and non-stiff systems. It is a variant of the LSODPK and LSODE solvers, intended mainly for large stiff systems. The main differences between LSODKR and LSODE are the following: a) for stiff systems, LSODKR uses a corrector iteration composed of Newton iteration and one of four preconditioned Krylov subspace iteration methods. The user must supply routines for the preconditioning operations, b) within the corrector iteration, LSODKR does automatic switching between functional (fix point) iteration and modified Newton iteration, The nonlinear iteration method-switching differs from the method-switching in LSODA and LSODAR, but provides similar savings by using the cheaper method in the non-stiff regions of the problem. c) LSODKR includes the ability to find roots of given functions of the solution during the integration. d) LSODKR also improves on the Krylov methods in LSODPK by offering the option to save and reuse the approximate Jacobian data underlying the pre-conditioner. The LSODKR source is commented extensively to facilitate modification. Both a single-precision version and a double-precision version are available. 2 - Methods: It is assumed that the ODEs are given explicitly, so that the system can be written in the form dy/dt = f(t,y), where y is the vector of dependent variables, and t is the independent variable. Integration is by Adams or BDF (Backward Differentiation Formula) methods, at user option. Corrector iteration is by Newton or fix point iteration, determined dynamically. Linear system solution is by a preconditioned Krylov iteration, selected by user from Incomplete Orthogonalization Method, Generalized Minimum Residual Method, and two variants of Preconditioned Conjugate Gradient Method. Preconditioning is to be supplied by the user

  9. Data archiving system implementation in ITER's CODAC Core System

    International Nuclear Information System (INIS)

    Castro, R.; Abadie, L.; Makushok, Y.; Ruiz, M.; Sanz, D.; Vega, J.; Faig, J.; Román-Pérez, G.; Simrock, S.; Makijarvi, P.

    2015-01-01

    Highlights: • Implementation of ITER's data archiving solution. • Integration of the solution into CODAC Core System. • Data archiving structure. • High efficient data transmission into fast plant system controllers. • Fast control and data acquisition in Linux. - Abstract: The aim of this work is to present the implementation of data archiving in ITER's CODAC Core System software. This first approach provides a client side API and server side software allowing the creation of a simplified version of ITERDB data archiving software, and implements all required elements to complete data archiving flow from data acquisition until its persistent storage technology. The client side includes all necessary components that run on devices that acquire or produce data, distributing and streaming to configure remote archiving servers. The server side comprises an archiving service that stores into HDF5 files all received data. The archiving solution aims at storing data coming for the data acquisition system, the conventional control and also processed/simulated data.

  10. Dhage Iteration Method for Generalized Quadratic Functional Integral Equations

    Directory of Open Access Journals (Sweden)

    Bapurao C. Dhage

    2015-01-01

    Full Text Available In this paper we prove the existence as well as approximations of the solutions for a certain nonlinear generalized quadratic functional integral equation. An algorithm for the solutions is developed and it is shown that the sequence of successive approximations starting at a lower or upper solution converges monotonically to the solutions of related quadratic functional integral equation under some suitable mixed hybrid conditions. We rely our main result on Dhage iteration method embodied in a recent hybrid fixed point theorem of Dhage (2014 in partially ordered normed linear spaces. An example is also provided to illustrate the abstract theory developed in the paper.

  11. Integration of IC/EC systems in ITER

    International Nuclear Information System (INIS)

    Gassmann, T.; Beaumont, B.; Baruah, U.K.; Bonicelli, T.; Chiocchio, S.; Cox, D.; Darbos, C.; Decamps, H.; Denisov, G.; Henderson, M.; Kazarian, F.; Lamalle, P.U.; Mukherjee, A.; Rasmussen, D.; Saibene, G.; Sartori, R.; Sakamoto, K.; Tanga, A.

    2010-01-01

    The RF heating and current drive (H and CD) systems that are to be installed in ITER during the construction phase, are the electron cyclotron (EC) and ion cyclotron (IC) systems. They are complex assemblies of high voltage power supplies (HVPS), RF generators, transmission lines and antennas. Their design and integration are constrained by many interfaces, both internal, between the subsystems, and external, with the other ITER systems. In addition, some components must be compatible with a nuclear environment and are classified as Safety Important Component. This paper describes the processes implemented in ITER to ensure proper integration.

  12. Validated design of the ITER main vacuum pumping systems

    International Nuclear Information System (INIS)

    Day, Chr.; Antipenkov, A.; Dremel, M.; Haas, H.; Hauer, V.; Mack, A.; Boissin, J.-C.; Class, G.; Murdoch, D.K.; Wykes, M.

    2005-01-01

    Forschungszentrum Karlsruhe is developing the ITER high vacuum cryogenic pumping systems (torus, cryostat, NBI) as well as the corresponding mechanical roughing pump trains. All force-cooled big cryopumps incorporate similar design of charcoal coated cryopanels cooled to 5 K with supercritical helium. A model of the torus exhaust cryopump was comprehensively characterised in the TIMO testbed at Forschungszentrum. This paper discusses the vacuum performance results of the model pump and outlines how these data were incorporated in a sound design of the whole ITER torus exhaust pumping system. To do this, the dedicated software package ITERVAC was developed which is able to describe gas flow in viscous, transitional and molecular flow regimes as needed for the gas coming through the divertor slots and along the pump ducts into the cryopumps. The entrance section between the divertor cassettes and each pumping duct was identified to be the bottleneck of the gas flow. The interrelation of achievable throughputs as a function of the divertor pressure and the cryopump pumping speed is discussed. The system design is completed by assessment of the NBI cryopump system and integrating performance curves for the roughing pump trains needed during the regeneration phases of the cryopumps. (author)

  13. Design of the RTO/RC ITER primary pumping system

    International Nuclear Information System (INIS)

    Ladd, P.; Ibbott, C; Janeschitz, G.; Martin, E.

    2000-01-01

    The primary pumping system is needed not only to exhaust helium ash resulting from the DT reaction but also excess fuelling gas injected during the fusion burn, which can extend for 100's to 1000's of seconds, and to perform a variety of other functions. The prevailing environmental conditions, principally nuclear radiation, tritium exposure, magnetic fields, and the need for containment, have a significant impact on the design and selection of equipment. This paper presents the design of the Reduced Technical Objectives/Reduced Cost (RTO/RC) ITER primary pumping system with particular emphasis on the nuclear aspects of the design. Component selection and equipment layout issues to meet established requirements for the system are reviewed together with the R and D that is being undertaken to support the design. In addition, serviceability and maintainability issues related to this system are also discussed

  14. Iterative group splitting algorithm for opportunistic scheduling systems

    KAUST Repository

    Nam, Haewoon; Alouini, Mohamed-Slim

    2014-01-01

    An efficient feedback algorithm for opportunistic scheduling systems based on iterative group splitting is proposed in this paper. Similar to the opportunistic splitting algorithm, the proposed algorithm adjusts (or lowers) the feedback threshold

  15. Iterative development of visual control systems in a research vivarium.

    Science.gov (United States)

    Bassuk, James A; Washington, Ida M

    2014-01-01

    The goal of this study was to test the hypothesis that reintroduction of Continuous Performance Improvement (CPI) methodology, a lean approach to management at Seattle Children's (Hospital, Research Institute, Foundation), would facilitate engagement of vivarium employees in the development and sustainment of a daily management system and a work-in-process board. Such engagement was implemented through reintroduction of aspects of the Toyota Production System. Iterations of a Work-In-Process Board were generated using Shewhart's Plan-Do-Check-Act process improvement cycle. Specific attention was given to the importance of detecting and preventing errors through assessment of the following 5 levels of quality: Level 1, customer inspects; Level 2, company inspects; Level 3, work unit inspects; Level 4, self-inspection; Level 5, mistake proofing. A functioning iteration of a Mouse Cage Work-In-Process Board was eventually established using electronic data entry, an improvement that increased the quality level from 1 to 3 while reducing wasteful steps, handoffs and queues. A visual workplace was realized via a daily management system that included a Work-In-Process Board, a problem solving board and two Heijunka boards. One Heijunka board tracked cage changing as a function of a biological kanban, which was validated via ammonia levels. A 17% reduction in cage changing frequency provided vivarium staff with additional time to support Institute researchers in their mutual goal of advancing cures for pediatric diseases. Cage washing metrics demonstrated an improvement in the flow continuum in which a traditional batch and queue push system was replaced with a supermarket-type pull system. Staff engagement during the improvement process was challenging and is discussed. The collective data indicate that the hypothesis was found to be true. The reintroduction of CPI into daily work in the vivarium is consistent with the 4P Model of the Toyota Way and selected Principles

  16. Iterative development of visual control systems in a research vivarium.

    Directory of Open Access Journals (Sweden)

    James A Bassuk

    Full Text Available The goal of this study was to test the hypothesis that reintroduction of Continuous Performance Improvement (CPI methodology, a lean approach to management at Seattle Children's (Hospital, Research Institute, Foundation, would facilitate engagement of vivarium employees in the development and sustainment of a daily management system and a work-in-process board. Such engagement was implemented through reintroduction of aspects of the Toyota Production System. Iterations of a Work-In-Process Board were generated using Shewhart's Plan-Do-Check-Act process improvement cycle. Specific attention was given to the importance of detecting and preventing errors through assessment of the following 5 levels of quality: Level 1, customer inspects; Level 2, company inspects; Level 3, work unit inspects; Level 4, self-inspection; Level 5, mistake proofing. A functioning iteration of a Mouse Cage Work-In-Process Board was eventually established using electronic data entry, an improvement that increased the quality level from 1 to 3 while reducing wasteful steps, handoffs and queues. A visual workplace was realized via a daily management system that included a Work-In-Process Board, a problem solving board and two Heijunka boards. One Heijunka board tracked cage changing as a function of a biological kanban, which was validated via ammonia levels. A 17% reduction in cage changing frequency provided vivarium staff with additional time to support Institute researchers in their mutual goal of advancing cures for pediatric diseases. Cage washing metrics demonstrated an improvement in the flow continuum in which a traditional batch and queue push system was replaced with a supermarket-type pull system. Staff engagement during the improvement process was challenging and is discussed. The collective data indicate that the hypothesis was found to be true. The reintroduction of CPI into daily work in the vivarium is consistent with the 4P Model of the Toyota Way and

  17. A Robust Threshold for Iterative Channel Estimation in OFDM Systems

    Directory of Open Access Journals (Sweden)

    A. Kalaycioglu

    2010-04-01

    Full Text Available A novel threshold computation method for pilot symbol assisted iterative channel estimation in OFDM systems is considered. As the bits are transmitted in packets, the proposed technique is based on calculating a particular threshold for each data packet in order to select the reliable decoder output symbols to improve the channel estimation performance. Iteratively, additional pilot symbols are established according to the threshold and the channel is re-estimated with the new pilots inserted to the known channel estimation pilot set. The proposed threshold calculation method for selecting additional pilots performs better than non-iterative channel estimation, no threshold and fixed threshold techniques in poor HF channel simulations.

  18. 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

  19. 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.

  20. Development of simulator for remote handling system of ITER blanket

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakanhira, Masataka; Matsumoto, Yasuhiro; Shibanuma, K.

    2007-01-01

    The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of gamma ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of gamma ray. It is expected that enough numbers of cameras and lights are not available because of arrangement restriction. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software ''ENVISION''. The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation. The developed simulator will be modified further adding other necessary functions and finally completed as a prototype of the actual simulator for the blanket remote handling system

  1. Multistep Hybrid Iterations for Systems of Generalized Equilibria with Constraints of Several Problems

    Directory of Open Access Journals (Sweden)

    Lu-Chuan Ceng

    2014-01-01

    Full Text Available We first introduce and analyze one multistep iterative algorithm by hybrid shrinking projection method for finding a solution of the system of generalized equilibria with constraints of several problems: the generalized mixed equilibrium problem, finitely many variational inclusions, the minimization problem for a convex and continuously Fréchet differentiable functional, and the fixed-point problem of an asymptotically strict pseudocontractive mapping in the intermediate sense in a real Hilbert space. We prove strong convergence theorem for the iterative algorithm under suitable conditions. On the other hand, we also propose another multistep iterative algorithm involving no shrinking projection method and derive its weak convergence under mild assumptions.

  2. Chevron beam dump for ITER edge Thomson scattering system

    International Nuclear Information System (INIS)

    Yatsuka, E.; Hatae, T.; Bassan, M.; Itami, K.; Vayakis, G.

    2013-01-01

    This paper contains the design of the beam dump for the ITER edge Thomson scattering system and mainly concerns its lifetime under the harsh thermal and electromagnetic loads as well as tight space allocation. The lifetime was estimated from the multi-pulse laser-induced damage threshold. In order to extend its lifetime, the structure of the beam dump was optimized. A number of bent sheets aligned parallel in the beam dump form a shape called a chevron which enables it to avoid the concentration of the incident laser pulse energy. The chevron beam dump is expected to withstand thermal loads due to nuclear heating, radiation from the plasma, and numerous incident laser pulses throughout the entire ITER project with a reasonable margin for the peak factor of the beam profile. Structural analysis was also carried out in case of electromagnetic loads during a disruption. Moreover, detailed issues for more accurate assessments of the beam dump's lifetime are clarified. Variation of the bi-directional reflection distribution function (BRDF) due to erosion by or contamination of neutral particles derived from the plasma is one of the most critical issues that needs to be resolved. In this paper, the BRDF was assumed, and the total amount of stray light and the absorbed laser energy profile on the beam dump were evaluated

  3. Chevron beam dump for ITER edge Thomson scattering system

    Energy Technology Data Exchange (ETDEWEB)

    Yatsuka, E.; Hatae, T.; Bassan, M.; Itami, K. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Vayakis, G. [ITER Organization, 13115 St Paul Lez Durance Cedex (France)

    2013-10-15

    This paper contains the design of the beam dump for the ITER edge Thomson scattering system and mainly concerns its lifetime under the harsh thermal and electromagnetic loads as well as tight space allocation. The lifetime was estimated from the multi-pulse laser-induced damage threshold. In order to extend its lifetime, the structure of the beam dump was optimized. A number of bent sheets aligned parallel in the beam dump form a shape called a chevron which enables it to avoid the concentration of the incident laser pulse energy. The chevron beam dump is expected to withstand thermal loads due to nuclear heating, radiation from the plasma, and numerous incident laser pulses throughout the entire ITER project with a reasonable margin for the peak factor of the beam profile. Structural analysis was also carried out in case of electromagnetic loads during a disruption. Moreover, detailed issues for more accurate assessments of the beam dump's lifetime are clarified. Variation of the bi-directional reflection distribution function (BRDF) due to erosion by or contamination of neutral particles derived from the plasma is one of the most critical issues that needs to be resolved. In this paper, the BRDF was assumed, and the total amount of stray light and the absorbed laser energy profile on the beam dump were evaluated.

  4. Chevron beam dump for ITER edge Thomson scattering system.

    Science.gov (United States)

    Yatsuka, E; Hatae, T; Vayakis, G; Bassan, M; Itami, K

    2013-10-01

    This paper contains the design of the beam dump for the ITER edge Thomson scattering system and mainly concerns its lifetime under the harsh thermal and electromagnetic loads as well as tight space allocation. The lifetime was estimated from the multi-pulse laser-induced damage threshold. In order to extend its lifetime, the structure of the beam dump was optimized. A number of bent sheets aligned parallel in the beam dump form a shape called a chevron which enables it to avoid the concentration of the incident laser pulse energy. The chevron beam dump is expected to withstand thermal loads due to nuclear heating, radiation from the plasma, and numerous incident laser pulses throughout the entire ITER project with a reasonable margin for the peak factor of the beam profile. Structural analysis was also carried out in case of electromagnetic loads during a disruption. Moreover, detailed issues for more accurate assessments of the beam dump's lifetime are clarified. Variation of the bi-directional reflection distribution function (BRDF) due to erosion by or contamination of neutral particles derived from the plasma is one of the most critical issues that needs to be resolved. In this paper, the BRDF was assumed, and the total amount of stray light and the absorbed laser energy profile on the beam dump were evaluated.

  5. 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.

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

    Science.gov (United States)

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

    2017-04-01

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

  7. Alternatives of ITER vacuum vessel support system

    International Nuclear Information System (INIS)

    Ohmori, Junji; Kitamura, Kazunori; Araki, Masanori; Ohno, Isamu; Shoji, Teruaki

    2002-07-01

    Optional designs of vacuum vessel (VV) support have been performed for the International Thermonuclear Experimental Reactor (ITER) to reduce stresses and buckling concern of the flexible plate structure in ITER-FDR. One of the optional designs is hanging type VV support concept that consists of top hanging supports at the top of VV and middle radial stoppers in the middle of outboard VV. The hanging supports are located at the toroidal field (TF) coil inboard top region (R∼5400 mm) using the narrow window space surrounded by a poloidal field coil (PF1) and TF coil. The radial stoppers are located inside TF coil cases in the TF coil outboard middle region (R∼9300 mm). The upper flange connection of the radial stoppers should slide in vertical direction to eliminate thermal stress produced by relative thermal displacement between VV wall and TF coil case. Both supports consist of flexible plates and are mounted on 18 locations in toroidal direction. The radial and toroidal reaction forces are shared with the hanging supports and the radial stoppers. However, the vertical force is sustained by only the hanging supports. The others are compressive type support concept that consists of nine VV supports located in alternate divertor port regions in toroidal direction. Two designs have been performed for the VV support concept. One is mounted on TF inter-coil structures (OIS) and the other is on cryostat ring. The compressive support on TF coil OIS is dependent on TF coil movement but that on cryostat is independent. In the optional designs, the bending stress due to the relative thermal displacement between TF coil and VV is classified to primary stress according to ASME Sec. III NF. The stress due to TF coil displacement is also considered as primary stress. The stress due to non-uniform temperature distribution of the flexible plate is classified to secondary stress. The preliminary structural assessments for the optional designs have been performed for all load

  8. 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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-15

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

  11. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

    Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

    1988-01-01

    A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

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

    International Nuclear Information System (INIS)

    Lee, D.

    1995-02-01

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

  13. Integration of the ITER diagnostic plant systems with CODAC

    International Nuclear Information System (INIS)

    Simrock, S.; Barnsley, R.; Bertalot, L.; Hansalia, C.; Klotz, W.D.; Makijarvi, P.; Reichle, R.; Vayakis, G.; Yonekawa, I.; Walker, C.; Wallander, A.; Walsh, M.; Winter, A.

    2011-01-01

    ITER requires extensive diagnostic systems in order to meet the requirements for machine operation, protection, plasma control and physics studies. The realization of these systems is a major challenge not only because of the harsh environment and the nuclear requirements but also with respect to Instrumentation and Control (I and C) of all the 59 diagnostics plants. The Plant Systems I and C are mostly 'in-kind', i.e. procured by the seven ITER Domestic Agencies (DAs), while the Central I and C Systems are 'in-fund', i.e. procured by ITER Organization (IO). Standardization of Plant Systems I and C is of primary importance and has been one of the highest priority tasks of CODAC. The standards are published in the Plant Control Design Handbook (PCDH) which will be followed to ensure a homogeneous design, guarantee high availability and simplify maintenance and support future upgrades. Most important for a successful commissioning and operation of the ITER facility are the concepts of interfacing the diagnostics plant systems with CODAC and the standards for instrumentation and control which must be followed all contributing parties. In this paper, we will elaborate on the concepts of interfacing the diagnostics plant systems with CODAC and the standards that must be followed for the design.

  14. Natural Preconditioning and Iterative Methods for Saddle Point Systems

    KAUST Repository

    Pestana, Jennifer

    2015-01-01

    © 2015 Society for Industrial and Applied Mathematics. The solution of quadratic or locally quadratic extremum problems subject to linear(ized) constraints gives rise to linear systems in saddle point form. This is true whether in the continuous or the discrete setting, so saddle point systems arising from the discretization of partial differential equation problems, such as those describing electromagnetic problems or incompressible flow, lead to equations with this structure, as do, for example, interior point methods and the sequential quadratic programming approach to nonlinear optimization. This survey concerns iterative solution methods for these problems and, in particular, shows how the problem formulation leads to natural preconditioners which guarantee a fast rate of convergence of the relevant iterative methods. These preconditioners are related to the original extremum problem and their effectiveness - in terms of rapidity of convergence - is established here via a proof of general bounds on the eigenvalues of the preconditioned saddle point matrix on which iteration convergence depends.

  15. Seismic Design of ITER Component Cooling Water System-1 Piping

    Science.gov (United States)

    Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

    2017-04-01

    The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

  16. Virtual fringe projection system with nonparallel illumination based on iteration

    International Nuclear Information System (INIS)

    Zhou, Duo; Wang, Zhangying; Gao, Nan; Zhang, Zonghua; Jiang, Xiangqian

    2017-01-01

    Fringe projection profilometry has been widely applied in many fields. To set up an ideal measuring system, a virtual fringe projection technique has been studied to assist in the design of hardware configurations. However, existing virtual fringe projection systems use parallel illumination and have a fixed optical framework. This paper presents a virtual fringe projection system with nonparallel illumination. Using an iterative method to calculate intersection points between rays and reference planes or object surfaces, the proposed system can simulate projected fringe patterns and captured images. A new explicit calibration method has been presented to validate the precision of the system. Simulated results indicate that the proposed iterative method outperforms previous systems. Our virtual system can be applied to error analysis, algorithm optimization, and help operators to find ideal system parameter settings for actual measurements. (paper)

  17. Period doubling for trapezoid function iteration: metric theory

    International Nuclear Information System (INIS)

    Beyer, W.A.; Stein, P.R.

    1982-01-01

    Iterations of a one-parameter family F(lambda,x) = lambda f(x) of endomorphisms of [0,2] having the form of a trapezoid f(x) = x/e for x belongs to [0,e], f(x) = 1 for x belongs to (e,2 - e) and f(x) = (2 - x)/e for x belongs to [2 - e,2], are investigated. Here lambda belongs to [1,2] and e belongs to (0,1). Let lambda/sub n/ be the smallest value of lambda > 1 for which x = 1 is a periodic point of period 2/sup n/. It is proved that for e < 0.99, lambda/sub n/ - lambda/sub n-1/ approx. = k(lambda/sub infinity//e)/sup -2n/, where k is some constant and lambda/sub infinity/ = lim/sub n→infinity/lambda/sub n/. The same conclusion probably holds for any e < 1. This behavior is substantially different from that found by Feigenbaum and others for the case where f(x) assumes its maximum value for a unique x. Numerical investigations are reported for functions related to the trapezoid function

  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. Arc detection for the ICRF system on ITER

    Science.gov (United States)

    D'Inca, R.

    2011-12-01

    The ICRF system for ITER is designed to respect the high voltage breakdown limits. However arcs can still statistically happen and must be quickly detected and suppressed by shutting the RF power down. For the conception of a reliable and efficient detector, the analysis of the mechanism of arcs is necessary to find their unique signature. Numerous systems have been conceived to address the issues of arc detection. VSWR-based detectors, RF noise detectors, sound detectors, optical detectors, S-matrix based detectors. Until now, none of them has succeeded in demonstrating the fulfillment of all requirements and the studies for ITER now follow three directions: improvement of the existing concepts to fix their flaws, development of new theoretically fully compliant detectors (like the GUIDAR) and combination of several detectors to benefit from the advantages of each of them. Together with the physical and engineering challenges, the development of an arc detection system for ITER raises methodological concerns to extrapolate the results from basic experiments and present machines to the ITER scale ICRF system and to conduct a relevant risk analysis.

  20. Iterated Hamiltonian type systems and applications

    Science.gov (United States)

    Tiba, Dan

    2018-04-01

    We discuss, in arbitrary dimension, certain Hamiltonian type systems and prove existence, uniqueness and regularity properties, under the independence condition. We also investigate the critical case, define a class of generalized solutions and prove existence and basic properties. Relevant examples and counterexamples are also indicated. The applications concern representations of implicitly defined manifolds and their perturbations, motivated by differential systems involving unknown geometries.

  1. Status of the ITER Ion Cyclotron H and CD system

    Energy Technology Data Exchange (ETDEWEB)

    Lamalle, P., E-mail: philippe.lamalle@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Beaumont, B.; Kazarian, F.; Gassmann, T. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Agarici, G. [Fusion for Energy, Carrer Josep Pla 2, Torres Diagonal Litoral Edificio B3, 08019 Barcelona (Spain); Ajesh, P. [ITER India, Institute for Plasma Research, Bhat, Gandhinagar 382424, Gujarat (India); Alonzo, T. [Solution F, Allée du Verdon, 13770 Venelles (France); Arambhadiya, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Argouarch, A. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); Bamber, R. [EURATOM/CCFE Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Berger-By, G.; Bernard, J.-M.; Brun, C. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); Carpentier, S. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Clairet, F.; Colas, L.; Courtois, X. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); Davis, A. [EURATOM/CCFE Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Dechelle, C.; Doceul, L. [CEA Cadarache, IRFM, F-13108 St-Paul-lez-Durance (France); and others

    2013-10-15

    Highlights: ► We summarize the progress and outstanding issues in the development of the ITER Ion Cyclotron Heating and Current Drive (IC H and CD) system. ► The system is designed to robustly couple 20 MW in quasi-CW operation for a broad range of plasma scenarios, and is upgradeable to up to 40 MW. ► The design is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject. ► R and D is ongoing to validate key antenna components, and to qualify the radio-frequency (RF) sources and the transmission and matching components. ► Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. -- Abstract: The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40–55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R and D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper

  2. Development of ITER in-vessel viewing and metrology systems

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation ({approx}30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  3. Development of ITER in-vessel viewing and metrology systems

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation (∼30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  4. Green function iterative solution of ground state wave function for Yukawa potential

    International Nuclear Information System (INIS)

    Zhang Zhao

    2003-01-01

    The newly developed single trajectory quadrature method is applied to solve central potentials. First, based on the series expansion method an exact analytic solution of the ground state for Hulthen potential and an approximate solution for Yukawa potential are obtained respectively. Second, the newly developed iterative method based on Green function defined by quadratures along the single trajectory is applied to solve Yukawa potential using the Coulomb solution and Hulthen solution as the trial functions respectively. The results show that a more proper choice of the trial function will give a better convergence. To further improve the convergence the iterative method is combined with the variational method to solve the ground state wave function for Yukawa potential, using variational solutions of the Coulomb and Hulthen potentials as the trial functions. The results give much better convergence. Finally, the obtained critical screen coefficient is applied to discuss the dissociate temperature of J/ψ in high temperature QGP

  5. Characterization of the radiation resistance of ITER-relevant and innovative fiber composites for the ITER magnet system

    International Nuclear Information System (INIS)

    Bittner-Rohrhofer, K.

    2003-06-01

    The application of glass-fiber reinforced composites for the insulation of the superconducting magnet coils of the ITER (International Thermonuclear Experimental Reactor ) fusion device requires high material performance. The mechanical integrity of the insulation is influenced by the neutron- and g-environment and by the high mechanical stresses of the magnet system over the entire plant lifetime of 20 years. Materials suggested as insulation have to be investigated in extensive test programs with respect to the present ITER design criteria. In particular, the ultimate tensile strength as well as the interlaminar shear behavior will change under static and dynamic load (tension-tension fatigue) at 77 K after irradiation to the ITER design fluence level of 1x1022 m-2 (E620.1 MeV). Therefore, a frequency of 10 Hz and a ratio of 0.1 were chosen, in order to simulate the pulsed TOKAMAK-operation as closely as possible. Furthermore, the fatigue behavior of the material is investigated over more than 3x104 cycles, which is the ITER- relevant design fatigue limit. Basically, these insulation systems are based on combined glass-fiber/Kapton tapes, which are impregnated with di-functional DGEBA epoxy resins. Several mechanical investigations showed that the radiation resistance of these organic resins is dramatically affected by radiation at a neutron fluence of 1x1022 m-2 (E620.1 MeV). Moreover, the material strength after irradiation is strongly influenced by these factors: the winding direction of the tapes, the quality of fabrication and the drastic delamination process of the whole compound. Furthermore, the radiation induced damage of adhesives applied for supporting the interfacial bonding between the glass-fiber tape and Kapton has an adverse effect on the material performance. In addition, the poor interlaminar shear behavior does not fulfil the requirements of ITER. These test-results motivated for the development of innovative resin systems with higher stability

  6. Hierarchical models and iterative optimization of hybrid systems

    Energy Technology Data Exchange (ETDEWEB)

    Rasina, Irina V. [Ailamazyan Program Systems Institute, Russian Academy of Sciences, Peter One str. 4a, Pereslavl-Zalessky, 152021 (Russian Federation); Baturina, Olga V. [Trapeznikov Control Sciences Institute, Russian Academy of Sciences, Profsoyuznaya str. 65, 117997, Moscow (Russian Federation); Nasatueva, Soelma N. [Buryat State University, Smolina str.24a, Ulan-Ude, 670000 (Russian Federation)

    2016-06-08

    A class of hybrid control systems on the base of two-level discrete-continuous model is considered. The concept of this model was proposed and developed in preceding works as a concretization of the general multi-step system with related optimality conditions. A new iterative optimization procedure for such systems is developed on the base of localization of the global optimality conditions via contraction the control set.

  7. On the automatic control of the ITER ion cyclotron system

    Energy Technology Data Exchange (ETDEWEB)

    Bosia, G. [Department of General Physics, University of Turin, Via P. Giuria 1, 10 125 Turin (Italy)], E-mail: giuseppe.bosia@to.infn.it

    2007-10-15

    The ITER ion cyclotron heating system requires an efficient control system capable of: (i) providing the desired array radiation spectrum, to optimize plasma coupling and absorption and to minimize parasitic power losses in the plasma edge; (ii) maintaining the RF power flow to the plasma against significant load variations, including fast fluctuations induced by ELMs; (iii) reliably detecting and suppressing RF voltage breakdowns in the array and/or in the transmission system, to avoid local equipment damage and (iv) implementing an accurate real time record of performance. In this paper specific aspects of the tuning control system, related to recent conceptual and engineering effort [K. Vulliez, et al., Design of the ITER ion cyclotron heating launcher based on in-vessel tuning system, Article ID106C, this conference] are addressed.

  8. Design of coolant distribution system (CDS) for ITER PF AC/DC converter

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

    2016-10-15

    Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

  9. Development of ITER CODAC compatible gyrotron local control system and its operation

    International Nuclear Information System (INIS)

    Ohshima, Katsumi; Oda, Yasuhisa; Takahashi, Koji; Terakado, Masayuki; Ikeda, Ryosuke; Moriyama, Shinichi; Kajiwara, Ken; Sakamoto, Keishi; Hayashi, Kazuo

    2016-03-01

    In Japan Atomic Energy Agency, an ITER relevant control system for ITER gyrotron was developed according to Plant Control Design Handbook. This control system was developed based on ITER CODAC Core System and implemented state machine control of gyrotron operation system, sequential timing control of gyrotron oscillation startup, and data acquisition. The operation of ITER 170 GHz gyrotron was demonstrated with ITER relevant power supply configuration. This system is utilized for gyrotron operation test for ITER procurement. This report describes the architecture of gyrotron local control system, its basic and detailed design, and recent operation results. (author)

  10. Complexions of Iterative Information Systems Development Approaches

    DEFF Research Database (Denmark)

    Hansen, Magnus Rotvit Perlt

    There is an abundance of methodologies and approaches of information Systems Development (ISD) Processes. We use the concept of complexions of development processes and Technology Use Mediation (TUM) from a client perspective to show how an instance of a planned project in the prehospital sector...... changes dynamically. We identify how the events in the ISD process unfold according to which actions are taken and responded to by users and management respectively. The study finds that depending on the types of actions that users take, managers will react and this can greatly change the complexion...

  11. The Tritium Systems Test Assembly applicability to ITER

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1988-01-01

    The Tritium Systems Test Assembly (TSTA), is operated by the Los Alamos National Laboratory (LANL) under the sponsorship of the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The objectives of the TSTA project are to develop, demonstrate, and evaluate the exhaust gas processing and tritium related safety systems for the magnetic fusion energy program. The applicability of these processes for the ITER Tokamak is discussed

  12. Comments on new iterative methods for solving linear systems

    Directory of Open Access Journals (Sweden)

    Wang Ke

    2017-06-01

    Full Text Available Some new iterative methods were presented by Du, Zheng and Wang for solving linear systems in [3], where it is shown that the new methods, comparing to the classical Jacobi or Gauss-Seidel method, can be applied to more systems and have faster convergence. This note shows that their methods are suitable for more matrices than positive matrices which the authors suggested through further analysis and numerical examples.

  13. Iterative algorithms for large sparse linear systems on parallel computers

    Science.gov (United States)

    Adams, L. M.

    1982-01-01

    Algorithms for assembling in parallel the sparse system of linear equations that result from finite difference or finite element discretizations of elliptic partial differential equations, such as those that arise in structural engineering are developed. Parallel linear stationary iterative algorithms and parallel preconditioned conjugate gradient algorithms are developed for solving these systems. In addition, a model for comparing parallel algorithms on array architectures is developed and results of this model for the algorithms are given.

  14. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  15. Verifying large modular systems using iterative abstraction refinement

    International Nuclear Information System (INIS)

    Lahtinen, Jussi; Kuismin, Tuomas; Heljanko, Keijo

    2015-01-01

    Digital instrumentation and control (I&C) systems are increasingly used in the nuclear engineering domain. The exhaustive verification of these systems is challenging, and the usual verification methods such as testing and simulation are typically insufficient. Model checking is a formal method that is able to exhaustively analyse the behaviour of a model against a formally written specification. If the model checking tool detects a violation of the specification, it will give out a counter-example that demonstrates how the specification is violated in the system. Unfortunately, sometimes real life system designs are too big to be directly analysed by traditional model checking techniques. We have developed an iterative technique for model checking large modular systems. The technique uses abstraction based over-approximations of the model behaviour, combined with iterative refinement. The main contribution of the work is the concrete abstraction refinement technique based on the modular structure of the model, the dependency graph of the model, and a refinement sampling heuristic similar to delta debugging. The technique is geared towards proving properties, and outperforms BDD-based model checking, the k-induction technique, and the property directed reachability algorithm (PDR) in our experiments. - Highlights: • We have developed an iterative technique for model checking large modular systems. • The technique uses BDD-based model checking, k-induction, and PDR in parallel. • We have tested our algorithm by verifying two models with it. • The technique outperforms classical model checking methods in our experiments

  16. Iterative Feedback Tuning in district heating systems; Iterative Feedback Tuning i vaermeproduktionsanlaeggningar

    Energy Technology Data Exchange (ETDEWEB)

    Raaberg, Martin; Velut, Stephane; Bari, Siavosh Amanat

    2010-10-15

    The project goal is to evaluate and describe how Iterative Feedback Tuning (IFT) can be used to tune controllers in the typical control loops in heat- and power plants. There are only a few practical studies carried out for IFT and they are not really relevant for power and heat processes. It is the practical problems in implementing the IFT and the result of trimming that is the focus of this project. The project will start with theoretical studies of the IFT-method, then realization and simple simulations in scilab. The IFT equations are then implemented in Freelance 2000, an ABB control system, for practical tests on a SISO- and a MIMO-process. By performing reproducible experiments on the process and analyze the results IFT can adjust the controller parameters to minimize a cost function that represents the control goal. The project selected for SISO experiments a pressure controller in an oil transportation system. By controlling the valve position of a control valve for the reversal to the supply tank, the pressure in the oil transport system is regulated. A disturbance in oil pressure can be achieved by changing the position of a valve that lets oil through to the day tank. The selected MIMO-process is a pre-heater in a degassing process. In this process, a valve on the secondary side is utilized to control the flow in the secondary system. A valve on the primary side is utilized to control the district heating water flow through the heat exchanger to control the temperature on the secondary side. An increased secondary flow increases the heat demand and thus requiring an increase in primary flow to maintain the secondary side outlet temperature. This is the cross-coupling responsible for why it is an advantage to consider the process as multi-variable. Using the IFT method, the two original PID-controllers and a feed-forward controller is tuned simultaneously. IFT-method was difficult to implement but worked well in both simulations and in real processes

  17. Iter in vessel viewing system design and assessment activities

    Energy Technology Data Exchange (ETDEWEB)

    Neri, C., E-mail: carlo.neri@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Costa, P.; Ferri De Collibus, M.; Florean, M.; Mugnaini, G.; Pillon, M.; Pollastrone, F.; Rossi, P. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2011-10-15

    The In Vessel Viewing System (IVVS) is fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A prototype of a laser In Vessel Viewing and ranging System was developed and tested at ENEA laboratories in Frascati under EFDA task agreements, it is able to perform sub-millimetric bi-dimensional and three-dimensional images inside ITER during maintenance procedure allowing the evaluation of the state and damages of the in-vessel surface. The present prototype has been designed to operate under room conditions and starting from springtime 2009 a Grant with F4E is in progress for the design and the assessment of the IVVS system for ITER, keeping in account all the environmental conditions and constraints.

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

    International Nuclear Information System (INIS)

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

    2008-12-01

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

  19. Progress and present status of ITER cryoline system

    International Nuclear Information System (INIS)

    Badgujar, S.; Bonneton, M.; Chalifour, M.; Forgeas, A.; Serio, L.; Sarkar, B.; Shah, N.

    2014-01-01

    The cryoline system at ITER forms a very complex network localized inside the Tokamak building, on a dedicated plant bridge and in cryoplant areas. The cooling power produced in the cryoplant is distributed via these lines with a total length of about 3.7 km and interconnecting all the cold boxes of the cryogenic system as well as the cold boxes of various clients (magnets, cryopumps and thermal shield). Distinct layouts and polygonal geometry, nuclear safety and confinement requirements, difficult installation and in-service inspection/repair demand very high reliability and availability for the cryolines. The finalization of the building-embedded plates for supporting the lines, before the detailed design, has made this project technologically more challenging. The conceptual design phase has been completed and procurement arrangements have been signed with India, responsible for providing the system of cryolines and warm lines to ITER, as in kind contribution. The prototype test for the design and performance validation has been planned on a representative cryoline section. After describing the basic features and general layout of the ITER cryolines, the paper presents key design requirements, conceptual design approach, progress and status of the cryolines project as well as challenges to build such a complex cryoline system

  20. Solution of the fully fuzzy linear systems using iterative techniques

    International Nuclear Information System (INIS)

    Dehghan, Mehdi; Hashemi, Behnam; Ghatee, Mehdi

    2007-01-01

    This paper mainly intends to discuss the iterative solution of fully fuzzy linear systems which we call FFLS. We employ Dubois and Prade's approximate arithmetic operators on LR fuzzy numbers for finding a positive fuzzy vector x-tilde which satisfies A-tildex-tilde=b, where A-tilde and b-tilde are a fuzzy matrix and a fuzzy vector, respectively. Please note that the positivity assumption is not so restrictive in applied problems. We transform FFLS and propose iterative techniques such as Richardson, Jacobi, Jacobi overrelaxation (JOR), Gauss-Seidel, successive overrelaxation (SOR), accelerated overrelaxation (AOR), symmetric and unsymmetric SOR (SSOR and USSOR) and extrapolated modified Aitken (EMA) for solving FFLS. In addition, the methods of Newton, quasi-Newton and conjugate gradient are proposed from nonlinear programming for solving a fully fuzzy linear system. Various numerical examples are also given to show the efficiency of the proposed schemes

  1. Progress on radio frequency auxiliary heating system designs in ITER

    International Nuclear Information System (INIS)

    Makowski, M.; Bosia, G.; Elio, F.

    1996-09-01

    ITER will require over 100 MW of auxiliary power for heating, on- and off-axis current drive, accessing the H-mode, and plasma shut-down. The Electron Cyclotron Range of Frequencies (ECRF) and Ion Cyclotron Range of Frequencies (ICRF) are two forms of Radio Frequency (RF) auxiliary power being developed for these applications. Design concepts for both the ECRF and ICRF systems are presented, key features and critical design issues are discussed, and projected performances outlined

  2. The ITER poloidal field system: control and power supplies

    International Nuclear Information System (INIS)

    Mondino, P.L.; Benfatto, I.; Gribov, Y.; Matsukawa, M.; Odajima, K.; Portone, A.; Roshal, A.; Bareyt, B.; Bertolini, E.; Bottereau, J.M.; Huart, M.; Maschio, A.; Bulgakov, S.; Kuchinski, V.

    1995-01-01

    The paper reports the preliminary scenario of the ITER Poloidal Field (PF) system operation, the method used to evaluate the installed power, the basic structure of the circuits and finally the concepts of the preliminary design of control and power supply. The superconducting coils are energized from the HV Grid with conventional AC/DC converters. R and D is required for circuit breakers, make switches and resistors, the basic components of both the switching networks and the discharge circuits. (orig.)

  3. Design options for an ITER ion cyclotron system

    International Nuclear Information System (INIS)

    Swain, D.W.; Baity, F.W.; Bigelow, T.S.; Ryan, P.M.; Goulding, R.H.; Carter, M.D.; Stallings, D.C.; Batchelor, D.B.; Hoffman, D.J.

    1995-01-01

    Recent changes have occurred in the design requirements for the ITER ion cyclotron system, requiring in-port launchers in four main horizontal ports to deliver 50 MW of power to the plasma. The design is complicated by the comparatively large antenna-separatrix distance of 10--20 cm. Designs of a conventional strap launcher and a folded waveguide launcher than can meet the new requirements are presented

  4. AZTEC: A parallel iterative package for the solving linear systems

    Energy Technology Data Exchange (ETDEWEB)

    Hutchinson, S.A.; Shadid, J.N.; Tuminaro, R.S. [Sandia National Labs., Albuquerque, NM (United States)

    1996-12-31

    We describe a parallel linear system package, AZTEC. The package incorporates a number of parallel iterative methods (e.g. GMRES, biCGSTAB, CGS, TFQMR) and preconditioners (e.g. Jacobi, Gauss-Seidel, polynomial, domain decomposition with LU or ILU within subdomains). Additionally, AZTEC allows for the reuse of previous preconditioning factorizations within Newton schemes for nonlinear methods. Currently, a number of different users are using this package to solve a variety of PDE applications.

  5. Status of development of functional materials with perspective on beyond ITER

    International Nuclear Information System (INIS)

    Shikama, T.; Knitter, R.; Moeslang, A.; Konys, J.; Deli, L.; Muroga, T.; Kawamura, H.; Kohyama, A.

    2007-01-01

    Any engineering system is composed of functional materials as well as of structural materials, and more advanced systems tend to demand a more important and versatile role to functional materials. In nuclear fusion systems, examples of principle functional materials will be breeders and neutron multipliers for tritium production, coatings on structural materials for corrosion-resistance, MHD-loss-reduction and control of tritium permeation, thermal insertions for heat transport control, and optical and electrical materials for plasma and environmental diagnostics. For incarnation of a nuclear fusion power plant, namely DEMO, development of the functional materials with appropriate properties is essential. A role of functional materials depends strongly on a specific design of DEMO, namely designs of systems for tritium-breeding, system-cooling and heat-transfer. In the framework of ITER project, development of tritium blanket modules (TBM) is underway. Also, in parallel with the ITER project, a complemental program called the BA (Broader Approach) is launched for realization of a DEMO nuclear fusion reactor in an appropriate time schedule, where key issues of the nuclear fusion engineering needed for the DEMO will be studied under EU/Japan collaboration. In the meantime, technologies and materials needed for diagnostics and control of burning plasma are extensively discussed under the framework of International Tokamak Physics Activity (ITPA). The present paper reviews a present status of development of functional materials from views of internationally coordinated activities based on fundamental aspects of the DEMO demands as well as from views of activities based on specific but currently dominant DEMO designs. Examples of functional materials reviewed here are solid breeders, beryllium and beryllium alloys, coating layers on structural materials, thermal inserts, and some electrical and optical materials. (orig.)

  6. Physics of the conceptual design of the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Snipes, J.A., E-mail: Joseph.Snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bremond, S. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Casper, T. [1166 Bordeaux St, Pleasanton, CA 94566 (United States); Douai, D. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Humphreys, D. [General Atomics, San Diego, CA 92186 (United States); Lister, J. [Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne (EPFL), CRPP, Lausanne CH-1015 (Switzerland); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sugihara, M., E-mail: Sugihara_ma@yahoo.co.jp [Japan (Japan); Winter, A.; Zabeo, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2014-05-15

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  7. Physics of the conceptual design of the ITER plasma control system

    International Nuclear Information System (INIS)

    Snipes, J.A.; Bremond, S.; Campbell, D.J.; Casper, T.; Douai, D.; Gribov, Y.; Humphreys, D.; Lister, J.; Loarte, A.; Pitts, R.; Sugihara, M.; Winter, A.; Zabeo, L.

    2014-01-01

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  8. Disruption Mitigation System Developments and Design for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Baylor, Larry R. [ORNL; Barbier, Charlotte N. [ORNL; Bull, Nora D. [ORNL; Combs, Stephen Kirk [ORNL; Fisher, Paul W. [ORNL; Kiss, Gabor [ITER Organization, Cadarache, France; Ericson, Milton Nance [ORNL; Wilgen, John B. [ORNL; Maruyama, So [ITER Organization, Cadarache, France; Meitner, Steven J. [ORNL; Lyttle, Mark S. [ORNL; Rasmussen, David A. [ORNL; Carmichael, Justin R. [ORNL; Smith, Stephen Fulton [ORNL

    2015-09-01

    A disruption mitigation system (DMS) is under design for ITER to inject sufficient material deeply into the plasma for rapid plasma thermal shutdown and collisional suppression of any resulting runaway electrons. Progress on the development and design of both a shattered pellet injector (SPI) that produces large solid cryogenic pellets to provide reliable deep penetration of material and a fast opening high flow rate gas valve for massive gas injection (MGI) is presented. Cryogenic pellets of deuterium and neon up to 25 mm in size have been formed and accelerated with a prototype injector and a full scale prototype MGI valve is now in testing. Implications of the design with respect to response time and reliability at the proposed injector locations on ITER are discussed.

  9. Measurement and control system for ITER remote maintenance equipment

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Kakudate, Satoshi; Takeda, Nobukazu; Takiguchi, Yuji; Akou, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    ITER in-vessel components such as blankets and divertors are categorized as scheduled maintenance components because they are subjected to severe plasma heat and particle loads. Blanket maintenance requires remote handling equipment and tools able to handle Heavy payloads of about 4 tons within a 2 mm precision tolerance. Divertor maintenance requires remote replacement of 60 cassettes with a dead weight of about 25 tons each. In the ITER R and D program, full-scale remote handling equipment for blanket and divertor maintenance has been designed and assembled for demonstration tests. This paper reviews the measurement and control system developed for full-scale remote handling equipment, the Japan Home Team contribution. (author)

  10. Measurement and control system for ITER remote maintenance equipment

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Kakudate, Satoshi; Takeda, Nobukazu; Takiguchi, Yuji; Akou, Kentaro

    1998-01-01

    ITER in-vessel components such as blankets and divertors are categorized as scheduled maintenance components because they are subjected to severe plasma heat and particle loads. Blanket maintenance requires remote handling equipment and tools able to handle Heavy payloads of about 4 tons within a 2 mm precision tolerance. Divertor maintenance requires remote replacement of 60 cassettes with a dead weight of about 25 tons each. In the ITER R and D program, full-scale remote handling equipment for blanket and divertor maintenance has been designed and assembled for demonstration tests. This paper reviews the measurement and control system developed for full-scale remote handling equipment, the Japan Home Team contribution. (author)

  11. Primary design and operation analysis of ITER air transfer system

    International Nuclear Information System (INIS)

    Wang Haitian; Li Ge; Qin Shijun

    2010-01-01

    Air transfer system (ATS) is a remote handling transfer, which can work in the nuclear radiation environment and can be driven by the electricity fully. Its motion power is provided by several servo motors. The remote control technology of ATS, which is China taking part in the plan of international Tokamak experimental reactor (ITER) and grasping this technology, is one of key technologies of ITER. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Because there is gamma irradiation and hazard material in these ITER parts, all required maintenance of port plugs and inner components are been transmitted by ATS. The pick-up or drop-off these components are completed by means of a remotely controlled TCS system between the Vacuum Vessel and the Hot Cell through the bridge-gallery. Tokamak building includes three floors, including upper port, equatorial port and lower port, linked by a lift. According to each port level configuration and safety requirement, the radius of curvature with ATS trajectory is optimized, and a trajectory of each level is determined by positioned guidance beacons. At last, the results of computer aided design (CAD) show single trajectory guidance of ATS in each level is available. (authors)

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. The conceptual design of the ITER CODAC system

    International Nuclear Information System (INIS)

    Farthing, J.; Greenwald, M.; Jo Lister; Izuru Yonekawa

    2006-01-01

    The COntrol Data Access and Communication (CODAC) system for ITER is presently under conceptual design, revising the previous design dating from 1998. The design concentrates on the major perceived challenges: 35-year life of the project for maintenance and evolution; harmonizing strict access security with world-wide participation in the exploitation of ITER; the complexity of CODAC which has to control a large number of disparate procurements systems, 24 hours/365 days; the particular '' in-kind '' procurement of all Plant Systems. The design has so-far concentrated on appropriate methods for combating these challenges. Concepts include: strict application and enforcement of standards for interfacing procured systems at a high '' black-box '' level; reliance on standard high performance networks; reliance on the self-description of the procured systems; maximizing the use of data-driven applications, rather than device-specific coding. The interfacing and procurement specifications will be presented, especially the self-description of '' black-box '' systems, and the boundaries of CODAC will be defined. The breakdown of CODAC into a number of manageable systems and their interfaces will be outlined. The data volumes and data rates will be estimated, suggesting an appropriate conceptual design of the various parts of the CODAC network. There are no required CODAC features which could not be provided with today's tools. However, one element of this conceptual design is to identify areas where ideal solutions are not clearly available for which appropriate R(and)D will be proposed. (author)

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

    International Nuclear Information System (INIS)

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

    1986-12-01

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

  15. General Large Deviations and Functional Iterated Logarithm Law for Multivalued Stochastic Differential Equations

    OpenAIRE

    Ren, Jiagang; Wu, Jing; Zhang, Hua

    2015-01-01

    In this paper, we prove a large deviation principle of Freidlin-Wentzell's type for the multivalued stochastic differential equations. As an application, we derive a functional iterated logarithm law for the solutions of multivalued stochastic differential equations.

  16. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    Stringer, J.

    1992-11-01

    This report describes further design issues concerning remote maintenance of torus vacuum pumping systems options for ITER. The key issues under investigation in this report are flask support systems for valve seal exchange operations for the compound cryopump scheme and remote maintenance of a proposed multiple turbomolecular pump (TMP) system, an alternative ITER torus exhaust pumping option. Previous studies have shown that the overhead support methods for seal exchange flask equipment could malfunction due to valve/flask misalignment. A floor-mounted support system is described in this report. This scheme provides a more rigid support system for seal exchange operations. An alternative torus pumping system, based on the use of multiple TMPs, is studied from a remote maintenance standpoint. In this concept, centre distance spacing for pump/valve assemblies is too restrictive for remote maintenance. Recommendations are made for adequate spacing of these assemblies based on commercially-available 0.8 m and 1.0 m diameter valves. Fewer pumps will fit in this arrangement, which implies a need for larger TMPs. Pumps of this size are not commercially available. Other concerns regarding the servicing and storage of remote handling equipment in cells are also identified. (9 figs.)

  17. From synthesis to function via iterative assembly of N-methyliminodiacetic acid boronate building blocks.

    Science.gov (United States)

    Li, Junqi; Grillo, Anthony S; Burke, Martin D

    2015-08-18

    The study and optimization of small molecule function is often impeded by the time-intensive and specialist-dependent process that is typically used to make such compounds. In contrast, general and automated platforms have been developed for making peptides, oligonucleotides, and increasingly oligosaccharides, where synthesis is simplified to iterative applications of the same reactions. Inspired by the way natural products are biosynthesized via the iterative assembly of a defined set of building blocks, we developed a platform for small molecule synthesis involving the iterative coupling of haloboronic acids protected as the corresponding N-methyliminodiacetic acid (MIDA) boronates. Here we summarize our efforts thus far to develop this platform into a generalized and automated approach for small molecule synthesis. We and others have employed this approach to access many polyene-based compounds, including the polyene motifs found in >75% of all polyene natural products. This platform further allowed us to derivatize amphotericin B, the powerful and resistance-evasive but also highly toxic last line of defense in treating systemic fungal infections, and thereby understand its mechanism of action. This synthesis-enabled mechanistic understanding has led us to develop less toxic derivatives currently under evaluation as improved antifungal agents. To access more Csp(3)-containing small molecules, we gained a stereocontrolled entry into chiral, non-racemic α-boryl aldehydes through the discovery of a chiral derivative of MIDA. These α-boryl aldehydes are versatile intermediates for the synthesis of many Csp(3) boronate building blocks that are otherwise difficult to access. In addition, we demonstrated the utility of these types of building blocks in accessing pharmaceutically relevant targets via an iterative Csp(3) cross-coupling cycle. We have further expanded the scope of the platform to include stereochemically complex macrocyclic and polycyclic molecules

  18. Iterative method of the parameter variation for solution of nonlinear functional equations

    International Nuclear Information System (INIS)

    Davidenko, D.F.

    1975-01-01

    The iteration method of parameter variation is used for solving nonlinear functional equations in Banach spaces. The authors consider some methods for numerical integration of ordinary first-order differential equations and construct the relevant iteration methods of parameter variation, both one- and multifactor. They also discuss problems of mathematical substantiation of the method, study the conditions and rate of convergence, estimate the error. The paper considers the application of the method to specific functional equations

  19. An overview of control system for the ITER electron cyclotron system

    International Nuclear Information System (INIS)

    Purohit, D.; Bigelow, T.; Billava, D.; Bonicelli, T.; Caughman, J.; Darbos, C.; Denisov, G.; Gandini, F.; Gassmann, T.; Henderson, M.; Journeux, J.Y.; Kajiwara, K.; Kobayashi, N.; Nazare, C.; Oda, Y.; Omori, T.; Rao, S.L.; Rasmussen, D.; Ronden, D.; Saibene, G.

    2011-01-01

    The ITER electron cyclotron (EC) system having capability of up to 26 MW generated power at 170 GHz is being procured by 5 domestic agencies via 10 procurement arrangements. This implies diverse types of equipment and complex interface management. It also places a challenge on control system architecture to entertain the constraints of procurement slicing and meeting the overall functional requirement. The envisioned architecture is to use the local control units (supplied with each procurement) and a supervisory plant controller (by ITER). This offers a reliable control configuration for such delicate and complex EC plant system. The control system is envisioned to monitor the whole plant and perform automated tasks that are today performed via direct human intervention. For example, the automated gyrotron conditioning and active control of the EC plant to respond to requests from the plasma control system (PCS). This later aspect requires rapid shut down of the gyrotrons and power supplies, deviation of the actuators to direct the power from an equatorial to upper launcher and then restart of the power generation for rapid stabilization of the magneto hydrodynamic (MHD) instabilities that occur in high performance plasma operation. The plant controller will be designed for optimized performance with the PCS and the feedback control system used to actively control the power (with modulation capability up to 5 kHz) and launching direction for MHD stabilization.

  20. Transfer function restoration in 3D electron microscopy via iterative data refinement

    International Nuclear Information System (INIS)

    Sorzano, C O S; Marabini, R; Herman, G T; Censor, Y; Carazo, J M

    2004-01-01

    Three-dimensional electron microscopy (3D-EM) is a powerful tool for visualizing complex biological systems. As with any other imaging device, the electron microscope introduces a transfer function (called in this field the contrast transfer function, CTF) into the image acquisition process that modulates the various frequencies of the signal. Thus, the 3D reconstructions performed with these CTF-affected projections are also affected by an implicit 3D transfer function. For high-resolution electron microscopy, the effect of the CTF is quite dramatic and limits severely the achievable resolution. In this work we make use of the iterative data refinement (IDR) technique to ameliorate the effect of the CTF. It is demonstrated that the approach can be successfully applied to noisy data

  1. Design and integration of lower ports for ITER diagnostic systems

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: Natalia.casal@iter.org [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Bertalot, Luciano; Cheng, Hao; Drevon, Jean Marc; Duckworth, Philip; Giacomin, Thibaud; Guirao, Julio; Iglesias, Silvia [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Kochergin, Mikhail [IOFFE Institute, Saint Petersburg (Russian Federation); Martin, Alex [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); McCarron, Eddie [Oxford Technologies Ltd., Abingdon (United Kingdom); Mokeev, Alexander [Russian Federation Domestic Agency, Moscow (Russian Federation); Mota, Fernando [CIEMAT, Madrid (Spain); Penot, Christophe; Portales, Mickael [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); Kitazawa, Sin-iti [Japanese Domestic Agency, Naka (Japan); Sky, Jack [Oxford Technologies Ltd., Abingdon (United Kingdom); Suarez, Alejandro; Udintsev, Victor; Utin, Yuri [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046 – 13067 St Paul Lez Durance Cedex (France); and others

    2015-10-15

    Highlights: • Lower port structures are in its conceptual design phase. • Electromagnetic and seismic loads, will dominate all other mechanical loads. • Design allows diagnostics support, neutron shielding while and signals transmission. • Installation and maintenance operations are fully remote handling compatible. - Abstract: All around the ITER vacuum vessel, forty-four ports will provide access to the vacuum vessel for remote handling operations, diagnostic systems, heating, and vacuum systems: 18 upper ports, 17 equatorial ports, and 9 lower ports. Among the lower ports, three of them will be used for the remote handling installation of the ITER divertor. Once the divertor is in place, these ports will host various diagnostic systems mounted in the so-called diagnostic racks. The diagnostic racks must allow the support and cooling of the diagnostics, extraction of the required diagnostic signals, and providing access and maintainability while minimizing the leakage of radiation toward the back of the port where the humans are allowed to enter. A fully integrated inner rack, carrying the near plasma diagnostic components, will be an stainless steel structure, 4.2 m long, with a maximum weight of 10 t. This structure brings water for cooling and baking at maximum temperature of 240 °C and provides connection with gas, vacuum and electric services. Additional racks (placed away from plasma and not requiring cooling) may be required for the support of some particular diagnostic components. The diagnostics racks and its associated ex vessel structures, which are in its conceptual design phase, are being designed to survive the lifetime of ITER of 20 years. This paper presents the current state of development including interfaces, diagnostic integration, operation and maintenance, shielding requirements, remote handling, loads cases and discussion of the main challenges coming from the severe environment and engineering requirements.

  2. Conceptual design of the ECH upper launcher system for ITER

    International Nuclear Information System (INIS)

    Heidinger, R.; Bertizzolo, R.; Bruschi, A.; Chavan, R.; Cirant, S.; Collazos, A.; de Baar, M.; Elzendoorn, B.; Farina, D.; Fischer, U.; Gafert, J.; Gandini, F.; Gantenbein, G.; Goede, A.; Goodman, T.; Hailfinger, G.; Henderson, M.; Kasparek, W.; Kleefeldt, K.; Landis, J.-D.

    2009-01-01

    The challenge of developing the conceptual design of the ECH Upper Launcher system for MHD control in the ITER plasmas has been tackled by team of European Associations together with the European Domestic Agency ('F4E'). The launcher system has to meet the following requirements: (a) a mm-wave system extending from the interface to the transmission line up to the target absorption zone in the plasma and performing as an intelligent antenna; (b) a structural system integrating the mm-wave system and ensuring sufficient thermal and nuclear shielding; (c) port plug remote handling and testing capability ensuring high port plug system availability. The paper describes the reference launcher design. The mm-wave system is composed of waveguide and quasi-optical sections with a front steering system. An automated feedback control system is developed as a concept based on an assimilation procedure between predicted and diagnosed absorption location. The structural system consists of the blanket shield module, the port plug frame, and the internal shield for appropriate neutron shielding towards the launcher back-end. The specific advantages of a double walled structure are discussed with respect to adequate baking, to rigidity towards launcher deflection under plasma-generated loads and to removal of thermal loads, including nuclear ones. Basic studies of remote handling (RH) to validate design development are initiated using a virtual reality simulation backed by experimental validation, for which a launcher handling test facility (LHT) is set up as a full scale experimental site allowing furthermore thermohydraulic studies with ITER blanket water parameters.

  3. Iterative Decoding for an Optical CDMA based Laser communication System

    International Nuclear Information System (INIS)

    Kim, Jin Young; Kim, Eun Cheol; Cha, Jae Sang

    2008-01-01

    An optical CDMA(code division multiple access)based Laser communication system has attracted much attention since it requires minimal optical Laser signal processing and it is virtually delay free, while from the theoretical point of view, its performance depends on the auto and cross correlation properties of employed sequences. Various kinds of channel coding schemes for optical CDMA based Laser communication systems have been proposed and analyzed to compensate nonideal channel and receiver conditions in impaired photon channels. In this paper, we propose and analyze an iterative decoding of optical CDMA based Laser communication signals for both shot noise limited and thermal noise limited systems. It is assumed that optical channel is an intensity modulated (IM)channel and direct detection scheme is employed to detect the received optical signal. The performance is evaluated in terms of bit error probability and throughput. It is demonstrated that the BER and throughput performance is substantially improved with interleaver length for a fixed code rate and with alphabet size of PPM (pulse position modulation). Also, the BER and throughput performance is significantly enhanced with the number of iterations for decoding process. The results in this paper can be applied to the optical CDMA based Laser communication network with multiple access applications

  4. Iterative Decoding for an Optical CDMA based Laser communication System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Young; Kim, Eun Cheol [Kwangwoon Univ., Seoul (Korea, Republic of); Cha, Jae Sang [Seoul National Univ. of Technology, Seoul (Korea, Republic of)

    2008-11-15

    An optical CDMA(code division multiple access)based Laser communication system has attracted much attention since it requires minimal optical Laser signal processing and it is virtually delay free, while from the theoretical point of view, its performance depends on the auto and cross correlation properties of employed sequences. Various kinds of channel coding schemes for optical CDMA based Laser communication systems have been proposed and analyzed to compensate nonideal channel and receiver conditions in impaired photon channels. In this paper, we propose and analyze an iterative decoding of optical CDMA based Laser communication signals for both shot noise limited and thermal noise limited systems. It is assumed that optical channel is an intensity modulated (IM)channel and direct detection scheme is employed to detect the received optical signal. The performance is evaluated in terms of bit error probability and throughput. It is demonstrated that the BER and throughput performance is substantially improved with interleaver length for a fixed code rate and with alphabet size of PPM (pulse position modulation). Also, the BER and throughput performance is significantly enhanced with the number of iterations for decoding process. The results in this paper can be applied to the optical CDMA based Laser communication network with multiple access applications.

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

    International Nuclear Information System (INIS)

    Koizumi, Koichi; Hasegawa, Mitsuru; Yoshida, Kiyoshi

    1991-08-01

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

  6. Iterative solution of large sparse systems of equations

    CERN Document Server

    Hackbusch, Wolfgang

    2016-01-01

    In the second edition of this classic monograph, complete with four new chapters and updated references, readers will now have access to content describing and analysing classical and modern methods with emphasis on the algebraic structure of linear iteration, which is usually ignored in other literature. The necessary amount of work increases dramatically with the size of systems, so one has to search for algorithms that most efficiently and accurately solve systems of, e.g., several million equations. The choice of algorithms depends on the special properties the matrices in practice have. An important class of large systems arises from the discretization of partial differential equations. In this case, the matrices are sparse (i.e., they contain mostly zeroes) and well-suited to iterative algorithms. The first edition of this book grew out of a series of lectures given by the author at the Christian-Albrecht University of Kiel to students of mathematics. The second edition includes quite novel approaches.

  7. Progress in the ITER electron cyclotron heating and current drive system design

    Energy Technology Data Exchange (ETDEWEB)

    Omori, Toshimichi, E-mail: toshimichi.omori@iter.org [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Albajar, Ferran; Bonicelli, Tullio; Carannante, Giuseppe; Cavinato, Mario; Cismondi, Fabio [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Darbos, Caroline [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Denisov, Grigory [Institute of Applied Physics Russian Academy of Sciences, 46 Ulyanov Street, Nizhny Novgorod 603950 (Russian Federation); Farina, Daniela [Istituto di Fisica del Plasma, Association EURATOM-ENEA-CNR, Milano (Italy); Gagliardi, Mario [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Gandini, Franco; Gassmann, Thibault [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Goodman, Timothy [CRPP, Association EURATOM-Confédération Suisse, EPFL Ecublens, CH-1015 Lausanne (Switzerland); Hanson, Gregory [US ITER Project Office, ORNL, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831 (United States); Henderson, Mark A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Kajiwara, Ken [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193 (Japan); McElhaney, Karen [US ITER Project Office, ORNL, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831 (United States); Nousiainen, Risto [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Oda, Yasuhisa [Japan Atomic Energy Agency (JAEA), 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193 (Japan); Oustinov, Alexander [Institute of Applied Physics Russian Academy of Sciences, 46 Ulyanov Street, Nizhny Novgorod 603950 (Russian Federation); and others

    2015-10-15

    Highlights: • EC system is designed with an ability to upgrade in power to 28 MW then 40 MW. • The TL is capable of 3 buildings movements; ±15 mm displacements at the worst. • Improved power deposition access injecting 20 MW across nearly the entire plasma. • Ensured nuclear safety by appropriate definition of confinement boundaries. • Proposed I&C architecture for the overall EC plant was successfully reviewed. - Abstract: An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion.

  8. Progress in the ITER electron cyclotron heating and current drive system design

    International Nuclear Information System (INIS)

    Omori, Toshimichi; Albajar, Ferran; Bonicelli, Tullio; Carannante, Giuseppe; Cavinato, Mario; Cismondi, Fabio; Darbos, Caroline; Denisov, Grigory; Farina, Daniela; Gagliardi, Mario; Gandini, Franco; Gassmann, Thibault; Goodman, Timothy; Hanson, Gregory; Henderson, Mark A.; Kajiwara, Ken; McElhaney, Karen; Nousiainen, Risto; Oda, Yasuhisa; Oustinov, Alexander

    2015-01-01

    Highlights: • EC system is designed with an ability to upgrade in power to 28 MW then 40 MW. • The TL is capable of 3 buildings movements; ±15 mm displacements at the worst. • Improved power deposition access injecting 20 MW across nearly the entire plasma. • Ensured nuclear safety by appropriate definition of confinement boundaries. • Proposed I&C architecture for the overall EC plant was successfully reviewed. - Abstract: An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion.

  9. Iterative learning control for multi-agent systems coordination

    CERN Document Server

    Yang, Shiping; Li, Xuefang; Shen, Dong

    2016-01-01

    A timely guide using iterative learning control (ILC) as a solution for multi-agent systems (MAS) challenges, this book showcases recent advances and industrially relevant applications. Readers are first given a comprehensive overview of the intersection between ILC and MAS, then introduced to a range of topics that include both basic and advanced theoretical discussions, rigorous mathematics, engineering practice, and both linear and nonlinear systems. Through systematic discussion of network theory and intelligent control, the authors explore future research possibilities, develop new tools, and provide numerous applications such as power grids, communication and sensor networks, intelligent transportation systems, and formation control. Readers will gain a roadmap of the latest advances in the fields and can use their newfound knowledge to design their own algorithms.

  10. Iterative group splitting algorithm for opportunistic scheduling systems

    KAUST Repository

    Nam, Haewoon

    2014-05-01

    An efficient feedback algorithm for opportunistic scheduling systems based on iterative group splitting is proposed in this paper. Similar to the opportunistic splitting algorithm, the proposed algorithm adjusts (or lowers) the feedback threshold during a guard period if no user sends a feedback. However, when a feedback collision occurs at any point of time, the proposed algorithm no longer updates the threshold but narrows down the user search space by dividing the users into multiple groups iteratively, whereas the opportunistic splitting algorithm keeps adjusting the threshold until a single user is found. Since the threshold is only updated when no user sends a feedback, it is shown that the proposed algorithm significantly alleviates the signaling overhead for the threshold distribution to the users by the scheduler. More importantly, the proposed algorithm requires a less number of mini-slots than the opportunistic splitting algorithm to make a user selection with a given level of scheduling outage probability or provides a higher ergodic capacity given a certain number of mini-slots. © 2013 IEEE.

  11. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  12. ITER hydrogen isotope separation system conceptual design description

    International Nuclear Information System (INIS)

    Busigin, A.; Sood, S.K.; Kveton, O.K.; Dinner, P.J.; Murdoch, D.K.; Leger, D.

    1990-01-01

    This paper presents integrated hydrogen Isotope Separation System (ISS) designs for ITER based on requirements for plasma exhaust processing, neutral beam injection deuterium cleanup, pellet injector propellant detritiation, waste water detritiation, and breeding blanket detritiation. Specific ISS designs are developed for a machine with an aqueous lithium salt blanket (ALSB) and a machine with a solid ceramic breeding blanket (SBB). The differences in the ISS designs arising from the different blanket concepts are highlighted. It is found that the ISS designs for the two blanket concepts considered are very similar with the only major difference being the requirement for an additional large water distillation column for ALSB water detritiation. The extraction of tritium from the ALSB is based on flash evaporation to separate the blanket water from the dissolved Li salt, with the tritiated water then being fed to the ISS for detritiation. This technology is considered to be relatively well understood in comparison to front-end processes for SBB detritiation. In the design of the cryogenic distillation portion of the ISS, it was found that the tritium inventory could be very large (> 600 g) unless specific design measures were taken to reduce it. In the designs which are presented, the tritium inventory has been reduced to about 180 g, which is less than the ITER single-failure release limit of 200 g. Further design optimization and isolation of components is expected to reduce the inventory further. (orig.)

  13. Geometric feasibility of flexible cask transportation system for ITER

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. Iterative Repair Planning for Spacecraft Operations Using the Aspen System

    Science.gov (United States)

    Rabideau, G.; Knight, R.; Chien, S.; Fukunaga, A.; Govindjee, A.

    2000-01-01

    This paper describes the Automated Scheduling and Planning Environment (ASPEN). ASPEN encodes complex spacecraft knowledge of operability constraints, flight rules, spacecraft hardware, science experiments and operations procedures to allow for automated generation of low level spacecraft sequences. Using a technique called iterative repair, ASPEN classifies constraint violations (i.e., conflicts) and attempts to repair each by performing a planning or scheduling operation. It must reason about which conflict to resolve first and what repair method to try for the given conflict. ASPEN is currently being utilized in the development of automated planner/scheduler systems for several spacecraft, including the UFO-1 naval communications satellite and the Citizen Explorer (CX1) satellite, as well as for planetary rover operations and antenna ground systems automation. This paper focuses on the algorithm and search strategies employed by ASPEN to resolve spacecraft operations constraints, as well as the data structures for representing these constraints.

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

    International Nuclear Information System (INIS)

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

    1991-08-01

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

  16. Progress in the conceptual design of the ITER cask and plug remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Locke, Darren, E-mail: darren.locke@f4e.europa.eu [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); González Gutiérrez, Carmen; Damiani, Carlo [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Friconneau, Jean-Pierre; Martins, Jean-Pierre [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    Highlights: • The CPRHS is a complex system with a significant number of complicated interfaces. • Significant effort is being made to ensure that the system requirements are clearly defined. • This solution relates to planned operations and also anticipation of rescue operations. • With the CPRHS performing a safety function process control is being put in place. • All these factors will have a significant impact on the success of the CPRHS. - Abstract: One function of the ITER remote maintenance system is the transportation of in-vessel components and remote handling systems to and from the vacuum vessel and docking stations in the Hot Cell via dedicated galleries and lift. The cask and plug remote handling system (CPRHS) has been adopted as the solution to provide this nuclear confinement and transportation. This paper discusses the development of the conceptual design to-date and presents the processes being implemented to effectively control the subsequent CPRHS development. The CPRHS is a complex suite of systems with a significant number of interfaces with other ITER systems. Significant effort is being made to ensure that the system requirements are comprehensively defined and carefully managed and a feasible solution is developed – including planned and rescue operations. With the CPRHS performing a critical confinement function appropriate processes are being put in place to control the system development of the CPRHS. The expectation is that the combination of these factors will have a significant impact on the successful implementation of the CPRHS.

  17. Progress in the conceptual design of the ITER cask and plug remote handling system

    International Nuclear Information System (INIS)

    Locke, Darren; González Gutiérrez, Carmen; Damiani, Carlo; Friconneau, Jean-Pierre; Martins, Jean-Pierre

    2014-01-01

    Highlights: • The CPRHS is a complex system with a significant number of complicated interfaces. • Significant effort is being made to ensure that the system requirements are clearly defined. • This solution relates to planned operations and also anticipation of rescue operations. • With the CPRHS performing a safety function process control is being put in place. • All these factors will have a significant impact on the success of the CPRHS. - Abstract: One function of the ITER remote maintenance system is the transportation of in-vessel components and remote handling systems to and from the vacuum vessel and docking stations in the Hot Cell via dedicated galleries and lift. The cask and plug remote handling system (CPRHS) has been adopted as the solution to provide this nuclear confinement and transportation. This paper discusses the development of the conceptual design to-date and presents the processes being implemented to effectively control the subsequent CPRHS development. The CPRHS is a complex suite of systems with a significant number of interfaces with other ITER systems. Significant effort is being made to ensure that the system requirements are comprehensively defined and carefully managed and a feasible solution is developed – including planned and rescue operations. With the CPRHS performing a critical confinement function appropriate processes are being put in place to control the system development of the CPRHS. The expectation is that the combination of these factors will have a significant impact on the successful implementation of the CPRHS

  18. Engineering challenges and development of the ITER Blanket System and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, Alphonse Rene; Chappuis, Philippe; Hirai, Takeshi; Gicquel, Stefan

    2015-10-15

    The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor.

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

    International Nuclear Information System (INIS)

    Hu Liqin; Yuan Run; Chen Hongli; Bai Yunqing

    2012-01-01

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

  20. R and D of atmosphere detritiation system for ITER in JAEA

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takumi, E-mail: hayashi.takumi@jaea.go.jp [Tritium Technol. Gr.: Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki Pref. 319-1195 (Japan); Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamura, Hirofumi; Yamanishi, Toshihiko [Tritium Technol. Gr.: Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki Pref. 319-1195 (Japan); Perevezentsev, Alexander [Tritium Plant Gr.: ITER Organization, Cadarache Site (France)

    2010-12-15

    In order to establish an effective ITER atmosphere detritiation system (DS), JAEA (Japan Atomic Energy Agency) has investigated the performance and the durability of the system at various incident/accident conditions in support of finalizing the DS conceptual design through the ITER design review. The current DS as the Safety Important Component (SIC) has been discussed and mainly consists of catalytic reactors, wet scrubber columns (SCs), and blowers. The functional failure of the DS designed with SC was evaluated using a database of failure experiences of various valves, controllers, and components. Regarding the tritium release into the biggest confinement sector of the Tokamak gallery, this design is improved by more than two orders of magnitude compared to the original DS designed with molecular sieves (MSs) dryer beds in the 2001 design report. This improvement is achieved mainly by the reduction of frequency of valve operation, like MS dryers requiring periodical regeneration, and by the standardized module arrangement of the DS with SC.

  1. Iterative Observer-based Estimation Algorithms for Steady-State Elliptic Partial Differential Equation Systems

    KAUST Repository

    Majeed, Muhammad Usman

    2017-07-19

    Steady-state elliptic partial differential equations (PDEs) are frequently used to model a diverse range of physical phenomena. The source and boundary data estimation problems for such PDE systems are of prime interest in various engineering disciplines including biomedical engineering, mechanics of materials and earth sciences. Almost all existing solution strategies for such problems can be broadly classified as optimization-based techniques, which are computationally heavy especially when the problems are formulated on higher dimensional space domains. However, in this dissertation, feedback based state estimation algorithms, known as state observers, are developed to solve such steady-state problems using one of the space variables as time-like. In this regard, first, an iterative observer algorithm is developed that sweeps over regular-shaped domains and solves boundary estimation problems for steady-state Laplace equation. It is well-known that source and boundary estimation problems for the elliptic PDEs are highly sensitive to noise in the data. For this, an optimal iterative observer algorithm, which is a robust counterpart of the iterative observer, is presented to tackle the ill-posedness due to noise. The iterative observer algorithm and the optimal iterative algorithm are then used to solve source localization and estimation problems for Poisson equation for noise-free and noisy data cases respectively. Next, a divide and conquer approach is developed for three-dimensional domains with two congruent parallel surfaces to solve the boundary and the source data estimation problems for the steady-state Laplace and Poisson kind of systems respectively. Theoretical results are shown using a functional analysis framework, and consistent numerical simulation results are presented for several test cases using finite difference discretization schemes.

  2. On differential operators generating iterative systems of linear ODEs of maximal symmetry algebra

    Science.gov (United States)

    Ndogmo, J. C.

    2017-06-01

    Although every iterative scalar linear ordinary differential equation is of maximal symmetry algebra, the situation is different and far more complex for systems of linear ordinary differential equations, and an iterative system of linear equations need not be of maximal symmetry algebra. We illustrate these facts by examples and derive families of vector differential operators whose iterations are all linear systems of equations of maximal symmetry algebra. Some consequences of these results are also discussed.

  3. Dynamic Performance of the ITER Reactive Power Compensation System

    International Nuclear Information System (INIS)

    Sheng Zhicai; Fu Peng; Xu Liuwei

    2011-01-01

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

  4. Fourier acceleration of iterative processes in disordered systems

    International Nuclear Information System (INIS)

    Batrouni, G.G.; Hansen, A.

    1988-01-01

    Technical details are given on how to use Fourier acceleration with iterative processes such as relaxation and conjugate gradient methods. These methods are often used to solve large linear systems of equations, but become hopelessly slow very rapidly as the size of the set of equations to be solved increases. Fourier acceleration is a method designed to alleviate these problems and result in a very fast algorithm. The method is explained for the Jacobi relaxation and conjugate gradient methods and is applied to two models: the random resistor network and the random central-force network. In the first model, acceleration works very well; in the second, little is gained. We discuss reasons for this. We also include a discussion of stopping criteria

  5. Re-design of ITER Glow Discharge Cleaning system based on a fixed electrode concept

    International Nuclear Information System (INIS)

    Yang, Y.; Maruyama, S.; Kiss, G.; O’Connor, M.; Zhang, Y.; Pitts, R.A.; Shimada, M.; Fang, T.; Wang, Y.; Wang, M.; Pan, Y.; Li, B.; Li, L.

    2014-01-01

    Highlights: •This paper summarizes the approved new design of ITER GDC. •It is based on the fixed electrode design instead of the previous movable concept. •Estimates were made on the glow current density. •R and D topics on initiation, steady state and heat load were presented. •Other relevant considerations were listed in an exhaustive manner. -- Abstract: A new design of ITER Glow Discharge Cleaning (GDC) system based on a fixed electrode concept replaces the previous design which was based on a movable electrode integrated with the ITER In-Vessel-Viewing-System. Recently the conceptual design of the GDC system was reviewed successfully on the functions, safety, operation and maintenance. The design proposed was checked against the requirements and found to be feasible. This paper gives an overall description of the requirements from physics and operation viewpoints and introduces the design at the conceptual level. Main R and D activities are listed and summarized. Further detailed studies are to be performed in the following design stage

  6. Development of laser cutting/welding system for remote maintenance of ITER manifold

    Energy Technology Data Exchange (ETDEWEB)

    Yamaoka, Hiroto; Tsuchiya, Kazuyuki; Awano, Toshihiko [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan); Oka, Kiyoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-09-01

    A remote pipe cutting/welding system using a YAG laser was designed and fabricated for the maintenance of the main structural parts of ITER (International Thermonuclear Experimental Reactor), and a mock-up test carried out. The functions of this system are to cut 100A x Sch 40 pipes of SUS316L by internal access, to adjust the core gap between the as-cut pipe and new pipe, and to weld the pipes automatically. The core gap of the pipes could be decreased within the proper welding conditions by the mock-up test, and sound beads were obtained. (author)

  7. Iterative Cellular Screening System for Nanoparticle Safety Testing

    Directory of Open Access Journals (Sweden)

    Franziska Sambale

    2015-01-01

    Full Text Available Nanoparticles have the potential to exhibit risks to human beings and to the environment; due to the wide applications of nanoproducts, extensive risk management must not be neglected. Therefore, we have constructed a cell-based, iterative screening system to examine a variety of nanoproducts concerning their toxicity during development. The sensitivity and application of various cell-based methods were discussed and proven by applying the screening to two different nanoparticles: zinc oxide and titanium dioxide nanoparticles. They were used as benchmarks to set up our methods and to examine their effects on mammalian cell lines. Different biological processes such as cell viability, gene expression of interleukin-8 and heat shock protein 70, as well as morphology changes were investigated. Within our screening system, both nanoparticle suspensions and coatings can be tested. Electric cell impedance measurements revealed to be a good method for online monitoring of cellular behavior. The implementation of three-dimensional cell culture is essential to better mimic in vivo conditions. In conclusion, our screening system is highly efficient, cost minimizing, and reduces the need for animal studies.

  8. Alternate design of ITER cryostat skirt support system

    International Nuclear Information System (INIS)

    Pandey, Manish Kumar; Gupta, Girish Kumar; Bhardwaj, Anil Kumar; Jha, Saroj Kumar

    2015-01-01

    The skirt support of ITER cryostat is a support system which takes all the load of cryostat cylinder and dome during normal and operational condition. The present design of skirt support has full penetration weld joints at the bottom (shell to horizontal plate joint). To fulfill the requirements of tolerances and control the welding distortions, we have proposed to change the full penetration weld into fillet weld. A detail calculation is done to check the feasibility and structural impact due to proposed design. The calculations provide the size requirements of fillet weld. To verify the structural integrity during most severe load case, finite element analysis (FEA) has been done in line with ASME section VIII division 2. By FEA 'Plastic Collapse' and 'Local Failure' modes has been assessed. 5° sector of skirt clamp has been modeled in CATIA V5 R21 and used in FEA. Fillet weld at shell to horizontal plate joint has been modeled and symmetry boundary condition at ± 2.5° applied. 'Elastic Plastic Analysis' has been performed for the most severe loading case i.e. Category IV loading. The alternate design of Cryostat Skirt support system has been found safe by analysis against Plastic collapse and Local Failure Modes with load proportionality factor 2.3. Alternate design of Cryostat skirt support system has been done and validated by FEA. As per alternate design, the proposal of fillet weld has been implemented in manufacturing. (author)

  9. The upgrade of the DIII-D EC system using 120 GHz ITER gyrotrons

    International Nuclear Information System (INIS)

    Callis, R.W.; Lohr, J.; Gorelov, I.A.; Ponce, D.; Kajiwara, K.; Tooker, J.F.

    2005-01-01

    The planned growth in the EC system on DIII-D over the next few years requires the installation of two depressed collector gyrotrons, a high voltage power supply, two low loss transmission lines, and the required support equipment. This new DIII-D EC equipment could be made identical to the ITER EC system requirements. By building the DIII-D hardware to the ITER specifications, it will allow ITER to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER

  10. Assistance tools for generic definition of ITER maintenance tasks and scenarios in advanced supervisory control systems

    International Nuclear Information System (INIS)

    Zieba, Stéphane; Russotto, François-Xavier; Da Silva Simoes, Max; Measson, Yvan

    2013-01-01

    Highlights: ► Improve supervisory control systems for ITER in-vessel and hot cell maintenance. ► Optimize remote handling operations effectiveness, reliability and safety. ► Provide a generic description of the maintenance tasks and scenarios. ► Development of context-based assistances for operators and supervisor. ► Improvement of operator's situation awareness. -- Abstract: This paper concerns the improvement of supervisory control systems in the context of remote handling for the maintenance tasks in ITER. This work aims at providing a single formalism and tools to define in a generic way the ITER maintenance tasks and scenarios for in-vessel and hot cell operations. A three-layered approach is proposed to model these tasks and scenarios. Physical actions are defined for the scene elements. From these physical actions, behaviours are defined to represent high-level functionalities. Finally, interaction modes define the way that behaviours are achieved in terms of human–machine interactions. Case study concerning the blanket maintenance procedure is discussed concerning the contributions of the descriptive model and the context-based assistances to the activities of supervisory control

  11. Design of fast tuning elements for the ITER ICH system

    International Nuclear Information System (INIS)

    Swain, D.W.; Goulding, R.H.

    1996-05-01

    The coupling between the ion cyclotron (IC) antenna and the ITER plasma (as expressed by the load resistance the antenna sees) will experience relatively fast variations due to plasma edge profile modifications. If uncompensated, these will cause an increase in the amount of power reflected back to the transmitter and ultimately a decrease in the amount of radio frequency (rf) power to the plasma caused by protective suppression of the amount of rf power generated by the transmitter. The goals of this task were to study several alternate designs for a tuning and matching (T ampersand M) system and to recommend some research and development (R ampersand D) tasks that could be carried out to test some of the most promising concepts. Analyses of five different T ampersand M configurations are presented in this report. They each have different advantages and disadvantages, and the choice among them must be made depending on the requirements for the IC system. Several general conclusions emerge from our study: The use of a hybrid splitter as a passive reflected-power dump [''edge localized mode (ELM)-dump''] appears very promising; this configuration will protect the rf power sources from reflected power during changes in plasma loading due to plasma motion or profile changes (e.g., ELM- induced changes in the plasma scrape-off region) and requires no active control of the rf system. Trade-offs between simplicity of design and capability of the system must be made. Simple system designs with few components near the antenna either have high voltages over considerable distances of transmission lines, or they are not easily tuned to operate at different frequencies. Designs using frequency shifts and/or fast tuning elements can provide fast matching over a wide range of plasma loading; however, the designs studied here require components near the antenna, complicating assembly and maintenance. Capacitor-tuned resonant systems may offer a good compromise

  12. Smooth values of the iterates of the Euler's Phi function

    OpenAIRE

    Lamzouri, Youness

    2005-01-01

    Let $\\phi(n)$ be the Euler-phi function, define $\\phi_0(n) = n$ and $\\phi_{k+1}(n)=\\phi(\\phi_{k}(n))$ for all $k\\geq 0$. We will determine an asymptotic formula for the set of integers $n$ less than $x$ for which $\\phi_k(n)$ is $y$-smooth, conditionally on a weak form of the Elliott-Halberstam conjecture.

  13. Iteration of certain exponential-like meromorphic functions

    Indian Academy of Sciences (India)

    68

    Complex dynamics studies the Fatou sets and the. Julia sets of meromorphic functions. The Fatou set is open by definition and each of its maximally connected subset is known as a Fatou component. A Fatou component U is called p-periodic if p is the smallest natural number satisfying f p(U) ⊆ U. If for all the points z in U, ...

  14. Japanese contribution of ITER PF systems design during CDA

    International Nuclear Information System (INIS)

    Azumi, Masafumi; Hasegawa, Mitsuru; Kameari, Akihisa

    1992-03-01

    This report presents the brief summary of major part of Japanese contributions to the poloidal field coil systems design during conceptual design activity (CDA). The followings are studied; layout of poloidal field coil system, effects of plasma triangularity and current profile, alternative configurations such as semi-double null and single null divertor plasmas apart from the reference double null plasma, magnetic line tracing study for the detailed examination of the divertor plasmas, repulsive force on the center solenoid for the volt-second saving operation scenario, general criteria of vertical stability by passive structure, stability performance of two types of passive structure, saddle loop type and twin loop type, in-vessel coil current and voltage evaluation required for active control of plasma verrical position, TOLFEX method for the identification of plasma shape and position for the control of ITER plasma, control of horizontal plasma position during breakdown, slow control algorithm for the plasma shape, means to reduce the effective divertor peak heat load, separatrix sweeping both by SC coils and in-vessel coils, feasibility to generate the ergodic magnetic layer by in-vessel coils for the vertical position control, simulation of plasma dynamic behavior after the thermal quench by TSC code. Many of the results are not fully complete, while should be sufficient o demonstrate the feasibility of the CDA design. (J.P.N.)

  15. A General Iterative Method for a Nonexpansive Semigroup in Banach Spaces with Gauge Functions

    Directory of Open Access Journals (Sweden)

    Kamonrat Nammanee

    2012-01-01

    Full Text Available We study strong convergence of the sequence generated by implicit and explicit general iterative methods for a one-parameter nonexpansive semigroup in a reflexive Banach space which admits the duality mapping Jφ, where φ is a gauge function on [0,∞. Our results improve and extend those announced by G. Marino and H.-K. Xu (2006 and many authors.

  16. Iterative solution of linear systems in the 20­th century

    NARCIS (Netherlands)

    Saad, Y.; Vorst, H.A. van der

    2000-01-01

    This paper sketches the main research developments in the area of iterative methods for solving linear systems during the 20th century. Although iterative methods for solving linear systems find their origin in the early nineteenth century (work by Gauss), the field has seen an explosion of

  17. An Overview Of The ITER In-Vessel Coil Systems

    International Nuclear Information System (INIS)

    Heitzenroeder, P.J.; Brooks, A.W.; Chrzanowski, J.H.; Dahlgren, F.; Hawryluk, R.J.; Loesser, G.D.; Neumeyer, C.; Mansfield, C.; Smith, J.P.; Schaffer, M.; Humphreys, D.; Cordier, J.J.; Campbell, D.; Johnson, G.A.; Martin, A.; Rebut, P.H.; Tao, J.O.; Fogarty, P.J.; Nelson, B.E.; Reed, R.P.

    2009-01-01

    ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable 'natural' small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

  18. Approximate inverse preconditioning of iterative methods for nonsymmetric linear systems

    Energy Technology Data Exchange (ETDEWEB)

    Benzi, M. [Universita di Bologna (Italy); Tuma, M. [Inst. of Computer Sciences, Prague (Czech Republic)

    1996-12-31

    A method for computing an incomplete factorization of the inverse of a nonsymmetric matrix A is presented. The resulting factorized sparse approximate inverse is used as a preconditioner in the iterative solution of Ax = b by Krylov subspace methods.

  19. Identifying an unknown function in a parabolic equation with overspecified data via He's variational iteration method

    International Nuclear Information System (INIS)

    Dehghan, Mehdi; Tatari, Mehdi

    2008-01-01

    In this research, the He's variational iteration technique is used for computing an unknown time-dependent parameter in an inverse quasilinear parabolic partial differential equation. Parabolic partial differential equations with overspecified data play a crucial role in applied mathematics and physics, as they appear in various engineering models. The He's variational iteration method is an analytical procedure for finding solutions of differential equations, is based on the use of Lagrange multipliers for identification of an optimal value of a parameter in a functional. To show the efficiency of the new approach, several test problems are presented for one-, two- and three-dimensional cases

  20. Towards a preliminary design of the ITER plasma control system architecture

    International Nuclear Information System (INIS)

    Treutterer, W.; Rapson, C.J.; Raupp, G.; Snipes, J.; Vries, P. de; Winter, A.; Humphreys, D.A.; Walker, M.; Tommasi, G. de; Cinque, M.; Bremond, S.; Moreau, P.; Nouailletas, R.; Felton, R.

    2017-01-01

    Highlights: • ITER control requirements and use scenarios for initial plasma operation have been analysed. • Basic choices from conceptual design could be confirmed. • Architectural design considers dynamic structure changes. • All PCS components are integrated in an exception handling hierarchy. - Abstract: Design of the ITER plasma control system is proceeding towards its next – preliminary design – stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is suitable for realistic use cases. This verification is conducted with focus on the concrete use cases of early operation and first plasma, since these phases are mandatory for ITER startup. In particular, detailed control requirements and functions for commissioning and first plasma operation including breakdown, burn-through and ramp-up in L-mode, as well as for planned or exceptional shutdown are identified. Control functions related to those operational phases and the underlying control system architecture are modeled. The goal is to check whether the flexibility of the conceptual architectural approach is adequate also in consideration of the more elaborate definitions for control functions and their interactions. In addition, architecture shall already be prepared for extension to H-mode operation and burn-control, even if the related control functions are only roughly defined at the moment. As a consequence, the architectural design is amended where necessary and converted into base components and infrastructure services allowing to deploy control and exception handling algorithms for the concrete first-plasma operation.

  1. Towards a preliminary design of the ITER plasma control system architecture

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Rapson, C.J.; Raupp, G. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Snipes, J.; Vries, P. de; Winter, A. [ITER Organization, Route de Vinon sur Verdon, 13067 St Paul Lez Durance (France); Humphreys, D.A.; Walker, M. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Tommasi, G. de; Cinque, M. [CREATE/Università di Napoli Federico II, Napoli (Italy); Bremond, S.; Moreau, P.; Nouailletas, R. [Association CEA pour la Fusion Contrôlée, CEA Cadarache, 13108 St Paul les Durance (France); Felton, R. [CCFE Fusion Association, Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire, OX14 3DB (United Kingdom)

    2017-02-15

    Highlights: • ITER control requirements and use scenarios for initial plasma operation have been analysed. • Basic choices from conceptual design could be confirmed. • Architectural design considers dynamic structure changes. • All PCS components are integrated in an exception handling hierarchy. - Abstract: Design of the ITER plasma control system is proceeding towards its next – preliminary design – stage. During the conceptual design in 2013 an overall assessment of high-level control tasks and their relationships has been conducted. The goal of the preliminary design is to show, that a reasonable implementation of the proposed concepts exists which fulfills the high-level requirements and is suitable for realistic use cases. This verification is conducted with focus on the concrete use cases of early operation and first plasma, since these phases are mandatory for ITER startup. In particular, detailed control requirements and functions for commissioning and first plasma operation including breakdown, burn-through and ramp-up in L-mode, as well as for planned or exceptional shutdown are identified. Control functions related to those operational phases and the underlying control system architecture are modeled. The goal is to check whether the flexibility of the conceptual architectural approach is adequate also in consideration of the more elaborate definitions for control functions and their interactions. In addition, architecture shall already be prepared for extension to H-mode operation and burn-control, even if the related control functions are only roughly defined at the moment. As a consequence, the architectural design is amended where necessary and converted into base components and infrastructure services allowing to deploy control and exception handling algorithms for the concrete first-plasma operation.

  2. Mechanical testing of a FW panel attachment system for ITER

    International Nuclear Information System (INIS)

    Oliva, Vladislav; Vaclavik, Jaroslav; Materna, Ales; Lorenzetto, Patrick; Furmanek, Andreas

    2009-01-01

    An objective of experiments and finite element simulations was to check the stiffness, the strength and the fatigue resistance of the attachment of the First Wall panels onto a shield block of blanket modules according to the ITER 2001 design. The panel has a poloidal key at the rear side (in so-called option A with the rear access bolting) and it is attached by means of special studs located on a key-way in the shield block. Special device for a test of stud tensile pre-load relaxation during a thermal cycling was developed. True-to-scale panels, the shield block mock-up and simplified studs were fabricated and the assembly was loaded alternatively by radial moment, poloidal force or poloidal moment simulating the loading during off-normal plasma operations. Thermal cycling led to an acceptable stud pre-load relaxation. Mechanical cycling caused neither the pre-load relaxation nor the loss of the contact in the key-way nor a damage of the attachment system. The combination of poloidal moment and radial force during vertical displacement events (VDEs) seems to be a most dangerous case because it could lead to the loss of the key-key-way contact.

  3. ITER - torus vacuum pumping system remote handling issues

    International Nuclear Information System (INIS)

    Stringer, J.

    1992-11-01

    This report describes design issues concerning remote maintenance of the ITER torus vacuum pumping system. Key issues under investigation in this report are bearings for inert gas operation, transporter integration options, cryopump access, gate valve maintenance frequency, tritium effects on materials, turbomolecular pump design, and remote maintenance. Alternative bearing materials are explored for inert gas operation. Encapsulated motors and rotary feedthroughs offer an alternative option where space requirements are restrictive. A number of transporter options are studied. The preferred scheme depends on the shielded reconfigured ducts to prevent streaming and activation of RH (remote handling) equipment. A radiation mapping of the cell is required to evaluate this concept. Valve seal and bellow life are critical issues and need to be evaluated, as they have a direct bearing on the provision of adequate RH equipment to meet scheduled and unscheduled maintenance outages. The limited space on the inboard side of the cryopumps for RH equipment access requires a reconfigured duct and manifold. A modified shielded duct arrangement is proposed, which would provide more access space, reduced activation of components, and the potential for improved valve seal life. Work at Mound Laboratories has shown the adverse effects of tritium on some bearing lubricants. Silicone-based lubricants should be avoided. (11 refs., 2 tabs., 31 figs.)

  4. 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.

  5. Design status of the ITER ECRH upper launcher mm-wave system

    Energy Technology Data Exchange (ETDEWEB)

    Landis, J.-D. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom-Confederation Suisse, 1015 Lausanne (Switzerland)], E-mail: jean-daniel.landis@epfl.ch; Chavan, R.; Bertizzolo, R.; Collazos, A.; Dolizy, F.; Felici, F.; Sanchez, F. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom-Confederation Suisse, 1015 Lausanne (Switzerland); Henderson, M. [ITER, Organization, Cadarache Centre, Saint Paul Lez Durance (France)

    2009-06-15

    The purpose of the ITER electron cyclotron resonance heating (ECRH) upper launcher (UL), or antennae will be to provide localised current drive by accurately directing mm-wave beams up to 2MW, out of the four allocated upper port plugs, at chosen rational magnetic flux surfaces in order to stabilise neoclassical tearing modes (NTMs). This paper will present an overview of the UL, with emphasis on the mm-wave components. The mm-wave layout includes corrugated waveguide sections and a quasi-optical path with both focusing mirrors and plane steering mirrors. One of the essential components of the UL is the Steering Mechanism Assembly (SMA), providing variable poloidal injection angles fulfilling high deposition accuracy requirements at the plasma location. The Actuator principle and rotor bearings are frictionless and backlash free, avoiding tribological difficulties such as stickslip and seizure. The underlying working principle is the use of mechanically compliant structures. Validation and proof testing of the steering principle is achieved with an uncooled first prototype demonstrator. A second prototype is currently being manufactured, comprising the functionalities needed for the ITER compatible system such as water cooling and high power mm-wave compatibility. In order to perform the fatigue tests of the actuator bellows, a test facility has been built, under ITER-like vacuum and temperature working conditions. Results of the cyclic fatigue tests are compared to the various manufacturer standards and codes, combining stress and strain controlled material fatigue properties.

  6. Fault Analysis of ITER Coil Power Supply System

    International Nuclear Information System (INIS)

    Song, In Ho; Jun, Tao; Benfatto, Ivone

    2009-01-01

    The ITER magnet coils are all designed using superconductors with high current carrying capability. The Toroidal Field (TF) coils operate in a steadystate mode with a current of 68 kA and discharge the stored energy in case of quench with using 9 interleaved Fast Discharge Units (FDUs). The Central Solenoid (CS) coils and Poloidal Field (PF) coils operate in a pulse mode with currents of up to 45 kA and require fast variation of currents inducing more than 10 kV during normal operation on the coil terminals using Switching Network (SN) systems (CSs, PF1 and 6) and Booster and VS converters (PF2 to 5), which are series connected to Main converters. SN and FDU systems comprise high current DC circuit breakers and resistors for generating high voltage (SN) and to dissipate magnetic energy (FDUs). High transient voltages can arise due to the switching operation of SN and FD and the characteristics of resistors and stray components of DC distribution systems. Also, faults in power supply control such as shorts or grounding faults can produce higher voltages between terminals and between terminal and ground. Therefore, the design of the coil insulation, coil terminal regions, feeders, feed throughs, pipe breaks and instrumentation must take account of these high voltages during normal and abnormal conditions. Voltage insulation level can be defined and it is necessary to test the coils at higher voltages, to be sure of reliable performance during the lifetime of operation. This paper describes the fault analysis of the TF, CS and PF coil power supply systems, taking account of the stray parameter of the power supply and switching systems and inductively coupled superconducting coil models. Resistor grounding systems are included in the simulation model and all fault conditions such as converter hardware and software faults, switching system hardware and software faults, DC short circuits and single grounding faults are simulated. The occurrence of two successive faults

  7. Overview of the preliminary design of the ITER plasma control system

    Science.gov (United States)

    Snipes, J. A.; Albanese, R.; Ambrosino, G.; Ambrosino, R.; Amoskov, V.; Blanken, T. C.; Bremond, S.; Cinque, M.; de Tommasi, G.; de Vries, P. C.; Eidietis, N.; Felici, F.; Felton, R.; Ferron, J.; Formisano, A.; Gribov, Y.; Hosokawa, M.; Hyatt, A.; Humphreys, D.; Jackson, G.; Kavin, A.; Khayrutdinov, R.; Kim, D.; Kim, S. H.; Konovalov, S.; Lamzin, E.; Lehnen, M.; Lukash, V.; Lomas, P.; Mattei, M.; Mineev, A.; Moreau, P.; Neu, G.; Nouailletas, R.; Pautasso, G.; Pironti, A.; Rapson, C.; Raupp, G.; Ravensbergen, T.; Rimini, F.; Schneider, M.; Travere, J.-M.; Treutterer, W.; Villone, F.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

    2017-12-01

    An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

  8. Enhanced configuration of a water detritiation system; impact on ITER Isotope Separation System based cryogenic distillation

    Energy Technology Data Exchange (ETDEWEB)

    Cristescu, Ion, E-mail: ion.cristescu@kit.edu

    2016-11-01

    Highlights: • An enhanced configuration of ITER WDS has been developed. • The proposed configuration allows minimization of hazards due to the reduction of tritium inventory. • The load on the tritium recovery system (ITER ISS) is minimized with benefits on mitigation of the explosion hazards. - Abstract: Tritiated water is generated in the ITER systems by various sources and may contain deuterium and tritium at various concentrations. The reference process for the ITER Water Detritiation System is based on Combined Electrolysis Catalytic Exchange (CECE) configuration. During long time operation of the CECE process, the accumulation of deuterium in the electrolysis unit and consequently along the Liquid Phase Catalytic Exchange (LPCE) column is unavoidable with consequences on the overall detritiation factor of the system. Beside the deuterium issue in the process, the large amount of the tritiated water with tritium activity up to 500 Ci/kg in the electrolysis cells is a concern from the safety aspect of the plant. The enhanced configuration of a system for processing tritiated water allows mitigation of the effects due to deuterium accumulation and also reduction of tritium inventory within the electrolysis system. In addition the benefits concerning to the interface between the water detritiation system and tritium recovery based cryogenic distillation are also presented.

  9. FMECA about pre-treatment system for purge gas of test blanket module in ITER

    International Nuclear Information System (INIS)

    Fu Wanfa; Luo Deli; Tang Tao

    2012-01-01

    The pre-treatment system for purge gas of TBM will be installed in Port Cell for installing TBM in ITER, the function of which includes filtering purge gas, removing HTO, cooling, and adjusting flow rate, etc. The purge gas treated will be conveyed into TES (Tritium Extraction System). The technological process and system components in pre-treatment system were introduced. Tritium releasing risk was regarded as failure criterion; failure mode, effects and criticality analysis (FMECA) were carried out and several weaknesses or failure mode in the system were found. Besides, risk priority number (RPN) and failure mode criticality were calculated. Finally, some design improvement measures and usage compensation measures were given. At last, four important potential failure modes were found out. The analysis will provide the design basis for reducing risk of excessive tritium releasing, which is also a useful assist for safety analysis about other tritium system. (authors)

  10. Parameter study on Japanese proposal of ITER hydrogen isotope separation system

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Enoeda, Mikio; Tanaka, Shigeru; Ohokawa, Yoshinao; Ohara, Atsushi; Nagakura, Masaaki; Naito, Taisei; Nagashima, Kazuhiro.

    1991-01-01

    As part of Japanese design contribution in the ITER activity, conceptual design of an entire ITER tritium system and their safety analysis have been carried out through the three-year period since 1988. The tritium system includes the following subsystems; - Fuelling (gas puffing and pellet injection) subsystem, - Torus vacuum pumping subsystem, - Plasma exhaust gas purification subsystem, - Hydrogen isotope separation subsystem, - NBI gas processing subsystem, - Blanket tritium recovery subsystem, - Tritiated water processing subsystem, - Tritium safety subsystem. Hydrogen isotope separation system is a key subsystem in the ITER tritium system because it is connected to all above subsystems. This report describes an analytical study on the Japanese concept of hydrogen isotope separation system. (author)

  11. The targeted heating and current drive applications for the ITER electron cyclotron system

    Energy Technology Data Exchange (ETDEWEB)

    Henderson, M.; Darbos, C.; Gandini, F.; Gassmann, T.; Loarte, A.; Omori, T.; Purohit, D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Saibene, G.; Gagliardi, M. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Farina, D.; Figini, L. [Istituto di Fisica del Plasma CNR, 20125 Milano (Italy); Hanson, G. [US ITER Project Office, ORNL, 1055 Commerce Park, PO Box 2008, Oak Ridge, Tennessee 37831 (United States); Poli, E. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Takahashi, K. [Japan Atomic Energy Agency (JAEA), Naka, Ibaraki 311-0193 (Japan)

    2015-02-15

    A 24 MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20 MW of this power to the plasma for Heating and Current Drive (H and CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H and CD systems. An initial attempt has been developed at prioritizing the allocated H and CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (∼12 MA), and Advanced (∼9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems.

  12. Current status of ITER I&C system as integration begins

    Energy Technology Data Exchange (ETDEWEB)

    Davis, William, E-mail: william.davis@iter.org [ITER Organisation, Route de Vinon-sur Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Wallander, Anders [ITER Organisation, Route de Vinon-sur Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Yonekawa, Izuru [Nippon Advanced Technology Ltd., 3129-45 Hibara Muramatsu, Tokai, Naka-gun, Ibaraki 319-1112 (Japan)

    2016-11-15

    Highlights: • The ITER I&C system is organisationally complicated and technically challenging. • Standard technologies for the ITER I&C systems have been selected. • Supply of non-standard technologies will cause serious issues. • Differing levels of design maturity of plant I&C systems is a serious challenge. • Systems are in the final stages of design and are being delivered to site. - Abstract: The ITER I&C system is organisationally complicated and technically challenging, and integrating its many sub-systems into a single coherent system is critical for the ITER project to meet its objectives. This paper explains the integration risks being faced now and anticipated in the near future. Standardisation initiatives by the ITER central team to mitigate these risks are described. The paper also presents the architecture of the ITER I&C system, the current status of design and manufacture key developments made in recent years, and the current and future activities of the central I&C teams. Finally, a short description is given of the plant I&C systems that will be delivered to ITER in the near future.

  13. More on Generalizations and Modifications of Iterative Methods for Solving Large Sparse Indefinite Linear Systems

    Directory of Open Access Journals (Sweden)

    Jen-Yuan Chen

    2014-01-01

    Full Text Available Continuing from the works of Li et al. (2014, Li (2007, and Kincaid et al. (2000, we present more generalizations and modifications of iterative methods for solving large sparse symmetric and nonsymmetric indefinite systems of linear equations. We discuss a variety of iterative methods such as GMRES, MGMRES, MINRES, LQ-MINRES, QR MINRES, MMINRES, MGRES, and others.

  14. Comparison results on preconditioned SOR-type iterative method for Z-matrices linear systems

    Science.gov (United States)

    Wang, Xue-Zhong; Huang, Ting-Zhu; Fu, Ying-Ding

    2007-09-01

    In this paper, we present some comparison theorems on preconditioned iterative method for solving Z-matrices linear systems, Comparison results show that the rate of convergence of the Gauss-Seidel-type method is faster than the rate of convergence of the SOR-type iterative method.

  15. Iterative Learning Control design for uncertain and time-windowed systems

    NARCIS (Netherlands)

    Wijdeven, van de J.J.M.

    2008-01-01

    Iterative Learning Control (ILC) is a control strategy capable of dramatically increasing the performance of systems that perform batch repetitive tasks. This performance improvement is achieved by iteratively updating the command signal, using measured error data from previous trials, i.e., by

  16. RAMI analysis of the ITER LFS CTS system

    DEFF Research Database (Denmark)

    Infante, V.; Henriques, E.; Gonçalves, B.

    2017-01-01

    with the methodology pursued. The Functional Analysis, developed both at system and sub-system level, are the major inputs for the RAMI analysis. A systematic approach has been used, and significant design assumptions have been made due to the lack of knowledge and definition inherent to preliminary design stages...

  17. Tomographic reconstruction by using FPSIRT (Fast Particle System Iterative Reconstruction Technique)

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Icaro Valgueiro M.; Melo, Silvio de Barros; Dantas, Carlos; Lima, Emerson Alexandre; Silva, Ricardo Martins; Cardoso, Halisson Alberdan C., E-mail: ivmm@cin.ufpe.br, E-mail: sbm@cin.ufpe.br, E-mail: rmas@cin.ufpe.br, E-mail: hacc@cin.ufpe.br, E-mail: ccd@ufpe.br, E-mail: eal@cin.ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    The PSIRT (Particle System Iterative Reconstruction Technique) is a method of tomographic image reconstruction primarily designed to work with configurations suitable for industrial applications. A particle system is an optimization technique inspired in real physical systems that associates to the reconstructing material a set of particles with certain physical features, subject to a force eld, which can produce movement. The system constantly updates the set of particles by repositioning them in such a way as to approach the equilibrium. The elastic potential along a trajectory is a function of the difference between the attenuation coefficient in the current configuration and the corresponding input data. PSIRT has been successfully used to reconstruct simulated and real objects subject to sets of parallel and fanbeam lines in different angles, representing typical gamma-ray tomographic arrangements. One of PSIRT's limitation was its performance, too slow for real time scenarios. In this work, it is presented a reformulation in PSIRT's computational model, which is able to grant the new algorithm, the FPSIRT - Fast System Iterative Reconstruction Technique, a performance up to 200-time faster than PSIRT's. In this work a comparison of their application to real and simulated data from the HSGT, High Speed Gamma Tomograph, is presented. (author)

  18. Tomographic reconstruction by using FPSIRT (Fast Particle System Iterative Reconstruction Technique)

    International Nuclear Information System (INIS)

    Moreira, Icaro Valgueiro M.; Melo, Silvio de Barros; Dantas, Carlos; Lima, Emerson Alexandre; Silva, Ricardo Martins; Cardoso, Halisson Alberdan C.

    2015-01-01

    The PSIRT (Particle System Iterative Reconstruction Technique) is a method of tomographic image reconstruction primarily designed to work with configurations suitable for industrial applications. A particle system is an optimization technique inspired in real physical systems that associates to the reconstructing material a set of particles with certain physical features, subject to a force eld, which can produce movement. The system constantly updates the set of particles by repositioning them in such a way as to approach the equilibrium. The elastic potential along a trajectory is a function of the difference between the attenuation coefficient in the current configuration and the corresponding input data. PSIRT has been successfully used to reconstruct simulated and real objects subject to sets of parallel and fanbeam lines in different angles, representing typical gamma-ray tomographic arrangements. One of PSIRT's limitation was its performance, too slow for real time scenarios. In this work, it is presented a reformulation in PSIRT's computational model, which is able to grant the new algorithm, the FPSIRT - Fast System Iterative Reconstruction Technique, a performance up to 200-time faster than PSIRT's. In this work a comparison of their application to real and simulated data from the HSGT, High Speed Gamma Tomograph, is presented. (author)

  19. Development of a virtual reality simulator for the ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi; Tesini, Alessandro

    2008-01-01

    The authors developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robotic simulation software, ENVISION. The simulator is connected to the control system of the manipulator, which was developed as part of the blanket maintenance system during the Engineering Design Activity (EDA), and can reconstruct the positions of the manipulator and blanket module using position data transmitted from motors through a LAN. In addition, it can provide virtual visual information (e.g., about the interface structures behind the blanket module) by making the module transparent on the screen. It can also be used for confirming a maintenance sequence before the actual operation. The simulator will be modified further, with addition of other necessary functions, and will finally serve as a prototype of the actual simulator for the blanket remote handling system, which will be procured as part of an in-kind contribution

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  2. Single image super-resolution based on approximated Heaviside functions and iterative refinement

    Science.gov (United States)

    Wang, Xin-Yu; Huang, Ting-Zhu; Deng, Liang-Jian

    2018-01-01

    One method of solving the single-image super-resolution problem is to use Heaviside functions. This has been done previously by making a binary classification of image components as “smooth” and “non-smooth”, describing these with approximated Heaviside functions (AHFs), and iteration including l1 regularization. We now introduce a new method in which the binary classification of image components is extended to different degrees of smoothness and non-smoothness, these components being represented by various classes of AHFs. Taking into account the sparsity of the non-smooth components, their coefficients are l1 regularized. In addition, to pick up more image details, the new method uses an iterative refinement for the residuals between the original low-resolution input and the downsampled resulting image. Experimental results showed that the new method is superior to the original AHF method and to four other published methods. PMID:29329298

  3. Design of ITER-FEAT RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.; Kobayashi, N.; Ioki, K.; Bibet, P.; Koch, R.; Chavan, R.; Tran, M.Q.; Takahashi, K.; Kuzikov, S.; Vdovin, V.

    2001-01-01

    Three radio frequency (RF) heating and current drive (H and CD) systems are being designed for ITER-FEAT: an electron cyclotron (EC), an ion cyclotron (IC) and a lower hybrid (LH) System. The launchers of the RF systems use four ITER equatorial ports and are fully interchangeable. They feature equal power outputs (20 MW/port), similar neutron shielding performance, and identical interfaces with the other machine components. An outline of the design is given in the paper. (author)

  4. On an Iteration Leading to a q-Analogue of the Digamma Function

    DEFF Research Database (Denmark)

    Berg, Christian; Petersen, Helle Bjerg

    2013-01-01

    We show that the q-Digamma function ψq for 0 < q < 1 appears in an iteration studied by Berg and Durán. This is connected with the determination of the probability measure νq on the unit interval with moments 1/n+1 k=1(1 − q)/(1 − qk), which are q-analogues of the reciprocals of the harmonic numb...

  5. Simulation-based design process for the verification of ITER remote handling systems

    International Nuclear Information System (INIS)

    Sibois, Romain; Määttä, Timo; Siuko, Mikko; Mattila, Jouni

    2014-01-01

    Highlights: •Verification and validation process for ITER remote handling system. •Simulation-based design process for early verification of ITER RH systems. •Design process centralized around simulation lifecycle management system. •Verification and validation roadmap for digital modelling phase. -- Abstract: The work behind this paper takes place in the EFDA's European Goal Oriented Training programme on Remote Handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. One of the projects of this programme focuses on the verification and validation (V and V) of ITER RH system requirements using digital mock-ups (DMU). The purpose of this project is to study and develop efficient approach of using DMUs in the V and V process of ITER RH system design utilizing a System Engineering (SE) framework. Complex engineering systems such as ITER facilities lead to substantial rise of cost while manufacturing the full-scale prototype. In the V and V process for ITER RH equipment, physical tests are a requirement to ensure the compliance of the system according to the required operation. Therefore it is essential to virtually verify the developed system before starting the prototype manufacturing phase. This paper gives an overview of the current trends in using digital mock-up within product design processes. It suggests a simulation-based process design centralized around a simulation lifecycle management system. The purpose of this paper is to describe possible improvements in the formalization of the ITER RH design process and V and V processes, in order to increase their cost efficiency and reliability

  6. 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

  7. Iterative Systems Biology for Medicine – time for advancing from network signature to mechanistic equations

    KAUST Repository

    Gomez-Cabrero, David; Tegner, Jesper

    2017-01-01

    The rise and growth of Systems Biology following the sequencing of the human genome has been astounding. Early on, an iterative wet-dry methodology was formulated which turned out as a successful approach in deciphering biological complexity

  8. Mechanical design of the ITER ion cyclotron heating launcher based on in-vessel tuning system

    Energy Technology Data Exchange (ETDEWEB)

    Vulliez, K. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul Lez Durance (France)], E-mail: karl.vulliez@cea.fr; Bosia, G. [Dipartimento di Fisica Generale, Universita di Torino (Italy); Agarici, G.; Beaumont, B.; Argouarch, A.; Mollard, P. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 St Paul Lez Durance (France); Testoni, P. [Electrical and Electronics Engineering Department, University of Cagliari (Italy); Maggiora, R.; Milanesio, D. [Dipartimento di Elettronica Politecnico di Torino (Italy)

    2007-10-15

    Since the release of the ITER ICRH system reference design report [ITER Final Design Report: DDD 5.1 -Ion Cyclotron and Current Drive System, July 2001], further design studies have been conducted. If the base of the reference design [Final Report on EFDA contract 04/1129, ITER ICRF antenna and Matching system design (Internalmatching), April 2005] is kept unchanged, several significant modifications have been proposed for a better efficiency and reliability. The increase of the poloidal order of the array and strong modifications of the matching system concept are the main changes. Technical aspects insufficiently covered in previous studies are also now worked out in detail, like the integration on a mid-plane port satisfying the constraints of the ITER environment.

  9. Electro-mechanical connection system for ITER in-vessel magnetic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Rizzolo, Andrea; Brombin, Matteo; Gonzalez, Winder [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Marconato, Nicolò, E-mail: nicolo.marconato@igi.cnr.it [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Peruzzo, Simone [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Arshad, Shakeib [Fusion for Energy, C/Josep Pla, 2, 08019 Barcelona (Spain); Ma, Yunxing; Vayakis, George [ITER Organization, Route de Vinon-sur-Verdon, 13067 St Paul Lez Durance (France); Williams, Adrian [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon, Oxon, OX14 1RL (United Kingdom)

    2016-11-01

    Highlights: • Latest status of the ITER “Generic In-Vessel Magnetic Platform” design activity. • Integration within the ITER In-Vessel configuration model. • Geometry optimization based on thermo-mechanical and magnetic field 3D calculation. • Assessment of the remote handling maintenance compatibility. - Abstract: This paper presents the preliminary design of the “In-Vessel Magnetic platform”, which is a subsystem of the magnetic diagnostics formed by all the components necessary for guaranteeing the thermo-mechanical interface of the actual magnetic sensors with the vacuum vessel (VV), their protection and the electrical connection to the in-vessel wiring for the transmission of the detected signal with a minimum level of noise. The design has been developed in order to comply with different functional requirements: the mechanical attachment to the VV; the electrical connection to the in-vessel wiring; efficient heat transfer to the VV; the compatibility with Remote Handling (RH) system for replacement; the integration of metrology features for post-installation control; the Electro Magnetic Interference (EMI) shielding from Electron Cyclotron Heating (ECH) stray radiation without compromising the sensor pass band (15 kHz). Significant effort has been dedicated to develop the CAD model, integrated within the ITER In-Vessel configuration model, taking care of the geometrical compliance with the Blanket modules (modified in order to accommodate the magnetic sensors in suitable grooves) and the RH compatibility. Thorough thermo-mechanical and electro-magnetic Finite Element Method (FEM) analyses have been performed to assess the reliability of the system in standard and off-normal operating conditions for the low frequency magnetic sensors.

  10. Scenario-based fitted Q-iteration for adaptive control of water reservoir systems under uncertainty

    Science.gov (United States)

    Bertoni, Federica; Giuliani, Matteo; Castelletti, Andrea

    2017-04-01

    Over recent years, mathematical models have largely been used to support planning and management of water resources systems. Yet, the increasing uncertainties in their inputs - due to increased variability in the hydrological regimes - are a major challenge to the optimal operations of these systems. Such uncertainty, boosted by projected changing climate, violates the stationarity principle generally used for describing hydro-meteorological processes, which assumes time persisting statistical characteristics of a given variable as inferred by historical data. As this principle is unlikely to be valid in the future, the probability density function used for modeling stochastic disturbances (e.g., inflows) becomes an additional uncertain parameter of the problem, which can be described in a deterministic and set-membership based fashion. This study contributes a novel method for designing optimal, adaptive policies for controlling water reservoir systems under climate-related uncertainty. The proposed method, called scenario-based Fitted Q-Iteration (sFQI), extends the original Fitted Q-Iteration algorithm by enlarging the state space to include the space of the uncertain system's parameters (i.e., the uncertain climate scenarios). As a result, sFQI embeds the set-membership uncertainty of the future inflow scenarios in the action-value function and is able to approximate, with a single learning process, the optimal control policy associated to any scenario included in the uncertainty set. The method is demonstrated on a synthetic water system, consisting of a regulated lake operated for ensuring reliable water supply to downstream users. Numerical results show that the sFQI algorithm successfully identifies adaptive solutions to operate the system under different inflow scenarios, which outperform the control policy designed under historical conditions. Moreover, the sFQI policy generalizes over inflow scenarios not directly experienced during the policy design

  11. On Green's function retrieval by iterative substitution of the coupled Marchenko equations

    Science.gov (United States)

    van der Neut, Joost; Vasconcelos, Ivan; Wapenaar, Kees

    2015-11-01

    Iterative substitution of the coupled Marchenko equations is a novel methodology to retrieve the Green's functions from a source or receiver array at an acquisition surface to an arbitrary location in an acoustic medium. The methodology requires as input the single-sided reflection response at the acquisition surface and an initial focusing function, being the time-reversed direct wavefield from the acquisition surface to a specified location in the subsurface. We express the iterative scheme that is applied by this methodology explicitly as the successive actions of various linear operators, acting on an initial focusing function. These operators involve multidimensional crosscorrelations with the reflection data and truncations in time. We offer physical interpretations of the multidimensional crosscorrelations by subtracting traveltimes along common ray paths at the stationary points of the underlying integrals. This provides a clear understanding of how individual events are retrieved by the scheme. Our interpretation also exposes some of the scheme's limitations in terms of what can be retrieved in case of a finite recording aperture. Green's function retrieval is only successful if the relevant stationary points are sampled. As a consequence, internal multiples can only be retrieved at a subsurface location with a particular ray parameter if this location is illuminated by the direct wavefield with this specific ray parameter. Several assumptions are required to solve the Marchenko equations. We show that these assumptions are not always satisfied in arbitrary heterogeneous media, which can result in incomplete Green's function retrieval and the emergence of artefacts. Despite these limitations, accurate Green's functions can often be retrieved by the iterative scheme, which is highly relevant for seismic imaging and inversion of internal multiple reflections.

  12. Preparation of acceptance tests and criteria for the Test Blanket Systems to be operated 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, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Cuquel, B. [AIRBUS Defence and Space S.A.S., 13115 Saint Paul Lez Durance (France); Demange, D.; Ghidersa, B.-E. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Giancarli, L.M.; Iseli, M.; Jourdan, T. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R.; Ring, W. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • Initial guideline for acceptance testing and acceptance criteria for Test Blanket Systems in ITER. • These tests complement those required by the applicable codes and standards, and regulations. • Completion of TBS manufacture will be followed by Factory Acceptance Testing, prior to shipment. • Next steps are “Reception Inspection Tests”, and on-site pre-installation and components tests. • This guideline allows the detailing of the TBS specific test plans and their scheduling. - Abstract: This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase.

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

    International Nuclear Information System (INIS)

    Slack, D.S.

    1991-01-01

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

  14. Variational Iteration Method for Volterra Functional Integrodifferential Equations with Vanishing Linear Delays

    Directory of Open Access Journals (Sweden)

    Ali Konuralp

    2014-01-01

    Full Text Available Application process of variational iteration method is presented in order to solve the Volterra functional integrodifferential equations which have multi terms and vanishing delays where the delay function θ(t vanishes inside the integral limits such that θ(t=qt for 0

  15. User requirements and conceptual design of the ITER Electron Cyclotron Control System

    Energy Technology Data Exchange (ETDEWEB)

    Carannante, Giuseppe, E-mail: Giuseppe.Carannante@F4E.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Cavinato, Mario [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Gandini, Franco [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Granucci, Gustavo [Istituto di Fisica del Plasma ENEA-CNR-EURATOM, via Cozzi 53, 20125 Milano (Italy); Henderson, Mark; Purohit, Dharmesh [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Saibene, Gabriella; Sartori, Filippo [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Sozzi, Carlo [Istituto di Fisica del Plasma ENEA-CNR-EURATOM, via Cozzi 53, 20125 Milano (Italy)

    2015-10-15

    The ITER Electron Cyclotron (EC) plant is a complex system, essential for plasma operation. The system is being designed to supply up to 20 MW of power at 170 GHz; it consists of 24 RF sources (or Gyrotrons) connected by switchable transmission lines to four upper and one equatorial launcher. The complexity of the EC plant requires a Plant Controller, which provides the functional and operational interface with CODAC and the Plasma Control System and coordinates the various Subsystem Control Units, i.e. the local controllers of power supplies, Gyrotrons, transmission lines and launchers. A conceptual design of the Electron Cyclotron Control System (ECCS) was developed, starting from the collection of the user requirements, which have then been organized as a set of operational scenarios exploiting the EC system. The design consists in a thorough functional analysis, including also protection functions, and in the development of a conceptual I&C architecture. The main aim of the work was to identify the physics requirements and to translate them into control system requirements, in order to define the interfaces within the components of the ECCS. The definition of these interfaces is urgent because some of the subsystems are already in an advanced design phase. The present paper describes both the methodology used and the resulting design.

  16. A comparative study of iterative solutions to linear systems arising in quantum mechanics

    International Nuclear Information System (INIS)

    Jing Yanfei; Huang Tingzhu; Duan Yong; Carpentieri, Bruno

    2010-01-01

    This study is mainly focused on iterative solutions with simple diagonal preconditioning to two complex-valued nonsymmetric systems of linear equations arising from a computational chemistry model problem proposed by Sherry Li of NERSC. Numerical experiments show the feasibility of iterative methods to some extent when applied to the problems and reveal the competitiveness of our recently proposed Lanczos biconjugate A-orthonormalization methods to other classic and popular iterative methods. By the way, experiment results also indicate that application specific preconditioners may be mandatory and required for accelerating convergence.

  17. The Iterative Solution to Discrete-Time H∞ Control Problems for Periodic Systems

    Directory of Open Access Journals (Sweden)

    Ivan G. Ivanov

    2016-03-01

    Full Text Available This paper addresses the problem of solving discrete-time H ∞ control problems for periodic systems. The approach for solving such a type of equations is well known in the literature. However, the focus of our research is set on the numerical computation of the stabilizing solution. In particular, two effective methods for practical realization of the known iterative processes are described. Furthermore, a new iterative approach is investigated and applied. On the basis of numerical experiments, we compare the presented methods. A major conclusion is that the new iterative approach is faster than rest of the methods and it uses less RAM memory than other methods.

  18. Identifying elementary iterated systems through algorithmic inference: The Cantor set example

    Energy Technology Data Exchange (ETDEWEB)

    Apolloni, Bruno [Dipartimento di Scienze dell' Informazione, Universita degli Studi di Milano, Via Comelico 39/41, 20135 Milan (Italy)]. E-mail: apolloni@dsi.unimi.it; Bassis, Simone [Dipartimento di Scienze dell' Informazione, Universita degli Studi di Milano, Via Comelico 39/41, 20135 Milan (Italy)]. E-mail: bassis@dsi.unimi.it

    2006-10-15

    We come back to the old problem of fractal identification within the new framework of algorithmic Inference. The key points are: (i) to identify sufficient statistics to be put in connection with the unknown values of the fractal parameters, and (ii) to manage the timing of the iterated process through spatial statistics. We fill these tasks successfully with the Cantor sets. We are able to compute confidence intervals for both the scaling parameter {theta} and the iteration number n at which we are observing a set. We both check numerically the coverage of these intervals and delineate a general strategy for affording more complex iterated systems.

  19. The Semianalytical Solutions for Stiff Systems of Ordinary Differential Equations by Using Variational Iteration Method and Modified Variational Iteration Method with Comparison to Exact Solutions

    Directory of Open Access Journals (Sweden)

    Mehmet Tarik Atay

    2013-01-01

    Full Text Available The Variational Iteration Method (VIM and Modified Variational Iteration Method (MVIM are used to find solutions of systems of stiff ordinary differential equations for both linear and nonlinear problems. Some examples are given to illustrate the accuracy and effectiveness of these methods. We compare our results with exact results. In some studies related to stiff ordinary differential equations, problems were solved by Adomian Decomposition Method and VIM and Homotopy Perturbation Method. Comparisons with exact solutions reveal that the Variational Iteration Method (VIM and the Modified Variational Iteration Method (MVIM are easier to implement. In fact, these methods are promising methods for various systems of linear and nonlinear stiff ordinary differential equations. Furthermore, VIM, or in some cases MVIM, is giving exact solutions in linear cases and very satisfactory solutions when compared to exact solutions for nonlinear cases depending on the stiffness ratio of the stiff system to be solved.

  20. Thermo-mechanical analysis of ITER first mirrors and its use for the ITER equatorial visible/infrared wide angle viewing system optical design

    International Nuclear Information System (INIS)

    Joanny, M.; Salasca, S.; Dapena, M.; Cantone, B.; Travère, J. M.; Thellier, C.; Fermé, J. J.; Marot, L.; Buravand, O.; Perrollaz, G.; Zeile, C.

    2012-01-01

    ITER first mirrors (FMs), as the first components of most ITER optical diagnostics, will be exposed to high plasma radiation flux and neutron load. To reduce the FMs heating and optical surface deformation induced during ITER operation, the use of relevant materials and cooling system are foreseen. The calculations led on different materials and FMs designs and geometries (100 mm and 200 mm) show that the use of CuCrZr and TZM, and a complex integrated cooling system can limit efficiently the FMs heating and reduce their optical surface deformation under plasma radiation flux and neutron load. These investigations were used to evaluate, for the ITER equatorial port visible/infrared wide angle viewing system, the impact of the FMs properties change during operation on the instrument main optical performances. The results obtained are presented and discussed.

  1. Mock-up test on key components of ITER blanket remote handling system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Koh; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, Alessandro

    2009-01-01

    The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high gamma-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency (then called as Japan Atomic Energy Research Institute) had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The ITER agreement, which entered into force in the last year, formally decided that Japan will procure the blanket remote handling system and the JAEA, as the Japanese Domestic Agency, is continuing several R and Ds so that the system can be procured smoothly. The residual key issues after the EDA are rail connection and cable handling. The mock-ups of the rail connection mechanism and the cable handling system were fabricated from the last year and installed at the JAEA Naka Site in this March. The former was composed of the rail connecting mechanism, two rail segments and their handling systems. The latter one utilized a slip ring, which implemented 80 lines for power and 208 lines for signal, because there is an electrical contact between the rotating spool and the fixed base. The basic function of these systems was confirmed through the mock-up test. The rail connection mechanism, for example, could accept misalignment of 1.5-2 mm at least. The future test plan is also mentioned in the paper.

  2. Conceptual design of an electron cyclotron wave system for NET/ITER

    International Nuclear Information System (INIS)

    Kasparek, W.; Kumric, H.; Mueller, G.A.; Pretterebner, J.; Schueller, P.G.; Wagner, D.

    1991-07-01

    Electron Cyclotron waves (ECWs) provide a scheme for electron heating, which, owing to the strong localization of the resonant interaction with the plasma, allows an efficient tailoring of the power deposition profile. In the proposed ITER reference scenario for current drive and heating, ECWs are considered to assist plasma formation, pre-heating, local current profile control near the q=2 surface and possibly for baking the first wall tiles. For these functions, a total power of 20 MW, CW, at a frequency around 120 GHz is needed. A higher frequency system (140 GHz, 20 MW, CW), is also considered to heat the plasma centre and provide burn control. The same system at increased power could be used for plasma heating to ignition. For NET, due to the higher magnetic field, the frequencies needed for the tasks mentioned above are approximately 140 GHz and 160 GHz, respectively. ECWs are also envisaged for bulk heating of the NET plasma. Here, frequencies of about 160 GHz are necessary. A detailed study for the 120 GHz/20 MW ITER reference system has been performed. Scaling rules as well as additional antenna designs for higher frequency systems have been developed. The design principle was to offer a high degree of flexibility for the wide range of envisaged uses of the ECWs. The ECW system should satisfy the physics requirements, advanced requirements of reliability and availability, and must be compatible with the nuclear environment (which requires radiation resistance as well as remote maintenance of at least the antenna part). Therefore, it has been tried to place the most critical components as far away from the machine as possible. To improve the availability, the installation of 15% spare tubes and transmission systems is proposed. (orig.)

  3. 2-D Reflectometer Modeling for Optimizing the ITER Low-field Side Reflectometer System

    International Nuclear Information System (INIS)

    Kramer, G.J.; Nazikian, R.; Valeo, E.J.; Budny, R.V.; Kessel, C.; Johnson, D.

    2005-01-01

    The response of a low-field side reflectometer system for ITER is simulated with a 2?D reflectometer code using a realistic plasma equilibrium. It is found that the reflected beam will often miss its launch point by as much as 40 cm and that a vertical array of receiving antennas is essential in order to observe a reflection on the low-field side of ITER

  4. Progress in Development of the ITER Plasma Control System Simulation Platform

    Science.gov (United States)

    Walker, Michael; Humphreys, David; Sammuli, Brian; Ambrosino, Giuseppe; de Tommasi, Gianmaria; Mattei, Massimiliano; Raupp, Gerhard; Treutterer, Wolfgang; Winter, Axel

    2017-10-01

    We report on progress made and expected uses of the Plasma Control System Simulation Platform (PCSSP), the primary test environment for development of the ITER Plasma Control System (PCS). PCSSP will be used for verification and validation of the ITER PCS Final Design for First Plasma, to be completed in 2020. We discuss the objectives of PCSSP, its overall structure, selected features, application to existing devices, and expected evolution over the lifetime of the ITER PCS. We describe an archiving solution for simulation results, methods for incorporating physics models of the plasma and physical plant (tokamak, actuator, and diagnostic systems) into PCSSP, and defining characteristics of models suitable for a plasma control development environment such as PCSSP. Applications of PCSSP simulation models including resistive plasma equilibrium evolution are demonstrated. PCSSP development supported by ITER Organization under ITER/CTS/6000000037. Resistive evolution code developed under General Atomics' Internal funding. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

  5. Iterative reconstruction using a Monte Carlo based system transfer matrix for dedicated breast positron emission tomography

    Energy Technology Data Exchange (ETDEWEB)

    Saha, Krishnendu [Ohio Medical Physics Consulting, Dublin, Ohio 43017 (United States); Straus, Kenneth J.; Glick, Stephen J. [Department of Radiology, University of Massachusetts Medical School, Worcester, Massachusetts 01655 (United States); Chen, Yu. [Department of Radiation Oncology, Columbia University, New York, New York 10032 (United States)

    2014-08-28

    To maximize sensitivity, it is desirable that ring Positron Emission Tomography (PET) systems dedicated for imaging the breast have a small bore. Unfortunately, due to parallax error this causes substantial degradation in spatial resolution for objects near the periphery of the breast. In this work, a framework for computing and incorporating an accurate system matrix into iterative reconstruction is presented in an effort to reduce spatial resolution degradation towards the periphery of the breast. The GATE Monte Carlo Simulation software was utilized to accurately model the system matrix for a breast PET system. A strategy for increasing the count statistics in the system matrix computation and for reducing the system element storage space was used by calculating only a subset of matrix elements and then estimating the rest of the elements by using the geometric symmetry of the cylindrical scanner. To implement this strategy, polar voxel basis functions were used to represent the object, resulting in a block-circulant system matrix. Simulation studies using a breast PET scanner model with ring geometry demonstrated improved contrast at 45% reduced noise level and 1.5 to 3 times resolution performance improvement when compared to MLEM reconstruction using a simple line-integral model. The GATE based system matrix reconstruction technique promises to improve resolution and noise performance and reduce image distortion at FOV periphery compared to line-integral based system matrix reconstruction.

  6. Measurement and control system for the ITER remote handling mock-up test

    International Nuclear Information System (INIS)

    Oka, K.; Kakudate, S.; Takiguchi, Y.; Ako, K.; Taguchi, K.; Tada, E.; Ozaki, F.; Shibanuma, K.

    1998-01-01

    The mock-up test platforms composed of full-scale remote handling (RH) equipment were developed for demonstrating remote replacement of the ITER blanket and divertor. In parallel, the measurement and control system for operating these RH equipment were constructed on the basis of open architecture with object oriented feature, aiming at realization of fully-remoted automatic operation required for ITER. This paper describes the design concept of the measurement and control system for the remote handling equipment of ITER, and outlines the measured performances of the fabricated measurement system for the remote handling mock-up tests, which includes Data Acquisition System (DAS), Visual Monitoring System (VMS) and Virtual Reality System (VRS). (authors)

  7. Progress in the integration of the ITER plant systems in auxiliary buildings

    International Nuclear Information System (INIS)

    Kotamäki, M.; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-01-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  8. Progress in the integration of the ITER plant systems in auxiliary buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kotamäki, M., E-mail: miikka.kotamaki@iter.org; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-11-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  9. ITER-like PAM launcher for tore supra's LHCD system

    International Nuclear Information System (INIS)

    Belo, J.H.; Bibet, Ph.; Missirlian, M.; Achard, J.; Beaumont, B.; Bertrand, B.; Chantant, M.; Chappuis, Ph.; Doceul, L.; Durocher, A.; Gargiulo, L.; Saille, A.; Samaille, F.; Villedieu, E.

    2004-01-01

    A new launcher based upon the PAM concept (Passive-Active Multijunction) already proposed for ITER has been developed and is currently under realisation at Tore Supra. It was designed for an injection power capability of 2.7 MW CW at 3.7 GHz, a power density of 25 MW/m 2 , to radiate a power spectrum peaked at N || = 1.7 with a maximum power directivity near the electron cut-off density and with very good coupling properties on a wide range of electron densities. In this paper an overall description of the antenna as well as the foreseen manufacturing and assembling processes are given, followed by the results from studies and optimizations of its RF components, and by a stress analysis of its thermomechanical behaviour. (authors)

  10. Improvement of the dynamic response of the ITER Reactive Power Compensation system

    International Nuclear Information System (INIS)

    Finotti, Claudio; Gaio, Elena; Song, Inho; Tao, Jun; Benfatto, Ivone

    2015-01-01

    Highlights: • The slow response reasons of the classic ITER Reactive Power Compensation (RPC) control are explained. • The dynamic behaviors of the ac/dc converter and of the RPC are characterized. • New control concept to speed up the RPC response is developed. • Good performance of the new RPC control is verified even during fast transient conditions. - Abstract: The ITER ac/dc conversion system can absorb a total active and reactive power up to 500 MW and 950 Mvar, respectively. The Reactive Power Compensation (RPC) system is rated for a nominal power of 750 Mvar necessary to comply with the allowable reactive power limit value from the grid of 200 Mvar. This system is currently under construction and is based on Static Var Compensation technology with Thyristor Controlled Reactor (TCR) and Tuned Filters. The RPC has to minimize the demand of reactive power from the grid; its control is based on a feed-forward method, where the corrective input is the measurement of the reactive power consumption of the ac/dc converters, derived from the 50 Hz component of the Fast Fourier Transform (FFT) of the three-phase voltages and currents. The delay introduced by the FFT calculation and the slow response of the TCR could make the response speed of the RPC not sufficient to face fast variations of the reactive power demand and therefore in this paper a new controller of the RPC able to overcome this shortcoming is proposed and evaluated. It is based on the calculation of the predicted consumption of the reactive power by using the voltage reference signals coming from the Plasma Control System and the measurements of the dc current of the ac/dc converters and of the 66 kV busbar voltage, and on the speed up of the RPC control by introducing a lead–lag transfer function.

  11. Improvement of the dynamic response of the ITER Reactive Power Compensation system

    Energy Technology Data Exchange (ETDEWEB)

    Finotti, Claudio, E-mail: claudio.finotti@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); Gaio, Elena [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); Song, Inho; Tao, Jun; Benfatto, Ivone [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • The slow response reasons of the classic ITER Reactive Power Compensation (RPC) control are explained. • The dynamic behaviors of the ac/dc converter and of the RPC are characterized. • New control concept to speed up the RPC response is developed. • Good performance of the new RPC control is verified even during fast transient conditions. - Abstract: The ITER ac/dc conversion system can absorb a total active and reactive power up to 500 MW and 950 Mvar, respectively. The Reactive Power Compensation (RPC) system is rated for a nominal power of 750 Mvar necessary to comply with the allowable reactive power limit value from the grid of 200 Mvar. This system is currently under construction and is based on Static Var Compensation technology with Thyristor Controlled Reactor (TCR) and Tuned Filters. The RPC has to minimize the demand of reactive power from the grid; its control is based on a feed-forward method, where the corrective input is the measurement of the reactive power consumption of the ac/dc converters, derived from the 50 Hz component of the Fast Fourier Transform (FFT) of the three-phase voltages and currents. The delay introduced by the FFT calculation and the slow response of the TCR could make the response speed of the RPC not sufficient to face fast variations of the reactive power demand and therefore in this paper a new controller of the RPC able to overcome this shortcoming is proposed and evaluated. It is based on the calculation of the predicted consumption of the reactive power by using the voltage reference signals coming from the Plasma Control System and the measurements of the dc current of the ac/dc converters and of the 66 kV busbar voltage, and on the speed up of the RPC control by introducing a lead–lag transfer function.

  12. High Heat flux (HHF) elements for Negative Ion Systems on ITER

    International Nuclear Information System (INIS)

    Milnes, J.; Chuilon, B.; Martin, D.; Waldon, Ch.; Yong Xue

    2006-01-01

    Negative Ion Neutral Beam systems on ITER will require actively cooled scrapers and dumps to process and shape the beam before injection into the tokamak. The scale of the systems is much larger than any presently operating, bringing challenges for designers in terms of available sub cooling, total pressure drop, deflection and mandatory remote maintenance. In common with Positive Ion systems, flux densities in the order of 15-20 MW/m 2 are commonplace but with much longer pulses. A pulse length in excess of 3000 seconds and the anticipated beam breakdown rate pose new challenges in terms of stress and fatigue life. The cooling system specification (up to 26 bar, 80 o C) adds further constraints impacting the material choice and operating temperature. The DDD designs, based on swirl tubes, have been reviewed as part of the design process and recommendations made. Additionally, alternative designs have been proposed based on the Hypervapotron high heat flux elements with modified geometry and drawing upon a vast background knowledge of large scale equipment procurement and integration. Existing operational and design experience has been applied to give a simple, robust and low maintenance alternative. A full thermomechanical analysis of all HHF components has been undertaken based on ITER design criteria and the limited material data available. The results of this analysis will be presented, highlighting areas where further R(and)D is necessary to reach the operating limits set out in the functional specification. Extensive comparison of these analyses is made with the large operational database of existing JET beamline components for benchmarking purposes. A particular feature of the thermo-mechanical analyses is a fully self-consistent description in which ageing characteristics are related to the local temperature, and the components' power loading takes into account the thermal distortion. The advantages and disadvantages of all designs will be presented and

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

    International Nuclear Information System (INIS)

    Slack, D.S.

    1989-01-01

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

  14. Implications of ITER requirements on R and D of RF heating and current drive systems

    International Nuclear Information System (INIS)

    Bosia, G.

    2002-01-01

    A strategic, rather than auxiliary role is assigned to H and CD systems in ITER-FEAT, as all operation phases are driven and controlled by heating and current drive (H and CD) systems. RF systems (Electron Cyclotron, Ion Cyclotron and Lower Hybrid), planned to contribute for ∼60% of ITER auxiliary power (72 MW), still require different level of pre-industrial technology development to operate in ITER at the required level of efficiency and religiosite. In this paper, RF H and CD systems technical and operational issues are reviewed and future R and D actions at CEA-Cadarache discussed, with the aim of providing a demonstration of all RF H and CD systems, within the current ITER construction time scale. The need and the economical advantage of an early on- and off- plasma design validation program for ITER-like RF devices (such as launcher and/or power sources), is also discussed with the aim of identifying and resolving operational issues. (author)

  15. Current Monitoring System for ITER Like ICRH Antenna

    International Nuclear Information System (INIS)

    Argouarch, A.

    2006-01-01

    On TS antennas, the power transfer optimization from ICRH antenna to Plasma load is performed using feedback internal matching system. Experimental handling is required to mach the reactive impedance accordingly to the fluctuant plasma loading. As part of the development of the new ICRH prototype antenna, an additional measurement system based on Rogowski coils was developed to monitor the current distribution in antenna straps. The objective is to control module and phase of the antenna current straps with measurement provided by the coil system. Matching capacitors values, generators power and phase can also be controlled using the output of the devices, improving the real time matching control of the array. This paper details the new measurement layout, the Rogowski coil, and the whole system connected on each strap design for RF currents measurement between 40 MHz - 60 MHz for maximum amplitude of 1 kA. On the new ICRH prototype antenna, the measurement coils are coupled to the point where the strap currents are short circuited to the frame. The module and phase measurements are performed with the coils by direct magnetic induction in a vacuum and high temperature environment. Also, the Rogowski coils were characterized at low level power with vector network analyzer and the design adapted in order to obtain a controlled and reproducible gain in the desire bandwidth. The transconductive function is established with an experimental gain near -80 dB between primary circuit and inductive signal generated by the Rogowski coil. In a second step, the system with its associated electronic was qualified under high RF power. First results with high RF current (closed to 500 A at 57 MHz) match the desire Rogowski coil response. Compromises with electrical response at low power level and the coil under thermal/RF stresses were the most challenging part of the development. The overall response of the system and the current module/phase measurements are promising. A proper

  16. Elliptic Preconditioner for Accelerating the Self-Consistent Field Iteration in Kohn--Sham Density Functional Theory

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Lin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Computational Research Division; Yang, Chao [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). Computational Research Division

    2013-10-28

    We discuss techniques for accelerating the self consistent field (SCF) iteration for solving the Kohn-Sham equations. These techniques are all based on constructing approximations to the inverse of the Jacobian associated with a fixed point map satisfied by the total potential. They can be viewed as preconditioners for a fixed point iteration. We point out different requirements for constructing preconditioners for insulating and metallic systems respectively, and discuss how to construct preconditioners to keep the convergence rate of the fixed point iteration independent of the size of the atomistic system. We propose a new preconditioner that can treat insulating and metallic system in a unified way. The new preconditioner, which we call an elliptic preconditioner, is constructed by solving an elliptic partial differential equation. The elliptic preconditioner is shown to be more effective in accelerating the convergence of a fixed point iteration than the existing approaches for large inhomogeneous systems at low temperature.

  17. Performance and capacity analysis of Poisson photon-counting based Iter-PIC OCDMA systems.

    Science.gov (United States)

    Li, Lingbin; Zhou, Xiaolin; Zhang, Rong; Zhang, Dingchen; Hanzo, Lajos

    2013-11-04

    In this paper, an iterative parallel interference cancellation (Iter-PIC) technique is developed for optical code-division multiple-access (OCDMA) systems relying on shot-noise limited Poisson photon-counting reception. The novel semi-analytical tool of extrinsic information transfer (EXIT) charts is used for analysing both the bit error rate (BER) performance as well as the channel capacity of these systems and the results are verified by Monte Carlo simulations. The proposed Iter-PIC OCDMA system is capable of achieving two orders of magnitude BER improvements and a 0.1 nats of capacity improvement over the conventional chip-level OCDMA systems at a coding rate of 1/10.

  18. Feasibility analysis of fuzzy logic control for ITER Poloidal field (PF) AC/DC converter system

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Mahmood Ul; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Chen, Xiaojiao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China (China); Zhang, Xiuqing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Humayun, Muhammad [Shanghai Jiaotong University (China)

    2017-05-15

    Highlights: • The implementation of the Fuzzy controller for the ITER PF converter system is presented. • The comparison of the FLC and PI simulation are investigated. • The FLC single and parallel bridge operation are presented. • Fuzzification and Defuzzification algorithms are presented using FLC controller. - Abstract: This paper describes the feasibility analysis of the fuzzy logic control to increase the performance of the ITER poloidal field (PF) converter systems. A fuzzy-logic-based controller is designed for ITER PF converter system, using the traditional PI controller and Fuzzy controller (FC), the dynamic behavior and transient response of the PF converter system are compared under normal operation by analysis and simulation. The analysis results show that the fuzzy logic control can achieve better operation performance than PI control.

  19. Magnetic field compatibility of components for ITER electrical distribution systems

    International Nuclear Information System (INIS)

    Hourtoule, J.; Van Houtte, D.; Hertout, P.; Fejoz, P.

    2004-01-01

    The electrical distribution boards and control cubicles, installed inside the ITER Tokamak building, can be subject to a constant or slowly variable magnetic field up to 70 mT, 10 mT/s induced by ITER coils. This is a very unusual environmental condition and there are almost no data available on static magnetic field compatibility for the standard industrial electrical and electronic components that might be used inside standard low voltage distribution boards and standard control boards. CEA has especially taken in charge the test of electronic, control and signal conditioning units. For this purpose a test bed composed of a solenoid and a 30 V, 800 A power supply, has been developed at CEA Cadarache. Its characteristics are the following: - Magnetic field capability: 40 mT; - Variation of field: up to 10 mT/s. Useful dimensions for equipment under test: 500 x 500 x 500 mm. The list of the components to be tested has been identified trying to find common and recent components of different manufacturers. Test procedures have been written following the most relevant IEC standards and manufacturing recommendations and procedures. Magnetic compatibility tests on the basic components of standard low voltage control boards has been performed. All components tested are more or less sensitive to magnetic fields and the effect varies from the simple perturbation of the output signals to the partial destruction of some electronic card. The most sensitive are the ferromagnetic components like small power supplies transformers, relays, on inductors. For power supplies the most important consequence is the increase of the consumption (primary current). For most of the conditioning units the operational limits were found in the range of 30 mT (destruction of materials and overload of the supply) For PLC (Programme Logic Control) central units tested, the limit is in the order of 40 mT (components permanently out of service) Relays are limited to 15 mT (outputs oscillations, no

  20. A policy iteration approach to online optimal control of continuous-time constrained-input systems.

    Science.gov (United States)

    Modares, Hamidreza; Naghibi Sistani, Mohammad-Bagher; Lewis, Frank L

    2013-09-01

    This paper is an effort towards developing an online learning algorithm to find the optimal control solution for continuous-time (CT) systems subject to input constraints. The proposed method is based on the policy iteration (PI) technique which has recently evolved as a major technique for solving optimal control problems. Although a number of online PI algorithms have been developed for CT systems, none of them take into account the input constraints caused by actuator saturation. In practice, however, ignoring these constraints leads to performance degradation or even system instability. In this paper, to deal with the input constraints, a suitable nonquadratic functional is employed to encode the constraints into the optimization formulation. Then, the proposed PI algorithm is implemented on an actor-critic structure to solve the Hamilton-Jacobi-Bellman (HJB) equation associated with this nonquadratic cost functional in an online fashion. That is, two coupled neural network (NN) approximators, namely an actor and a critic are tuned online and simultaneously for approximating the associated HJB solution and computing the optimal control policy. The critic is used to evaluate the cost associated with the current policy, while the actor is used to find an improved policy based on information provided by the critic. Convergence to a close approximation of the HJB solution as well as stability of the proposed feedback control law are shown. Simulation results of the proposed method on a nonlinear CT system illustrate the effectiveness of the proposed approach. Copyright © 2013 ISA. All rights reserved.

  1. A faster ordered-subset convex algorithm for iterative reconstruction in a rotation-free micro-CT system

    International Nuclear Information System (INIS)

    Quan, E; Lalush, D S

    2009-01-01

    We present a faster iterative reconstruction algorithm based on the ordered-subset convex (OSC) algorithm for transmission CT. The OSC algorithm was modified such that it calculates the normalization term before the iterative process in order to save computational cost. The modified version requires only one backprojection per iteration as compared to two required for the original OSC. We applied the modified OSC (MOSC) algorithm to a rotation-free micro-CT system that we proposed previously, observed its performance, and compared with the OSC algorithm for 3D cone-beam reconstruction. Measurements on the reconstructed images as well as the point spread functions show that MOSC is quite similar to OSC; in noise-resolution trade-off, MOSC is comparable with OSC in a regular-noise situation and it is slightly worse than OSC in an extremely high-noise situation. The timing record shows that MOSC saves 25-30% CPU time, depending on the number of iterations used. We conclude that the MOSC algorithm is more efficient than OSC and provides comparable images.

  2. EPICS device support module as ATCA system manager for the ITER fast plant system controller

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Paulo F., E-mail: pricardofc@ipfn.ist.utl.pt [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico – Universidade Técnica de Lisboa, Lisboa (Portugal); Santos, Bruno; Gonçalves, Bruno; Carvalho, Bernardo B.; Sousa, Jorge; Rodrigues, A.P.; Batista, António J.N.; Correia, Miguel; Combo, Álvaro [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico – Universidade Técnica de Lisboa, Lisboa (Portugal); Correia, Carlos M.B.A. [Centro de Instrumentação, Departamento de Física, Universidade de Coimbra, Coimbra (Portugal); Varandas, Carlos A.F. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico – Universidade Técnica de Lisboa, Lisboa (Portugal)

    2013-10-15

    Highlights: ► In Nuclear Fusion, demanding security and high-availability requirements call for redundancy to be available. ► ATCA based Nuclear Fusion Systems are composed by several electronic and mechanical component. ► Control and monitoring of ATCA electronic systems are recommended. ► ITER Fast Plant System Controller Project CODAC system prototype. ► EPICS device support module as External ATCA system manager solution. -- Abstract: This paper presents an Enhanced Physics and Industrial Control System (EPICS) device support module for the International Thermonuclear Experimental Reactor (ITER) Fast Plant System Controller (FPSC) project based in Advanced Telecommunications Computing Architecture (ATCA) specification. The developed EPICS device support module provides an External System Manager (ESM) solution for monitoring and control the ITER FPSC ATCA shelf system and data acquisition boards in order to take proper action and report problems to a control room operator or high level management unit in case of any system failure occurrence. EPICS device support module acts as a Channel Access (CA) server to report problems and publish ATCA system data information to the control room operator, high level management unit or other CA network clients such as Control System Studio Operator Interfaces (CSS OPIs), Best Ever Alarm System Toolkit (BEAST), Best Ever Archive Utility (BEAUTY) or other CA client applications. EPICS device support module communicates with the ATCA Shelf manager (ShM) using HTTP protocol to send and receive commands through POST method in order to get and set system and shelf components properties such as fan speeds measurements, temperatures readings, module status and ATCA boards acquisition and configuration parameters. All system properties, states, commands and parameters are available through the EPICS device support module CA server in EPICS Process Variables (PV) and signals format. ATCA ShM receives the HTTP protocol

  3. ITER in-vessel system design and performance

    Science.gov (United States)

    Parker, R. R.

    2000-03-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

  4. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    1999-01-01

    This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

  5. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2001-01-01

    This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

  6. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2000-01-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events. (author)

  7. Progress in the design of the ITER Neutral Beam cell Remote Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Shuff, R., E-mail: robin.shuff@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Van Uffelen, M.; Damiani, C. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Tesini, A.; Choi, C.-H. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Meek, R. [Oxford Technologies Limited, 7 Nuffield Way, Abingdon OX14 1RL (United Kingdom)

    2014-10-15

    The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper.

  8. Iterative solution of a nonlinear system arising in phase change problems

    International Nuclear Information System (INIS)

    Williams, M.A.

    1987-01-01

    We consider several iterative methods for solving the nonlinear system arising from an enthalpy formulation of a phase change problem. We present the formulation of the problem. Implicit discretization of the governing equations results in a mildly nonlinear system at each time step. We discuss solving this system using Jacobi, Gauss-Seidel, and SOR iterations and a new modified preconditioned conjugate gradient (MPCG) algorithm. The new MPCG algorithm and its properties are discussed in detail. Numerical results are presented comparing the performance of the SOR algorithm and the MPCG algorithm with 1-step SSOR preconditioning. The MPCG algorithm exhibits a superlinear rate of convergence. The SOR algorithm exhibits a linear rate of convergence. Thus, the MPCG algorithm requires fewer iterations to converge than the SOR algorithm. However in most cases, the SOR algorithm requires less total computation time than the MPCG algorithm. Hence, the SOR algorithm appears to be more appropriate for the class of problems considered. 27 refs., 11 figs

  9. Progress in the design of the ITER Neutral Beam cell Remote Handling System

    International Nuclear Information System (INIS)

    Shuff, R.; Van Uffelen, M.; Damiani, C.; Tesini, A.; Choi, C.-H.; Meek, R.

    2014-01-01

    The ITER Neutral Beam cell will include a suite of Remote Handling equipment for maintenance tasks. This paper summarises the current status and recent developments in the design of the ITER Neutral Beam Remote Handling System. Its concept design was successfully completed in July 2012 by CCFE in the frame of a grant agreement with F4E, in collaboration with the ITER Organisation, including major systems like monorail crane, Beam Line Transporter, beam source equipment, upper port and neutron shield equipment and associated tooling. Research and development activities are now underway on the monorail crane radiation hardened on-board control system and first of a kind remote pipe and lip seal maintenance tooling for the beam line vessel, reported in this paper

  10. Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

    Science.gov (United States)

    Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

    2017-12-01

    The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

  11. Data archiving system implementation in ITER's CODAC Core System

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R., E-mail: rodrigo.castro@visite.es [CIEMAT Fusion Program, Avda. Complutense 40, Madrid (Spain); Abadie, L. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Makushok, Y. [Sgenia, C/Chile, 4 Edificio II, Las Rozas, Madrid (Spain); Ruiz, M.; Sanz, D. [Instrumentation and Applied Acoustic Research Group, Technical University of Madrid, Madrid (Spain); Vega, J. [CIEMAT Fusion Program, Avda. Complutense 40, Madrid (Spain); Faig, J. [INDRA Sistemas, S.A. Unid. de Sistemas de Control, Dirección de Tecnología Energética, Madrid (Spain); Román-Pérez, G. [Sgenia, C/Chile, 4 Edificio II, Las Rozas, Madrid (Spain); Simrock, S.; Makijarvi, P. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Implementation of ITER's data archiving solution. • Integration of the solution into CODAC Core System. • Data archiving structure. • High efficient data transmission into fast plant system controllers. • Fast control and data acquisition in Linux. - Abstract: The aim of this work is to present the implementation of data archiving in ITER's CODAC Core System software. This first approach provides a client side API and server side software allowing the creation of a simplified version of ITERDB data archiving software, and implements all required elements to complete data archiving flow from data acquisition until its persistent storage technology. The client side includes all necessary components that run on devices that acquire or produce data, distributing and streaming to configure remote archiving servers. The server side comprises an archiving service that stores into HDF5 files all received data. The archiving solution aims at storing data coming for the data acquisition system, the conventional control and also processed/simulated data.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  13. Contractive function systems, their attractors and metrization

    Czech Academy of Sciences Publication Activity Database

    Banakh, T.; Kubiś, Wieslaw; Novosad, N.; Nowak, M.; Strobin, F.

    2015-01-01

    Roč. 46, č. 2 (2015), s. 1029-1066 ISSN 1230-3429 R&D Projects: GA ČR(CZ) GA14-07880S Institutional support: RVO:67985840 Keywords : fractal * attractor * iterated function system * contracting function system Subject RIV: BA - General Mathematics Impact factor: 0.717, year: 2015 http://www.apcz.pl/czasopisma/index.php/TMNA/article/view/TMNA.2015.076

  14. Progress in design and integration of the ITER Electron Cyclotron H and CD system

    International Nuclear Information System (INIS)

    Darbos, C.; Henderson, M.; Albajar, F.; Bigelow, T.; Bonicelli, T.; Chavan, R.; Denisov, G.G.; Fasel, D.; Heidinger, R.; Hogge, J.P.; Kobayashi, N.; Piosczyk, B.; Rao, S.L.; Rasmussen, D.; Saibene, G.; Sakamoto, K.; Takahashi, K.; Thumm, M.

    2009-01-01

    The Electron Cyclotron system for ITER is an in-kind procurement shared between five parties and the total installed power will be 24 MW, corresponding to a nominal injected power of 20 MW to the plasma, with a possible upgrade up to 48 MW (corresponding to 40 MW injected). Some critical issues have been raised and changes are proposed to simplify these procurements and to facilitate the integration into ITER. The progress in the design and the integration of the EC system into the whole project is presented in this paper, as well as some issues still under studies and some recommendations made by external expert committees.

  15. On a new iterative method for solving linear systems and comparison results

    Science.gov (United States)

    Jing, Yan-Fei; Huang, Ting-Zhu

    2008-10-01

    In Ujevic [A new iterative method for solving linear systems, Appl. Math. Comput. 179 (2006) 725-730], the author obtained a new iterative method for solving linear systems, which can be considered as a modification of the Gauss-Seidel method. In this paper, we show that this is a special case from a point of view of projection techniques. And a different approach is established, which is both theoretically and numerically proven to be better than (at least the same as) Ujevic's. As the presented numerical examples show, in most cases, the convergence rate is more than one and a half that of Ujevic.

  16. An iterative two-threshold analysis for single-subject functional MRI of the human brain

    Energy Technology Data Exchange (ETDEWEB)

    Auer, Tibor; Schweizer, Renate; Frahm, Jens [Biomedizinische NMR Forschungs GmbH am Max-Planck-Institut fuer Biophysikalische Chemie, Goettingen (Germany)

    2011-11-15

    Current thresholding strategies for the analysis of functional MRI (fMRI) datasets may suffer from specific limitations (e.g. with respect to the required smoothness) or lead to reduced performance for a low signal-to-noise ratio (SNR). Although a previously proposed two-threshold (TT) method offers a promising solution to these problems, the use of preset settings limits its performance. This work presents an optimised TT approach that estimates the required parameters in an iterative manner. The iterative TT (iTT) method is compared with the original TT method, as well as other established voxel-based and cluster-based thresholding approaches and spatial mixture modelling (SMM) for both simulated data and fMRI of a hometown walking task at different experimental settings (spatial resolution, filtering and SNR). In general, the iTT method presents with remarkable sensitivity and good specificity that outperforms all conventional approaches tested except for SMM in a few cases. This also holds true for challenging conditions such as high spatial resolution, the absence of filtering, high noise level, or a low number of task repetitions. Thus, iTT emerges as a good candidate for both scientific fMRI studies at high spatial resolution and more routine applications for clinical purposes. (orig.)

  17. Industrial cost assessment for ITER tritium plant system (water distillation, VPCE and ISS)

    International Nuclear Information System (INIS)

    Sood, S.K.; Kalyanam, K.M.; Fong, C.

    1995-01-01

    The objective of this Industrial Cost Assessment Task for ITER Tritium Plant System consists of providing and order of magnitude cost estimate for the following major subsystems, as outlined in the Scope of Task Agreement and Work Program: water distillation (WD) system, vapour phase catalytic exchange (VPCE) system and the isotope separation system (ISS). The methodology adopted in preparing the order of magnitude cost estimate for the above three subsystems of the ITER tritium plant system is based on building the estimate from the ground up, starting with equipment cost estimates, and adding labour activities separately for engineering, fabrication, assembly, testing installation commissioning, etc. The estimate has been developed assuming that the systems are to be engineered, fabricated and constructed in Canada, (to comply with the Codes, Standards, QA and Seismic Classification applicable in Canada) since information on ITER siting is not currently available. The estimate is based on Ontario Hydro in house cost data on similar systems and equipment, such as the heavy water upgrading plants. The cost estimates are not based on quotations from suppliers for specific ITER components, since this would require completion of detailed design and specifications. 4 refs., 9 tabs., 7 figs

  18. Design and fabrication of the 'ITER-like' SINGAP D- acceleration system

    International Nuclear Information System (INIS)

    Massmann, P.; Esch, H.P.L. de; Hemsworth, R.S.; Svensson, L.

    2005-01-01

    To demonstrate ITER NBI (1 MV, 40 A) relevant beam optics in the Cadarache 1 MV, 100 mA test bed, a new D - beam source system has been put into operation. The system retains a maximum of the ITER SINGAP key parameters, e.g. the perveance matched D - current density at 1 MeV is 20 mA/cm 2 . The accelerator parameters are identical to the ITER SINGAP design, aiming at a near parallel 1 MeV beam of 5 mrad divergence. The design is aimed at also demonstrating SINGAP 'on to off-axis' beam steering by a simple transverse displacement of the post-acceleration electrode. First beams up to 850 keV have been obtained after only 4 weeks of commissioning

  19. A family of small-world network models built by complete graph and iteration-function

    Science.gov (United States)

    Ma, Fei; Yao, Bing

    2018-02-01

    Small-world networks are popular in real-life complex systems. In the past few decades, researchers presented amounts of small-world models, in which some are stochastic and the rest are deterministic. In comparison with random models, it is not only convenient but also interesting to study the topological properties of deterministic models in some fields, such as graph theory, theorem computer sciences and so on. As another concerned darling in current researches, community structure (modular topology) is referred to as an useful statistical parameter to uncover the operating functions of network. So, building and studying such models with community structure and small-world character will be a demanded task. Hence, in this article, we build a family of sparse network space N(t) which is different from those previous deterministic models. Even though, our models are established in the same way as them, iterative generation. By randomly connecting manner in each time step, every resulting member in N(t) has no absolutely self-similar feature widely shared in a large number of previous models. This makes our insight not into discussing a class certain model, but into investigating a group various ones spanning a network space. Somewhat surprisingly, our results prove all members of N(t) to possess some similar characters: (a) sparsity, (b) exponential-scale feature P(k) ∼α-k, and (c) small-world property. Here, we must stress a very screming, but intriguing, phenomenon that the difference of average path length (APL) between any two members in N(t) is quite small, which indicates this random connecting way among members has no great effect on APL. At the end of this article, as a new topological parameter correlated to reliability, synchronization capability and diffusion properties of networks, the number of spanning trees on a representative member NB(t) of N(t) is studied in detail, then an exact analytical solution for its spanning trees entropy is also

  20. Comparison between iterative wavefront control algorithm and direct gradient wavefront control algorithm for adaptive optics system

    International Nuclear Information System (INIS)

    Cheng Sheng-Yi; Liu Wen-Jin; Chen Shan-Qiu; Dong Li-Zhi; Yang Ping; Xu Bing

    2015-01-01

    Among all kinds of wavefront control algorithms in adaptive optics systems, the direct gradient wavefront control algorithm is the most widespread and common method. This control algorithm obtains the actuator voltages directly from wavefront slopes through pre-measuring the relational matrix between deformable mirror actuators and Hartmann wavefront sensor with perfect real-time characteristic and stability. However, with increasing the number of sub-apertures in wavefront sensor and deformable mirror actuators of adaptive optics systems, the matrix operation in direct gradient algorithm takes too much time, which becomes a major factor influencing control effect of adaptive optics systems. In this paper we apply an iterative wavefront control algorithm to high-resolution adaptive optics systems, in which the voltages of each actuator are obtained through iteration arithmetic, which gains great advantage in calculation and storage. For AO system with thousands of actuators, the computational complexity estimate is about O(n 2 ) ∼ O(n 3 ) in direct gradient wavefront control algorithm, while the computational complexity estimate in iterative wavefront control algorithm is about O(n) ∼ (O(n) 3/2 ), in which n is the number of actuators of AO system. And the more the numbers of sub-apertures and deformable mirror actuators, the more significant advantage the iterative wavefront control algorithm exhibits. (paper)

  1. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

  2. Quench detection and behaviour in case of quench in the ITER magnet systems

    International Nuclear Information System (INIS)

    Coatanea-Gouachet, M.

    2012-02-01

    The quench of one of the ITER magnet system is an irreversible transition from superconducting to normal resistive state, of a conductor. This normal zone propagates along the cable in conduit conductor dissipating a large power. The detection has to be fast enough to dump out the magnetic energy and avoid irreversible damage of the systems. The primary quench detection in ITER is based on voltage detection, which is the most rapid detection. The very magnetically disturbed environment during the plasma scenario makes the voltage detection particularly difficult, inducing large inductive components in the coils and voltage compensations have to be designed to discriminate the resistive voltage associated with the quench. A conceptual design of the quench detection based on voltage measurements is proposed for the three majors magnet systems of ITER. For this, a clear methodology was developed. It includes the classical hot spot criterion, the quench propagation study using the commercial code Gandalf and the careful estimation of the inductive disturbances by developing the TrapsAV code. Specific solutions have been proposed for the compensation in the three ITER magnet systems and for the quench detection parameters, which are the voltage threshold (in the range of 0.1 V - 0.55 V) and the holding time (in the range of 1-1.4 s). The selected values, in particular the holding time, are sufficiently high to ensure the reliability of the system and avoid fast safety discharges not induced by a quench, which is a classical problem. (author)

  3. An Empirical Comparison of Seven Iterative and Evolutionary Function Optimization Heuristics

    Science.gov (United States)

    Baluja, Shumeet

    1995-01-01

    This report is a repository of the results obtained from a large scale empirical comparison of seven iterative and evolution-based optimization heuristics. Twenty-seven static optimization problems, spanning six sets of problem classes which are commonly explored in genetic algorithm literature, are examined. The problem sets include job-shop scheduling, traveling salesman, knapsack, binpacking, neural network weight optimization, and standard numerical optimization. The search spaces in these problems range from 2368 to 22040. The results indicate that using genetic algorithms for the optimization of static functions does not yield a benefit, in terms of the final answer obtained, over simpler optimization heuristics. Descriptions of the algorithms tested and the encodings of the problems are described in detail for reproducibility.

  4. Bi-directional reflectance distribution function of a tungsten block for ITER divertor

    International Nuclear Information System (INIS)

    Iwamae, Atsushi; Ogawa, Hiroaki; Sugie, Tatsuo; Kusama, Yoshinori

    2012-02-01

    In order to investigate reflection properties on plasma-facing material in ITER, the bi-directional reflectance distribution function (BRDF) of a tungsten block sample has been measured. On the machining surface of the block, one-directional machining lines are engraved. Two laser diodes λ652 nm and λ473 nm were used to simulate H α and H β emissions, respectively. The reflected light is affected by the machining surface. The reflected light traces an arc when the incident light is injected in the parallel direction to the engraved line. On the other hand the reflected light traces a line shape when the incident light is injected in the perpendicular direction to the engraved lines. Ray tracing simulation qualitatively explains the experimental results. (author)

  5. Green`s function of Maxwell`s equations and corresponding implications for iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Singer, B.S. [Macquarie Univ., Sydney (Australia); Fainberg, E.B. [Inst. of Physics of the Earth, Moscow (Russian Federation)

    1996-12-31

    Energy conservation law imposes constraints on the norm and direction of the Hilbert space vector representing a solution of Maxwell`s equations. In this paper, we derive these constrains and discuss the corresponding implications for the Green`s function of Maxwell`s equations in a dissipative medium. It is shown that Maxwell`s equations can be reduced to an integral equation with a contracting kernel. The equation can be solved using simple iterations. Software based on this algorithm have successfully been applied to a wide range of problems dealing with high contrast models. The matrix corresponding to the integral equation has a well defined spectrum. The equation can be symmetrized and solved using different approaches, for instance one of the conjugate gradient methods.

  6. Tests of dry mechanical forepumps for use in the ITER vacuum pumping system

    International Nuclear Information System (INIS)

    Kirchhof, U.; Kammerer, B.; Perinic, D.

    1995-04-01

    This report is a description of the design and construction of FORTE (Forepumps Test Facility) which has been built in order to enable testing of the pumping speeds of prototypical mechanical forepumps connected in series, as proposed for the ITER forepump system. Three NORMETEX pumps (1300, 600, 60 m 3 /h) and one METAL BELLOWS pump (6m 3 /h) have been integrated into the test bench. Measurements of the pumping characteristics were performed, both with the single pumps and with trains of series connected pumps, using the gases N 2 , H 2 , D 2 , He as well as ITER typical gas mixture. The results of the tests are presented. (orig.)

  7. Strong and Weak Convergence Criteria of Composite Iterative Algorithms for Systems of Generalized Equilibria

    Directory of Open Access Journals (Sweden)

    Lu-Chuan Ceng

    2014-01-01

    Full Text Available We first introduce and analyze one iterative algorithm by using the composite shrinking projection method for finding a solution of the system of generalized equilibria with constraints of several problems: a generalized mixed equilibrium problem, finitely many variational inequalities, and the common fixed point problem of an asymptotically strict pseudocontractive mapping in the intermediate sense and infinitely many nonexpansive mappings in a real Hilbert space. We prove a strong convergence theorem for the iterative algorithm under suitable conditions. On the other hand, we also propose another iterative algorithm involving no shrinking projection method and derive its weak convergence under mild assumptions. Our results improve and extend the corresponding results in the earlier and recent literature.

  8. An iterative method for the solution of nonlinear systems using the Faber polynomials for annular sectors

    Energy Technology Data Exchange (ETDEWEB)

    Myers, N.J. [Univ. of Durham (United Kingdom)

    1994-12-31

    The author gives a hybrid method for the iterative solution of linear systems of equations Ax = b, where the matrix (A) is nonsingular, sparse and nonsymmetric. As in a method developed by Starke and Varga the method begins with a number of steps of the Arnoldi method to produce some information on the location of the spectrum of A. This method then switches to an iterative method based on the Faber polynomials for an annular sector placed around these eigenvalue estimates. The Faber polynomials for an annular sector are used because, firstly an annular sector can easily be placed around any eigenvalue estimates bounded away from zero, and secondly the Faber polynomials are known analytically for an annular sector. Finally the author gives three numerical examples, two of which allow comparison with Starke and Varga`s results. The third is an example of a matrix for which many iterative methods would fall, but this method converges.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. The Potential of User Feedback Through the Iterative Refining of Queries in an Image Retrieval System

    NARCIS (Netherlands)

    Ben Moussa, Maher; Pasch, Marco; Hiemstra, Djoerd; van der Vet, P.E.; Huibers, Theo W.C.; Marchand-Maillet, Stephane; Bruno, Eric; Nürnberger, Andreas; Detyniecki, Marcin

    2007-01-01

    Inaccurate or ambiguous expressions in queries lead to poor results in information retrieval. We assume that iterative user feedback can improve the quality of queries. To this end we developed a system for image retrieval that utilizes user feedback to refine the user’s search query. This is done

  11. High heat flux (HHF) elements for negative ion systems on ITER

    International Nuclear Information System (INIS)

    Milnes, J.; Chuilon, B.; Xue, Y.; Martin, D.; Waldon, C.

    2007-01-01

    Negative Ion Neutral Beam systems on ITER will require actively cooled scrapers and dumps to process and shape the beam before injection into the tokamak. The scale of the systems is much larger than any presently operating, bringing challenges for designers in terms of available sub cooling, total pressure drop, deflection and mandatory remote maintenance. High heat fluxes (∼15-20 MW/m 2 ), pulse lengths in excess of 3000 s and high number of cycles pose new challenges in terms of stress and fatigue life. The designs outlined in the Design Description Document for the ITER Neutral Beam System [N53 DDD 29 01-07-03 R 0.1. ITER Design Description Document, DDD 5.3, Neutral Beam H and CD system (including Appendices).], based on swirl tubes, have been reviewed as part of the design process and recommendations made. Additionally, alternative designs have been proposed based on the Hypervapotron high heat flux elements with modified geometry and drawing upon a vast background knowledge of large scale equipment procurement and integration. A full thermo-mechanical analysis of all HHF components has also been undertaken based on ITER design criteria and the limited material data available. The advantages and disadvantages of all designs are presented and recommendations for improvements discussed

  12. The European contribution to the procurement of the ITER Remote Handling systems

    International Nuclear Information System (INIS)

    Damiani, Carlo; Irving, Mike; Semeraro, Luigi

    2009-01-01

    Fusion for Energy (F4E) will manage the European in-kind contribution of various remote handling (RH) systems for the maintenance of ITER components: (i) the divertor cassette movers, end effectors, manipulator arms and tooling; (ii) 50% of the transfer casks, in particular the air transfer systems and some in-cask devices; (iii) the in-vessel viewing and metrology system (IVVS); (iv) the Neutral Beam (NB) Cell crane, manipulator arms, tooling, Caesium Oven replacement tooling, NB source installation/removal trolley, auxiliary vehicles. A wide range of technologies is involved: special monorail crane, movers, manipulator arms, pipe cutting/welding tooling, special cameras, laser-based metrology devices, control systems, virtual reality. An important aspect to consider is the resistance to radiation levels that range from max ∼10 KGy/h for IVVS down to ∼1 Gy/h for the RH devices operating in the NB cell. Given the unprecedented complexity of the ITER maintenance scenario, a development strategy is being implemented that includes prototyping and testing of RH subsystems before proceeding with the final production for ITER. This paper presents an overview of the various procurement packages, the status of development for each of them, the validation and procurement strategy, including issues like radiation resistance and standardisation policy, and the organisational and managerial challenges in relation with the complex ITER Organisation (IO).

  13. Low complexity variational bayes iterative reviver for MIMO-OFDM systems

    DEFF Research Database (Denmark)

    Xiong, Chunlin; Wang, Hua; Zhang, Xiaoying

    2009-01-01

    A low complexity iterative receiver is proposed in this paper for MIMO-OFDM systems in time-varying multi-path channel based on the variational Bayes (VB) method. According to the VB method, the estimation algorithms of the signal distribution and the channel distribution are derived for the rece...

  14. Solving Ratio-Dependent Predatorprey System with Constant Effort Harvesting Using Variational Iteration Method

    DEFF Research Database (Denmark)

    Ghotbi, Abdoul R; Barari, Amin

    2009-01-01

    Due to wide range of interest in use of bio-economic models to gain insight in to the scientific management of renewable resources like fisheries and forestry, variational iteration method (VIM) is employed to approximate the solution of the ratio-dependent predator-prey system with constant effort...

  15. Proposed configuration for ITER hydrogen isotope separation system (ISS)

    International Nuclear Information System (INIS)

    Lazar, A.; Brad, S.; Sofalca, N.; Vijulie, M.; Cristescu, I.; Doer, L; Wurster, W.

    2008-01-01

    Full text: The isotope separation system utilizes cryogenic distillation and catalytic reaction for isotope exchange to separate elemental hydrogen isotope gas mixtures. The ISS shall separate hydrogen isotope mixtures from two sources to produce up to five different products. These are: protium, effluent for discharge to the atmosphere, deuterium for fuelling, deuterium for NB injector (NBI) source gas, 50 % and 90% T fuelling streams. The concept of equipment 3D layout for the ISS main components were developed using the Part Design, Assembly Design, Piping Design, Equipment Arrangement and Plant Layout application from CATIA V5. The 3D conceptual layouts for ISS system were created having as reference the DDD -32-B report, the drawings 0028.0001.2D. 0100. R 'Process Flow Diagram'; 0029.0001.2D. 0200.R 'Process Instrumentation Diagram -1' (in the cold box); 0030.0001.2D. 0100. R 'Process Instrumentation Diagram -2' (in the hard shell confinement) and imputes from TLK team. The main components designed for ISS are: ISS cold box system (CB) with cryogenic distillation columns (CD) and recovery heat exchangers (HX), ISS hard shell containment (HSC) system with metals bellow pumps (MB) and chemical equilibrators (RC), valve box system, instrumentation box system, vacuum system and hydrogen expansion vessels. Work related to these topics belongs to the contract FU06-CT-2006-00508 (EFDA 06-1511) from the EFDA Technology Workprogramm 2006 and was done in collaboration with FZK Association team during the period January 2007 - September 2008. (authors)

  16. A new 2-d approach to iterative , learning control system

    International Nuclear Information System (INIS)

    Ashraf, S.; Muhammad, E.; Tasleem, M.

    2004-01-01

    The well known two-dimensional system theory is used to analyze and develop a class of learning control system. In this paper we first explore and test a method given by ZHENG and JAMSHIDI. In that paper all the input samples are treated at once. In comparison our paper presents a scheme in which one sample at a time is treated. The 2- D state-space model of proposed learning control scheme is given. An important consequence of the proposed scheme is that given the right choice of gain matrix and sampling time the system's output can be made to converge to any degree of accuracy. (author)

  17. NetCDF based data archiving system applied to ITER Fast Plant System Control prototype

    International Nuclear Information System (INIS)

    Castro, R.; Vega, J.; Ruiz, M.; De Arcas, G.; Barrera, E.; López, J.M.; Sanz, D.; Gonçalves, B.; Santos, B.; Utzel, N.; Makijarvi, P.

    2012-01-01

    Highlights: ► Implementation of a data archiving solution for a Fast Plant System Controller (FPSC) for ITER CODAC. ► Data archiving solution based on scientific NetCDF-4 file format and Lustre storage clustering. ► EPICS control based solution. ► Tests results and detailed analysis of using NetCDF-4 and clustering technologies on fast acquisition data archiving. - Abstract: EURATOM/CIEMAT and Technical University of Madrid (UPM) have been involved in the development of a FPSC (Fast Plant System Control) prototype for ITER, based on PXIe (PCI eXtensions for Instrumentation). One of the main focuses of this project has been data acquisition and all the related issues, including scientific data archiving. Additionally, a new data archiving solution has been developed to demonstrate the obtainable performances and possible bottlenecks of scientific data archiving in Fast Plant System Control. The presented system implements a fault tolerant architecture over a GEthernet network where FPSC data are reliably archived on remote, while remaining accessible to be redistributed, within the duration of a pulse. The storing service is supported by a clustering solution to guaranty scalability, so that FPSC management and configuration may be simplified, and a unique view of all archived data provided. All the involved components have been integrated under EPICS (Experimental Physics and Industrial Control System), implementing in each case the necessary extensions, state machines and configuration process variables. The prototyped solution is based on the NetCDF-4 (Network Common Data Format) file format in order to incorporate important features, such as scientific data models support, huge size files management, platform independent codification, or single-writer/multiple-readers concurrency. In this contribution, a complete description of the above mentioned solution is presented, together with the most relevant results of the tests performed, while focusing in the

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  19. Natural Preconditioning and Iterative Methods for Saddle Point Systems

    KAUST Repository

    Pestana, Jennifer; Wathen, Andrew J.

    2015-01-01

    or the discrete setting, so saddle point systems arising from the discretization of partial differential equation problems, such as those describing electromagnetic problems or incompressible flow, lead to equations with this structure, as do, for example

  20. Conceptual design of the radial gamma ray spectrometers system for α particle and runaway electron measurements at ITER

    Science.gov (United States)

    Nocente, M.; Tardocchi, M.; Barnsley, R.; Bertalot, L.; Brichard, B.; Croci, G.; Brolatti, G.; Di Pace, L.; Fernandes, A.; Giacomelli, L.; Lengar, I.; Moszynski, M.; Krasilnikov, V.; Muraro, A.; Pereira, R. C.; Perelli Cippo, E.; Rigamonti, D.; Rebai, M.; Rzadkiewicz, J.; Salewski, M.; Santosh, P.; Sousa, J.; Zychor, I.; Gorini, G.

    2017-07-01

    We here present the principles and main physics capabilities behind the design of the radial gamma ray spectrometers (RGRS) system for alpha particle and runaway electron measurements at ITER. The diagnostic benefits from recent advances in gamma-ray spectrometry for tokamak plasmas and combines space and high energy resolution in a single device. The RGRS system as designed can provide information on α ~ particles on a time scale of 1/10 of the slowing down time for the ITER 500 MW full power DT scenario. Spectral observations of the 3.21 and 4.44 MeV peaks from the 9\\text{Be}≤ft(α,nγ \\right){{}12}\\text{C} reaction make the measurements sensitive to α ~ particles at characteristic resonant energies and to possible anisotropies of their slowing down distribution function. An independent assessment of the neutron rate by gamma-ray emission is also feasible. In case of runaway electrons born in disruptions with a typical duration of 100 ms, a time resolution of at least 10 ms for runaway electron studies can be achieved depending on the scenario and down to a current of 40 kA by use of external gas injection. We find that the bremsstrahlung spectrum in the MeV range from confined runaways is sensitive to the electron velocity space up to E≈ 30 -40 MeV, which allows for measurements of the energy distribution of the runaway electrons at ITER.

  1. Constructing Frozen Jacobian Iterative Methods for Solving Systems of Nonlinear Equations, Associated with ODEs and PDEs Using the Homotopy Method

    Directory of Open Access Journals (Sweden)

    Uswah Qasim

    2016-03-01

    Full Text Available A homotopy method is presented for the construction of frozen Jacobian iterative methods. The frozen Jacobian iterative methods are attractive because the inversion of the Jacobian is performed in terms of LUfactorization only once, for a single instance of the iterative method. We embedded parameters in the iterative methods with the help of the homotopy method: the values of the parameters are determined in such a way that a better convergence rate is achieved. The proposed homotopy technique is general and has the ability to construct different families of iterative methods, for solving weakly nonlinear systems of equations. Further iterative methods are also proposed for solving general systems of nonlinear equations.

  2. Magnetic analysis of the magnetic field reduction system of the ITER neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, Germán, E-mail: german.barrera@ciemat.es [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Ahedo, Begoña; Alonso, Javier; Ríos, Luis [CIEMAT, Laboratorio Nacional de Fusión, Avda. Complutense 22, 28040 Madrid (Spain); Chareyre, Julien; El-Ouazzani, Anass [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 neutral beam system for ITER consists of two heating and current drive neutral beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. For the correct operation of the beam, the ion source and the ion path until it is neutralized must operate under a very low magnetic field environment. To prevent the stray ITER field from penetrating inside those mentioned critical areas, a magnetic field reduction system (MFRS) will envelop the beam vessels and the high voltage transmission lines to ion source. This system comprises the passive magnetic shield (PMS), a box like assembly of thick low carbon steel plates, and the Active Correction and Compensation Coils (ACCC), a set of coils carrying a current which depends on the tokamak stray field. This paper describes the magnetic model and analysis results presented at the PMS and ACCC preliminary design review held in ITER organization in April 2013. The paper focuses on the magnetic model description and on the description of the analysis results. The iterative process for obtaining optimized currents in the coils is presented. The set of coils currents chosen among the many possible solutions, the magnetic field results in the interest regions and the fulfillment of the magnetic field requirements are described.

  3. The iteration formula of the Maslov-type index theory with applications to nonlinear Hamiltonian systems

    International Nuclear Information System (INIS)

    Di Dong; Yiming Long.

    1994-10-01

    In this paper, the iteration formula of the Maslov-type index theory for linear Hamiltonian systems with continuous periodic and symmetric coefficients is established. This formula yields a new method to determine the minimality of the period for solutions of nonlinear autonomous Hamiltonian systems via their Maslov-type indices. Applications of this formula give new results on the existence of periodic solutions with prescribed minimal period for such systems. (author). 40 refs

  4. Iterative absolute electroanalytical approach to characterization of bulk redox conducting systems.

    Science.gov (United States)

    Lewera, Adam; Miecznikowski, Krzysztof; Chojak, Malgorzata; Makowski, Oktawian; Golimowski, Jerzy; Kulesza, Pawel J

    2004-05-15

    A novel electroanalytical approach is proposed here, and it is demonstrated with the direct and simultaneous determination of two unknowns: the concentration of redox sites and the apparent diffusion coefficient for charge propagation in a single crystal of dodecatungstophosphoric acid. This Keggin-type polyoxometalate serves as a model bulk redox conducting inorganic material for solid-state voltammetry. The system has been investigated using an ultramicrodisk working electrode in the absence of external liquid supporting electrolyte. The analytical method requires numerical solution of the combination of two equations in which the first one describes current (or charge) in a well-defined (either spherical or linear) diffusional regime and the second general equation describes chronoamperometric (or normal pulse voltammetric current) under mixed (linear-spherical) conditions. The iterative approach is based on successive approximations through calculation and minimizing the least-squares error function. The method is fairly universal, and in principle, it can be extended to the investigation of other bulk systems including sol-gel processed materials, redox melts, and solutions on condition that they are electroactive and well behaved, they contain redox centers at sufficiently high level, and a number of electrons for the redox reaction considered is known.

  5. High voltage power supplies for ITER RF heating and current drive systems

    International Nuclear Information System (INIS)

    Gassmann, T.; Arambhadiya, B.; Beaumont, B.; Baruah, U.K.; Bonicelli, T.; Darbos, C.; Purohit, D.; Decamps, H.; Albajar, F.; Gandini, F.; Henderson, M.; Kazarian, F.; Lamalle, P.U.; Omori, T.; Parmar, D.; Patel, A.; Rathi, D.; Singh, N.P.

    2011-01-01

    The RF heating and current drive (H and CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation. These HVPS should deliver around 50 MW electrical power to each of the RF H and CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems. This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.

  6. Variational Iterative Methods for Nonsymmetric Systems of Linear Equations.

    Science.gov (United States)

    1981-08-01

    With a third matrix-vector product, b(i) can be computed as i j ( ATAr i+l’pj)/ApjpApj), and the previous (Apj) need not be saved. Page 8 I OCR I Orthomin... Economics and Mathematical Systems, Volume 134, Springer-Verlag, Berlin, 1976. [51 Paul Concus, Gene H. Golub, and Dianne P. O’Leary. A generalized

  7. ANAEROBIC COMPOST CONSTRUCTED WETLANDS SYSTEM TECHNOLOGY - SITE ITER

    Science.gov (United States)

    In Fall 1994, anaerobic compost wetlands in both upflow and downflow configurations were constructed adjacent to and received drainage from the Burleigh Tunnel, which forms part of the Clear Creek/Central City Superfund site. The systems were operated over a 3 year period. The e...

  8. Advanced Data Acquisition System Implementation for the ITER Neutron Diagnostic Use Case Using EPICS and FlexRIO Technology on a PXIe Platform

    Science.gov (United States)

    Sanz, D.; Ruiz, M.; Castro, R.; Vega, J.; Afif, M.; Monroe, M.; Simrock, S.; Debelle, T.; Marawar, R.; Glass, B.

    2016-04-01

    To aid in assessing the functional performance of ITER, Fission Chambers (FC) based on the neutron diagnostic use case deliver timestamped measurements of neutron source strength and fusion power. To demonstrate the Plant System Instrumentation & Control (I&C) required for such a system, ITER Organization (IO) has developed a neutron diagnostics use case that fully complies with guidelines presented in the Plant Control Design Handbook (PCDH). The implementation presented in this paper has been developed on the PXI Express (PXIe) platform using products from the ITER catalog of standard I&C hardware for fast controllers. Using FlexRIO technology, detector signals are acquired at 125 MS/s, while filtering, decimation, and three methods of neutron counting are performed in real-time via the onboard Field Programmable Gate Array (FPGA). Measurement results are reported every 1 ms through Experimental Physics and Industrial Control System (EPICS) Channel Access (CA), with real-time timestamps derived from the ITER Timing Communication Network (TCN) based on IEEE 1588-2008. Furthermore, in accordance with ITER specifications for CODAC Core System (CCS) application development, the software responsible for the management, configuration, and monitoring of system devices has been developed in compliance with a new EPICS module called Nominal Device Support (NDS) and RIO/FlexRIO design methodology.

  9. Using Functional Electrical Stimulation Mediated by Iterative Learning Control and Robotics to Improve Arm Movement for People With Multiple Sclerosis.

    Science.gov (United States)

    Sampson, Patrica; Freeman, Chris; Coote, Susan; Demain, Sara; Feys, Peter; Meadmore, Katie; Hughes, Ann-Marie

    2016-02-01

    Few interventions address multiple sclerosis (MS) arm dysfunction but robotics and functional electrical stimulation (FES) appear promising. This paper investigates the feasibility of combining FES with passive robotic support during virtual reality (VR) training tasks to improve upper limb function in people with multiple sclerosis (pwMS). The system assists patients in following a specified trajectory path, employing an advanced model-based paradigm termed iterative learning control (ILC) to adjust the FES to improve accuracy and maximise voluntary effort. Reaching tasks were repeated six times with ILC learning the optimum control action from previous attempts. A convenience sample of five pwMS was recruited from local MS societies, and the intervention comprised 18 one-hour training sessions over 10 weeks. The accuracy of tracking performance without FES and the amount of FES delivered during training were analyzed using regression analysis. Clinical functioning of the arm was documented before and after treatment with standard tests. Statistically significant results following training included: improved accuracy of tracking performance both when assisted and unassisted by FES; reduction in maximum amount of FES needed to assist tracking; and less impairment in the proximal arm that was trained. The system was well tolerated by all participants with no increase in muscle fatigue reported. This study confirms the feasibility of FES combined with passive robot assistance as a potentially effective intervention to improve arm movement and control in pwMS and provides the basis for a follow-up study.

  10. An iterative algorithm for fuzzy mixed production planning based on the cumulative membership function

    Directory of Open Access Journals (Sweden)

    Juan Carlos Figueroa García

    2011-12-01

    The presented approach uses an iterative algorithm which finds stable solutions to problems with fuzzy parameter sinboth sides of an FLP problem. The algorithm is based on the soft constraints method proposed by Zimmermann combined with an iterative procedure which gets a single optimal solution.

  11. 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.)

  12. ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates

    International Nuclear Information System (INIS)

    Eriksson, John; Sjoeberg, A.; Sponton, L.L.

    2001-05-01

    An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls

  13. Functional electrical stimulation mediated by iterative learning control and 3D robotics reduces motor impairment in chronic stroke

    Directory of Open Access Journals (Sweden)

    Meadmore Katie L

    2012-06-01

    Full Text Available Abstract Background Novel stroke rehabilitation techniques that employ electrical stimulation (ES and robotic technologies are effective in reducing upper limb impairments. ES is most effective when it is applied to support the patients’ voluntary effort; however, current systems fail to fully exploit this connection. This study builds on previous work using advanced ES controllers, and aims to investigate the feasibility of Stimulation Assistance through Iterative Learning (SAIL, a novel upper limb stroke rehabilitation system which utilises robotic support, ES, and voluntary effort. Methods Five hemiparetic, chronic stroke participants with impaired upper limb function attended 18, 1 hour intervention sessions. Participants completed virtual reality tracking tasks whereby they moved their impaired arm to follow a slowly moving sphere along a specified trajectory. To do this, the participants’ arm was supported by a robot. ES, mediated by advanced iterative learning control (ILC algorithms, was applied to the triceps and anterior deltoid muscles. Each movement was repeated 6 times and ILC adjusted the amount of stimulation applied on each trial to improve accuracy and maximise voluntary effort. Participants completed clinical assessments (Fugl-Meyer, Action Research Arm Test at baseline and post-intervention, as well as unassisted tracking tasks at the beginning and end of each intervention session. Data were analysed using t-tests and linear regression. Results From baseline to post-intervention, Fugl-Meyer scores improved, assisted and unassisted tracking performance improved, and the amount of ES required to assist tracking reduced. Conclusions The concept of minimising support from ES using ILC algorithms was demonstrated. The positive results are promising with respect to reducing upper limb impairments following stroke, however, a larger study is required to confirm this.

  14. Fault-tolerant design of local controller for the poloidal field converter control system on ITER

    International Nuclear Information System (INIS)

    Shen, Jun; Fu, Peng; Gao, Ge; He, Shiying; Huang, Liansheng; Zhu, Lili; Chen, Xiaojiao

    2016-01-01

    Highlights: • The requirements on the Local Control Cubicles (LCC) for ITER Poloidal Field Converter are analyzed. • Decoupled service-based software architecture is proposed to make control loops on LCC running at varying cycle-time. • Fault detection and recovery methods for the LCC are developed to enhance the system. • The performance of the LCC with or without fault-tolerant feature is tested and compared. - Abstract: The control system for the Poloidal Field (PF) on ITER is a synchronously networked control system, which has several kinds of computational controllers. The Local Control Cubicles (LCC) play a critical role in the networked control system for they are the interface to all input and output signals. Thus, some additional work must be done to guarantee the LCCs proper operation under influence of faults. This paper mainly analyzes the system demands of the LCCs and faults which have been encountered recently. In order to handle these faults, decoupled service-based software architecture has been proposed. Based on this architecture, fault detection and system recovery methods, such as redundancy and rejuvenation, have been incorporated to achieve a fault-tolerant private network with the aid of QNX operating system. Unlike the conventional method, this method requires no additional hardware and can be achieved relatively easily. To demonstrate effectiveness the LCCs have been successfully tested during the recent PF Converter Unit performance tests for ITER.

  15. Fault-tolerant design of local controller for the poloidal field converter control system on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Jun; Fu, Peng; Gao, Ge; He, Shiying; Huang, Liansheng, E-mail: huangls@ipp.ac.cn; Zhu, Lili; Chen, Xiaojiao

    2016-11-15

    Highlights: • The requirements on the Local Control Cubicles (LCC) for ITER Poloidal Field Converter are analyzed. • Decoupled service-based software architecture is proposed to make control loops on LCC running at varying cycle-time. • Fault detection and recovery methods for the LCC are developed to enhance the system. • The performance of the LCC with or without fault-tolerant feature is tested and compared. - Abstract: The control system for the Poloidal Field (PF) on ITER is a synchronously networked control system, which has several kinds of computational controllers. The Local Control Cubicles (LCC) play a critical role in the networked control system for they are the interface to all input and output signals. Thus, some additional work must be done to guarantee the LCCs proper operation under influence of faults. This paper mainly analyzes the system demands of the LCCs and faults which have been encountered recently. In order to handle these faults, decoupled service-based software architecture has been proposed. Based on this architecture, fault detection and system recovery methods, such as redundancy and rejuvenation, have been incorporated to achieve a fault-tolerant private network with the aid of QNX operating system. Unlike the conventional method, this method requires no additional hardware and can be achieved relatively easily. To demonstrate effectiveness the LCCs have been successfully tested during the recent PF Converter Unit performance tests for ITER.

  16. Commissioning of Water Detritiation and Cryogenic Distillation Systems at TLK in View of ITER Design

    International Nuclear Information System (INIS)

    Cristescu, I.; Doerr, L.; Glugla, M.; Hellriegel, G.; Schaefer, P.; Welte, St.; Wurster, W.; Murdoch, D.

    2006-01-01

    The Water Detritiation System (WDS) of ITER is one of the key systems to control the tritium content in the effluents streams, to recover as much tritium as possible and consequently to minimize the impact on the environment. In order to mitigate the concern over tritium release into the environment during pulsed operation of the Torus, the WDS and Isotope Separation System (ISS) will operate in such way that WDS will be a final barrier for the processed protium waste gas stream discharged from ISS. The ITER ISS consists of a cascade of four cryogenic distillation columns with the aim to process mainly two gas streams, one from Torus exhaust and other from WDS mixed with the returned stream from Neutral Beam Injectors (NBI). The behavior of the CD cascade has to be characterized with high accuracy with respect to thermal and isotopic fluctuations during Torus pulses. To support the research activities needed to characterize the performances of various components for WDS and ISS processes in various working conditions and configurations as needed for ITER design, an experimental facility called TRENTA based on the combination Combined Electrolysis Catalytic Exchange (CECE) - Cryogenic Distillation (CD), representative of the ITER WDS and ISS protium separation column, is under full commissioning at TLK. The CECE process consists of a solid polymer electrolyser unit as envisaged to be used in ITER WDS, and an 8 m Liquid Phase Catalytic Exchange Column (LPCE). The Electrolysis unit was commissioned with tritiated water and the enrichment factor was measured. The experimental program on the Cryogenic distillation facility at TLK is conducted to provide the necessary design and operation information for ITER ISS. It is focused on two major issues: - To investigate the separation performances and liquid hold up of different packings in cryogenic distillation process and to validate the steady-state mathematical modeling of the process. - To investigate the CD process

  17. The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

    International Nuclear Information System (INIS)

    Carvalho, B.B.; Santos, B.; Carvalho, P.F.; Neto, A.; Boncagni, L.; Batista, A.J.N.; Correia, M.; Sousa, J.; Gonçalves, B.

    2013-01-01

    Highlights: ► High performance ATCA systems for fast control and data acquisition. ► IEEE1588 timing system and synchronization. ► Plasma control algorithms. ► Real-time control software frameworks. ► Targeted for nuclear fusion experiments with long duration discharges. -- Abstract: IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I and C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards. The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management. This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed

  18. Application of the variational iteration method for system of initial value problems delay differential equations

    Science.gov (United States)

    Yousef, Hamood. M.; Ismail, A. I. B. MD.

    2017-08-01

    Many attempts have been presented to solve the system of Delay Differential Equations (DDE) with Initial Value Problem. As a result, it has shown difficulties when getting the solution or cannot be solved. In this paper, a Variational Iteration Method is employed to find out an approximate solution for the system of DDE with initial value problems. The example illustrates convenient and an efficiency comparison with the exact solution.

  19. The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, B.B., E-mail: bernardo@ipfn.ist.utl.pt [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Santos, B.; Carvalho, P.F.; Neto, A. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Boncagni, L. [Associazione Euratom-ENEA sulla Fusione, Frascati Research Centre, Division of Fusion Physics, Frascati, Rome (Italy); Batista, A.J.N.; Correia, M.; Sousa, J.; Gonçalves, B. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► High performance ATCA systems for fast control and data acquisition. ► IEEE1588 timing system and synchronization. ► Plasma control algorithms. ► Real-time control software frameworks. ► Targeted for nuclear fusion experiments with long duration discharges. -- Abstract: IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I and C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards. The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management. This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed.

  20. Criticality calculations by source-collision iteration technique for cylindrical systems

    International Nuclear Information System (INIS)

    Sundaram, V.K.; Gopinath, D.V.

    1977-01-01

    A fast-converging iterative technique is presented which uses first collision probabilities developed for obtaining criticality parameters in two-region cylindrical systems with multigroup structure in energy of the neutrons. The space transmission matrix is obtained part analytically and part numerically through evaluation of a single-fold integral. Critical dimensions for condensed systems of uranium and plutonium computed using this method are presented and compared with published values

  1. Status of the IPP RF Negative Ion Source Development for the ITER NBI System

    International Nuclear Information System (INIS)

    Peter Franzen, P.; Falter, H.-D.; Fantz, U.

    2006-01-01

    For heating and current drive the ITER neutral beam system requires negative hydrogen ion sources capable of delivering above 40 A of D - ions from a 1.5 x 0.6 m 2 source for up to one hour pulses with an accelerated current density of 200 A/m 2 . In order to reduce the losses by electron stripping in the acceleration system and the power loading of the grids, the source pressure is required to be 0.3 Pa at an electron/ion ratio 2 H - / 230 A/m 2 D - ) in excess of the ITER requirements have been already achieved on the small test facility '' BATMAN '' (Bavarian Test Machine for Negative Ions) at the required source pressure (0.3 Pa) and electron/ion ratio ( 2 ) and limited pulse length ( 2 and the pulse length up to 3600 s, using the same source as it is used at BATMAN. In order to demonstrate the required homogeneity of a large RF plasma source as well as the operation of an ITER relevant RF circuit, a so called '' half-size source '' - with roughly the width and half the height of the ITER source - was designed and went into operation on a dedicated plasma source test bed ('' RADI ''). An extensive diagnostic and modelling programme is accompanying those activities. The paper will present as an overview a summary of the latest results of the RF source development, with an emphasis on the first results of the operation of the half size ITER source and on the status of the long pulse operation. The details will be presented in several other papers. (author)

  2. Performance and Complexity Evaluation of Iterative Receiver for Coded MIMO-OFDM Systems

    Directory of Open Access Journals (Sweden)

    Rida El Chall

    2016-01-01

    Full Text Available Multiple-input multiple-output (MIMO technology in combination with channel coding technique is a promising solution for reliable high data rate transmission in future wireless communication systems. However, these technologies pose significant challenges for the design of an iterative receiver. In this paper, an efficient receiver combining soft-input soft-output (SISO detection based on low-complexity K-Best (LC-K-Best decoder with various forward error correction codes, namely, LTE turbo decoder and LDPC decoder, is investigated. We first investigate the convergence behaviors of the iterative MIMO receivers to determine the required inner and outer iterations. Consequently, the performance of LC-K-Best based receiver is evaluated in various LTE channel environments and compared with other MIMO detection schemes. Moreover, the computational complexity of the iterative receiver with different channel coding techniques is evaluated and compared with different modulation orders and coding rates. Simulation results show that LC-K-Best based receiver achieves satisfactory performance-complexity trade-offs.

  3. The ITER Equatorial Visible/Infra-Red Wide Angle Viewing System: Status of design and R&D

    Energy Technology Data Exchange (ETDEWEB)

    Salasca, Sophie, E-mail: sophie.salasca@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Aumeunier, Marie-Helene; Benoit, Fabrice; Cantone, Bruno; Corre, Yann; Delchambre, Elise; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Houtte, Didier van; Keller, Delphine; Labasse, Florence; Larroque, Sebastien; Loarer, Thierry; Micolon, Frederic; Peluso, Bertrand; Proust, Maxime [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Blanchet, David; Peneliau, Yannick [CEA, DEN/DER, F-13108 Saint-Paul-lez-Durance (France); Alonso, Javier [CIEMAT, Avda. Complutense, 40, Madrid 28040 (Spain); and others

    2015-10-15

    Highlights: • The status of Equatorial Visible/Infra-Red Wide Angle Viewing System is presented. • An assessment of measurement parameters relevant for machine protection has been done. • Remaining uncertainties will be clarified during the System Level Design (SLD). • WAVS design is not considered mature enough to launch prototypes of subcomponents. • Mandatory prototypes and qualification tests are already identified. • Next stage (SLD) will enable to do trade-offs and address pending design issues. - Abstract: The Equatorial Visible/Infra-Red Wide Angle Viewing System (WAVS) is one of the ITER key diagnostics owing to its role in machine investment protection through the monitoring of Plasma Facing Components (PFCs) by Infra-Red thermography and visible imaging. Foreseen to be installed in 4 equatorial port plugs to maximize the coverage of divertor, first wall, heating antennas and upper strike zone, the WAVS will likely be composed of 15 lines of sight and 15 optical systems transferring the light along several meters from the PFCs through the port plug and interspace up to detectors located in the port cell. After a conceptual design phase led by ITER Organization, the design is being further developed through a Framework Partnership Agreement signed between the European Domestic Agency, Fusion for Energy, and a consortium gathering CEA, CIEMAT (with INTA as third party) and Bertin Technologies company. The next design step is the System Level Design (SLD) which will enable to consolidate the WAVS specifications as well as the performance realistically achievable (taking into account ITER and project constraints). The SLD has been preceded by a preparatory phase aiming at clarifying the WAVS functions and identifying critical prototyping. The outcomes of this preparatory phase are reported in this paper. First a review by the consortium of the WAVS measurement specifications is presented, for the purpose of a clearer separation of measurement

  4. A fast iterative recursive least squares algorithm for Wiener model identification of highly nonlinear systems.

    Science.gov (United States)

    Kazemi, Mahdi; Arefi, Mohammad Mehdi

    2017-03-01

    In this paper, an online identification algorithm is presented for nonlinear systems in the presence of output colored noise. The proposed method is based on extended recursive least squares (ERLS) algorithm, where the identified system is in polynomial Wiener form. To this end, an unknown intermediate signal is estimated by using an inner iterative algorithm. The iterative recursive algorithm adaptively modifies the vector of parameters of the presented Wiener model when the system parameters vary. In addition, to increase the robustness of the proposed method against variations, a robust RLS algorithm is applied to the model. Simulation results are provided to show the effectiveness of the proposed approach. Results confirm that the proposed method has fast convergence rate with robust characteristics, which increases the efficiency of the proposed model and identification approach. For instance, the FIT criterion will be achieved 92% in CSTR process where about 400 data is used. Copyright © 2016 ISA. Published by Elsevier Ltd. All rights reserved.

  5. Integration of an iterative methodology for exergoeconomic improvement of thermal systems with a process simulator

    International Nuclear Information System (INIS)

    Vieira, Leonardo S.; Donatelli, Joao L.; Cruz, Manuel E.

    2004-01-01

    In this paper, we present the development and automated implementation of an iterative methodology for exergoeconomic improvement of thermal systems integrated with a process simulator, so as to be applicable to real, complex plants. The methodology combines recent available exergoeconomic techniques with new qualitative and quantitative criteria for the following tasks: (i) identification of decision variables that affect system total cost and exergetic efficiency; (ii) hierarchical classification of components; (iii) identification of predominant terms in the component total cost; and (iv) choice of main decision variables in the iterative process. To show the strengths and potential advantages of the proposed methodology, it is here applied to the benchmark CGAM cogeneration system. The results obtained are presented and discussed in detail and are compared to those reached using a mathematical optimization procedure

  6. Design and RF Test of Broadband Coaxial Hybrid Splitter for ITER ICRF System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. J.; Wang, S. J.; Park, B. H.; Yang, H. L.; Kwak, J. G. [National Fusion Research Institute, Daejeon (Korea, Republic of); Choi, J. J. [Kwangwoon Univ., Seoul (Korea, Republic of)

    2013-10-15

    The ICRF system of the ITER is required to couple 20 MW to the plasma in the 40∼55 MHz frequency band for RF heating and current drive operation. The corresponding matching system of ICRF antenna must be load-resilient for a wide range of antenna load variations due to mode transitions or edge localized modes. Indeed the use of hybrid splitters ensures that no reflections occur at the generator when the reflections on the adjacent lines are equal both in magnitude and in phase, in which case all reflected power will not be seen by the generators and will be returned to the dummy loads. Most 3 dB coaxial hybrid circuits installed and implemented on the ICRF system is single section coupler providing best performance at the design frequency with narrow bandwidth. The bandwidth of such a single-section 3 dB hybrid coupler is limited to less than 20% due to the quarter wavelength transmission line requirement. The amplitude balance becomes rapidly degraded away from the center frequency. We designed, fabricated and tested a high power, ultra-wideband two-section 3 dB coaxial hybrid coupler over all frequencies from 40 MHz to 55 MHz for ITER ICRF system by configuring asymmetric impedance matching. We have designed, fabricated, and tested a 3-dB wideband hybrid coupler for stable and load resilient operation of the ITER ICRF system. The wideband two section 3-dB coaxial hybrid coupler was well designed by configuring asymmetric impedance matching using HFSS. In the rf measurements, we found that wideband hybrid splitter has an amplitude imbalance of 0.1 dB over all frequencies from 40 MHz to 55 MHz. We expect that wideband hybrid splitter will be applicable to ITER ICRF matching system for load resilient operation at fusion plasmas.

  7. Comparison between iterative wavefront control algorithm and direct gradient wavefront control algorithm for adaptive optics system

    Science.gov (United States)

    Cheng, Sheng-Yi; Liu, Wen-Jin; Chen, Shan-Qiu; Dong, Li-Zhi; Yang, Ping; Xu, Bing

    2015-08-01

    Among all kinds of wavefront control algorithms in adaptive optics systems, the direct gradient wavefront control algorithm is the most widespread and common method. This control algorithm obtains the actuator voltages directly from wavefront slopes through pre-measuring the relational matrix between deformable mirror actuators and Hartmann wavefront sensor with perfect real-time characteristic and stability. However, with increasing the number of sub-apertures in wavefront sensor and deformable mirror actuators of adaptive optics systems, the matrix operation in direct gradient algorithm takes too much time, which becomes a major factor influencing control effect of adaptive optics systems. In this paper we apply an iterative wavefront control algorithm to high-resolution adaptive optics systems, in which the voltages of each actuator are obtained through iteration arithmetic, which gains great advantage in calculation and storage. For AO system with thousands of actuators, the computational complexity estimate is about O(n2) ˜ O(n3) in direct gradient wavefront control algorithm, while the computational complexity estimate in iterative wavefront control algorithm is about O(n) ˜ (O(n)3/2), in which n is the number of actuators of AO system. And the more the numbers of sub-apertures and deformable mirror actuators, the more significant advantage the iterative wavefront control algorithm exhibits. Project supported by the National Key Scientific and Research Equipment Development Project of China (Grant No. ZDYZ2013-2), the National Natural Science Foundation of China (Grant No. 11173008), and the Sichuan Provincial Outstanding Youth Academic Technology Leaders Program, China (Grant No. 2012JQ0012).

  8. Determination of an effective scoring function for RNA-RNA interactions with a physics-based double-iterative method.

    Science.gov (United States)

    Yan, Yumeng; Wen, Zeyu; Zhang, Di; Huang, Sheng-You

    2018-05-18

    RNA-RNA interactions play fundamental roles in gene and cell regulation. Therefore, accurate prediction of RNA-RNA interactions is critical to determine their complex structures and understand the molecular mechanism of the interactions. Here, we have developed a physics-based double-iterative strategy to determine the effective potentials for RNA-RNA interactions based on a training set of 97 diverse RNA-RNA complexes. The double-iterative strategy circumvented the reference state problem in knowledge-based scoring functions by updating the potentials through iteration and also overcame the decoy-dependent limitation in previous iterative methods by constructing the decoys iteratively. The derived scoring function, which is referred to as DITScoreRR, was evaluated on an RNA-RNA docking benchmark of 60 test cases and compared with three other scoring functions. It was shown that for bound docking, our scoring function DITScoreRR obtained the excellent success rates of 90% and 98.3% in binding mode predictions when the top 1 and 10 predictions were considered, compared to 63.3% and 71.7% for van der Waals interactions, 45.0% and 65.0% for ITScorePP, and 11.7% and 26.7% for ZDOCK 2.1, respectively. For unbound docking, DITScoreRR achieved the good success rates of 53.3% and 71.7% in binding mode predictions when the top 1 and 10 predictions were considered, compared to 13.3% and 28.3% for van der Waals interactions, 11.7% and 26.7% for our ITScorePP, and 3.3% and 6.7% for ZDOCK 2.1, respectively. DITScoreRR also performed significantly better in ranking decoys and obtained significantly higher score-RMSD correlations than the other three scoring functions. DITScoreRR will be of great value for the prediction and design of RNA structures and RNA-RNA complexes.

  9. Iterative learning-based decentralized adaptive tracker for large-scale systems: a digital redesign approach.

    Science.gov (United States)

    Tsai, Jason Sheng-Hong; Du, Yan-Yi; Huang, Pei-Hsiang; Guo, Shu-Mei; Shieh, Leang-San; Chen, Yuhua

    2011-07-01

    In this paper, a digital redesign methodology of the iterative learning-based decentralized adaptive tracker is proposed to improve the dynamic performance of sampled-data linear large-scale control systems consisting of N interconnected multi-input multi-output subsystems, so that the system output will follow any trajectory which may not be presented by the analytic reference model initially. To overcome the interference of each sub-system and simplify the controller design, the proposed model reference decentralized adaptive control scheme constructs a decoupled well-designed reference model first. Then, according to the well-designed model, this paper develops a digital decentralized adaptive tracker based on the optimal analog control and prediction-based digital redesign technique for the sampled-data large-scale coupling system. In order to enhance the tracking performance of the digital tracker at specified sampling instants, we apply the iterative learning control (ILC) to train the control input via continual learning. As a result, the proposed iterative learning-based decentralized adaptive tracker not only has robust closed-loop decoupled property but also possesses good tracking performance at both transient and steady state. Besides, evolutionary programming is applied to search for a good learning gain to speed up the learning process of ILC. Copyright © 2011 ISA. Published by Elsevier Ltd. All rights reserved.

  10. A design study of hydrogen isotope separation system for ITER-FEAT

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka

    2001-03-01

    Preliminary design study of the hydrogen isotope separation system (ISS) for the fuel cycle of the ITER-FEAT, a fusion experimental reactor, was carried out based on the substantial reduction of hydrogen flow to the ISS resulting from the design study for scale reduction of the formerly-designed ITER. Three feed streams (plasma exhaust gas stream, streams from the water detritiation system and that from the neutral beam injectors) are fed to the ISS, and three product streams (high purity tritium gas, high purity deuterium gas and hydrogen gas) are made in it by the method of cryogenic distillation. In this study, an original four-column cascade was proposed to the ISS cryogenic distillation column system considering simplification and the operation scenario of the ITER-FEAT. Substantial reduction of tritium inventory in the ISS was found to be possible in the progress of investigation concerning of the corresponding flow rate of tritium product stream (T>90 %) for pellet injector which depends upon the operation condition. And it was found that tritium concentration in the released hydrogen stream into environment from the ISS could easily fluctuate with current design of column arrangement due to the small disturbance in mass flow balance in the ISS. To solve this problem, two-column system for treatment of this flow was proposed. (author)

  11. Towards the conceptual design of the ITER real-time plasma control system

    International Nuclear Information System (INIS)

    Winter, A.; Makijarvi, P.; Simrock, S.; Snipes, J.A.; Wallander, A.; Zabeo, L.

    2014-01-01

    Highlights: • We describe the main control areas and interfaces for the ITER real-time plasma control system and the current state of their design. • An overview is given for the implementation strategy for the plasma control system as part of the ITER control, data access and communication system. • Current efforts on the creation of simulation and development tools are presented. - Abstract: ITER will be the world's largest magnetic confinement tokamak fusion device and is currently under construction in southern France. The ITER Plasma Control System (PCS) is a fundamental component of the ITER Control, Data Access and Communication system (CODAC). It will control the evolution of all plasma parameters that are necessary to operate ITER throughout all phases of the discharge. The design and implementation of the PCS poses a number of unique challenges. The timescales of phenomena to be controlled spans three orders of magnitude, ranging from a few milliseconds to seconds. Novel control schemes, which have not been implemented at present-day machines need to be developed, and control schemes that are only done as demonstration experiments today will have to become routine. In addition, advances in computing technology and available physics models make the implementation of real-time or faster-than-real-time predictive calculations to forecast and subsequently to avoid disruptions or undesired plasma regimes feasible. This requires the PCS design to be adaptable in real-time to the results of these forecasting algorithms. A further novel feature is a sophisticated event handling system, which provides a means to deal with plasma related events (such as MHD instabilities or L-H transitions) or component failure. Finally, the schedule for design and implementation poses another challenge. The beginning of ITER operation will be in late 2020, but the conceptual design activity of the PCS has already commenced as required by the on-going development of

  12. A second-order iterative implicit-explicit hybrid scheme for hyperbolic systems of conservation laws

    International Nuclear Information System (INIS)

    Dai, Wenlong; Woodward, P.R.

    1996-01-01

    An iterative implicit-explicit hybrid scheme is proposed for hyperbolic systems of conservation laws. Each wave in a system may be implicitly, or explicitly, or partially implicitly and partially explicitly treated depending on its associated Courant number in each numerical cell, and the scheme is able to smoothly switch between implicit and explicit calculations. The scheme is of Godunov-type in both explicit and implicit regimes, is in a strict conservation form, and is accurate to second-order in both space and time for all Courant numbers. The computer code for the scheme is easy to vectorize. Multicolors proposed in this paper may reduce the number of iterations required to reach a converged solution by several orders for a large time step. The feature of the scheme is shown through numerical examples. 38 refs., 12 figs

  13. Summarized results of the cryosorption panel test programme for the ITER cryopumping system

    International Nuclear Information System (INIS)

    Day, C.; Haas, H.; Mack, A.; Kazakovsky, N.T.; Murdoch, D.K.; Roehrig, D.; Saksagansky, G.L.

    2001-01-01

    A reliable but versatile primary cryopumping system is required for high vacuum pumping of the ITER torus during all phases of plasma operation. To achieve that goal, an extensive R and D programme has been performed within the framework of the Nuclear Fusion Project of FZK, supported by the European Communities under the European Fusion Technology Programme. The present paper covers that part of the programme, which focuses on the pumping speed of the recommended cryopanel type and the various aspects of the charcoal-bonding system in the cryogenic temperature range. It is demonstrated that the investigated cryosorption panels exhibit a very good behaviour with respect to pumping efficiency, long-term thermomechanical endurance and compatibility with tritium. The recommended cryopump design was therefore chosen as point design for ITER. (author)

  14. Robust Monotonically Convergent Iterative Learning Control for Discrete-Time Systems via Generalized KYP Lemma

    Directory of Open Access Journals (Sweden)

    Jian Ding

    2014-01-01

    Full Text Available This paper addresses the problem of P-type iterative learning control for a class of multiple-input multiple-output linear discrete-time systems, whose aim is to develop robust monotonically convergent control law design over a finite frequency range. It is shown that the 2 D iterative learning control processes can be taken as 1 D state space model regardless of relative degree. With the generalized Kalman-Yakubovich-Popov lemma applied, it is feasible to describe the monotonically convergent conditions with the help of linear matrix inequality technique and to develop formulas for the control gain matrices design. An extension to robust control law design against systems with structured and polytopic-type uncertainties is also considered. Two numerical examples are provided to validate the feasibility and effectiveness of the proposed method.

  15. Progress in the design, R and D and procurement preparation of the ITER Divertor Remote Handling System

    Energy Technology Data Exchange (ETDEWEB)

    Esqué, Salvador, E-mail: Salvador.Esque@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Hille, Carine van; Ranz, Roberto; Damiani, Carlo [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Palmer, Jim; Hamilton, David [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France)

    2014-10-15

    Highlights: •The ITER Divertor Remote Handling System (DRHS) reference design is presented. •Different R and D activities that have contributed to the development and validation of the current reference design are reported. •The DRHS turns to be a unique system in terms of complexity due to size of the to-be-handled components, the novelty of the remote operations and the operational conditions. -- Abstract: The ITER Divertor Remote Handling System (DRHS) consists of a number of dedicated remote handling equipment and tooling that will provide the means to perform the exchange of the divertor system in a full-remote way. In order to achieve this objective the DRHS will need to perform a number of novel and complex remote operations in a contaminated and space-constrained environment, in rather poor lightening conditions. Fusion for Energy has recently launched the tendering phase for the in-kind procurement of the DRHS. The procurement is based on a set of system requirements and functional specifications supported by a reference design which are presented and discussed in this paper along with the main outcomes of the different R and D activities that have contributed to the development and validation of the current reference design.

  16. 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

  17. Proposal of a torus pumping and fuel recycling system for ITER

    International Nuclear Information System (INIS)

    Perinic, D.; Mack, A.; Perinic, G.; Murdoch, D.

    1995-01-01

    A universal torus pumping and fuel recycling system is proposed for all operation modes of ITER. It comprises primary cryopumps and secondary fuel separating cryopumps located inside the cryostat and a common mechanical forepump station located outside the cryostat. In this paper two different primary cryopump options are compared. The results of Monte Carlo calculations of pumping probabilities for helium show a significant difference leading to a distinct preference for the concept of a co-pumping cryopump. (orig.)

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

    International Nuclear Information System (INIS)

    Huget, M.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Huguet, M.

    2001-01-01

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

  20. Lithium test module on ITER: engineering design of the tritium recovery system

    International Nuclear Information System (INIS)

    Finn, P.A.

    1988-01-01

    The design presented is an overview of the tritium recovery system for a lithium module on an ITER type reactor. The design of a tritium recovery system for larger blanket units, sectors, etc. could use the information developed in this report. A goal of this design was to ensure that a reliable, integrated performance of the tritium recovery system could be demonstrated. An equally important goal was to measure and account for the tritium in the liquid lithium blanket module and its recovery system in order to validate the operation of the blanket module

  1. Flexible path optimization for the Cask and Plug Remote Handling System in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Fonte, Daniel; Valente, Filipe; Ferreira, João [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Gonzalez, Carmen [Fusion for Energy Agency (F4E), Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: ► Complementary approach for path optimization named free roaming that takes full advantage of the rhombic like kinematics of the Cask and Plug Remote Handling System (CPRHS). ► Possibility to find trajectories not possible in the past using the line guidance developed in a previous work, in particular when moving the Cask Transfer System (CTS) beneath the pallet or in rescue missions. ► Methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. -- Abstract: The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed.

  2. Conceptual design of volume reduction system for ITER low level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Chon, Je Keun, E-mail: jekeun.chon@iter.org; Beaudoin, Virginie; Pitcher, Charles S.

    2016-11-01

    For ITER Type A (low-level radwaste) radioactive waste process system, the super-compaction of solid wastes is introduced to reduce the waste volume in order to increase the quantity of waste to be enclosed in a final package. A study has been conducted to develop the conceptual design of super-compaction system by exploring the optimum design features of available equipment. The principal finding of the study is that a compactor with a vertical extrusion die that can apply a compressive force of up to 7000 kN will be fit for purpose of ITER Type A waste treatment. The confinement box has been proposed to limit the spread of aerial effluents that might escape from the compacted pellets. In order to increase the packing efficiency of the disposal containers, the height of each pellet will be recorded, enabling the disposal containers to be filled as close as possible to their limit if pellets of an appropriate size are available. The proposed conceptual design of super-compaction system is capable of meeting the objectives and constraints of target waste streams from ITER operation.

  3. 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

  4. A computational procedure for the investigation of whipping effect on ITER High Energy Piping and its application to the ITER divertor primary heat transfer system

    Energy Technology Data Exchange (ETDEWEB)

    Spagnuolo, G.A., E-mail: Alessandro.Spagnuolo@kraftanlagen.com [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany); Dell’Orco, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Mazzei, M. [Kraftanlagen Heidelberg Gmbh, Im Breitspiel 7, D-69126 Heidelberg (Germany)

    2015-10-15

    Highlights: • High Energy Piping (HEP) are components containing water or steam with P ≥ 2.0 MPa and/or T ≥ 100 °C. • The whipping effect in HEP may cause dangerous domino effect with relative rupture propagation. • The rapture is envisaged or postulated according to the stress state of piping. • A FEM analysis is performed in order to study the dynamic of whipping effect. • Study of special support to avoid and/or mitigate the whipping effect. - Abstract: The Tokamak Cooling Water System of nuclear facility has the function to remove heat from plasma facing components maintaining coolant temperatures, pressures and flow rates as required and, depending on thermal-hydraulic requirements, its systems are defined as High Energy Piping (HEP) because they contain fluids, such as water or steam, at a pressure greater than or equal to 2.0 MPa and/or at a temperature greater than or equal to 100 °C, or even gas at pressure above the atmospheric one. The French standards contemplate the need to consider the whipping effect on HEP design. This effect happens when, after a double ended guillotine break, the reaction force could create a displacement of the piping which might affect adjacent components. A research campaign has been performed, in cooperation by ITER Organization and University of Palermo, to outline the procedure to check whether whipping effect might occur and assess its potential damage effects so to allow their mitigation. This procedure is based on the guidelines issued by U.S. Nuclear Regulatory Commission. The proposed procedure has been applied to the analysis of the whipping effect of divertor primary heat transfer system HEP, using a theoretical–computational approach based on the finite element method.

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  8. Conceptual design on interface between ITER and tritium extraction system of Chinese helium-cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Zhang Long; Luo Tianyong; Feng Kaiming

    2010-01-01

    Tritium extraction system is essential for CN HCSB TBM for safety and technical reasons. Based on the assessments of system functions, integration issues and safety considerations, two main modifications of the system from previous design (Feng et al., 2007 ; Chen et al., 2008 ) are adopted: a)the TES has been split to 2 parts with one in port cell and another in tritium building. Q 2 O in the purge gas is reduced to Q 2 in a hot metal bed located in port cell; Q 2 is separated from the stream by a pair of cryogenic molecular sieve beds and a Pd/Ag diffuser located in tritium building. b)isotope separation process has been excluded. TES components sizes are estimated and space allocations are estimated. Required services and where and when they are needed are preliminary defined. Fluids delivered towards ITER tritium system are analyzed.

  9. ITER SAFETY TASK NID-5D: Operational tritium loss and accident investigation for heat transport and water detritiation systems

    International Nuclear Information System (INIS)

    Kalyanam, K.M.; Fong, C.; Moledina, M.; Natalizio, A.

    1995-02-01

    The task objectives are to: a) determine major pathways for tritium loss during normal operation of the cooling systems and water detritiation system, b) estimate operational losses and environmental tritium releases from the heat transport and water detritiation systems of ITER, and c) prepare a preliminary Failure Modes and Effects Analysis (FMEA) for the ITER Water Detritiation System. The analysis will be used to estimate chronic environmental tritium releases (airborne and waterborne) for the ITER Cooling Systems and Water Detritiation System. The assessment will form the basis for demonstrating the acceptability of ITER for siting in the Early Safety and Environmental Characterization Study (ESECS), to be issued in early 1995. (author). 7 refs., 10 tabs., 11 figs

  10. Overpower transient in the first wall cooling system of NET/ITER

    International Nuclear Information System (INIS)

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

    1993-09-01

    The overpower transient from a plasma power excursion. The overpower transient considered in this report results from a postulated linear increase of the plasma power from the nominal generated power to four times this nominal power in 30 s. The Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design will be cooled by a number of separate cooling systems. The most important cooling systems are: The first wall cooling system, the blanket cooling system, the divertor cooling system, and the shield cooling system. In this report, the thermal-hydraulic analysis of the above-mentioned overpower transient will be presented for the first wall cooling system of NET/ITER. During overpower transients, the fusion power will increase to less than four times the nominal power. For this reason, the overpower transient considered in this report is the worst case scenario. The analysis of the thermal-hydraulic system behaviour during the considered overpower transient has been performed for a coolant temperature of 333 K (60 C) in the first wall inlet manifolds and 433 K (160 C) in the first wall outlet manifolds. The analysis has been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analysis, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. (orig.)

  11. Alignment Condition-Based Robust Adaptive Iterative Learning Control of Uncertain Robot System

    Directory of Open Access Journals (Sweden)

    Guofeng Tong

    2014-04-01

    Full Text Available This paper proposes an adaptive iterative learning control strategy integrated with saturation-based robust control for uncertain robot system in presence of modelling uncertainties, unknown parameter, and external disturbance under alignment condition. An important merit is that it achieves adaptive switching of gain matrix both in conventional PD-type feedforward control and robust adaptive control in the iteration domain simultaneously. The analysis of convergence of proposed control law is based on Lyapunov's direct method under alignment initial condition. Simulation results demonstrate the faster learning rate and better robust performance with proposed algorithm by comparing with other existing robust controllers. The actual experiment on three-DOF robot manipulator shows its better practical effectiveness.

  12. An Iterative Multiuser Detector for Turbo-Coded DS-CDMA Systems

    Directory of Open Access Journals (Sweden)

    Takawira Fambirai

    2005-01-01

    Full Text Available We propose an iterative multiuser detector for turbo-coded synchronous and asynchronous direct-sequence CDMA (DS-CDMA systems. The receiver is derived from the maximum a posteriori (MAP estimation of the single user's transmitted data, conditioned on information about the estimate of the multiple-access interference (MAI and the received signal from the channel. This multiple-access interference is reconstructed by making hard decisions on the users' detected bits at the preceding iteration. The complexity of the proposed receiver increases linearly with the number of users. The proposed detection scheme is compared with a previously developed one. The multiuser detector proposed in this paper has a better performance when the transmitted powers of all active users are equal in the additive white Gaussian noise (AWGN channel. Also, the detector is found to be resilient against the near-far effect.

  13. Toward a design for the ITER plasma shape and stability control system

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Leuer, J.A.; Kellman, A.G.; Haney, S.W.; Bulmer, R.H.; Pearlstein, L.D.; Portone, A.

    1994-07-01

    A design strategy for an integrated shaping and stability control algorithm for ITER is described. This strategy exploits the natural multivariable nature of the system so that all poloidal field coils are used to simultaneously control all regulated plasma shape and position parameters. A nonrigid, flux-conserving linearized plasma response model is derived using a variational procedure analogous to the ideal MHD Extended Energy Principle. Initial results are presented for the non-rigid plasma response model approach applied to an example DIII-D equilibrium. For this example, the nonrigid model is found to yield a higher passive growth rate than a rigid current-conserving plasma response model. Multivariable robust controller design methods are discussed and shown to be appropriate for the ITER shape control problem

  14. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    Energy Technology Data Exchange (ETDEWEB)

    Pérez, Germán, E-mail: German.Perez@iter.org; Mitteau, Raphaël; Furmanek, Andreas; Martin, Alex; Raffray, René; Merola, Mario; Sabourin, Flavien

    2014-10-15

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented.

  15. Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

    International Nuclear Information System (INIS)

    Pérez, Germán; Mitteau, Raphaël; Furmanek, Andreas; Martin, Alex; Raffray, René; Merola, Mario; Sabourin, Flavien

    2014-01-01

    Highlights: • Optimized water distribution in ITER blanket modules is presented. • All key challenging constraints are included. • The methodology and the successful result are presented. - Abstract: This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements. The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented

  16. Draining and drying process development of the Tokamak Cooling Water System of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seokho, E-mail: kims@ornl.gov [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Van Hove, Walter; Ferrada, Juan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Di Maio, Pietro Alessandro [University of Palermo, Viale delle Scienze, Palermo 90128 (Italy); Felde, David [Reactor and Nuclear Systems Division, ORNL, Oak Ridge, TN (United States); Raphael, Mitteau; Dell’Orco, Giovanni [ITER Organization, 13067 St Paul Lez Durance (France); Berry, Jan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2016-11-01

    Highlights: • A thermal-hydraulic model using RELAP was developed for the ITER FW/BLK modules to determine design parameters for the nitrogen blowout flow rate and pressure. • The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will sufficiently evacuate the water in blankets. • A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation. - Abstract: The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation.

  17. Primary design and operation analysis of the ITER air transfer system

    International Nuclear Information System (INIS)

    Wang Haitian; Li Ge; Qin Shijun

    2010-01-01

    Air transfer system (ATS) is a remote handling transfer, which can work in the nuclear radiation environment and can be driven by the electricity fully. Its motion power is provided by several servo motors. The remote control technology of ATS, which is China taking part in the plan of international Tokamak experimental reactor (ITER) and grasping this technology, is one of key technologies of ITER. The remote handling technology can lay the foundation for developing demonstration nuclear fusion power plant in China on self-reliance. Because there is gamma irradiation and hazard material in these ITER parts, all required maintenance of port plugs and inner components are been transmitted by ATS. The pick-up or drop-off these components are completed by means of a remotely controlled TCS system between the Vacuum Vessel and the Hot Cell through the bridge-gallery. Tokamak building includes three floors, including upper port, equatorial port and lower port, linked by a lift. According to each port level configuration and safety requirement, the radius of curvature with ATS trajectory is optimized, and a trajectory of each level is determined by positioned guidance beacons. At last, the results of computer aided design (CAD) show single trajectory guidance of ATS in each level is available. (authors)

  18. Progress on the interface between UPP and CPRHS (Cask and Plug Remote Handling System) tractor/gripping tool for ITER

    International Nuclear Information System (INIS)

    Rosa, Elena V.; Rios, Luis; Queral, Vicente

    2013-01-01

    Highlights: ► UPP interface requirements in the plug RH extraction/insertion for ITER. ► Analyze of maximum misalignment between port duct and port cell. ► Friction study between plug skids and VV port/ramp rails during the plug transfer. ► Definition of the tolerance in the plug skids to avoid the plug jamming. ► Concepts of gripping tools based on one gripping point and avoiding force feedback. -- Abstract: EFDA finances a training programme called Goal Oriented Training Programme for Remote Handling (GOT RH), whose goal is to train engineers in Remote Handling for ITER. As part of this training programme, the conceptual design of the mechanical interface between Upper Port Plug (UPP) and Cask and Plug Remote Handling System (CPRHS) as well as the conceptual design of the needed tools for UPP Remote Handling is carried out. The paper presents the conceptual design of the UPP/Gripping Tool Interface. This includes the conceptual design of the gripping tool for introducing/removing the UPP in/from the ITER port and the mechanical features on both sides of the UPP/Gripping Tool Interface (e.g. alignment features, mechanical connectors, fasteners). In order to develop the design of the interface between UPP and CPRHS it is necessary to first identify the functional requirements of the Transfer Cask System (TCS) and the CPRHS, such as required degrees of freedom (DoF), required performances of system, geometrical constraints, loading conditions, alignment requirements, RAMI requirements. These requirements are the input data for the design of the interface between UPP and gripping tool and some of them are also described in the paper

  19. Concept development for the ITER equatorial port visible/infrared wide angle viewing system

    International Nuclear Information System (INIS)

    Reichle, R.; Beaumont, B.; Boilson, D.; Bouhamou, R.; Direz, M.-F.; Encheva, A.; Henderson, M.; Kazarian, F.; Lamalle, Ph.; Lisgo, S.; Mitteau, R.; Patel, K. M.; Pitcher, C. S.; Pitts, R. A.; Prakash, A.; Raffray, R.; Schunke, B.; Snipes, J.; Diaz, A. Suarez; Udintsev, V. S.

    2012-01-01

    The ITER equatorial port visible/infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R and D topics are outlined.

  20. Plasma position and current control system enhancements for the JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    De Tommasi, G. [Associazione EURATOM-ENEA-CREATE, Univ. di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Via Claudio 21, 80125 Napoli (Italy); Neto, A.C. [Ass. EURATOM-IST, Instituto de Plasmas e Fusão Nuclear, IST, 1049-001 Lisboa (Portugal); Lomas, P.J.; McCullen, P.; Rimini, F.G. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom)

    2014-03-15

    Highlights: • JET plasma position and current control system enhanced for the JET ITER like wall. • Vertical stabilization system enhanced to speed up its response and to withstand larger perturbations. • Improved termination management system. • Implementation of the current limit avoidance system. • Implementation of PFX-on-early-task. - Abstract: The upgrade of Joint European Torus (JET) to a new all-metal wall, the so-called ITER-like wall (ILW), has posed a set of new challenges regarding both machine operation and protection. The plasma position and current control (PPCC) system plays a crucial role in minimizing the possibility that the plasma could permanently damage the ILW. The installation of the ILW has driven a number of upgrades of the two PPCC components, namely the Vertical Stabilization (VS) system and the Shape Controller (SC). The VS system has been enhanced in order to speed up its response and to withstand larger perturbations. The SC upgrade includes three new features: an improved termination management system, the current limit avoidance system, and the PFX-on-early-task. This paper describes the PPCC upgrades listed above, focusing on the implementation issues and on the experimental results achieved during the 2011–12 JET experimental campaigns.

  1. Localization of cask and plug remote handling system in ITER using multiple video cameras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Vale, Alberto [Instituto de Plasmas e Fusão Nuclear - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ribeiro, Isabel [Laboratório de Robótica e Sistemas em Engenharia e Ciência - Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building.

  2. Localization of cask and plug remote handling system in ITER using multiple video cameras

    International Nuclear Information System (INIS)

    Ferreira, João; Vale, Alberto; Ribeiro, Isabel

    2013-01-01

    Highlights: ► Localization of cask and plug remote handling system with video cameras and markers. ► Video cameras already installed on the building for remote operators. ► Fiducial markers glued or painted on cask and plug remote handling system. ► Augmented reality contents on the video streaming as an aid for remote operators. ► Integration with other localization systems for enhanced robustness and precision. -- Abstract: The cask and plug remote handling system (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell building and the Tokamak building in ITER. Different CPRHS typologies will be autonomously guided following predefined trajectories. Therefore, the localization of any CPRHS in operation must be continuously known in real time to provide the feedback for the control system and also for the human supervision. This paper proposes a localization system that uses the video streaming captured by the multiple cameras already installed in the ITER scenario to estimate with precision the position and the orientation of any CPRHS. In addition, an augmented reality system can be implemented using the same video streaming and the libraries for the localization system. The proposed localization system was tested in a mock-up scenario with a scale 1:25 of the divertor level of Tokamak building

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  4. Progress in the integration of Test Blanket Systems in ITER equatorial port cells and in the interfaces definition

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Beloglazov, S.; Bonagiri, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Commin, L. [CEA, IRFM, Cadarache (France); Cortes, P.; Giancarli, L.M.; Gliss, C.; Iseli, M.; Lanza, R.; Levesy, B.; Martins, J.-P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Neviere, J.-C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L.; Plutino, D.; Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Swami, H.L. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The design integration of two test blanket systems in ITER port cell is addressed. Black-Right-Pointing-Pointer Definition of interfaces of TBSs with building and other ITER systems is done. Black-Right-Pointing-Pointer Designs of pipe forest, bioshield plug and ancillary equipment unit are described. Black-Right-Pointing-Pointer The maintenance of the two test blanket systems in ITER port cell is considered. Black-Right-Pointing-Pointer The management of the heat and tritium releases in the TBM port cell is described. - Abstract: In the framework of the TBM Program, three ITER vacuum vessel equatorial ports (no. 16, no. 18 and no. 02) have been allocated for the testing of up to six mock-ups of six different DEMO tritium breeding blankets. Each one is called a Test Blanket System (TBS). A TBS consists mainly of the Test Blanket Module (TBM), the in-vessel component facing the plasma, and several ancillary systems, in particular the cooling system and the tritium extraction system. Each port accommodates two TBMs and therefore the two TBSs have to share the corresponding port cell. This paper deals with the design integration aspects of the two TBSs in each port cell performed at ITER Organization (IO) with the corresponding definition of interfaces with other ITER systems. The performed activities have raised several issues that are discussed in the paper and for which design solutions are proposed.

  5. Final design of the Korean AC/DC converters for the ITER coil power supply system

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jong-Seok, E-mail: jsoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Choi, Jungwan; Suh, Jae-Hak; Choi, Jihyun [ITER Korea, National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Lee, Lacksang; Kim, Changwoo; Park, Hyungjin; Jo, Seongman; Lee, Seungyun; Hwang, Kwangcheol; Liu, Hyoyol [Dawonsys Corp., Siheung 429-450 (Korea, Republic of); Hong, Ki-Don; Sim, Dong-Joon; Lee, Jang-Soo [Hyosung Corp., Gongdeok-Dong, Seoul 121-720 (Korea, Republic of); Lee, Eui-Jae; Kwon, Yang-Hae; Lee, Dae-Yeol; Ko, Ki-Won; Kim, Jong-Min [Mobiis Corp., Yangjae-dong, Seoul 137-888 (Korea, Republic of); Song, Inho [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); and others

    2015-10-15

    The final design of the ITER TF, CS, CC and VS AC/DC converters has been completed to implement ITER requirements following the detailed design and refinements of the preliminary design. The number of parallel thyristors and the rating of fuses are coordinated to keep those devices within the explosion limit even under most severe fault conditions. The impedance of the converter transformer has been optimized taking into account the energization inrush current, short circuit current, reactive power consumption and the available DC voltage. To ensure system integrity, AC/DC converters are mechanically divided into transformers, AC busbars, 6-pulse bridges, DC interconnecting busbars and DC reactors, and then all subsystems are decoupled by flexible links. To provide stable real time network communication down to the converters, a one GbE link is deployed between master controllers and local controllers. IEEE 1588 is implemented to the embedded controllers for precision time synchronization. This paper describes the detailed solutions implemented in the final design for the ITER AC/DC converters with R&D results of converter prototypes.

  6. A Parametric Learning and Identification Based Robust Iterative Learning Control for Time Varying Delay Systems

    Directory of Open Access Journals (Sweden)

    Lun Zhai

    2014-01-01

    Full Text Available A parametric learning based robust iterative learning control (ILC scheme is applied to the time varying delay multiple-input and multiple-output (MIMO linear systems. The convergence conditions are derived by using the H∞ and linear matrix inequality (LMI approaches, and the convergence speed is analyzed as well. A practical identification strategy is applied to optimize the learning laws and to improve the robustness and performance of the control system. Numerical simulations are illustrated to validate the above concepts.

  7. Contact dose rates and relevant radioactive inventory in ITER TBM systems

    International Nuclear Information System (INIS)

    Zucchetti, M.; Guerrini, L.; Poitevin, Y.; Ricapito, I.; Zmitko, M.

    2011-01-01

    The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated.

  8. Contact dose rates and relevant radioactive inventory in ITER TBM systems

    Energy Technology Data Exchange (ETDEWEB)

    Zucchetti, M., E-mail: massimo.zucchetti@polito.it [EURATOM/ENEA Fusion Association Politecnico di Torino, Torino (Italy); Guerrini, L., E-mail: Laurent.Guerrini@f4e.europa.eu [Fusion for Energy, ITER Department, Test Blanket Modules Group, Barcelona (Spain); Poitevin, Y.; Ricapito, I.; Zmitko, M. [Fusion for Energy, ITER Department, Test Blanket Modules Group, Barcelona (Spain)

    2011-10-15

    The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated.

  9. Linear quadratic Gaussian controller design for plasma current, position and shape control system in ITER

    International Nuclear Information System (INIS)

    Belyakov, V.; Kavin, A.; Rumyantsev, E.; Kharitonov, V.; Misenov, B.; Ovsyannikov, A.; Ovsyannikov, D.; Veremei, E.; Zhabko, A.; Mitrishkin, Y.

    1999-01-01

    This paper is focused on the linear quadratic Gaussian (LQG) controller synthesis methodology for the ITER plasma current, position and shape control system as well as power derivative management system. It has been shown that some poloidal field (PF) coils have less influence on reference plasma-wall gaps control during plasma disturbances and hence they have been used to reduce total control power derivative by means of the additional non-linear feedback. The design has been done on the basis of linear models. Simulation was provided for non-linear model and results are presented and discussed. (orig.)

  10. Overview of the ITER Tokamak complex building and integration of plant systems toward construction

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2015-10-15

    The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

  11. Overview of the ITER EC H and CD system and its capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Omori, T., E-mail: toshimichi.omori@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Henderson, M.A. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Albajar, F. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Alberti, S. [CRPP-Association EURATOM-Confederation Suisse, EPFL Ecublens, CH-1015 Lausanne (Switzerland); Baruah, U. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Bigelow, T.S. [US ITER Project Office, ORNL, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831 (United States); Beckett, B. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Bertizzolo, R. [CRPP-Association EURATOM-Confederation Suisse, EPFL Ecublens, CH-1015 Lausanne (Switzerland); Bonicelli, T. [Fusion for Energy, C/Josep Pla 2, Torres Diagonal Litoral-B3, E-08019 Barcelona (Spain); Bruschi, A. [Istituto di Fisica del Plasma, Association EURATOM-ENEA-CNR, Milano (Italy); Caughman, J.B. [US ITER Project Office, ORNL, 055 Commerce Park, PO Box 2008, Oak Ridge, TN 37831 (United States); Chavan, R. [CRPP-Association EURATOM-Confederation Suisse, EPFL Ecublens, CH-1015 Lausanne (Switzerland); Cirant, S. [Istituto di Fisica del Plasma, Association EURATOM-ENEA-CNR, Milano (Italy); Collazos, A. [CRPP-Association EURATOM-Confederation Suisse, EPFL Ecublens, CH-1015 Lausanne (Switzerland); Cox, D.; Darbos, C. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baar, M.R. de [Association EURATOM-FOM, 3430 BE Nieuwegein (Netherlands); Denisov, G. [Institute of Applied Physics, 46 Ulyanov Street, Nizhny Novgorod 603950 (Russian Federation); Farina, D. [Istituto di Fisica del Plasma, Association EURATOM-ENEA-CNR, Milano (Italy); Gandini, F. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject {>=}20 MW RF power into the plasma for Heating and Current Drive (H and CD) applications. The EC system consists of up to 26 gyrotrons (between 1 and 2 MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MHD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20 MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities. The overall EC system has been modified during the past 3 years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past 2 years and how the design is compliant with the principal objectives.

  12. Overview of the ITER EC H and CD system and its capabilities

    International Nuclear Information System (INIS)

    Omori, T.; Henderson, M.A.; Albajar, F.; Alberti, S.; Baruah, U.; Bigelow, T.S.; Beckett, B.; Bertizzolo, R.; Bonicelli, T.; Bruschi, A.; Caughman, J.B.; Chavan, R.; Cirant, S.; Collazos, A.; Cox, D.; Darbos, C.; Baar, M.R. de; Denisov, G.; Farina, D.; Gandini, F.

    2011-01-01

    The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject ≥20 MW RF power into the plasma for Heating and Current Drive (H and CD) applications. The EC system consists of up to 26 gyrotrons (between 1 and 2 MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MHD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20 MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities. The overall EC system has been modified during the past 3 years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past 2 years and how the design is compliant with the principal objectives.

  13. Upgrade of DC power supply system in ITER CS model coil test facility

    International Nuclear Information System (INIS)

    Shimono, Mitsugu; Uno, Yasuhiro; Yamazaki, Keita; Kawano, Katsumi; Isono, Takaaki

    2014-03-01

    Objective of the ITER CS Model Coil Test Facility is to evaluate a large scale superconducting conductor for fusion using the Central Solenoid (CS) Model Coil, which can generate a 13T magnetic field in the inner bore with a 1.5 m diameter. The facility is composed of a helium refrigerator / liquefier system, a DC power supply system, a vacuum system and a data acquisition system. The DC power supply system supplies currents to two superconducting coils, the CS Model Coil and an insert coil. A 50-kA DC power supply is installed for the CS Model Coil and two 30 kA DC power supplies are installed for an insert coil. In order to evaluate superconducting performance of a conductor used for ITER Toroidal Field (TF) coils whose operating current is 68 kA, the line for an insert coil is upgraded. A 10 kA DC power supply was added, DC circuit breakers were upgraded, bus bars and current measuring instrument were replaced. In accordance to the upgrade, operation manual was revised. (author)

  14. Numerical simulation of draining and drying procedure for the ITER Generic Equatorial Port Plug cooling system

    International Nuclear Information System (INIS)

    Tanchuk, Victor; Grigoriev, Sergey; Lyublin, Boris; Maquet, Philippe; Senik, Konstantin; Pak, Sunil; Udintsev, Victor

    2016-01-01

    Highlights: • The cooling system of the ITER Generic Equatorial Port Plug (GEPP) is of a complicated combination of horizontal and vertical channels. • The calculation model for the entire GEPP cooling circuit comprising 12 sub-circuits and built up of 2421 finite-volume elements has been developed. • Transient analysis of this model simulating the draining procedure by the KORSAR/B1 code has been performed. • Water in amount of 263 g of initial 531 kg in the GEPP remains in the dead-ends of the DSM and DFW channels in 150 s of draining procedure. • Almost 3 h are required to boil off 263 g of water trapped in the dead-ends. - Abstract: For effective vacuum leak testing all cooling circuits serving the ITER vessel and in-vessel components shall be drained and dried so that after this procedure taking less than 100 h the purge gas passing through a component has water content less than 100 ppm. This process is four-stage, with the first stage using a short blast of compressed nitrogen to blow most of water in the coolant channels out of the circuit. This process is hindered by volumes which trap water due to gravity. To remove the trapped water, it is necessary, first, to heat up the structure by hot and compressed nitrogen, and then water is evaporated by depressurized nitrogen. The cooling system of the ITER Diagnostic Equatorial Port Plugs is of a complicated hydraulic configuration. The system branching might make difficult removal of water from the piping in the scheduled draining mode. The authors have proposed the KORSAR computation code to simulate draining of the GEPP cooling circuit. The numerical simulation performed has made it possible to describe the process dynamics during draining of the entire GEPP cooling circuit and to define the process time, amount and location of residual water and evolution of two-phase flow regime.

  15. Numerical simulation of draining and drying procedure for the ITER Generic Equatorial Port Plug cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Grigoriev, Sergey; Lyublin, Boris [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Maquet, Philippe [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Senik, Konstantin [JSC “D.V. Efremov Institute of Electrophysical Apparatus”, 196641 St. Petersburg (Russian Federation); Pak, Sunil [National Fusion Research Institute, Daejeon (Korea, Republic of); Udintsev, Victor [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2016-11-01

    Highlights: • The cooling system of the ITER Generic Equatorial Port Plug (GEPP) is of a complicated combination of horizontal and vertical channels. • The calculation model for the entire GEPP cooling circuit comprising 12 sub-circuits and built up of 2421 finite-volume elements has been developed. • Transient analysis of this model simulating the draining procedure by the KORSAR/B1 code has been performed. • Water in amount of 263 g of initial 531 kg in the GEPP remains in the dead-ends of the DSM and DFW channels in 150 s of draining procedure. • Almost 3 h are required to boil off 263 g of water trapped in the dead-ends. - Abstract: For effective vacuum leak testing all cooling circuits serving the ITER vessel and in-vessel components shall be drained and dried so that after this procedure taking less than 100 h the purge gas passing through a component has water content less than 100 ppm. This process is four-stage, with the first stage using a short blast of compressed nitrogen to blow most of water in the coolant channels out of the circuit. This process is hindered by volumes which trap water due to gravity. To remove the trapped water, it is necessary, first, to heat up the structure by hot and compressed nitrogen, and then water is evaporated by depressurized nitrogen. The cooling system of the ITER Diagnostic Equatorial Port Plugs is of a complicated hydraulic configuration. The system branching might make difficult removal of water from the piping in the scheduled draining mode. The authors have proposed the KORSAR computation code to simulate draining of the GEPP cooling circuit. The numerical simulation performed has made it possible to describe the process dynamics during draining of the entire GEPP cooling circuit and to define the process time, amount and location of residual water and evolution of two-phase flow regime.

  16. Failure Mode and Effect Analysis for remote handling transfer systems of ITER

    International Nuclear Information System (INIS)

    Pinna, T.; Caporali, R.; Tesini, A.

    2008-01-01

    A Failure Mode and Effect Analysis (FMEA) at component level was done to study safety-relevant implications arising from possible failures in performing remote handling (RH) operations at ITER facility . Autonomous air cushion transporter, pallet, sealed casks and tractor movers needed for port plug mounting/dismantling operation were analysed. For each sub-system, the breakdown of significant components was outlined and, for each component, possible failure modes have been investigated pointing out possible causes, possible actions to prevent the causes, consequences and actions to prevent or mitigate consequences. Off-normal events which may result in hazardous consequences to the public and the environment have been defined as Postulated Initiating Events (PIEs). Two safety-relevant PIEs have been defined by assessing elementary failures related to the analysed system. Each PIE has been discussed in order to qualitatively identify accident sequences arising from each of them. As an output of this FMEA study, possible incidental scenarios, where the intervention of rescue RH equipments is required to overcome critical situations determined by fault of RH components, were defined as well. Being rescue scenarios of main concern for ITER remote handling activities, such families could be helpful in defining the design requirements of port handling systems in general and on RH transfer system in particular. Furthermore, they could be useful in defining casks and vehicles to be used for rescue activities

  17. The steady performance prediction of propeller-rudder-bulb system based on potential iterative method

    International Nuclear Information System (INIS)

    Liu, Y B; Su, Y M; Ju, L; Huang, S L

    2012-01-01

    A new numerical method was developed for predicting the steady hydrodynamic performance of propeller-rudder-bulb system. In the calculation, the rudder and bulb was taken into account as a whole, the potential based surface panel method was applied both to propeller and rudder-bulb system. The interaction between propeller and rudder-bulb was taken into account by velocity potential iteration in which the influence of propeller rotation was considered by the average influence coefficient. In the influence coefficient computation, the singular value should be found and deleted. Numerical results showed that the method presented is effective for predicting the steady hydrodynamic performance of propeller-rudder system and propeller-rudder-bulb system. Comparing with the induced velocity iterative method, the method presented can save programming and calculation time. Changing dimensions, the principal parameter—bulb size that affect energy-saving effect was studied, the results show that the bulb on rudder have a optimal size at the design advance coefficient.

  18. Reduced Complexity Detection in MIMO Systems with SC-FDE Modulations and Iterative DFE Receivers

    Directory of Open Access Journals (Sweden)

    Filipe Casal Ribeiro

    2018-04-01

    Full Text Available This paper considers a Multiple-Input Multiple-Output (MIMO system with P transmitting and R receiving antennas and different overall noise characteristics on the different receiver antennas (e.g., due to nonlinear effects at the receiver side. Each communication link employs a Single-Carrier with Frequency-Domain Equalization (SC-FDE modulation scheme, and the receiver is based on robust iterative frequency-domain multi-user detectors based on the Iterative Block Decision Feedback Equalization (IB-DFE concept. We present low complexity efficient receivers that can employ low resolution Analog-to-Digital Converters (ADCs and require the inversion of matrices with reduced dimension when the number of receive antennas is larger than the number of independent data streams. The advantages of the proposed techniques are particularly high for highly unbalanced MIMO systems, such as in the uplink of Base Station (BS cooperation systems that aim for Single-Frequency Network (SFN operation or massive MIMO systems with much more antennas at the receiver side.

  19. ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

    International Nuclear Information System (INIS)

    Berry, Jan; Ferrada, Juan J.; Curd, Warren; Dell Orco, Giovanni; Barabash, Vladimir; Kim, Seokho H.

    2011-01-01

    During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

  20. A noise power spectrum study of a new model-based iterative reconstruction system: Veo 3.0.

    Science.gov (United States)

    Li, Guang; Liu, Xinming; Dodge, Cristina T; Jensen, Corey T; Rong, X John

    2016-09-08

    The purpose of this study was to evaluate performance of the third generation of model-based iterative reconstruction (MBIR) system, Veo 3.0, based on noise power spectrum (NPS) analysis with various clinical presets over a wide range of clinically applicable dose levels. A CatPhan 600 surrounded by an oval, fat-equivalent ring to mimic patient size/shape was scanned 10 times at each of six dose levels on a GE HD 750 scanner. NPS analysis was performed on images reconstructed with various Veo 3.0 preset combinations for comparisons of those images reconstructed using Veo 2.0, filtered back projection (FBP) and adaptive statistical iterative reconstruc-tion (ASiR). The new Target Thickness setting resulted in higher noise in thicker axial images. The new Texture Enhancement function achieved a more isotropic noise behavior with less image artifacts. Veo 3.0 provides additional reconstruction options designed to allow the user choice of balance between spatial resolution and image noise, relative to Veo 2.0. Veo 3.0 provides more user selectable options and in general improved isotropic noise behavior in comparison to Veo 2.0. The overall noise reduction performance of both versions of MBIR was improved in comparison to FBP and ASiR, especially at low-dose levels. © 2016 The Authors.

  1. A noise power spectrum study of a new model‐based iterative reconstruction system: Veo 3.0

    Science.gov (United States)

    Li, Guang; Liu, Xinming; Dodge, Cristina T.; Jensen, Corey T.

    2016-01-01

    The purpose of this study was to evaluate performance of the third generation of model‐based iterative reconstruction (MBIR) system, Veo 3.0, based on noise power spectrum (NPS) analysis with various clinical presets over a wide range of clinically applicable dose levels. A CatPhan 600 surrounded by an oval, fat‐equivalent ring to mimic patient size/shape was scanned 10 times at each of six dose levels on a GE HD 750 scanner. NPS analysis was performed on images reconstructed with various Veo 3.0 preset combinations for comparisons of those images reconstructed using Veo 2.0, filtered back projection (FBP) and adaptive statistical iterative reconstruction (ASiR). The new Target Thickness setting resulted in higher noise in thicker axial images. The new Texture Enhancement function achieved a more isotropic noise behavior with less image artifacts. Veo 3.0 provides additional reconstruction options designed to allow the user choice of balance between spatial resolution and image noise, relative to Veo 2.0. Veo 3.0 provides more user selectable options and in general improved isotropic noise behavior in comparison to Veo 2.0. The overall noise reduction performance of both versions of MBIR was improved in comparison to FBP and ASiR, especially at low‐dose levels. PACS number(s): 87.57.‐s, 87.57.Q‐, 87.57.C‐, 87.57.nf, 87.57.C‐, 87.57.cm PMID:27685118

  2. Selection of a quench detection system for the ITER CS magnet

    International Nuclear Information System (INIS)

    Coatanea, Marc; Duchateau, Jean-Luc; Lacroix, Benoit; Nicollet, Sylvie; Rodriguez-Mateos, Felix; Topin, Frederic

    2011-01-01

    At variance with most of the existing superconducting systems operating in the world, the ITER central solenoid (CS) magnet is a fast pulsed system. This peculiarity creates a specific situation regarding the quench detection system, as a small resistive signal associated with a quench has to be discriminated from the high inductive signals imposed by the plasma scenario. The quench detection is based on an inductive compensation built from three adjacent double pancakes. The ITER protection rules for a superconducting magnet impose to respect the so-called maximum hot spot temperature criterion of 250 K in the quenched cable at the end of the fast discharge. A careful analysis of the residual inductive signals in the detection voltage shows that a blanking of the quench detection cannot be avoided during the early times of the plasma discharge (i.e. during 3.5 s). It is demonstrated that this blanking is, however, acceptable while fulfilling the hot spot criterion because the plasma initiation phase (PIP) is very similar to a fast safety discharge and corresponds to a fast decrease of the modules currents, which is favourable for the magnet protection.

  3. Improved Iterative Parallel Interference Cancellation Receiver for Future Wireless DS-CDMA Systems

    Directory of Open Access Journals (Sweden)

    Andrea Bernacchioni

    2005-04-01

    Full Text Available We present a new turbo multiuser detector for turbo-coded direct sequence code division multiple access (DS-CDMA systems. The proposed detector is based on the utilization of a parallel interference cancellation (PIC and a bank of turbo decoders. The PIC is broken up in order to perform interference cancellation after each constituent decoder of the turbo decoding scheme. Moreover, in the paper we propose a new enhanced algorithm that provides a more accurate estimation of the signal-to-noise-plus-interference-ratio used in the tentative decision device and in the MAP decoding algorithm. The performance of the proposed receiver is evaluated by means of computer simulations for medium to very high system loads, in AWGN and multipath fading channel, and compared to recently proposed interference cancellation-based iterative MUD, by taking into account the number of iterations and the complexity involved. We will see that the proposed receiver outperforms the others especially for highly loaded systems.

  4. Proposed high voltage power supply for the ITER relevant lower hybrid current drive system

    International Nuclear Information System (INIS)

    Sharma, P.K.; Kazarian, F.; Garibaldi, P.; Gassman, T.; Artaud, J.F.; Bae, Y.S.; Belo, J.; Berger-By, G.; Bernard, J.M.; Cara, Ph.; Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Cesario, R.; Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Goniche, M.; Guilhem, D.

    2011-01-01

    In the framework of the EFDA task HCD-08-03-01, the ITER lower hybrid current drive (LHCD) system design has been reviewed. The system aims to generate 24 MW of RF power at 5 GHz, of which 20 MW would be coupled to the plasmas. The present state of the art does not allow envisaging a unitary output of the klystrons exceeding 500 kW, so the project is based on 48 klystron units, leaving some margin when the transmission lines losses are taken into account. A high voltage power supply (HVPS), required to operate the klystrons, is proposed. A single HVPS would be used to feed and operate four klystrons in parallel configuration. Based on the above considerations, it is proposed to design and develop twelve HVPS, based on pulse step modulator (PSM) technology, each rated for 90 kV/90 A. This paper describes in details, the typical electrical requirements and the conceptual design of the proposed HVPS for the ITER LHCD system.

  5. A protection system for the JET ITER-like wall based on imaging diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Arnoux, G.; Balboa, I.; Balshaw, N.; Beldishevski, M.; Cramp, S.; Felton, R.; Goodyear, A.; Horton, A.; Kinna, D.; McCullen, P.; Obrejan, K.; Patel, K.; Lomas, P. J.; Rimini, F.; Stamp, M.; Stephen, A.; Thomas, P. D.; Williams, J.; Wilson, J.; Zastrow, K.-D. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2012-10-15

    The new JET ITER-like wall (made of beryllium and tungsten) is more fragile than the former carbon fiber composite wall and requires active protection to prevent excessive heat loads on the plasma facing components (PFC). Analog CCD cameras operating in the near infrared wavelength are used to measure surface temperature of the PFCs. Region of interest (ROI) analysis is performed in real time and the maximum temperature measured in each ROI is sent to the vessel thermal map. The protection of the ITER-like wall system started in October 2011 and has already successfully led to a safe landing of the plasma when hot spots were observed on the Be main chamber PFCs. Divertor protection is more of a challenge due to dust deposits that often generate false hot spots. In this contribution we describe the camera, data capture and real time processing systems. We discuss the calibration strategy for the temperature measurements with cross validation with thermal IR cameras and bi-color pyrometers. Most importantly, we demonstrate that a protection system based on CCD cameras can work and show examples of hot spot detections that stop the plasma pulse. The limits of such a design and the associated constraints on the operations are also presented.

  6. 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

  7. Functional System Dynamics

    OpenAIRE

    Ligterink, N.E.

    2007-01-01

    Functional system dynamics is the analysis, modelling, and simulation of continuous systems usually described by partial differential equations. From the infinite degrees of freedom of such systems only a finite number of relevant variables have to be chosen for a practical model description. The proper input and output of the system are an important part of the relevant variables.

  8. Iterative functionalism and climate management regimes: From intergovernmental panel on climate change to intergovernmental negotiating committee

    International Nuclear Information System (INIS)

    Feldman, D.L.; Tennessee Univ., Knoxville, TN

    1992-01-01

    This paper contends that an iterative ''functionalist'' regime -- comprised of international organizations that monitor the global climate and perform scientific and policy research on prevention, mitigation, and adaptation strategies for response to possible global warming -- has developed over the past decade. A common global effort by scientists, diplomats, and others to negotiate a framework convention that would reduce emissions of carbon dioxide and other ''greenhouse gases'' has been brought about by this regime. Individuals that participate in this regime are engaged in several cooperative activities including: (1) international research on the causes and consequences of global change; (2) global environmental monitoring and standard-setting for analyses of climate data; and (3) negotiating a framework convention that places limits on greenhouse gas emissions by countries. The implications of this iterative approach for successful implementation of a treaty to forestall global climate change are discussed

  9. 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.

  10. ATCA Shelf Manager EPICS device support for ITER CODAC Core System

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Bruno, E-mail: bsantos@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Carvalho, Paulo F.; Rodrigues, A.P.; Carvalho, Bernardo B.; Sousa, Jorge; Batista, António J.N.; Correia, Miguel; Combo, Álvaro M.; Cruz, Nuno [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Correia, Carlos M.B.A. [Centro de Instrumentação, Departamento de Física, Universidade de Coimbra, 3004-516 Coimbra (Portugal); Gonçalves, Bruno [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal)

    2015-10-15

    Highlights: • This architecture targets the health management integration into the NDS. • The developed solution supports the ShM redundancy features, specified by ATCA. • The average RTT was around 59 ms and in 99.9% of the cases was less than 130 ms. • Without losing any update cycle, can monitor a system shelf with approximately 400 sensors. • This solution enables the user to configure the entire system in DB files and st.cmd. - Abstract: The ITER CODAC Core System (CCS) is responsible for plant Instrumentation and Control (I&C) supervising and monitoring. This system uses the Enhanced Physics and Industrial Control System (EPICS) Channel Access (CA) protocol as the interface with the Plant Operation Network (PON). This paper presents a generic EPICS device support developed for the integration of the ATCA Shelf Manager (ShM) into the ITER CCS, providing scalability and easy configuration. The device support uses the available HTTP interface on Shelf Manager in the communication layer. Both HTTP server and sensors/actuators definitions can be configured using the EPICS database file and the Input/Output Controller (IOC) initialization file. A proposal based on this device is also presented, targeting the Nominal Device Support (NDS) for health management. The EPICS device support running in an IOC provides Process Variables (PV) to the PON network with the system information and these PVs can be used by all CA clients, such as EPICS user interface clients, alarm systems and archive systems. Operation with redundant ATCA ShMs and device support scalability tests were performed and the results are presented.

  11. 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

  12. Integrated approach to the development of the ITER control system configuration data

    International Nuclear Information System (INIS)

    Stepanov, D.; Abadie, L.; Bertin, J.; Bourguignon, G.; Darcourt, G.; Liotard, O.

    2012-01-01

    ITER control system (CODAC) will rely on a large number of configuration data, coming from different sources. This information is being created using different tools, stored in various databases and, generally, has different life-cycle. In many cases it is difficult for instrumentation and control (IC) engineers to have a common view on this information or to check data consistency. The plant system profile database, described in this paper, tries to address these issues by gathering all IC-specific information in the same database and providing means to analyze these data. At the design phase, the current ITER infrastructure and CODAC practices have been evaluated, and the following architectural and software decisions have been made: -1) the database back-end should be Microsoft SQL Server; -2) the application should have web interface; -3) business logic and the user interface should be written in Java. Prime Faces should be used for the user interface, Spring for transactional support and Java Hibernate for interaction with the database; and -4) data integration tool should be Talend. The task was launched in September 2010; in February 2011 the first version was put in production, and it was gradually introduced into CODAC processes in the following months

  13. Low-Bit Rate Feedback Strategies for Iterative IA-Precoded MIMO-OFDM-Based Systems

    Science.gov (United States)

    Teodoro, Sara; Silva, Adão; Dinis, Rui; Gameiro, Atílio

    2014-01-01

    Interference alignment (IA) is a promising technique that allows high-capacity gains in interference channels, but which requires the knowledge of the channel state information (CSI) for all the system links. We design low-complexity and low-bit rate feedback strategies where a quantized version of some CSI parameters is fed back from the user terminal (UT) to the base station (BS), which shares it with the other BSs through a limited-capacity backhaul network. This information is then used by BSs to perform the overall IA design. With the proposed strategies, we only need to send part of the CSI information, and this can even be sent only once for a set of data blocks transmitted over time-varying channels. These strategies are applied to iterative MMSE-based IA techniques for the downlink of broadband wireless OFDM systems with limited feedback. A new robust iterative IA technique, where channel quantization errors are taken into account in IA design, is also proposed and evaluated. With our proposed strategies, we need a small number of quantization bits to transmit and share the CSI, when comparing with the techniques used in previous works, while allowing performance close to the one obtained with perfect channel knowledge. PMID:24678274

  14. Block-Iterative Frequency-Domain Equalizations for SC-IDMA Systems

    Directory of Open Access Journals (Sweden)

    Salah Awad Salman

    2015-07-01

    Full Text Available In wireless broadband communications using single-carrier interleave division multiple access (SC-IDMA systems, efficient multiuser detection (MUD classes that make use of joint hybrid decision feedback equalization (HDFE/ frequency decision-feedback equalization (FDFE and interference cancellation (IC techniques, are proposed in conjunction with channel coding to deal with several users accessing the multipath fading channels. In FDFE-IDMA, the feedforward (FF and feedback (FB filtering operations of FDFE, which use to remove intersymbol interference (ISI, are implemented by Fast Fourier Transforms (FFTs, while in HDFE-IDMA the only FF filter is implemented by FFTs. Further, the parameters involved in the FDFE/HDFE filtering are derived according to the minimum mean square error (MMSE criteria, and the feedback symbol decisions are directly designed from soft detection of the decoded signals at the previous iteration. The simulation results including comparisons with those of frequency domain equalization (FDE, SC-FDE-IDMA and multi-carrier OFDM-IDMA schemes, with cyclic prefixing (CP and zero padding (ZP techniques, show that the combination of FDFE-IC/HDFE-IC provides an efficient solution with good performance for IDMA systems in ISI channels. Moreover, these iterative structures with block equalization yield a much lower complexity than equivalent existing structures, making them attractive for a wealth of applications.

  15. Conceptual design finalisation of the ITER In-Vessel Viewing and Metrology System (IVVS)

    Energy Technology Data Exchange (ETDEWEB)

    Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Damiani, Carlo; Van Uffelen, Marco; Lo Bue, Alessandro; Izquierdo, Jesus; Semeraro, Luigi [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Martins, Jean-Pierre; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R and D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis.

  16. 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)

  17. Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

    International Nuclear Information System (INIS)

    Wu, Y.

    2007-01-01

    A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

  18. Conceptual design of the Radial Gamma Ray Spectrometers system for α particle and runaway electron measurements at ITER

    DEFF Research Database (Denmark)

    Nocente, Massimo; Tardocchi, Marco; Barnsley, Robin

    2017-01-01

    We here present the principles and main physics capabilities behind the design of the radial gamma ray spectrometers (RGRS) system for alpha particle and runaway electron measurements at ITER. The diagnostic benefits from recent advances in gamma-ray spectrometry for tokamak plasmas and combines...... the measurements sensitive to α particles at characteristic resonant energies and to possible anisotropies of their slowing down distribution function. An independent assessment of the neutron rate by gamma-ray emission is also feasible. In case of runaway electrons born in disruptions with a typical duration...... of 100ms, a time resolution of at least 10ms for runaway electron studies can be achieved depending on the scenario and down to a current of 40 kA by use of external gas injection. We find that the bremsstrahlung spectrum in the MeV range from confined runaways is sensitive to the electron velocity space...

  19. Guidelines for remote handling maintenance of ITER neutral beam line components: Proposal of an alternate supporting system

    International Nuclear Information System (INIS)

    Cordier, J.J.; Bayetti, P.; Hemsworth, R.; David, O.; Friconneau, J.P.

    2007-01-01

    Remote handling (R/H) maintenance of ITER components is one of the main challenges of the ITER project. This type of maintenance shall be operational for the assembly and nuclear phase of exploitation of ITER. It must be considered at a very early stage since it significantly impacts on the components design, interfaces management, assembly, maintenance and integration aspects. A large part of the R/H equipment will be procured by the EU Participating Team, including the whole Neutral Beam R/H Equipment. The Neutral Beam Heating and Current Drive system (NB and CD) design is being revisited by the ITER project. A vertical maintenance scheme is presently considered which may significantly impact on the reference design and associated components and lead to a new design of the NB and CD vacuum tank. In addition, NB line components remote handling solutions are being studied. The neutral beam test facility ITER to be built in Europe in the near future is also based on the vertical NB maintenance scheme of beam line components. New design guidelines compliant for both the ITER NB and CD system and the NB test facility proposed by the CEA association are described in the paper

  20. An iterative method for the canard explosion in general planar systems

    DEFF Research Database (Denmark)

    Brøns, Morten

    2013-01-01

    The canard explosion is the change of amplitude and period of a limit cycle born in a Hopf bifurcation in a very narrow parameter interval. The phenomenon is well understood in singular perturbation problems where a small parameter controls the slow/fast dynamics. However, canard explosions are a...... on the van der Pol equation, showing that the asymptotics of the method is correct, and on a templator model for a self-replicating system....... are also observed in systems where no such parameter can obviously be identied. Here we show how the iterative method of Roussel and Fraser, devised to construct regular slow manifolds, can be used to determine a canard point in a general planar system of nonlinear ODEs. We demonstrate the method...

  1. A Block Iterative Finite Element Model for Nonlinear Leaky Aquifer Systems

    Science.gov (United States)

    Gambolati, Giuseppe; Teatini, Pietro

    1996-01-01

    A new quasi three-dimensional finite element model of groundwater flow is developed for highly compressible multiaquifer systems where aquitard permeability and elastic storage are dependent on hydraulic drawdown. The model is solved by a block iterative strategy, which is naturally suggested by the geological structure of the porous medium and can be shown to be mathematically equivalent to a block Gauss-Seidel procedure. As such it can be generalized into a block overrelaxation procedure and greatly accelerated by the use of the optimum overrelaxation factor. Results for both linear and nonlinear multiaquifer systems emphasize the excellent computational performance of the model and indicate that convergence in leaky systems can be improved up to as much as one order of magnitude.

  2. A description of the ITER's gas injection systems and current R and D activities

    International Nuclear Information System (INIS)

    Li, W.; Li, B.; Maruyama, S.; Jiang, T.; Yang, Y.; Xia, Z.W.; Zhang, Y.X.; Lu, J.

    2012-01-01

    The gas injection system (GIS) is an indispensable part of ITER fueling system. It deliveries the necessary gas species from tritium plant to vacuum vessel, pellet injection system or neutral beam for plasma operation and fusion power shutdown. In this paper, the current design status of GIS, including the previous design changes, is briefly described. As the GIS design justification and support, the experimental study on GIS response time is illustrated. The factors delayed the GIS response time are identified, and two kinds of control mode are proved to be effective for improving the GIS response time. The exploration on magnetic shield design shows the discrepancy of shielding performance occurs in the case of the paralleling external magnetic field to the sample cylinder. These R and D works prove the design feasibility in some ways, and support possible solutions for design challenges as alternative design options.

  3. Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

    International Nuclear Information System (INIS)

    Bittner-Rohrhofer, K.; Humer, K.; Weber, H.W.; Hamada, K.; Sugimoto, M.; Okuno, K.

    2002-01-01

    The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2x10 22 m -2 (E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite

  4. Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

    Science.gov (United States)

    Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Hamada, K.; Sugimoto, M.; Okuno, K.

    2002-12-01

    The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2×10 22 m -2 ( E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite.

  5. 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

  6. Enhancement of the use of digital mock-ups in the verification and validation process for ITER remote handling systems

    Energy Technology Data Exchange (ETDEWEB)

    Sibois, R., E-mail: romain.sibois@vtt.fi [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Salminen, K.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland); Mattila, J. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T. [VTT Technical Research Centre of Finland, P.O. Box 1300, 33101 Tampere (Finland)

    2013-10-15

    Highlights: • Verification and validation process for ITER remote handling system. • Verification and validation framework for complex engineering systems. • Verification and validation roadmap for digital modelling phase. • Importance of the product life-cycle management in the verification and validation framework. -- Abstract: The paper is part of the EFDA's programme of European Goal Oriented Training programme on remote handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. This paper is written based on the results of a project “verification and validation (V and V) of ITER RH system using digital mock-ups (DMUs)”. The purpose of this project is to study efficient approach of using DMU for the V and V of the ITER RH system design utilizing a system engineering (SE) framework. This paper reviews the definitions of DMU and virtual prototype and overviews the current trends of using virtual prototyping in the industry during the early design phase. Based on the survey of best industrial practices, this paper proposes ways to improve the V and V process for ITER RH system utilizing DMUs.

  7. Efficient methodologies for system matrix modelling in iterative image reconstruction for rotating high-resolution PET

    Energy Technology Data Exchange (ETDEWEB)

    Ortuno, J E; Kontaxakis, G; Rubio, J L; Santos, A [Departamento de Ingenieria Electronica (DIE), Universidad Politecnica de Madrid, Ciudad Universitaria s/n, 28040 Madrid (Spain); Guerra, P [Networking Research Center on Bioengineering, Biomaterials and Nanomedicine (CIBER-BBN), Madrid (Spain)], E-mail: juanen@die.upm.es

    2010-04-07

    A fully 3D iterative image reconstruction algorithm has been developed for high-resolution PET cameras composed of pixelated scintillator crystal arrays and rotating planar detectors, based on the ordered subsets approach. The associated system matrix is precalculated with Monte Carlo methods that incorporate physical effects not included in analytical models, such as positron range effects and interaction of the incident gammas with the scintillator material. Custom Monte Carlo methodologies have been developed and optimized for modelling of system matrices for fast iterative image reconstruction adapted to specific scanner geometries, without redundant calculations. According to the methodology proposed here, only one-eighth of the voxels within two central transaxial slices need to be modelled in detail. The rest of the system matrix elements can be obtained with the aid of axial symmetries and redundancies, as well as in-plane symmetries within transaxial slices. Sparse matrix techniques for the non-zero system matrix elements are employed, allowing for fast execution of the image reconstruction process. This 3D image reconstruction scheme has been compared in terms of image quality to a 2D fast implementation of the OSEM algorithm combined with Fourier rebinning approaches. This work confirms the superiority of fully 3D OSEM in terms of spatial resolution, contrast recovery and noise reduction as compared to conventional 2D approaches based on rebinning schemes. At the same time it demonstrates that fully 3D methodologies can be efficiently applied to the image reconstruction problem for high-resolution rotational PET cameras by applying accurate pre-calculated system models and taking advantage of the system's symmetries.

  8. Functional System Dynamics

    NARCIS (Netherlands)

    Ligterink, N.E.

    2007-01-01

    Functional system dynamics is the analysis, modelling, and simulation of continuous systems usually described by partial differential equations. From the infinite degrees of freedom of such systems only a finite number of relevant variables have to be chosen for a practical model description. The

  9. 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)

  10. Manufacture and testing of the CTB&SBB thermal shield for the ITER magnet feeder system

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Kun; Liu, Chen; Song, Yuntao; Feng, Hansheng; Ding, Kaizhong, E-mail: kzding@ipp.ac.cn; Wang, Tanbin; Ji, Hui

    2015-10-15

    The system of International Thermonuclear Experimental Reactor (ITER) feeders is responsible for the power, helium cooling, and instrumentation of the magnets of the coil terminal box and S-bend box (CTB&SBB) thermal shield outside the cryostat. An 80-K rectangular Al thermal shield is hung inside the CTB&SBB to reduce the thermal radiation heat loads of 4.5-K helium. The American Society of Interventional Pain Physicians (ASIPP) will supply all the 31 sets of feeders for ITER. A manufactured prototype of CTB&SBB thermal shield is first quality-tested before the commencement of the series production. First, a detailed configuration of the rectangular Al thermal shield is presented in this article. The paper also presents more information on the manufacturing process of the thermal shield, especially the welding process, the procedure for ensuring good weld quality, and the use of a specially designed tool to ensure <5-mm deformation on such a 7.3-m-long thermal shield during welding. In addition, the cold test and results, including the cooling process with 13-bar and 17.5-g/s 80-K He gas, and the temperature distribution on different panels of the thermal shield are presented. The whole process of manufacture and testing lays a good foundation for the series production of the thermal shield.

  11. MPPT-Based Control Algorithm for PV System Using iteration-PSO under Irregular shadow Conditions

    Directory of Open Access Journals (Sweden)

    M. Abdulkadir

    2017-02-01

    Full Text Available The conventional maximum power point tracking (MPPT techniques can hardly track the global maximum power point (GMPP because the power-voltage characteristics of photovoltaic (PV exhibit multiple local peaks in irregular shadow, and therefore easily fall into the local maximum power point. These conditions make it very challenging, and to tackle this deficiency, an efficient Iteration Particle Swarm Optimization (IPSO has been developed to improve the quality of solution and convergence speed of the traditional PSO, so that it can effectively track the GMPP under irregular shadow conditions. This proposed technique has such advantages as simple structure, fast response and strong robustness, and convenient implementation. It is applied to MPPT control of PV system in irregular shadow to solve the problem of multi-peak optimization in partial shadow. In order to verify the rationality of the proposed algorithm, however, recently the dynamic MPPT performance under varying irradiance conditions has been given utmost attention to the PV society. As the European standard EN 50530 which defines the recommended varying irradiance profiles, was released lately, the corresponding researchers have been required to improve the dynamic MPPT performance. This paper tried to evaluate the dynamic MPPT performance using EN 50530 standard. The simulation results show that iterative-PSO method can fast track the global MPP, increase tracking speed and higher dynamic MPPT efficiency under EN 50530 than the conventional PSO.

  12. 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.

  13. On iterative solution of nonlinear functional equations in a metric space

    Directory of Open Access Journals (Sweden)

    Rabindranath Sen

    1983-01-01

    Full Text Available Given that A and P as nonlinear onto and into self-mappings of a complete metric space R, we offer here a constructive proof of the existence of the unique solution of the operator equation Au=Pu, where u∈R, by considering the iterative sequence Aun+1=Pun (u0 prechosen, n=0,1,2,…. We use Kannan's criterion [1] for the existence of a unique fixed point of an operator instead of the contraction mapping principle as employed in [2]. Operator equations of the form Anu=Pmu, where u∈R, n and m positive integers, are also treated.

  14. A practical iterative PID tuning method for mechanical systems using parameter chart

    Science.gov (United States)

    Kang, M.; Cheong, J.; Do, H. M.; Son, Y.; Niculescu, S.-I.

    2017-10-01

    In this paper, we propose a method of iterative proportional-integral-derivative parameter tuning for mechanical systems that possibly possess hidden mechanical resonances, using a parameter chart which visualises the closed-loop characteristics in a 2D parameter space. We employ a hypothetical assumption that the considered mechanical systems have their upper limit of the derivative feedback gain, from which the feasible region in the parameter chart becomes fairly reduced and thus the gain selection can be extremely simplified. Then, a two-directional parameter search is carried out within the feasible region in order to find the best set of parameters. Experimental results show the validity of the assumption used and the proposed parameter tuning method.

  15. Iterative solution of general sparse linear systems on clusters of workstations

    Energy Technology Data Exchange (ETDEWEB)

    Lo, Gen-Ching; Saad, Y. [Univ. of Minnesota, Minneapolis, MN (United States)

    1996-12-31

    Solving sparse irregularly structured linear systems on parallel platforms poses several challenges. First, sparsity makes it difficult to exploit data locality, whether in a distributed or shared memory environment. A second, perhaps more serious challenge, is to find efficient ways to precondition the system. Preconditioning techniques which have a large degree of parallelism, such as multicolor SSOR, often have a slower rate of convergence than their sequential counterparts. Finally, a number of other computational kernels such as inner products could ruin any gains gained from parallel speed-ups, and this is especially true on workstation clusters where start-up times may be high. In this paper we discuss these issues and report on our experience with PSPARSLIB, an on-going project for building a library of parallel iterative sparse matrix solvers.

  16. 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.

  17. ITER fast ion collective Thomson scattering, conceptual design of 60 GHz system

    International Nuclear Information System (INIS)

    Meo, F.; Bindslev, H.; Korsholm, S.B.

    2007-08-01

    The collective Thomson scattering diagnostic for ITER at the 60 GHz range is capable of measuring the fast ion distribution parallel and perpendicular to the magnetic field at different radial locations simultaneously. The design is robust technologically with no moveable components near the plasma. The fast ion CTS diagnostic consists of two separate systems. Each system has its own RF launcher and separate set of detectors. The first system measures the perpendicular component of the fast ion velocity distribution. It consists of radially directed RF launcher and receiver, both located in the equatorial port on the low field side (LFS). This system will be referred to by the acronym LFS-BS system referring to the location of the receiver and the fact that it measures backscattered radiation. The second part of the CTS diagnostic measures the parallel component of the fast ion distribution. It consists of an RF launcher located in the mid-plane port on the LFS and a receiver mounted on the inner vacuum vessel wall that views the plasma from between two blanket modules. This system will be referred to as HFS-FS referring to the location of the receivers and that they measure forward scattered radiation. The design of both LFS-BS and HFS-FS receivers is aimed at measuring at different spatial locations simultaneously with no moveable components near the plasma. This report is a preliminary study of the hardware design and engineering constraints for this frequency range. Section 2 conceptually describes the two systems and their main components. Section 3 clarifies the impact of design parameters such as beam widths and scattering angle on the CTS measurements. With this in hand, the ITER measurement requirements are translated into constraints on the CTS system designs. An important result in this section is that systems can be designed inside these constraints. Section 4 outlines the technical feasibility and describes in more detail the design and the engineering

  18. 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

  19. Protection measures for selected ITER magnet system off-normal conditions

    International Nuclear Information System (INIS)

    Yoshida, K.; Iida, F.; Gallix, R.; Britousov, N.; Mitchell, N.; Thome, R.J.

    1998-01-01

    The International Thermonuclear Experimental Reactor (ITER) magnet systems provide the magnetic field intensity and field geometry to contain and control plasma during the various phases of pulsed operation. During these pulses, the toroidal field (TF) coils operate with a constant current. The central solenoid (CS) and poloidal field (PF) coils, on the other hand, are each independently powered. The maximum terminal voltages during plasma operation and protective discharges are 15 kV for CS and 10 kV for TF and PF. The energy stored in the 20 TF coil system is 103 GJ; in each of the other coils it is approximately 10 GJ or less. This paper describes the protection requirements and selected design concepts being considered for the large superconducting coils for the ITER. Ground faults, short circuits and helium leaks are the major serious accidents to be prevented in the coils. All coils use a solid insulation system to avoid ground faults. The electrical circuits including coil and power supply are grounded through resistors that limit current in the event of a ground fault. In the case of a short circuit within the coil winding, a large energy would be dissipated close to the small shorted volume. The impact of the short circuit can be reduced by using a potential screen. Inside the cryostat, helium leakage is most likely at the electrical insulating breaks in the cryogenic cooling lines between the coils and helium manifolds. A double containment (metallic shield and glass-epoxy) is therefore provided for the insulation breaks to allow for the detection of small leaks and to limit the spread of helium to other locations. (orig.)

  20. Evaluation of iterative algorithms for tomography image reconstruction: A study using a third generation industrial tomography system

    Energy Technology Data Exchange (ETDEWEB)

    Velo, Alexandre F.; Carvalho, Diego V.; Alvarez, Alexandre G.; Hamada, Margarida M.; Mesquita, Carlos H., E-mail: afvelo@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The greatest impact of the tomography technology currently occurs in medicine. The success is due to the human body presents standardized dimensions with well-established composition. These conditions are not found in industrial objects. In industry, there is much interest in using the tomography in order to know the inner of (1) the manufactured industrial objects or (2) the machines and their means of production. In these cases, the purpose of the tomography is to (a) control the quality of the final product and (b) to optimize production, contributing to the pilot phase of the projects and analyzing the quality of the means of production. This scan system is a non-destructive, efficient and fast method for providing sectional images of industrial objects and is able to show the dynamic processes and the dispersion of the materials structures within these objects. In this context, it is important that the reconstructed image presents a great spatial resolution with a satisfactory temporal resolution. Thus the algorithm to reconstruct the images has to meet these requirements. This work consists in the analysis of three different iterative algorithm methods, such Maximum Likelihood Estimation Method (MLEM), Maximum Likelihood Transmitted Method (MLTR) and Simultaneous Iterative Reconstruction Method (SIRT. The analysis consists on measurement of the contrast to noise ratio (CNR), the root mean square error (RMSE) and the Modulation Transfer Function (MTF), to know which algorithm fits better the conditions in order to optimize system. The algorithms and the image quality analysis were performed by the Matlab® 2013b. (author)

  1. Evaluation of iterative algorithms for tomography image reconstruction: A study using a third generation industrial tomography system

    International Nuclear Information System (INIS)

    Velo, Alexandre F.; Carvalho, Diego V.; Alvarez, Alexandre G.; Hamada, Margarida M.; Mesquita, Carlos H.

    2017-01-01

    The greatest impact of the tomography technology currently occurs in medicine. The success is due to the human body presents standardized dimensions with well-established composition. These conditions are not found in industrial objects. In industry, there is much interest in using the tomography in order to know the inner of (1) the manufactured industrial objects or (2) the machines and their means of production. In these cases, the purpose of the tomography is to (a) control the quality of the final product and (b) to optimize production, contributing to the pilot phase of the projects and analyzing the quality of the means of production. This scan system is a non-destructive, efficient and fast method for providing sectional images of industrial objects and is able to show the dynamic processes and the dispersion of the materials structures within these objects. In this context, it is important that the reconstructed image presents a great spatial resolution with a satisfactory temporal resolution. Thus the algorithm to reconstruct the images has to meet these requirements. This work consists in the analysis of three different iterative algorithm methods, such Maximum Likelihood Estimation Method (MLEM), Maximum Likelihood Transmitted Method (MLTR) and Simultaneous Iterative Reconstruction Method (SIRT. The analysis consists on measurement of the contrast to noise ratio (CNR), the root mean square error (RMSE) and the Modulation Transfer Function (MTF), to know which algorithm fits better the conditions in order to optimize system. The algorithms and the image quality analysis were performed by the Matlab® 2013b. (author)

  2. Remote maintenance of a combined regeneration-isolation valve for the ITER Torus vacuum pumping system

    International Nuclear Information System (INIS)

    Stringer, J.; Blevins, J.

    1992-01-01

    A large diameter valve suitable for high vacuum operation is under study for ITER Torus evacuation. The valves must comply with specifications for leak-tightness, radiation resistance, dust tolerance, overpressure, and thermal gradients. Remote maintenance of the seal and valve moving parts without disturbance to the rest of the valve system is a requirement. This paper describes tow methods of seal exchange by remote means. In the first method, a flask is proposed for the valve moving parts exchange in inert gas, when the machine is shut down. In the second method a novel concept is described for seal exchange while under vacuum, without having to bring the machine up to atmosphere. The advantages of this method are that scheduled remote handling (RH) operations and outages for seal replacement are not required. Also, the need for a flask is avoided

  3. An alpha particle measurement system using an energetic neutral helium beam in ITER (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Sasao, M.; Tanaka, N.; Terai, K.; Kaneko, O. [Graduate school of Engineering, Tohoku University, Sendai 980-8579 (Japan); Kisaki, M.; Kobuchi, T.; Tsumori, K.; Okamoto, A.; Kitajima, S. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Shinto, K. [IFMIF R and D Center, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Wada, M. [Graduate School of Engineering, Doshisha University, Kyotanabe, Kyoto 610-0321 (Japan)

    2012-02-15

    An energetic helium neutral beam is involved in the beam neutralization measurement system of alpha particles confined in a DT fusion plasma. A full size strong-focusing He{sup +} ion source (2 A, the beam radius of 11.3 mm, the beam energy less than 20 keV). Present strong-focusing He{sup +} ion source shows an emittance diagram separated for each beamlet of multiple apertures without phase space mixing, despite the space charge of a beamlet is asymmetric and the beam flow is non-laminar. The emittance of beamlets in the peripheral region was larger than that of center. The heat load to the plasma electrode was studied to estimate the duty factor for the ITER application.

  4. Modeling of complex gas distribution systems operating under any vacuum conditions: Simulations of the ITER divertor pumping system

    International Nuclear Information System (INIS)

    Vasileiadis, N.; Tatsios, G.; Misdanitis, S.; Valougeorgis, D.

    2016-01-01

    Highlights: • An integrated s/w for modeling complex rarefied gas distribution systems is presented. • Analysis is based on kinetic theory of gases. • Code effectiveness is demonstrated by simulating the ITER divertor pumping system. • The present s/w has the potential to support design work in large vacuum systems. - Abstract: An integrated software tool for modeling and simulation of complex gas distribution systems operating under any vacuum conditions is presented and validated. The algorithm structure includes (a) the input geometrical and operational data of the network, (b) the definition of the fundamental set of network loops and pseudoloops, (c) the formulation and solution of the mass and energy conservation equations, (d) the kinetic data base of the flow rates for channels of any length in the whole range of the Knudsen number, supporting, in an explicit manner, the solution of the conservation equations and (e) the network output data (mainly node pressures and channel flow rates/conductance). The code validity is benchmarked under rough vacuum conditions by comparison with hydrodynamic solutions in the slip regime. Then, its feasibility, effectiveness and potential are demonstrated by simulating the ITER torus vacuum system with the six direct pumps based on the 2012 design of the ITER divertor. Detailed results of the flow patterns and paths in the cassettes, in the gaps between the cassettes and along the divertor ring, as well as of the total throughput for various pumping scenarios and dome pressures are provided. A comparison with previous results available in the literature is included.

  5. Function integrated track system

    OpenAIRE

    Hohnecker, Eberhard

    2010-01-01

    The paper discusses a function integrated track system that focuses on the reduction of acoustic emissions from railway lines. It is shown that the combination of an embedded rail system (ERS), a sound absorbing track surface, and an integrated mini sound barrier has significant acoustic advantages compared to a standard ballast superstructure. The acoustic advantages of an embedded rail system are particularly pronounced in the case of railway bridges. Finally, it is shown that a...

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

    International Nuclear Information System (INIS)

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

    1989-03-01

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

  7. Iterative reconstruction of SiPM light response functions in a square-shaped compact gamma camera

    Science.gov (United States)

    Morozov, A.; Alves, F.; Marcos, J.; Martins, R.; Pereira, L.; Solovov, V.; Chepel, V.

    2017-05-01

    Compact gamma cameras with a square-shaped monolithic scintillator crystal and an array of silicon photomultipliers (SiPMs) are actively being developed for applications in areas such as small animal imaging, cancer diagnostics and radiotracer guided surgery. Statistical methods of position reconstruction, which are potentially superior to the traditional centroid method, require accurate knowledge of the spatial response of each photomultiplier. Using both Monte Carlo simulations and experimental data obtained with a camera prototype, we show that the spatial response of all photomultipliers (light response functions) can be parameterized with axially symmetric functions obtained iteratively from flood field irradiation data. The study was performed with a camera prototype equipped with a 30  ×  30  ×  2 mm3 LYSO crystal and an 8  ×  8 array of SiPMs for 140 keV gamma rays. The simulations demonstrate that the images, reconstructed with the maximum likelihood method using the response obtained with the iterative approach, exhibit only minor distortions: the average difference between the reconstructed and the true positions in X and Y directions does not exceed 0.2 mm in the central area of 22  ×  22 mm2 and 0.4 mm at the periphery of the camera. A similar level of image distortions is shown experimentally with the camera prototype.

  8. Rapid prototyping of the Central Safety System for Nuclear Risk in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Scibile, L. [ITER Organization, CS 90 046, St. Paul-lez-Durance, Cedex (France); Ambrosino, G. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); De Tommasi, G., E-mail: detommas@unina.i [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); Pironti, A. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy)

    2010-07-15

    The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.

  9. Iterative Systems Biology for Medicine – time for advancing from network signature to mechanistic equations

    KAUST Repository

    Gomez-Cabrero, David

    2017-05-09

    The rise and growth of Systems Biology following the sequencing of the human genome has been astounding. Early on, an iterative wet-dry methodology was formulated which turned out as a successful approach in deciphering biological complexity. Such type of analysis effectively identified and associated molecular network signatures operative in biological processes across different systems. Yet, it has proven difficult to distinguish between causes and consequences, thus making it challenging to attack medical questions where we require precise causative drug targets and disease mechanisms beyond a web of associated markers. Here we review principal advances with regard to identification of structure, dynamics, control, and design of biological systems, following the structure in the visionary review from 2002 by Dr. Kitano. Yet, here we find that the underlying challenge of finding the governing mechanistic system equations enabling precision medicine remains open thus rendering clinical translation of systems biology arduous. However, stunning advances in raw computational power, generation of high-precision multi-faceted biological data, combined with powerful algorithms hold promise to set the stage for data-driven identification of equations implicating a fundamental understanding of living systems during health and disease.

  10. Rapid prototyping of the Central Safety System for Nuclear Risk in ITER

    International Nuclear Information System (INIS)

    Scibile, L.; Ambrosino, G.; De Tommasi, G.; Pironti, A.

    2010-01-01

    The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.

  11. Rapid Prototyping of the Central Safety System for Nuclear Risk in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Scibile, L. [ITER Organization, 13 - St. Paul lez Durance (France); Ambrosino, G.; De Tommasi, G.; Pironti, A. [Euratom-ENEA-CREATE, Universita di Napoli Federico II, Napoli (Italy)

    2009-07-01

    Full text of publication follows: In the current ITER Baseline design, the Central Safety System for Nuclear Risk (CSS-N) is the safety control system in charge to assure nuclear safety for the plant, personnel and environment. In particular it is envisaged that the CSS shall interface to the plant safety systems for nuclear risk and shall coordinate the individual protection provided by the intervention of these systems by the activation, where required, of additional protections. The design of such a system, together with its implementation, strongly depends on the requirements, particularly in terms of reliability. The CSS-N is a safety critical system, thus its validation and commissioning play a very important role, since the required level of reliability must be demonstrated. In such a scenario, where a new and non-conventional system has to be deployed, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the system requirements, and they will be used to test and validate the control logic. Furthermore these tools can be used to rapid design the safety system and to carry out hardware-in-the-loop (HIL) simulations, which permit to assess the performance of the control hardware against a plant simulator. Both a control system prototype and a safety system oriented plant simulator have been developed to assess first the requirements and then the performance of the CSS-N. In particular the presented SW/HW framework permits to design and verify the CSS protection logics and to test and validate these logics by means of HIL simulations. This work introduces both the prototype and plant simulator architectures, together with the methodology adopted to design and implement these validation tools. (authors)

  12. Design of single-longitudinal-mode laser oscillator for edge Thomson scattering system in ITER

    International Nuclear Information System (INIS)

    Hatae, Takaki; Kusama, Yoshinori; Kubomura, Hiroyuki; Matsuoka, Shin-ichi

    2006-06-01

    A high output energy (5J) and high repetition rate (100 Hz) laser system is required for the edge Thomson scattering system in ITER. A YAG laser (Nd:YAG laser) is a first candidate for the laser system satisfying the requirements. It is important to develop a high beam quality and single longitudinal mode (SLM) laser oscillator in order to realize this high power laser system. In this design work, following activities relating to the SLM laser oscillator have been carried out: design of the laser head and the resonator, estimation of the output power for the SLM laser oscillator, consideration of the feedback control scheme and consideration of interface for amplification system to achieve required performance (5J, 100 Hz). It is expected that the designed laser diode (LD) pumped SLM laser oscillator realizes: 100 Hz of repetition rate, 10 mJ of output energy, 10 ns of pulse width, single longitudinal mode, TEM 00 of transversal mode, divergence less than 4 times of the diffraction limit, energy stability within 5%. (author)

  13. Adaptive optimal control of unknown constrained-input systems using policy iteration and neural networks.

    Science.gov (United States)

    Modares, Hamidreza; Lewis, Frank L; Naghibi-Sistani, Mohammad-Bagher

    2013-10-01

    This paper presents an online policy iteration (PI) algorithm to learn the continuous-time optimal control solution for unknown constrained-input systems. The proposed PI algorithm is implemented on an actor-critic structure where two neural networks (NNs) are tuned online and simultaneously to generate the optimal bounded control policy. The requirement of complete knowledge of the system dynamics is obviated by employing a novel NN identifier in conjunction with the actor and critic NNs. It is shown how the identifier weights estimation error affects the convergence of the critic NN. A novel learning rule is developed to guarantee that the identifier weights converge to small neighborhoods of their ideal values exponentially fast. To provide an easy-to-check persistence of excitation condition, the experience replay technique is used. That is, recorded past experiences are used simultaneously with current data for the adaptation of the identifier weights. Stability of the whole system consisting of the actor, critic, system state, and system identifier is guaranteed while all three networks undergo adaptation. Convergence to a near-optimal control law is also shown. The effectiveness of the proposed method is illustrated with a simulation example.

  14. Performance of multi-aperture grid extraction systems for an ITER-relevant RF-driven negative hydrogen ion source

    Science.gov (United States)

    Franzen, P.; Gutser, R.; Fantz, U.; Kraus, W.; Falter, H.; Fröschle, M.; Heinemann, B.; McNeely, P.; Nocentini, R.; Riedl, R.; Stäbler, A.; Wünderlich, D.

    2011-07-01

    The ITER neutral beam system requires a negative hydrogen ion beam of 48 A with an energy of 0.87 MeV, and a negative deuterium beam of 40 A with an energy of 1 MeV. The beam is extracted from a large ion source of dimension 1.9 × 0.9 m2 by an acceleration system consisting of seven grids with 1280 apertures each. Currently, apertures with a diameter of 14 mm in the first grid are foreseen. In 2007, the IPP RF source was chosen as the ITER reference source due to its reduced maintenance compared with arc-driven sources and the successful development at the BATMAN test facility of being equipped with the small IPP prototype RF source ( {\\sim}\\frac{1}{8} of the area of the ITER NBI source). These results, however, were obtained with an extraction system with 8 mm diameter apertures. This paper reports on the comparison of the source performance at BATMAN of an ITER-relevant extraction system equipped with chamfered apertures with a 14 mm diameter and 8 mm diameter aperture extraction system. The most important result is that there is almost no difference in the achieved current density—being consistent with ion trajectory calculations—and the amount of co-extracted electrons. Furthermore, some aspects of the beam optics of both extraction systems are discussed.

  15. Performance of multi-aperture grid extraction systems for an ITER-relevant RF-driven negative hydrogen ion source

    International Nuclear Information System (INIS)

    Franzen, P.; Gutser, R.; Fantz, U.; Kraus, W.; Falter, H.; Froeschle, M.; Heinemann, B.; McNeely, P.; Nocentini, R.; Riedl, R.; Staebler, A.; Wuenderlich, D.

    2011-01-01

    The ITER neutral beam system requires a negative hydrogen ion beam of 48 A with an energy of 0.87 MeV, and a negative deuterium beam of 40 A with an energy of 1 MeV. The beam is extracted from a large ion source of dimension 1.9 x 0.9 m 2 by an acceleration system consisting of seven grids with 1280 apertures each. Currently, apertures with a diameter of 14 mm in the first grid are foreseen. In 2007, the IPP RF source was chosen as the ITER reference source due to its reduced maintenance compared with arc-driven sources and the successful development at the BATMAN test facility of being equipped with the small IPP prototype RF source ( ∼ 1/8 of the area of the ITER NBI source). These results, however, were obtained with an extraction system with 8 mm diameter apertures. This paper reports on the comparison of the source performance at BATMAN of an ITER-relevant extraction system equipped with chamfered apertures with a 14 mm diameter and 8 mm diameter aperture extraction system. The most important result is that there is almost no difference in the achieved current density-being consistent with ion trajectory calculations-and the amount of co-extracted electrons. Furthermore, some aspects of the beam optics of both extraction systems are discussed.

  16. [Biotechnological functional systems].

    Science.gov (United States)

    Bokser, O Ia

    1999-01-01

    Based on the theory of functional systems and a concept of the quantum system of behavior, studies of the quantumsystems were conducted. Their structure, the interaction of biological and technical sections were analyzed. Mathematical, biophysical, and experimental models were designed. The paper shows that biotechnical quantumsystems are involved in the formation of biological feedback. A system with imperative feedback from the programmed and introduced current results of efforts has been developed and put into practice for the self-regulation of muscle tension. Training by using this biological feedback system causes a stable increase in the perception rate of proprioceptive stimulus in examinees (operates, sportsmen, neurological patients).

  17. The pre-compression system of the toroidal field coils in ITER

    International Nuclear Information System (INIS)

    Knaster, J.; Ferrari, M.; Jong, C.; Vollmann, T.

    2007-01-01

    The toroidal field (TF) coils of ITER will undergo out-of-plane forces caused by the poloidal fields required to confine the plasma. These forces will be supported against overturning moments by links between the coils. In turn, these links consist of the inner intercoil structure (IIC), which is composed of two pairs (placed at the top and bottom part of the inboard leg) of four sets of poloidal shear keys inserted in slots between adjacent coils, and the outer intercoil structure (OIS) formed by four bands of shear panels on the outboard leg. The magnetic forces during energization of ITER would cause at IIC locations at toroidal gap between adjacent TF coils of 0.35 mm; during plasma operation this value could reach >1 mm causing a loosening of the keys and intensifying stress concentrations. This undesired effect will be suppressed by the application of a centripetal force of 70 MN/coil (35 MN at both the bottom and top part of the inboard leg of each of the 18 TF coils) that will be provided by two sets of three glass fibre/epoxy composite rings submitted to a toroidal hoop force of 100 MN/set. The calculated maximum stress in the rings will occur during the installation phase at room temperature, where the maximum radial elongation (∼25 mm) is required, and it reaches 1/5 of the composite presently estimated ultimate stress. The imposed elongation to reach that force and the lower Young's modulus of the composite compared with that of stainless steel will ease component tolerances and/or settlement effects in the final assembly. The paper describes the evolution in the design of the pre-compression system, from the conceptual phase when two circular cross-sections rings were considered to the present definitive one with three rectangular cross-section rings

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

    Energy Technology Data Exchange (ETDEWEB)

    Langeslag, S.A.E., E-mail: stefanie.langeslag@cern.ch [CERN, CH-1211 Genève 23 (Switzerland); Sgobba, S. [CERN, CH-1211 Genève 23 (Switzerland); Libeyre, P. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul lez Durance Cedex (France); Marcinek, D.J. [Cracow University of Technology, Warszawska 24, 30-962 Kraków (Poland); Zhang, Z. [CERN, CH-1211 Genève 23 (Switzerland); EPFL, CH-1015 Lausanne (Switzerland)

    2015-10-15

    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 procurement of these 15 m long tie plates. A welded solution originates from individual head- and slab-forgings, welded together by Gas Metal Arc Welding (GMAW). In parallel, a single piece forged solution is proven feasible, impressively forged in one piece by applying successive open die forging steps, followed by final machining. Maximum internal stress is experienced during cool-down to 4 K as a result of a large difference in thermal contraction between the support system and the coils. Furthermore, the varying magnetic fields in the independently driven coils introduce cyclic loading. Therefore, assessment of the two manufacturing solutions, in terms of both static and dynamic mechanical behaviour, is performed at ambient as well as cryogenic temperature. An extensive characterisation including microstructural and mechanical examination is conducted, evaluating the comparative performance of both solutions, reporting, amongst others, yield strength reaching the

  19. Nuclear systems and testing programs for ITER. Progress report for FY 1998

    International Nuclear Information System (INIS)

    1998-01-01

    The effort during this performance period focused on a number of TBWG activities (including test module design and analysis) that were identified and agreed upon (in the presence of the ITER Director and Deputy Director) at TBWG-4. These include: (a) DEMO test module design and performance analysis under pulsed operation; (b) Test program operation plan; (c) Test port design and analysis; (d) Decay heat calculations and safety analysis; (e) Further discussion among the parties to define collaboratory on R and D for the test program as well as possible collaboration on the construction and operation of test articles; (f) Remote handling and ancillary equipment; (g) Criteria for qualifying a blanket module or submodule for actual insertion and testing in ITER; (h) Definition of test module instrumentation and verification of capability to perform in the ITER fusion environment (magnetic field, radiation, heating, etc.); and (i) Analysis to show that the results to be obtained from the test modules as designed can be extrapolated to DEMO and reactor conditions (e.g., higher wall loads and the need to demonstrate tritium self-sufficiency). The main achievements during this performance period include: (1) updating and finalizing the US DDDs for the ITER Blanket Program to form part of the ITER Final Design Report (FDR). Specific revisions were in response to the minimal lithium volume test blanket design requirements and safety impact and (2) evaluating the feasibility of the US test program, including instrumentation and the benefits of the ITER test program. Details of this assessment, including solid breeder and liquid breeder blanket test plans, are documented in UCLA-IFNT-13 (attached). In addition, dose mapping calculations were performed for the ITER Building, including equipment and layout of coolant pipes/heat exchangers. A report on ITER Building dose calculations was sent to UD ITER management and to the Garching Task Coordinator in April, 1998. The report

  20. Safety-factor profile tailoring by improved electron cyclotron system for sawtooth control and reverse shear scenarios in ITER

    International Nuclear Information System (INIS)

    Zucca, C.; Sauter, O.; Fable, E.; Henderson, M. A.; Polevoi, A.; Farina, D.; Ramponi, G.; Saibene, G.; Zohm, H.

    2008-01-01

    The effect of the predicted local electron cyclotron current driven by the optimized electron cyclotron system on ITER is discussed. A design variant was recently proposed to enlarge the physics program covered by the upper and equatorial launchers. By extending the functionality range of the upper launcher, significant control capabilities of the sawtooth period can be obtained. The upper launcher improvement still allows enough margin to exceed the requirements for neoclassical tearing mode stabilization, for which it was originally designed. The analysis of the sawtooth control is carried on with the ASTRA transport code, coupled with the threshold model by Por-celli, to study the control capabilities of the improved upper launcher on the sawtooth instability. The simulations take into account the significant stabilizing effect of the fusion alpha particles. The sawtooth period can be increased by a factor of 1.5 with co-ECCD outside the q = 1 surface, and decreased by at least 30% with co-ECCD inside q = 1. The present ITER base-line design has the electron cyclotron launchers providing only co-ECCD. The variant for the equatorial launcher proposes the possibility to drive counter-ECCD with 1 of the 3 rows of mirrors: the counter-ECCD can then be balanced with co-ECCD and provide pure ECH with no net driven current. The difference between full co-ECCD off-axis using all 20MW from the equatorial launcher and 20MW co-ECCD driven by 2/3 from the equatorial launcher and 1/3 from the upper launcher is shown to be negligible. Cnt-ECCD also offers greater control of the plasma current density, therefore this analysis addresses the performance of the equatorial launcher to control the central q profile. The equatorial launcher is shown to control very efficiently the value of q 0.2 -q min in advanced scenarios, if one row provides counter-ECCD.

  1. 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...

  2. Matlab modeling of ITER CODAC

    International Nuclear Information System (INIS)

    Pangione, L.; Lister, J.B.

    2008-01-01

    The ITER CODAC (COntrol, Data Access and Communication) conceptual design resulted from 2 years of activity. One result was a proposed functional partitioning of CODAC into different CODAC Systems, each of them partitioned into other CODAC Systems. Considering the large size of this project, simple use of human language assisted by figures would certainly be ineffective in creating an unambiguous description of all interactions and all relations between these Systems. Moreover, the underlying design is resident in the mind of the designers, who must consider all possible situations that could happen to each system. There is therefore a need to model the whole of CODAC with a clear and preferably graphical method, which allows the designers to verify the correctness and the consistency of their project. The aim of this paper is to describe the work started on ITER CODAC modeling using Matlab/Simulink. The main feature of this tool is the possibility of having a simple, graphical, intuitive representation of a complex system and ultimately to run a numerical simulation of it. Using Matlab/Simulink, each CODAC System was represented in a graphical and intuitive form with its relations and interactions through the definition of a small number of simple rules. In a Simulink diagram, each system was represented as a 'black box', both containing, and connected to, a number of other systems. In this way it is possible to move vertically between systems on different levels, to show the relation of membership, or horizontally to analyse the information exchange between systems at the same level. This process can be iterated, starting from a global diagram, in which only CODAC appears with the Plant Systems and the external sites, and going deeper down to the mathematical model of each CODAC system. The Matlab/Simulink features for simulating the whole top diagram encourage us to develop the idea of completing the functionalities of all systems in order to finally have a full

  3. Fusion Power Measurement Using a Combined Neutron Spectrometer-Camera System at ITER

    International Nuclear Information System (INIS)

    Sjoestrand, Henrik; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Giacomelli, L.; Hellesen, C.; Hjalmarsson, A.; Ronchi, E.; Weiszflog, M.; Kaellne, J.

    2008-01-01

    A central task for fusion plasma diagnostics is to measure the 2.5 and 14 MeV neutron emission rate in order to determine the fusion power. A new method for determining the neutron yield has been developed at JET. It makes use of the magnetic proton recoil neutron spectrometer and a neutron camera and provides the neutron yield with small systematic errors. At ITER a similar system could operate if a high-resolution, high-performance neutron spectrometer similar to the MPR was installed. In this paper, we present how such system could be implemented and how well it would perform under different assumption of plasma scenarios and diagnostic capabilities. It is found that the systematic uncertainty for using such a system as an absolute calibration reference is as low as 3% and hence it would be an excellent candidate for the calibration of neutron monitors such as fission chambers. It is also shown that the system could provide a 1 ms time resolved estimation of the neutron rate with a total uncertainty of 5%

  4. GoldenBraid: An Iterative Cloning System for Standardized Assembly of Reusable Genetic Modules

    Science.gov (United States)

    Sarrion-Perdigones, Alejandro; Falconi, Erica Elvira; Zandalinas, Sara I.; Juárez, Paloma; Fernández-del-Carmen, Asun; Granell, Antonio; Orzaez, Diego

    2011-01-01

    Synthetic Biology requires efficient and versatile DNA assembly systems to facilitate the building of new genetic modules/pathways from basic DNA parts in a standardized way. Here we present GoldenBraid (GB), a standardized assembly system based on type IIS restriction enzymes that allows the indefinite growth of reusable gene modules made of standardized DNA pieces. The GB system consists of a set of four destination plasmids (pDGBs) designed to incorporate multipartite assemblies made of standard DNA parts and to combine them binarily to build increasingly complex multigene constructs. The relative position of type IIS restriction sites inside pDGB vectors introduces a double loop (“braid”) topology in the cloning strategy that allows the indefinite growth of composite parts through the succession of iterative assembling steps, while the overall simplicity of the system is maintained. We propose the use of GoldenBraid as an assembly standard for Plant Synthetic Biology. For this purpose we have GB-adapted a set of binary plasmids for A. tumefaciens-mediated plant transformation. Fast GB-engineering of several multigene T-DNAs, including two alternative modules made of five reusable devices each, and comprising a total of 19 basic parts are also described. PMID:21750718

  5. Physical performance analysis and progress of the development of the negative ion RF source for the ITER NBI system

    International Nuclear Information System (INIS)

    Fantz, U.; Franzen, P.; Kraus, W.; Berger, M.; Christ-Koch, S.; Falter, H.; Froeschle, M.; Gutser, R.; Heinemann, B.; Martens, C.; McNeely, P.; Riedl, R.; Speth, E.; Staebler, A.; Wuenderlich, D.

    2009-01-01

    For heating and current drive the neutral beam injection (NBI) system for ITER requires a 1 MeV deuterium beam for up to 1 h pulse length. In order to inject the required 17 MW the large area source (1.9 m x 0.9 m) has to deliver 40 A of negative ion current at the specified source pressure of 0.3 Pa. In 2007, the IPP RF driven negative hydrogen ion source was chosen by the ITER board as the new reference source for the ITER NBI system due to, in principle, its maintenance free operation and the progress in the RF source development. The performance analysis of the IPP RF sources is strongly supported by an extensive diagnostic program and modelling of the source and beam extraction. The control of the plasma chemistry and the processes in the plasma region near the extraction system are the most critical topics for source optimization both for long pulse operation as well as for the source homogeneity. The long pulse stability has been demonstrated at the test facility MANITU which is now operating routinely at stable pulses of up to 10 min with parameters near the ITER requirements. A quite uniform plasma illumination of a large area source (0.8 m x 0.8 m) has been demonstrated at the ion source test facility RADI. The new test facility ELISE presently planned at IPP is being designed for long pulse plasma operation and short pulse, but large-scale extraction from a half-size ITER source which is an important intermediate step towards ITER NBI.

  6. 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'

  7. Modeling and analysis of alternative concept of ITER vacuum vessel primary heat transfer system

    International Nuclear Information System (INIS)

    Carbajo, Juan; Yoder, Graydon; Dell'Orco, G.; Curd, Warren; Kim, Seokho

    2010-01-01

    A RELAP5-3D model of the ITER (Latin for 'the way') vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

  8. Development of radiation hard components for ITER blanket remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Makiko, E-mail: saito.makiko@jaea.go.jp; Anzai, Katsunori; Maruyama, Takahito; Noguchi, Yuto; Ueno, Kenichi; Takeda, Nobukazu; Kakudate, Satoshi

    2016-11-01

    Highlights: • Clarify the components that will degrade by gamma ray irradiation. • Perform the irradiation tests to BRHS components. • Optimize the materials to increase the radiation hardness. - Abstract: The ITER blanket remote handling system (BRHS) will be operated in a high radiation environment (250 Gy/h max.) and must stably handle the blanket modules, which weigh 4.5 t and are more than 1.5 m in length, with a high degree of position and posture accuracy. The reliability of the system can be improved by reviewing the failure events of the system caused by high radiation. A failure mode and effects analysis (FMEA) identified failure modes and determined that lubricants, O-rings, and electric insulation cables were the dominant components affecting radiation hardness. Accordingly, we tried to optimize the lubricants and cables of the AC servo motors by using polyphenyl ether (PPE)-based grease and polyether ether ketone (PEEK), respectively. Materials containing radiation protective agents were also selected for the cable sheaths and O-rings to improve radiation hardness. Gamma ray irradiation tests were performed on these components and as a result, a radiation hardness of 8 MGy was achieved for the AC servo motors. On the other hand, to develop the radiation hardness and BRHS compatibility furthermore, the improvement of materials of cable and O ring were performed.

  9. Engineering analysis of ITER In-Vessel Viewing System guide tube

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@iter.org [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Bates, Philip [Fusion for Energy, Barcelona (Spain); Bede, Ottó [Oxford Technologies Ltd., Abingdon (United Kingdom); Damiani, Carlo; Dubus, Gregory [Fusion for Energy, Barcelona (Spain); Omran, Hassan [Oxford Technologies Ltd., Abingdon (United Kingdom); Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Puiu, Adrian [Fusion for Energy, Barcelona (Spain); Reichle, Roger; Suárez, Alejandro; Walker, Christopher; Walsh, Michael [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Conceptual design of IVVS Loads action on IVVS Dominant loads. • Seismic and baking conditions. • No active cooling needed for IVVS. • IVVS requires additional support points to avoid excessive deformation. - Abstract: The In Vessel Viewing System (IVVS) will be one of the essential machine diagnostic systems at ITER to provide information about the status of in-vessel and plasma facing components and to evaluate the dust inside the Vacuum Vessel. The current design consists of six scanning probes and their deployment systems, which are placed in dedicated ports at the divertor level. These units are located in resident guiding tubes 10 m long, which allow the IVVS probes to go from their storage location to the scanning position by means of a simple straight translation. Moreover, each resident tube is supported inside the corresponding Vacuum Vessel and Cryostat port extensions, which are part of the primary confinement barrier. As the Vacuum Vessel and the Cryostat will move with respect to each other during operation (especially during baking) and during incidents and accidents (disruptions, vertical displacement events, seismic events), the structural integrity of the resident tube and the surrounding vacuum boundaries would be compromised if the required flexibility and supports are not appropriately assured. This paper focuses on the integration of the present design of the IVVS into the Vacuum Vessel and Cryostat environment. It presents the adopted strategy to withstand all the main interfacing loads without damaging the confinement barriers and the corresponding analysis supporting it.

  10. Efficacy of variational iteration method for chaotic Genesio system - Classical and multistage approach

    International Nuclear Information System (INIS)

    Goh, S.M.; Noorani, M.S.M.; Hashim, I.

    2009-01-01

    This is a case study of solving the Genesio system by using the classical variational iteration method (VIM) and a newly modified version called the multistage VIM (MVIM). VIM is an analytical technique that grants us a continuous representation of the approximate solution, which allows better information of the solution over the time interval. Unlike its counterpart, numerical techniques, such as the Runge-Kutta method, provide solutions only at two ends of the time interval (with condition that the selected time interval is adequately small for convergence). Furthermore, it offers approximate solutions in a discretized form, making it complicated in achieving a continuous representation. The explicit solutions through VIM and MVIM are compared with the numerical analysis of the fourth-order Runge-Kutta method (RK4). VIM had been successfully applied to linear and nonlinear systems of non-chaotic in nature and this had been testified by numerous scientists lately. Our intention is to determine whether VIM is also a feasible method in solving a chaotic system like Genesio. At the same time, MVIM will be applied to gauge its accuracy compared to VIM and RK4. Since, for most situations, the validity domain of the solutions is often an issue, we will consider a reasonably large time frame in our work.

  11. A Study of an Iterative Channel Estimation Scheme of FS-FBMC System

    Directory of Open Access Journals (Sweden)

    YongJu Won

    2017-01-01

    Full Text Available A filter bank multicarrier on offset-quadrature amplitude modulation (FBMC/OQAM system is an alternative multicarrier modulation scheme that does not need cyclic prefix (CP even in the presence of a multipath fading channel by the properties of prototype filter. The FBMC/OQAM system can be implemented either by using the poly-phase network with fast fourier transform (PPN-FFT or by using the extended FFT on a frequency-spreading (FS domain. In this paper, we propose an iterative channel estimation scheme for each sub channel of a FBMC/OQAM system over a frequency-spreading domain. The proposed scheme first estimates the channel using the received pilot signal in the subchannel domain and interpolates the estimated channel to fine frequency-spreading domain. Then the channel compensated FS domain pilot is despread again to modify the channel state information (CSI estimation. Computer simulation shows that the proposed method outperforms the conventional FBMC/OQAM channel estimator in a frequency selective channel.

  12. Iterative methods for the solution of very large complex symmetric linear systems of equations in electrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Clemens, M.; Weiland, T. [Technische Hochschule Darmstadt (Germany)

    1996-12-31

    In the field of computational electrodynamics the discretization of Maxwell`s equations using the Finite Integration Theory (FIT) yields very large, sparse, complex symmetric linear systems of equations. For this class of complex non-Hermitian systems a number of conjugate gradient-type algorithms is considered. The complex version of the biconjugate gradient (BiCG) method by Jacobs can be extended to a whole class of methods for complex-symmetric algorithms SCBiCG(T, n), which only require one matrix vector multiplication per iteration step. In this class the well-known conjugate orthogonal conjugate gradient (COCG) method for complex-symmetric systems corresponds to the case n = 0. The case n = 1 yields the BiCGCR method which corresponds to the conjugate residual algorithm for the real-valued case. These methods in combination with a minimal residual smoothing process are applied separately to practical 3D electro-quasistatical and eddy-current problems in electrodynamics. The practical performance of the SCBiCG methods is compared with other methods such as QMR and TFQMR.

  13. Issues in developing parallel iterative algorithms for solving partial differential equations on a (transputer-based) distributed parallel computing system

    International Nuclear Information System (INIS)

    Rajagopalan, S.; Jethra, A.; Khare, A.N.; Ghodgaonkar, M.D.; Srivenkateshan, R.; Menon, S.V.G.

    1990-01-01

    Issues relating to implementing iterative procedures, for numerical solution of elliptic partial differential equations, on a distributed parallel computing system are discussed. Preliminary investigations show that a speed-up of about 3.85 is achievable on a four transputer pipeline network. (author). 2 figs., 3 a ppendixes., 7 refs

  14. Discrete-Time Local Value Iteration Adaptive Dynamic Programming: Admissibility and Termination Analysis.

    Science.gov (United States)

    Wei, Qinglai; Liu, Derong; Lin, Qiao

    In this paper, a novel local value iteration adaptive dynamic programming (ADP) algorithm is developed to solve infinite horizon optimal control problems for discrete-time nonlinear systems. The focuses of this paper are to study admissibility properties and the termination criteria of discrete-time local value iteration ADP algorithms. In the discrete-time local value iteration ADP algorithm, the iterative value functions and the iterative control laws are both updated in a given subset of the state space in each iteration, instead of the whole state space. For the first time, admissibility properties of iterative control laws are analyzed for the local value iteration ADP algorithm. New termination criteria are established, which terminate the iterative local ADP algorithm with an admissible approximate optimal control law. Finally, simulation results are given to illustrate the performance of the developed algorithm.In this paper, a novel local value iteration adaptive dynamic programming (ADP) algorithm is developed to solve infinite horizon optimal control problems for discrete-time nonlinear systems. The focuses of this paper are to study admissibility properties and the termination criteria of discrete-time local value iteration ADP algorithms. In the discrete-time local value iteration ADP algorithm, the iterative value functions and the iterative control laws are both updated in a given subset of the state space in each iteration, instead of the whole state space. For the first time, admissibility properties of iterative control laws are analyzed for the local value iteration ADP algorithm. New termination criteria are established, which terminate the iterative local ADP algorithm with an admissible approximate optimal control law. Finally, simulation results are given to illustrate the performance of the developed algorithm.

  15. An ITER-relevant evacuated waveguide transmission system for the JET-EP ECRH project

    International Nuclear Information System (INIS)

    Henderson, M.A.; Alberti, S.; Bird, J.

    2003-01-01

    An over-moded evacuated waveguide line was chosen for use in the transmission system for the proposed JET-enhanced performance project (JET-EP) electron cyclotron resonance heating (ECRH) system. A comparison between the quasi-optical, atmospheric waveguide and evacuated waveguide systems was performed for the project with a strong emphasis placed on the technical and financial aspects. The evacuated waveguide line was chosen as the optimal system in light of the above criteria. The system includes six lines of 63.5mm wave guide for transmitting 6.0 MW(10 s) at 113.3 GHz from the gyrotrons to the launching antenna. The designed lines are on average 72m in length and consist of nine mitre bends, for an estimated transmission efficiency of ∼90%. Each line is designed to include an evacuated switch leading to a calorimetric load, two dc breaks, two gate valves, one pump out tee, a power monitor mitre bend and a double-disc CVD window near the torus. The location of waveguide support is positioned to minimize the power converted to higher-order modes from waveguide sagging and misalignment. The two gate valves and CVD window are designed to be used as tritium barriers at the torus and between the J1T and J1D buildings. The last leg of the waveguide leading to the torus has to be designed to accommodate the torus movement during disruptions and thermal cycles. All lines are also designed to be compatible with the ITER ECRH system operating at 170 GHz. (author)

  16. Experimental proof of a load resilient external matching solution for the ITER ICRH system

    International Nuclear Information System (INIS)

    Vervier, M.; Messiaen, A.; Dumortier, P.; Lamalle, P.

    2005-01-01

    A reliable load resilient external matching scheme for the ITER ICRH system has been successfully tested on the mock-up of the external matching system with variable plasma load simulation. To avoid the deleterious mutual coupling effects the power has been passively distributed among the upper half and the bottom half of the 24 radiating straps of the antenna plug. In this plug the straps are grouped in 8 triplets by 4-ports junctions. The 4 top and 4 bottom triplets are respectively put in parallel outside the antenna plug near a voltage anti-node by means of T junctions. The load resilient matching is then obtained by a 4 parameters single 'conjugate T' (CT) configuration. For an antenna loading variation of about 1 to 8 Ω/m the VSWR at the power source remains below 1.3. The maximum voltage along the line remains equal to the one in the antenna plug and there is a fair power share between the straps. A π0π0 toroidal phasing is easily obtained. The poloidal phasing between the top and bottom triplets is determined by the loading. A straightforward matching procedure is described. Good load resilience is also obtained by replacing the CT by one hybrid

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  18. Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

    International Nuclear Information System (INIS)

    Turner, Andrew; Pampin, Raul; Loughlin, M.J.; Ghani, Zamir; Hurst, Gemma; Lo Bue, Alessandro; Mangham, Samuel; Puiu, Adrian; Zheng, Shanliang

    2014-01-01

    The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge

  19. Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

    Energy Technology Data Exchange (ETDEWEB)

    Turner, Andrew, E-mail: andrew.turner@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Loughlin, M.J. [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghani, Zamir; Hurst, Gemma [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lo Bue, Alessandro [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mangham, Samuel [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Puiu, Adrian [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Zheng, Shanliang [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2014-10-15

    The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.

  20. Study of ITER equatorial port plug handling system and vacuum sealing interface

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Jean-Pierre [Association Euratom CEA, CEA/DSM/IRFM, Cadarache, F-13108 Saint-Paul-lez-Durance (France)], E-mail: jean-pierre.martins@cea.fr; Doceul, Louis; Marol, Sebastien; Delchie, Elise [Association Euratom CEA, CEA/DSM/IRFM, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Cordier, Jean-Jacques; Levesy, Bruno; Tesini, Alessandro [ITER International Organization, F-13108 Saint-Paul-lez-Durance cedex (France); Ciattaglia, Emanuela [EFDA CSU - Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Tivey, Richard [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Gillier, Rene; Abbes, Christophe [GARLOCK - Sealing Technologies - 90, rue de la roche du Geai, F-42029 St-Etienne cedex 1 (France)

    2009-06-15

    In the field of the ITER port plug engineering and integration task, CEA has contributed to define proposals concerning the port plugs vacuum sealing interface with the vessel flange and the equatorial plug handling. The 2001 baseline vacuum flange sealing consisted of TIG welding of a 316L strip plate on to U shapes. This arrangement presented some issues like welding access, implementation of tools, lip consumption, complex local leak test, continuous leak checking. Therefore, an alternate sealing solution based on the use of metallic gaskets is proposed. The different technical aspects are discussed to explain how this design can simplify the maintenance and deal with safety and vacuum requirements. The design of the mechanical attachment and vacuum sealing of the plug has constantly evolved, but the associated remote handling equipment was not systematically reviewed. An update of the cask and maintenance procedure was studied in order to design it in accordance with the last generic plug flange design. This includes a concept of a gripping system that uses the plug flange bolting area and, to help the remote handling process, a cantilever assisting system is suggested to increase the reliability of the transfer operation between vacuum vessel and cask.

  1. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  2. Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lasnier, C. J., E-mail: lasnier@LLNL.gov; Allen, S. L.; Ellis, R. E.; Fenstermacher, M. E.; McLean, A. G.; Meyer, W. H.; Morris, K.; Seppala, L. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551-0808 (United States); Crabtree, K. [College of Optics, University of Arizona, Tucson, Arizona 85721 (United States); Van Zeeland, M. A. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2014-11-15

    An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

  3. Lower hybrid heating and current drive in Iter operation scenarios and outline system design

    International Nuclear Information System (INIS)

    1994-11-01

    Lower Hybrid Waves (LHW) are considered a valid method of plasma heating and the best demonstrated current drive method. Current drive by LHW possesses the unique feature, as compared to the other methods, to retain a good current drive efficiency in plasma regions of low to medium temperature, or in low-β phases of the discharges. This makes them an essential element to realize the so called 'advanced steady-state Tokamak scenarios' in which a hollow current density profile (deep shear reversal) - established during the ramp-up of the plasma current - offers the prospects of improved confinement and an MHD-stable route to continuous burn. This report contains both modelling and design studies of an LHW system for ITER. It aims primarily at the definition of concepts and parameters for steady-state operation using LHW combined with Fast Waves (FW), or other methods of generating a central seed current for high bootstrap current operation. However simulations addressing the use of LHW for current profile control in the high current pulsed operation scenario are also presented. The outline design of a LHW system which covers the needs for both pulsed and steady-state operation is described in detail. (author). 28 refs., 49 figs

  4. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  5. Iterative MIMO Turbo Multiuser Detection and Equalization for STTrC-Coded Systems with Unknown Interference

    Directory of Open Access Journals (Sweden)

    Veselinovic Nenad

    2004-01-01

    Full Text Available Iterative multiuser detection in a single-carrier broadband multiple-input multiple-output (MIMO system is studied in this paper. A minimum mean squared error (MMSE low-complexity multiuser receiver is derived for space-division multiple-access (SDMA space-time trellis-coded (STTrC systems in frequency-selective fading channels. The receiver uses MMSE filtering to jointly detect several transmit antennas of the user of interest, while the interference from the undetected transmit antennas, cochannel interference (CCI, and intersymbol interference (ISI are all cancelled by the soft cancellation. The performances of two extreme receiver cases are evaluated. In the first case, only one transmit antenna of the user of interest is detected at a time and the remaining ones are cancelled by soft cancellation. In the second case, all the transmit antennas are detected jointly. The comparison of the two cases shows improvement with the latter one, both in single-user and multiuser communications and in the presence of unknown cochannel interference (UCCI. It is further shown that in the multiuser case, the proposed receivers approach the corresponding single-user bounds. The number of receive antenna elements required to achieve single-user bound is thereby equal to the number of users and not to the total number of transmit antennas.

  6. Analysis of active and passive magnetic field reduction systems (MFRS) of the ITER NBI

    International Nuclear Information System (INIS)

    Roccella, M.; Lucca, F.; Roccella, R.; Pizzuto, A.; Ramogida, G.; Portone, A.; Tanga, A.; Formisano, A.; Martone, R.

    2007-01-01

    In ITER two heating (HNBI) and one diagnostic neutral beam injectors (DNBI) are foreseen. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some G along the ion path and below 20 G in the neutralizing regions). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils (PFC), both passive and active shielding systems have been foreseen. The present design of the magnetic field reduction systems (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates. The electromagnetic analyses of the effectiveness of these shields have been performed by a 3D FEM model using ANSYS code for the HNBI. The ANSYS models of the ferromagnetic box and of the active coils are fully parametric, thus any size change of the ferromagnetic box and coils (linear dimension or thickness) preserving the overall box shape could be easily reproduced by simply changing some parameter in the model

  7. Analysis of active and passive magnetic field reduction systems (MFRS) of the ITER NBI

    Energy Technology Data Exchange (ETDEWEB)