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

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.

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.

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)

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.

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.

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.

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)

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)

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.

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

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)

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

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

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.

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

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

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

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

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...

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.

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.

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)

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)

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...

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)

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

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

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

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.

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

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.

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.

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.)

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.

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.

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.

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

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.

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.)

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.

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.

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

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)

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

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.

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

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

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

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.

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.

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.

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)

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.

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

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.

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.

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.

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

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

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.

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

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)

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.

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.

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

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

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.

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.

Plan of ITER remote experimentation center

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

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)

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.

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

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.

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.

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

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)

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)

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.

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

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.

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

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)

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

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)

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.

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.)

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

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

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.

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.

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.

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

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

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

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.

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.

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

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)

ITER Fast Ion Collective Thomson Scattering

DEFF Research Database (Denmark)

Bindslev, Henrik; Larsen, Axel Wright; Meo, Fernando

2005-01-01

The EFDA Contract 04-1213 with Risø National Laboratory concerning a detailed integrated design of a Fast Ion Collective Thomson Scattering (CTS) diagnostic for ITER was signed on 31 December 2004. In 2003 the Risø CTS group finished a feasibility study and a conceptual design of an ITER Fast Ion...... Collective Thomson Scattering System (Contract 01.654) [1, 2]. The purpose of the CTS diagnostic is to measure the distribution function of fast ions in the plasma. The feasibility study demonstrated that the only system that can fully meet the ITER measurement requirements for confined fusion alphas is a 60...... the blanket gap, and calculations of diagnosing fuel ion ratio and rotation velocity by CTS....

Local control unit for ITER-India gyrotron test facility (IIGTF)

Energy Technology Data Exchange (ETDEWEB)

Rathod, Vipal, E-mail: vipal.rathod@iter-india.org; Shah, Ronak; Mandge, Deepak; Parmar, Rajvi; Rao, S.L.

2016-11-15

Highlights: • A dedicated full scale ITER prototype Local Control Unit for ITER-India Gyrotron test facility. • National Instruments® make PXIe system for real time control & data acquisition and Siemens® PLC for sequence control function. • Hardwired FPGA based fast protection interlock system. • High speed analog fiber optical transmission link using V/F and F/V technique. • Software framework based on LabVIEW™ platform and ITER CODAC Core System. - Abstract: Electron Cyclotron system on ITER, is one of the important RF ancillary systems based on high power Gyrotron RF sources, that is used for plasma heating and current drive applications. To operate a Gyrotron source, various auxiliary systems and services such as Super Conducting Magnet set, High Voltage Power Supplies, Auxiliary Power Supplies, Waveguide components, Cooling water system and a Local Control Unit (LCU) are required. The LCU plays a very crucial role for the safe and reliable operation of Gyrotron system. A dedicated full scale ITER prototype LCU is being developed for testing and commissioning of an ITER like Test Gyrotron at ITER-India Gyrotron Test facility (IIGTF). The main functions of LCU include Sequence Control, Local Interlock Protection and Real Time Data Acquisition. PLC based slow controller is used for implementing the Sequence Control & Slow Interlock functions. Critical Protection Interlocks are required to have a response time of <10 μs and are implemented using custom built hardware and PXIe based fast controller. Also PXIe system is used for implementing Real Time Data Acquisition function that is required to have slow and fast acquisition with online visualization and off line analysis facility. A Signal Conditioning Unit (SCU) is used to interface and faithfully transmit the field signals to the remote control systems. Necessary controller hardware is procured and several pre-prototype developments have been taken up to establish the critical subsystems such as

Local control unit for ITER-India gyrotron test facility (IIGTF)

International Nuclear Information System (INIS)

Rathod, Vipal; Shah, Ronak; Mandge, Deepak; Parmar, Rajvi; Rao, S.L.

2016-01-01

Highlights: • A dedicated full scale ITER prototype Local Control Unit for ITER-India Gyrotron test facility. • National Instruments® make PXIe system for real time control & data acquisition and Siemens® PLC for sequence control function. • Hardwired FPGA based fast protection interlock system. • High speed analog fiber optical transmission link using V/F and F/V technique. • Software framework based on LabVIEW™ platform and ITER CODAC Core System. - Abstract: Electron Cyclotron system on ITER, is one of the important RF ancillary systems based on high power Gyrotron RF sources, that is used for plasma heating and current drive applications. To operate a Gyrotron source, various auxiliary systems and services such as Super Conducting Magnet set, High Voltage Power Supplies, Auxiliary Power Supplies, Waveguide components, Cooling water system and a Local Control Unit (LCU) are required. The LCU plays a very crucial role for the safe and reliable operation of Gyrotron system. A dedicated full scale ITER prototype LCU is being developed for testing and commissioning of an ITER like Test Gyrotron at ITER-India Gyrotron Test facility (IIGTF). The main functions of LCU include Sequence Control, Local Interlock Protection and Real Time Data Acquisition. PLC based slow controller is used for implementing the Sequence Control & Slow Interlock functions. Critical Protection Interlocks are required to have a response time of <10 μs and are implemented using custom built hardware and PXIe based fast controller. Also PXIe system is used for implementing Real Time Data Acquisition function that is required to have slow and fast acquisition with online visualization and off line analysis facility. A Signal Conditioning Unit (SCU) is used to interface and faithfully transmit the field signals to the remote control systems. Necessary controller hardware is procured and several pre-prototype developments have been taken up to establish the critical subsystems such as

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

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

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)

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)

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

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

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

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)

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.

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

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.

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.

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

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

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

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.

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)

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)

Preliminary RAMI analysis of DFLL TBS for ITER

Energy Technology Data Exchange (ETDEWEB)

Wang, Dagui [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230031 (China); Yuan, Run [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Jiaqun, E-mail: jiaqun.wang@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Fang; Wang, Jin [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2016-11-15

Highlights: • We performed the functional analysis of the DFLL TBS. • We performed a failure mode analysis of the DFLL TBS. • We estimated the reliability and availability of the DFLL TBS. • The ITER RAMI approach was applied to the DFLL TBS for technical risk control in the design phase. - Abstract: ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) 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.

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.

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.

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

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.

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.

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

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.

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

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

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

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

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.

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

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

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.

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.

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

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)

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)

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

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.

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)

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)

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.

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.

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

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.

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

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.

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.

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.)

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.

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

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)

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.

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

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.

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.

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

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.

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

ITER tokamak device

International Nuclear Information System (INIS)

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

1991-01-01

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

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)

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.

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

Iterative Schemes for Convex Minimization Problems with Constraints

Directory of Open Access Journals (Sweden)

Lu-Chuan Ceng

2014-01-01

Full Text Available We first introduce and analyze one implicit iterative algorithm for finding a solution of the minimization problem for a convex and continuously Fréchet differentiable functional, with constraints of several problems: the generalized mixed equilibrium problem, the system of generalized equilibrium problems, and finitely many variational inclusions 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 implicit iterative algorithm for finding a fixed point of infinitely many nonexpansive mappings with the same constraints, and derive its strong convergence under mild assumptions.

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.

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.

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.

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.

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.

Diagnostic integration solutions in the ITER first wall

International Nuclear Information System (INIS)

Martínez, Gonzalo; Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger; Shigin, Pavel; Sabourin, Flavien; Gicquel, Stefan; Mitteau, Raphael; González, Jorge

2015-01-01

Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

Diagnostic integration solutions in the ITER first wall

Energy Technology Data Exchange (ETDEWEB)

Martínez, Gonzalo, E-mail: gonzalo.martinez@iter.org [Technical University of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Martin, Alex; Watts, Christopher; Veshchev, Evgeny; Reichle, Roger [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Shigin, Pavel [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); National Research Nuclear University (MEPhI), Kashirskoe shosse, 115409 Moscow (Russian Federation); Sabourin, Flavien [ABMI-Groupe, Parc du Relais BatD 201 Route de SEDS, 13127 Vitrolles (France); Gicquel, Stefan; Mitteau, Raphael [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); González, Jorge [RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat (Spain)

2015-10-15

Highlights: • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets. • An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW. - Abstract: ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems.

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

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.

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.

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.

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.

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

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.

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.

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

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.

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.

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.

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

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

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

Progress in XRCS-Survey plant instrumentation and control design for ITER

International Nuclear Information System (INIS)

Varshney, Sanjeev; Jha, Shivakant; Simrock, Stefan; Barnsley, Robin; Martin, Vincent; Mishra, Sapna; Patil, Prabhakant; Patel, Shreyas; Kumar, Vinay

2016-01-01

Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

Progress in XRCS-Survey plant instrumentation and control design for ITER

Energy Technology Data Exchange (ETDEWEB)

Varshney, Sanjeev, E-mail: sanjeev.varshney@iter-india.org [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Jha, Shivakant [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Simrock, Stefan; Barnsley, Robin; Martin, Vincent [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Mishra, Sapna [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Patil, Prabhakant [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Patel, Shreyas; Kumar, Vinay [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India)

2016-11-15

Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

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.

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).

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

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

International Nuclear Information System (INIS)

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

2017-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-15

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

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.

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.

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.

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)

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.)

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.

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

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)

On the Convergence of Iterative Receiver Algorithms Utilizing Hard Decisions

Directory of Open Access Journals (Sweden)

Jürgen F. Rößler

2009-01-01

Full Text Available The convergence of receivers performing iterative hard decision interference cancellation (IHDIC is analyzed in a general framework for ASK, PSK, and QAM constellations. We first give an overview of IHDIC algorithms known from the literature applied to linear modulation and DS-CDMA-based transmission systems and show the relation to Hopfield neural network theory. It is proven analytically that IHDIC with serial update scheme always converges to a stable state in the estimated values in course of iterations and that IHDIC with parallel update scheme converges to cycles of length 2. Additionally, we visualize the convergence behavior with the aid of convergence charts. Doing so, we give insight into possible errors occurring in IHDIC which turn out to be caused by locked error situations. The derived results can directly be applied to those iterative soft decision interference cancellation (ISDIC receivers whose soft decision functions approach hard decision functions in course of the iterations.

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)

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)

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

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

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

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.

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

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

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

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.

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.

ITER primary cryopump test facility

International Nuclear Information System (INIS)

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

1998-01-01

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

ITER instrumentation and control-Status and plans

International Nuclear Information System (INIS)

Wallander, Anders; Abadie, Lana; Dave, Haresh; Di Maio, Franck; Gulati, Hitesh Kumar; Hansalia, Chandresh; Joonekindt, Didier; Journeaux, Jean-Yves; Klotz, Wolf-Dieter; Mahajan, Kirti; Makijarvi, Petri; Scibile, Luigi; Stepanov, Denis; Utzel, Nadine; Yonekawa, Izuru

2010-01-01

The ITER instrumentation and control (I and C) system is the term encompassing all hardware and software required to operate ITER. It has two levels of hierarchy: the central I and C systems and the plant systems I and C. The central I and C systems comprise CODAC (Control, Data Access and Communication), the central interlock system (CIS) and the central safety systems (CSS). The central I and C systems are 'in-fund', i.e. procured by ITER Organization (IO), while plant systems I and C are 'in-kind', i.e. procured by the seven ITER domestic agencies. This procurement model, together with the current estimate of 161 plant systems I and C, poses a major challenge for the realization and integration of the ITER I and C system. To address this challenge a main strategic focus of the CODAC group, formed in 2008, has been to establish good relations with the domestic agencies. By distributing the required R and D tasks and contracts fairly between the domestic agencies we build collaborations for the future at the same time as technical work proceed. The primary goal of ITER I and C system is to provide a fully integrated and automated control system for ITER. Standardization of plant systems I and C is of primary importance and has been the highest priority task during the last year. The target of associated R and D activities is to survey, benchmark and prototype main stream technologies, in order to choose the best and most widely used technology standards for plant systems I and C. In this paper we elaborate on our approach, both from a technical and a non-technical perspective, explain technology evaluation and decisions and finally present the way forward to ensure ITER I and C system will contribute and be instrumental in making ITER a success.

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

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)

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...

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

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)

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

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.

Integration of remote refurbishment performed on ITER components

Energy Technology Data Exchange (ETDEWEB)

Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Antola, L. [AMEC, 31 Parc du Golf, CS 90519, 13596 Aix en Provence (France); Beaudoin, V. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dremel, C. [Westinghouse, Electrique France/Astare, 122 Avenue de Hambourg, 13008 Marseille (France); Evrard, D. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Levesy, B.; Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2015-10-15

Highlights: • System engineering approach to consolidate requirements to modify the layout of the Hot Cell. • Illustration of the loop between requirement and design. • Verification process. - Abstract: Internal components of the ITER Tokamak are replaced and transferred to the Hot Cell by remote handling equipment. These components include port plugs, cryopumps, divertor cassettes, blanket modules, etc. They are brought to the refurbishment area of the ITER Hot Cell Building for cleaning and maintenance, using remote handling techniques. The ITER refurbishment area will be unique in the world, when considering combination of size, quantity of complex component to refurbish in presence of radiation, activated dust and tritium. The refurbishment process to integrate covers a number of workstations to perform specific remote operations fully covered by a mast on crane system. This paper describes the integration of the Refurbishment Area, explaining the functions, the methodology followed, some illustrations of trade-off and safety improvements.

ITER blanket designs

International Nuclear Information System (INIS)

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

1995-01-01

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

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

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

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.)

Three-dimensional tolerance investigation on main ITER components

International Nuclear Information System (INIS)

Reich, J.; Chiocchio, S.; Cordier, J.-J.; Gallix, R.; Guerin, O.; Halcrow, T.

2009-01-01

ITER has to focus on all processes that ensure the permanent consistency between the requirements of ITER and the performance attributes of its components. This includes integration tolerance studies. One of the main goals of this work is to establish a sufficient tolerance scheme for all main components. The investigation in sufficient tolerance studies at a very early stage of the project will result in cost savings during the installation process. Due to the complexity of the ITER components and their several interfaces to their surroundings, it is advantageous to perform tolerance studies with a specialised tool like 'three-dimensional control systems' (3DCS) that is compatible with the ITER CATIA-V5 CAD engineering system and Enovia PRC environment. On single components (e.g. Magnet TF Coils) detailed two-dimensional tolerance schemes have been developed from the beginning. Using them as a starting point, functional or key interface tolerances have to be defined. Furthermore the tolerance studies have to consider the different configurations of each component (e.g. manufacturing stages, assembly plan, integration with surrounding, operation conditions). Especially for assembly it is necessary to analyse the final ranges which have to be achieved during the installation process. From the integration point of view, the key tolerances of all main in-cryostat ITER components have to be brought together in a complete and consistent manner.

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.

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)

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.

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)

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.

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

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.

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

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

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

ITER containment design-assist analysis

International Nuclear Information System (INIS)

Nguyen, T.H.

1992-03-01

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

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.

Iterated interactions method. Realistic NN potential

International Nuclear Information System (INIS)

Gorbatov, A.M.; Skopich, V.L.; Kolganova, E.A.

1991-01-01

The method of iterated potential is tested in the case of realistic fermionic systems. As a base for comparison calculations of the 16 O system (using various versions of realistic NN potentials) by means of the angular potential-function method as well as operators of pairing correlation were used. The convergence of genealogical series is studied for the central Malfliet-Tjon potential. In addition the mathematical technique of microscopical calculations is improved: new equations for correlators in odd states are suggested and the technique of leading terms was applied for the first time to calculations of heavy p-shell nuclei in the basis of angular potential functions

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

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.

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)

Wet scrubber technology for tritium confinement at ITER

Energy Technology Data Exchange (ETDEWEB)

Perevezentsev, A.N., E-mail: alexander.perevezentsev@iter.org [ITER Organization, CS 90 046, 13067 St Paul lez Durance Cedex (France); Andreev, B.M.; Rozenkevich, M.B.; Pak, Yu.S.; Ovcharov, A.V.; Marunich, S.A. [Mendeleev University of Chemical Technology, 125047 Miusskaya Sq. 9, Moscow (Russian Federation)

2010-12-15

Operation of the ITER machine with tritium plasma requires tritium confinement systems to protect workers and the environment. Tritium confinement at ITER is based on multistage approach. The final stage provides tritium confinement in building sectors and consists of building's walls as physical barriers and control of sub-atmospheric pressure in those volumes as a dynamic barrier. The dynamic part of the confinement function shall be provided by safety important components that are available all the time when required. Detritiation of air prior to its release to the environment is based on catalytic conversion of tritium containing gaseous species to water vapour followed by their isotopic exchange with liquid water in scrubber column of packed bed type. Wet scrubber technology has been selected because of its advantages over conventional air detritiation technique based on gas drying by water adsorption. The most important design target of system availability was very difficult to meet with conventional water adsorption driers. This paper presents results of experimental trial for validation of wet scrubber technology application in the ITER tritium confinement system and process evaluation using developed simulation computer code.

Final design of ITER port plug test facility

Energy Technology Data Exchange (ETDEWEB)

Cerisier, Thierry, E-mail: thierry.cerisier@yahoo.fr [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Levesy, Bruno [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Romannikov, Alexander [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation); Rumyantsev, Yuri [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Cordier, Jean-Jacques; Dammann, Alexis [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Minakov, Victor; Rosales, Natalya; Mitrofanova, Elena [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Portone, Sergey; Mironova, Ekaterina [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation)

2016-11-01

Highlights: • We introduce the port plug test facility (purpose and status of the design). • We present the PPTF sub-systems. • We present the environmental and functional tests. • We present the occupational and nuclear safety functions. • We conclude on the achievements and next steps. - Abstract: To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs. The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented.

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Dynamical behaviour of neuronal networks iterated with memory

International Nuclear Information System (INIS)

Melatagia, P.M.; Ndoundam, R.; Tchuente, M.

2005-11-01

We study memory iteration where the updating consider a longer history of each site and the set of interaction matrices is palindromic. We analyze two different ways of updating the networks: parallel iteration with memory and sequential iteration with memory that we introduce in this paper. For parallel iteration, we define Lyapunov functional which permits us to characterize the periods behaviour and explicitly bounds the transient lengths of neural networks iterated with memory. For sequential iteration, we use an algebraic invariant to characterize the periods behaviour of the studied model of neural computation. (author)

The European contribution to the ITER Remote Maintenance

Energy Technology Data Exchange (ETDEWEB)

Damiani, C., E-mail: carlo.damiani@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Annino, C.; Balagué, S.; Bates, P.; Ceccanti, F.; Di Mascio, T.; Dubus, G.; Esqué, S.; Gonzalez, C.; Lewczanin, M.; Locke, D.; Mont, L.; Olajos, K.; Ranz, R.; Shuff, R.; Puiu, A.; Van Hille, C.; Van Uffelen, M. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Choi, C.H.; Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

2014-10-15

Highlights: •The article introduces the needs for remote maintenance in ITER. •It also discusses some of the issues related to the cultural transition from tokamaks as plasma physics to nuclear reactors. •It highlights the related cultural change and the implications on plant topology and maintenance. •Then, it presents those remote handling systems that will be procured by Europe. •The article emphasises the need of a major involvement of industries from now on. -- Abstract: For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source. The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R and D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages. Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations. After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems. Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.

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)

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.

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.

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...

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.

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

Existence test for asynchronous interval iterations

DEFF Research Database (Denmark)

Madsen, Kaj; Caprani, O.; Stauning, Ole

1997-01-01

In the search for regions that contain fixed points ofa real function of several variables, tests based on interval calculationscan be used to establish existence ornon-existence of fixed points in regions that are examined in the course ofthe search. The search can e.g. be performed...... as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... on thecomponentwise calculations done in the course of the iteration. These componentwisetests are useful for parallel implementation of the search, sincethe tests can then be performed local to each processor and only when a test issuccessful do a processor communicate this result to other processors....

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.

Status of ITER neutron diagnostic development

Science.gov (United States)

Krasilnikov, A. V.; Sasao, M.; Kaschuck, Yu. A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V. S.; Popovichev, S.; Iguchi, T.; Jarvis, O. N.; Källne, J.; Fiore, C. L.; Roquemore, A. L.; Heidbrink, W. W.; Fisher, R.; Gorini, G.; Prosvirin, D. V.; Tsutskikh, A. Yu.; Donné, A. J. H.; Costley, A. E.; Walker, C. I.

2005-12-01

Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented.

Status of ITER neutron diagnostic development

International Nuclear Information System (INIS)

Krasilnikov, A.V.; Sasao, M.; Kaschuck, Yu.A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V.S.; Popovichev, S.; Iguchi, T.; Jarvis, O.N.; Kaellne, J.; Fiore, C.L.; Roquemore, A.L.; Heidbrink, W.W.; Fisher, R.; Gorini, G.; Prosvirin, D.V.; Tsutskikh, A.Yu.; Donne, A.J.H.; Costley, A.E.; Walker, C.I.

2005-01-01

Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented

Status of ITER neutron diagnostic development

International Nuclear Information System (INIS)

Sasao, M.; Krasilnikov, A.V.; Kaschuck, Yu.A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V.S.; Popovichev, S.; Jarvis, O.N.; Iguchi, T.; Kaellne, J.; Fiore, C.L.; Roquemore, A.L.; Heidbrink, W.W.; Fisher, R.; Gorini, G.; Donne, A.J.H.; Costley, A.E.; Walker, C.I.

2005-01-01

Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be well measured by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include: radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors, neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The neutron flux monitors need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented. (author)

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.

U.S. Contributions to ITER

International Nuclear Information System (INIS)

Sauthoff, Ned R.

2005-01-01

The United States participates in the ITER project and program to enable the study of the science and technology of burning plasmas, a key programmatic element missing from the world fusion program. The 2003 U.S. decision to enter the ITER negotiations followed an extensive series of community and governmental reviews of the benefits, readiness, and approaches to the study of burning plasmas. This paper describes both the technical and the organizational preparations and plans for U.S. participation in the ITER construction activity: in-kind contributions, staff contributions, and cash contributions as well as supporting physics and technology research. Near-term technical activities focus on the completion of R and D and design and mitigation of risks in the areas of the central solenoid magnet, shield/blanket, diagnostics, ion cyclotron system, electron cyclotron system, pellet fueling system, vacuum system, tritium processing system, and conventional systems. Outside the project, the U .S. is engaged in preparations for the test blanket module program. Organizational activities focus on preparations of the project management arrangements to maximize the overall success of the ITER Project; elements include refinement of U.S. directions on the international arrangements, the establishment of the U.S. Domestic Agency, progress along the path of the U.S. Department of Energy's Project Management Order, and overall preparations for commencement of the fabrication of major items of equipment and for provision of staff and cash as specified in the upcoming ITER agreement

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)

A convergent iterative solution of the quantum double-well potential

International Nuclear Information System (INIS)

Friedberg, R.; Lee, T.D.; Zhao, W.Q.; Cimenser, A.

2001-01-01

We present a new convergent iterative solution for the two lowest quantum wave functions ψ ev and ψ od of the Hamiltonian with a quartic double-well potential V in one dimension. By starting from a trial function, which is by itself the exact lowest even or odd eigenstate of a different Hamiltonian with a modified potential V+δV, we construct the Green's function for the modified potential. The true wave functions, ψ ev or ψ od , then satisfy a linear inhomogeneous integral equation, in which the inhomogeneous term is the trial function, and the kernel is the product of the Green's function times the sum of δV, the potential difference, and the corresponding energy shift. By iterating this equation we obtain successive approximations to the true wave function; furthermore, the approximate energy shift is also adjusted at each iteration so that the approximate wave function is well behaved everywhere. We are able to prove that this iterative procedure converges for both the energy and the wave function at all x. The effectiveness of this iterative process clearly depends on how good the trial function is, or equivalently, how small the potential difference δV is. Although each iteration brings a correction smaller than the previous one by a factor proportional to the parameter that characterizes the smallness of δV, it is not a power series expansion in the parameter. The exact tunneling information of the modified potential is, of course, contained in the Green's function; by adjusting the kernel of the integral equation via the energy shift at each iteration, we bring enough of this information into the calculation so that each approximate wave function is exponentially tuned. This is the underlying reason why the present method converges, while the usual power series expansion does not

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

The ITER CODAC conceptual design

International Nuclear Information System (INIS)

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

2007-01-01

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

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.

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

Iterative-Transform Phase Retrieval Using Adaptive Diversity

Science.gov (United States)

Dean, Bruce H.

2007-01-01

A phase-diverse iterative-transform phase-retrieval algorithm enables high spatial-frequency, high-dynamic-range, image-based wavefront sensing. [The terms phase-diverse, phase retrieval, image-based, and wavefront sensing are defined in the first of the two immediately preceding articles, Broadband Phase Retrieval for Image-Based Wavefront Sensing (GSC-14899-1).] As described below, no prior phase-retrieval algorithm has offered both high dynamic range and the capability to recover high spatial-frequency components. Each of the previously developed image-based phase-retrieval techniques can be classified into one of two categories: iterative transform or parametric. Among the modifications of the original iterative-transform approach has been the introduction of a defocus diversity function (also defined in the cited companion article). Modifications of the original parametric approach have included minimizing alternative objective functions as well as implementing a variety of nonlinear optimization methods. The iterative-transform approach offers the advantage of ability to recover low, middle, and high spatial frequencies, but has disadvantage of having a limited dynamic range to one wavelength or less. In contrast, parametric phase retrieval offers the advantage of high dynamic range, but is poorly suited for recovering higher spatial frequency aberrations. The present phase-diverse iterative transform phase-retrieval algorithm offers both the high-spatial-frequency capability of the iterative-transform approach and the high dynamic range of parametric phase-recovery techniques. In implementation, this is a focus-diverse iterative-transform phaseretrieval algorithm that incorporates an adaptive diversity function, which makes it possible to avoid phase unwrapping while preserving high-spatial-frequency recovery. The algorithm includes an inner and an outer loop (see figure). An initial estimate of phase is used to start the algorithm on the inner loop, wherein

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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.

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.

Application of MCAM in generating Monte Carlo model for ITER port limiter

International Nuclear Information System (INIS)

Lu Lei; Li Ying; Ding Aiping; Zeng Qin; Huang Chenyu; Wu Yican

2007-01-01

On the basis of the pre-processing and conversion functions supplied by MCAM (Monte-Carlo Particle Transport Calculated Automatic Modeling System), this paper performed the generation of ITER Port Limiter MC (Monte-Carlo) calculation model from the CAD engineering model. The result was validated by using reverse function of MCAM and MCNP PLOT 2D cross-section drawing program. the successful application of MCAM to ITER Port Limiter demonstrates that MCAM is capable of dramatically increasing the efficiency and accuracy to generate MC calculation models from CAD engineering models with complex geometry comparing with the traditional manual modeling method. (authors)

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.

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.

ITER ITA newsletter. No. 20, February-March 2005

International Nuclear Information System (INIS)

2005-03-01

This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including interview on the occasion of Academician E.P. Velikhov' 70th birthday conducted by Dr. Lev Golubbchikov, former ITER Contact Person of the Russian Federation and a new document management system of ITER called IDM (ITER Document Management), which supersedes the old IDoMS

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.

Tritium inventories and tritium safety design principles for the fuel cycle of ITER

International Nuclear Information System (INIS)

Cristescu, I.R.; Cristescu, I.; Doerr, L.; Glugla, M.; Murdoch, D.

2007-01-01

Within the tritium plant of ITER a total inventory of about 2-3 kg will be necessary to operate the machine in the DT phase. During plasma operation, tritium will be distributed in the different sub-systems of the fuel cycle. A tool for tritium inventory evaluation within each sub-system of the fuel cycle is important with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems; however, tritium accounting may be achieved by modelling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the sub-systems. To get reliable results, an accurate dynamic modelling of the tritium content in each sub-system is necessary. A dynamic model (TRIMO) for tritium inventory calculation reflecting the design of each fuel cycle sub-systems was developed. The amount of tritium needed for ITER operation has a direct impact on the tritium inventories within the fuel cycle sub-systems. As ITER will function in pulses, the main characteristics that influence the rapid tritium recovery from the fuel cycle as necessary for refuelling are discussed. The confinement of tritium within the respective sub-systems of the fuel cycle is one of the most important safety objectives. The design of the deuterium/tritium fuel cycle of ITER includes a multiple barrier concept for the confinement of tritium. The buildings are equipped with a vent detritiation system and re-circulation type room atmosphere detritiation systems, required for tritium confinement barrier during possible tritium spillage events. Complementarily to the atmosphere detritiation systems, in ITER a water detritiation system for tritium recovery from various sources will also be operated

Failure mode analysis of preliminary design of ITER divertor impurity monitor

International Nuclear Information System (INIS)

Kitazawa, Sin-iti; Ogawa, Hiroaki

2016-01-01

Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • Failure mode of DIM was prepared for RAMI analysis. • RAMI analysis on DIM was performed to reduce technical risks. - Abstract: The objective of the divertor impurity influx monitor (DIM) for ITER is to measure the parameters of impurities and hydrogen isotopes (tritium, deuterium, and hydrogen) in divertor plasma using visible and UV spectroscopic techniques in the 200–1000 nm wavelength range. In ITER, special provisions are required to ensure accuracy and full functionality of the diagnostic components under harsh conditions (high temperature, high magnetic field, high vacuum condition, and high radiation field). Japan Domestic Agency is preparing the preliminary design of the ITER DIM system, which will be installed in the upper, equatorial and lower ports. The optical and mechanical designs of the DIM are conducted to fit ITER’s requirements. The optical and mechanical designs meet the requirements of spatial resolution. Some auxiliary systems were examined via prototyping. The preliminary design of the ITER DIM system was evaluated by RAMI analysis. The availability of the designed system is adequately high to satisfy the project requirements. However, some equipment does not have certain designs, and this may cause potential technical risks. The preliminary design should be modified to reduce technical risks and to prepare the final design.

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.)

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.

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.

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

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.

Low Complexity V-BLAST MIMO-OFDM Detector by Successive Iterations Reduction

Directory of Open Access Journals (Sweden)

AHMED, K.

2015-02-01

Full Text Available V-BLAST detection method suffers large computational complexity due to its successive detection of symbols. In this paper, we propose a modified V-BLAST algorithm to decrease the computational complexity by reducing the number of detection iterations required in MIMO communication systems. We begin by showing the existence of a maximum number of iterations, beyond which, no significant improvement is obtained. We establish a criterion for the number of maximum effective iterations. We propose a modified algorithm that uses the measured SNR to dynamically set the number of iterations to achieve an acceptable bit-error rate. Then, we replace the feedback algorithm with an approximate linear function to reduce the complexity. Simulations show that significant reduction in computational complexity is achieved compared to the ordinary V-BLAST, while maintaining a good BER performance.

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

Iterative Refinement Methods for Time-Domain Equalizer Design

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Evans Brian L

2006-01-01

Full Text Available Commonly used time domain equalizer (TEQ design methods have been recently unified as an optimization problem involving an objective function in the form of a Rayleigh quotient. The direct generalized eigenvalue solution relies on matrix decompositions. To reduce implementation complexity, we propose an iterative refinement approach in which the TEQ length starts at two taps and increases by one tap at each iteration. Each iteration involves matrix-vector multiplications and vector additions with matrices and two-element vectors. At each iteration, the optimization of the objective function either improves or the approach terminates. The iterative refinement approach provides a range of communication performance versus implementation complexity tradeoffs for any TEQ method that fits the Rayleigh quotient framework. We apply the proposed approach to three such TEQ design methods: maximum shortening signal-to-noise ratio, minimum intersymbol interference, and minimum delay spread.

ITER assembly and maintenance

International Nuclear Information System (INIS)

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

1991-01-01

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

NUMERICAL WITHOUT ITERATION METHOD OF MODELING OF ELECTROMECHANICAL PROCESSES IN ASYNCHRONOUS ENGINES

Directory of Open Access Journals (Sweden)

D. G. Patalakh

2018-02-01

Full Text Available Purpose. Development of calculation of electromagnetic and electromechanic transients is in asynchronous engines without iterations. Methodology. Numeral methods of integration of usual differential equations, programming. Findings. As the system of equations, describing the dynamics of asynchronous engine, contents the products of rotor and stator currents and product of rotation frequency of rotor and currents, so this system is nonlinear one. The numeral solution of nonlinear differential equations supposes an iteration process on every step of integration. Time-continuing and badly converging iteration process may be the reason of calculation slowing. The improvement of numeral method by the way of an iteration process removing is offered. As result the modeling time is reduced. The improved numeral method is applied for integration of differential equations, describing the dynamics of asynchronous engine. Originality. The improvement of numeral method allowing to execute numeral integrations of differential equations containing product of functions is offered, that allows to avoid an iteration process on every step of integration and shorten modeling time. Practical value. On the basis of the offered methodology the universal program of modeling of electromechanics processes in asynchronous engines could be developed as taking advantage on fast-acting.

Engineering design and analysis of an ITER-like first mirror test assembly on JET

DEFF Research Database (Denmark)

Vizvary, Z.; Bourdel, B.; Garcia-Carrasco, A.

2017-01-01

is underway on JET, under contract to ITER, with primary objective to test if, under realistic plasma and wall material conditions and with ITER-like first mirror aperture geometry, deposits do grow on first mirrors. This paper describes the engineering design and analysis of this mirror test assembly......The ITER first mirrors are the components of optical diagnostic systems closest to the plasma. Deposition may build up on the surfaces of the mirror affecting their ability to fulfil their function. However, physics modelling of this layer growth is fraught with uncertainty. A new experiment...

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'

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.

Chapter 8: Plasma operation and control [Progress in the ITER Physics Basis (PIPB)

International Nuclear Information System (INIS)

Gribov, Y.; Humphreys, D.; Kajiwara, K.; Lazarus, E.A.; Lister, J.B.; Ozeki, T.; Portone, A.; Shimada, M.; Sips, A.C.C.; Wesley, J.C.

2007-01-01

The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m -1 ), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation

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

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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.

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.

A linear iterative unfolding method

International Nuclear Information System (INIS)

László, András

2012-01-01

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

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.

Low-memory iterative density fitting.

Science.gov (United States)

Grajciar, Lukáš

2015-07-30

A new low-memory modification of the density fitting approximation based on a combination of a continuous fast multipole method (CFMM) and a preconditioned conjugate gradient solver is presented. Iterative conjugate gradient solver uses preconditioners formed from blocks of the Coulomb metric matrix that decrease the number of iterations needed for convergence by up to one order of magnitude. The matrix-vector products needed within the iterative algorithm are calculated using CFMM, which evaluates them with the linear scaling memory requirements only. Compared with the standard density fitting implementation, up to 15-fold reduction of the memory requirements is achieved for the most efficient preconditioner at a cost of only 25% increase in computational time. The potential of the method is demonstrated by performing density functional theory calculations for zeolite fragment with 2592 atoms and 121,248 auxiliary basis functions on a single 12-core CPU workstation. © 2015 Wiley Periodicals, Inc.

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

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.

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.

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.

Recent Progress on ECH Technology for ITER

Science.gov (United States)

Sirigiri, Jagadishwar

2005-10-01

The Electron Cyclotron Heating and Current Drive (ECH&CD) system for ITER is a critical ITER system that must be available for use on Day 1 of the ITER experimental program. The applications of the system include plasma start-up, plasma heating and suppression of Neoclassical Tearing Modes (NTMs). These applications are accomplished using 27 one megawatt continuous wave gyrotrons: 24 at a frequency of 170 GHz and 3 at a frequency of 120 GHz. There are DC power supplies for the gyrotrons, a transmission line system, one launcher at the equatorial plane and three upper port launchers. The US will play a major role in delivering parts of the ECH&CD system to ITER. The present state-of-the-art includes major advances in all areas of ECH technology. In the US, a major effort is underway to supply gyrotrons of up to 1.5 MW power level at 110 GHz to General Atomics for use in heating the DIII-D tokamak. This presentation will include a brief review of the state-of-the-art, worldwide, in ECH technology. The requirements for the ITER ECH&CD system will then be reviewed. ITER calls for gyrotrons capable of operating from a 50 kV power supply, after potential depression, with a minimum of 50% overall efficiency. This is a very significant challenge and some approaches to meeting this goal will be presented. Recent experimental results at MIT showing improved efficiency of high frequency, 1.5 MW gyrotrons will be described. These results will be incorporated into the planned development of gyrotrons for ITER. The ITER ECH&CD system will also be a challenge to the transmission lines, which must operate at high average power at up to 1000 seconds and with high efficiency. The technology challenges and efforts in the US and other ITER parties to solve these problems will be reviewed. *In collaboration with E. Choi, C. Marchewka, I. Mastovosky, M. A. Shapiro and R. J. Temkin. This work is supported by the Office of Fusion Energy Sciences of the U. S. Department of Energy.

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.)

Conceptual design Fusion Experimental Reactor (FER/ITER)

International Nuclear Information System (INIS)

Uehara, Kazuya; Nagashima, Takashi; Ikeda, Yoshitaka

1991-11-01

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

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

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.

Non-Characteristic Harmonics Analysis of the ITER Pulsed Power Supply

International Nuclear Information System (INIS)

Yang Wei; Xu Liuwei; Fu Peng; Lu Huawei; Sheng Zhicai

2009-01-01

The ITER pulsed power supply system will be operated in non-ideal conditions including an asymmetric firing angle, an unbalanced AC supply and an unbalanced AC side impedance of the transformer. In this study the switching functions approach is used to calculate non-characteristic harmonics in ITER, possibly caused by an AC-DC convertor in non-ideal conditions. A PSCAD simulation model is set up to study the non-characteristic harmonics in those non-ideal conditions. It is found that the non-characteristic harmonic does appear and the simulation result is in accordance with the calculating strategy. (fusion engineering)

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.

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)

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)

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.)

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.

The ITER reduced cost design

International Nuclear Information System (INIS)

Aymar, R.

2000-01-01

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

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.

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.

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.

IWR-solution for the ITER vacuum vessel assembly

Energy Technology Data Exchange (ETDEWEB)

Wu, H., E-mail: huapeng@lut.fi [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Handroos, H. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Pela, P. [Tekes (Finland); Wang, Y. [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland)

2011-10-15

The assembly of ITER vacuum vessel (VV) is still a very big challenge as the process can only be done from inside the VV. The welding of the VV assembly is carried out using the dedicated robotic systems. The main functions of the robots are: (i) measuring the actual space between every two sectors, (ii) positioning of the 150 kg splice plates between the sector shells, (iii) welding the splice plates to the sector shells, (iv) NDT of the welds, (v) repairing, including machining of the welds, (vi) He-leak tests of the welds, and (vii) the non-planned functions that may turn out. This paper presents a reasonable method to assemble the ITER VV. In this article, one parallel mobile robot, running on the track rail fixed on the wall inside the VV, is designed and tested. The assembling process, carried out by the mobile robot together with the welding robot, is presented.

Electromagnetic analysis of ITER diagnostic equatorial port plugs during plasma disruptions

International Nuclear Information System (INIS)

Zhai, Y.; Feder, R.; Brooks, A.; Ulrickson, M.; Pitcher, C.S.; Loesser, G.D.

2013-01-01

Highlights: ► Disruption loads on ITER diagnostic equatorial port plugs are extracted. ► Upward major disruption produces the largest radial moment and radial force on diagnostic first walls and diagnostic shield modules. ► Large eddy currents on supporting rails, keys and water pipes are observed during disruption. -- Abstract: ITER diagnostic port plugs perform many functions including structural support of diagnostic systems under high electromagnetic loads while allowing for diagnostic access to the plasma. The design of diagnostic equatorial port plugs (EPP) are largely driven by electromagnetic loads and associate responses of EPP structure during plasma disruptions and VDEs. This paper summarizes results of transient electromagnetic analysis using Opera 3d in support of the design activities for ITER diagnostic EPP. A complete distribution of disruption loads on the diagnostic first walls (DFWs), diagnostic shield modules (DSMs) and the EPP structure, as well as impact on the system design integration due to electrical contact among various EPP structural components are discussed

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)

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)

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

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

Overview and status of ITER internal components

International Nuclear Information System (INIS)

Merola, Mario; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

2014-01-01

Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine

Overview and status of ITER internal components

Energy Technology Data Exchange (ETDEWEB)

Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

2014-10-15

Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.

News from ITER controls - a status report

International Nuclear Information System (INIS)

Wallander, A.; Abadie, L.; Di Maio, F.; Evrard, B.; Fourneron, J.M.; Gulati, H.; Hansalia, C.; Journeaux, J.Y.; Kim, C.; Klotz, W.D.; Mahajan, K.; Makijarvi, P; Matsumoto, Y.; Pande, S.; Simrock, S.; Stepanov, D.; Utzel, N.; Vergara, A.; Winter, A.; Yonekawa, I.

2012-01-01

Construction of ITER has started at the Cadarache site in southern France. The first buildings are taking shape and more than 60 % of the in-kind procurement has been committed by the seven ITER member states (China, Europe, India, Japan, Korea, Russia and United States). The design and manufacturing of the main components of the machine is now underway all over the world. Each of these components comes with a local control system, which must be integrated in the central control system. The control group at ITER has developed two products to facilitate it; the plant control design handbook (PCDH) and the control, data access and communication (CODAC) core system. PCDH is a document which prescribes the technologies and methods to be used in developing local control systems and sets the rules applicable to the in-kind procurements. CODAC core system is a software package, distributed to all in-kind procurement developers, which implements the PCDH and facilitates the compliance of the local control system. In parallel, the ITER control group is proceeding with the design of the central control system to allow fully integrated and automated operation of ITER. In this paper we report on the progress of the design and technology choices and we discuss justifications of those choices. We also report on the results of some pilot projects aimed at validating the design and technologies. (authors)

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)

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.

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).

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.

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.

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.)

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)

Potential failure mode and effects analysis for the ITER NB injector

International Nuclear Information System (INIS)

Boldrin, M.; De Lorenzi, A.; Fiorentin, A.; Grando, L.; Marcuzzi, D.; Peruzzo, S.; Pomaro, N.; Rigato, W.; Serianni, G.

2009-01-01

The failure mode and effects analysis (FMEA) is a widely used analytical technique that helps in identifying and reducing the risks of failure in a system, component or process. The application of a systematic method like the FMEA was deemed necessary and adequate to support the design process of the ITER NBI (neutral beam injector). The approach adopted was to develop a FMEA at a general 'system level', focusing the study on the main functions of the system and ensuring that all the interfaces and interactions are covered among the various subsystems. The FMEA was extended to the whole NBI system taking into account the present design status. The FMEA procedure will be then applied to the detailed design phase at the component level, in particular to identify (or define) the ITER Class of Risk. Several important failure modes were evidenced, and estimates of subsystems and components reliability are now available. FMEA procedure resulted essential to identify and confirm the diagnostic systems required for protection and control, and the outcome of this analysis will represent the baseline document for the design of the NBI and NBTF integrated protection system. In the paper, rationale and background of the FMEA for ITER NBI are presented, methods employed are described and most interesting results are reported and discussed.

Investigation into the pumping characteristics of ITER cryopumps

International Nuclear Information System (INIS)

Day, C.; Mack, A.

1998-01-01

Within the framework of the European fusion technology programme, a cryopump system for ITER is being developed. It is based on combined sorption and condensation of gases at SK-surfaces, which are coated with activated charcoal. For verification of the design conditions an experimental programme has been launched. The tested cryopanels followed a quilted design, which is currently being discussed for its use in ITER. According to the composition range of the ITER exhaust gas in the various operation modes foreseen, pure gases (protium, deuterium, helium and neon) and gas mixtures (pseudobinaries of a D 2 -based mixture and one noble gas out of helium, neon or argon) were investigated. Quantitative measurements of pumping speed and equilibrium pressures at zero flow conditions were performed as a function of gas load; relative pumping probabilities were also derived. It is revealed that protium is pumped by sorption whereas neon is pumped by sublimation and deuterium is subjected to both mechanisms. The results demonstrate that the required pump ultimate pressure can be achieved. It is further shown that for the gases investigated the pumping characteristics will not be a limiting factor; the ITER requirements are well achieved. The saturation capacity will not be reached, except if pure helium is pumped. (orig.)

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.

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)

Magnet design technical report---ITER definition phase

International Nuclear Information System (INIS)

Henning, C.

1989-01-01

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

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

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.

The ITER activity

International Nuclear Information System (INIS)

Glass, A.J.

1991-01-01

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

Status of the R&D activities to the design of an ITER core CXRS diagnostic system

Energy Technology Data Exchange (ETDEWEB)

Mertens, Philippe, E-mail: ph.mertens@fz-juelich.de [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Castaño Bardawil, David A. [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Baross, Tétény [Wigner Research Centre for Physics (Wigner RCP), HU-1121 Budapest (Hungary); Biel, Wolfgang; Friese, Sebastian [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Hawkes, Nick [Culham Centre for Fusion Energy (CCFE), Culham OX14 3DB (United Kingdom); Jaspers, Roger J.E. [Eindhoven University of Technology (TU/e), PO Box 513, NL-5600 MB Eindhoven (Netherlands); Kotov, Vladislav; Krasikov, Yury; Krimmer, Andreas; Litnovsky, Andrey; Marchuk, Oleksander; Neubauer, Olaf [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); Offermanns, Guido [Zentralinstitut für Engineering, Elektronik und Analytik ZEA-1 (Engineering and Technology), FZJ, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Panin, Anatoly [Institute of Energy and Climate Research IEK-4 (Plasma Physics), Forschungszentrum Jülich (FZJ), Trilateral Euregio Cluster, D-52425 Jülich (Germany); and others

2015-10-15

Highlights: • The CXRS diagnostic for the core plasma of ITER will provide observation of the dedicated diagnostic beam (DNB) over a wide radial range, roughly r/a = 0.7 to 0. • A high performance (étendue × transmission, dynamic range) is expected for the port plug system since the beam attenuation is large and the background light omnipresent. • The design is particularly challenging in view of the ITER environment, especially with respect to the first mirror which faces the plasma. • The current status of development is presented by detailing several sub-systems before a four years design phase under an FPA between F4E and the ITER core CXRS Consortium (IC3). - Abstract: The CXRS (Charge-eXchange Recombination Spectroscopy) diagnostic for the core plasma of ITER will be designed to provide observation of the dedicated diagnostic beam (DNB) over a wide radial range, roughly from a normalised radius r/a = 0.7 to close to the plasma axis. The collected light will be transported through the Upper Port Plug #3 (UPP3) to a bundle of fibres and ultimately to a set of remote spectrometers. The design is particularly challenging in view of the ITER environment of particle, heat and neutron fluxes, temperature cycles, electromagnetic loads, vibrations, expected material degradation and fatigue, constraints against tritium penetration, integration in the plug and limited opportunities for maintenance. Moreover, a high performance (étendue × transmission, dynamic range) is expected for the port plug system since the beam attenuation is large and the background light omnipresent, especially in terms of bremsstrahlung, line radiation and reflections. The present contribution will give an overview of the current status and activities which deal with the core CXRS system, summarising the investigations which have taken place before entering the actual development and design phase.

ITER driver blanket, European Community design

International Nuclear Information System (INIS)

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

1993-01-01

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

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

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.

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.

Fabrication of divertor cassette for ITER

International Nuclear Information System (INIS)

Sanguinetti, G.P.

2008-01-01

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

Advances in optical thermometry for the ITER divertor

Energy Technology Data Exchange (ETDEWEB)

Lott, F. [CEA, IRFM, F-13108 St Paul lez Durance (France)], E-mail: fraser.lott@gmail.com; Netchaieff, A. [Laboratoire National de Metrologie et d' Essais (LNE), ZA de Trappes-Elancourt, 29 avenue Roger Hennequin, 78197 TRAPPES Cedex (France); Escourbiac, F. [CEA, IRFM, F-13108 St Paul lez Durance (France); Jouvelot, J.-L.; Constans, S. [AREVA NP, Centre Technique-FE200, Porte Magenta BP 181, 71205 Le Creusot (France); Hernandez, D. [Procedes, Materiaux et Energie Solaire (PROMES), Centre National de la Recherche Scientifique (CNRS), B.P. 5, 66125 Font-Romeu Cedex (France)

2010-01-15

Thermography will be an important diagnostic on the ITER tokamak, but the inclusion of reflective materials such as tungsten in the design for ITER's first wall and divertor region presents problems for optical temperature measurement. The ongoing testing of ITER plasma facing components (PFCs) provides an excellent opportunity to resolve such problems. This has focused on the variation of PFC emissivity with temperature and time, as well as environmental influence on thermography. The sensitivity of these systems to ambient temperature, due primarily to modification of the transmission of the optical path, has been established and minimised. The accuracy of the system is then sufficient to measure the variation of emissivity in heated material samples, by comparing its front-face luminance measured with an infrared camera to the temperature given by an implanted thermocouple. Measurements on both tungsten and carbon fibre composite are in broad agreement with theory, and thus give the material's function of emissivity with temperature at the start of its life. To determine its evolution, a bicolour pyroreflectometer was then installed. This uses two lasers to measure the reflectivity in addition to the luminance at two wavelengths, and thus the true temperature can be calculated. This was validated against the instrumented sample, then used along with the camera to observe an ITER mock-up during {approx}50,000 s of 5 MW/m{sup 2} testing. Emissivity was seen to vary little in the 500 deg. C region. Higher temperature tests are ongoing.

CFTSIM-ITER dynamic fuel cycle model

International Nuclear Information System (INIS)

Busigin, A.; Gierszewski, P.

1998-01-01

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

Plasma control concepts for ITER

International Nuclear Information System (INIS)

Lister, J.B.; Nieswand, C.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Iterative solutions of finite difference diffusion equations

International Nuclear Information System (INIS)

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

1981-01-01

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

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.

Multicore Performance of Block Algebraic Iterative Reconstruction Methods

DEFF Research Database (Denmark)

Sørensen, Hans Henrik B.; Hansen, Per Christian

2014-01-01

Algebraic iterative methods are routinely used for solving the ill-posed sparse linear systems arising in tomographic image reconstruction. Here we consider the algebraic reconstruction technique (ART) and the simultaneous iterative reconstruction techniques (SIRT), both of which rely on semiconv......Algebraic iterative methods are routinely used for solving the ill-posed sparse linear systems arising in tomographic image reconstruction. Here we consider the algebraic reconstruction technique (ART) and the simultaneous iterative reconstruction techniques (SIRT), both of which rely...... on semiconvergence. Block versions of these methods, based on a partitioning of the linear system, are able to combine the fast semiconvergence of ART with the better multicore properties of SIRT. These block methods separate into two classes: those that, in each iteration, access the blocks in a sequential manner...... a fixed relaxation parameter in each method, namely, the one that leads to the fastest semiconvergence. Computational results show that for multicore computers, the sequential approach is preferable....

Activation and afterheat analyses for the HCPB test blanket module in ITER

International Nuclear Information System (INIS)

Pereslavtsev, P.; Fischer, U.

2008-01-01

To provide a sound data basis for the safety analyses of the HCPB TBM system in ITER, the afterheat and activity inventories were assessed making use of a code system that allows performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. A suitable MCNP model of a 20 deg. ITER torus sector with an integrated TBM of the HCPB PI (plant integration) type in the horizontal test blanket port was developed and adapted to the requirements for coupled 3D neutron transport and activation calculations. Two different irradiation scenarios were considered in the coupled 3D neutron transport and activation calculations. The first one is representative for the TBM irradiation in ITER with a total of 9000 neutron pulses over a 3 (calendar) years period. The second (conservative) irradiation scenario assumes an extended irradiation time over the full anticipated lifetime of ITER. The radioactivity inventories, the afterheat and the contact gamma dose were calculated as function of the decay time. Data were processed for the total activity, afterheat and contact dose rates of the TBM, its constituting components and materials

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...

ITER Dynamic Tritium Inventory Modeling Code

International Nuclear Information System (INIS)

Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

2005-01-01

A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

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.

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

ITER safety and operational scenario

International Nuclear Information System (INIS)

Shimomura, Y.; Saji, G.

1998-01-01

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