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A number of models for the critical surface of Nb3 Sn, and in general A15 superconductors, have been developed in the past years. This paper compares the most common parameterizations using consistent notation. Although the parameterizations appear dissimilar at first sight, they are in reality all based on a fit of the normalized pinning force vs. the reduced field, and have similar scalings for the critical field and critical temperature based on a Unified Scaling Law. In this paper we take the various parameterizations as a basis for a generic scaling proposed for the characterization and production follow-up of the ITER Nb3 Sn strands. The accuracy of the scaling is estimated using the fitting residuals on various sets of Ic (B, T, ϵ) data available in literature. We discuss the results, and give our view of the work towards a unified, practical parameterization. [ABSTRACT FROM AUTHOR]

The measurement of the current sharing temperature of a Cable in Conduit Conductor is a complex task. The voltage traces measured on the conduit (jacket) of these conductors during tests, give V∙I characteristics that significantly differ from the typical transition from the superconducting to the resistive state measured on single strands. In these measurements, after an initial ramping of the transport current up to the test value, the temperature is increased by steps until the resistive transition occurs. However, even in the first phase of the current ramping, when the cable is still fully or almost superconductive, early voltages are measured along the jacket at different angular positions around the cable. In particular, this was observed in the ITER TF conductors recently tested in the SULTAN facility at Villigen PSI, Switzerland. These samples showed significant ramps of voltages taken in the proximity of the joints and terminations. The present paper gives a possible qualitative and quantitative explanation for this mechanism. Numerical simulations are also shown which reproduce the scattering of the voltages measured around the jacket at a given location, showing the effect of the voltage taps position along the jacket in a six sub-cable model. [ABSTRACT FROM AUTHOR]

In this paper we report the main test results obtained on the Poloidal Field Conductor Insert coil (PFI) for the International Thermonuclear Experimental Reactor (ITER), built jointly by the EU and RF ITER parties, recently installed and tested in the CS Model Coil facility, at JAEA-Naka. During the test we (a) verified the DC and AC operating margin of the NbTi Cable-in-Conduit Conductor in conditions representative of the operation of the ITER PF coils, (b) measured the intermediate conductor joint resistance, margin and loss, and (c) measured the AC loss of the conductor and its changes once subjected to a significant number of Lorentz force cycles. We compare the results obtained to expectations from strand and cable characterization, which were studied extensively earlier. We finally discuss the implications for the ITER PF system. [ABSTRACT FROM AUTHOR]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. The use of High Temperature Superconductor current leads (HTS-CL) will reduce the resultant large refrigerator load considerably. Within the EU Fusion Technology Programme, the Forschungszentrum Karlsruhe and CRPP have developed and built a 70 kA HTS-CL demonstrator optimized for 50 K Helium operation. In 2004-2005, the CL was successfully tested with 50 K as well as 80 K Helium, and finally with LN2. In the meantime, ITER has decided to use HTS-CLs for the magnet system and China has taken the responsibility to provide the sc feeders for ITER including the current leads. In this paper, the performance of a 70 kA HTS-CL in fault conditions will be analysed using the FZK-CRPP demonstrator as a model. In particular, the quench performance and the safety behavior in case of a loss-of-flow accident (LOFA) will be described and an extrapolation of the leads performance in real ITER conditions will be given. Finally the effect of different shunt materials is discussed. [ABSTRACT FROM AUTHOR]

The ITER Organization (10) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly developing the 68 kA current leads using High Temperature Superconductors for the superconducting Toroidal Field (TF) magnet system of the International Thermonuclear Experimental Reactor, ITER. The proposed design consists of a conventional helium cooled heat exchanger operating between 65 K and 320 K and an HTS module covering the low temperature end using Bi-2223 tapes. The first HTS current lead prototypes will be tested in the EAST tokamak facility at ASIPP. This paper discusses the design of the first, 68 kA, TF HTS lead, which is the main emphasis of this first stage of the ITER current lead development program. [ABSTRACT FROM AUTHOR]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. To reduce the resultant large refrigerator load at 4.5 K, High Temperature Superconductor current leads (HTS-CL) could be used. Therefore, EFDA CSU Garching had launched a development program for a 70 kA HTS-CL demonstrator. The Forschungszentrum Karlsruhe and CRPP developed and built this CL optimized for 50 K Helium operation. In 2004, the CL was successfully tested in the TOSKA facility at the Forschungszentrum Karlsruhe. The very encouraging results lead to testing this CL with 80 K Helium because ITER provides a large 80 K Helium cooling capacity for the thermal shields. At the end of last year, the test could be successfully performed demonstrating that high current capacity current leads can be stably operated at about 80–85 K. Recently, the CL was retested using liquid nitrogen which would be an interesting alternative option. In this paper, the test results for the 80 K He cooling operation as well as for the operation with LN2 are presented and compared to the results obtained before for nominal conditions. [ABSTRACT FROM AUTHOR]

Abstract: In this paper, a complete framework including modeling, identification and control design for an experimental Helium refrigerator facility submitted to high pulsed loads is proposed. Such high pulsed loads are expected to take place in future Tokamaks such as the International Thermonuclear Experimental Reactor (ITER) or the Japan Torus-60 Super Advanced (JT60SA). The control architecture is split into two loops: The first concerns the warm compression system while the second manages the remaining part of the refrigerator (often called the cold box) in which the set-points of the first loop are viewed as manipulated variables. The efficiency of the proposed solution is experimentally assessed under several scenarios including those involving realistic high pulsed loads. [Copyright &y& Elsevier]

Abstract: In the ITER edge Thomson scattering measurement system, polychromators with optical band-pass filters and avalanche photodiodes are planned to be used. In this paper, we optimized the transmission wavelength ranges of the optical filters by a numerical method. Since measurements in a high electron temperature range are required for the measurement system, the Thomson spectrum could overlap a strong line emission of when the electron temperature is rather high. It is shown that a filter whose transmission wavelength range is shorter than becomes important to decrease the measurement error in the high temperature range. Moreover, it is found that a filter whose transmission wavelength range is above 1064nm (laser wavelength) is useful to improve the measurement accuracy, in particular, when the number of filters is more than six. [Copyright &y& Elsevier]

Abstract: The present paper deals with the detailed investigation of the helium-cooled lithium lead test blanket module (HCLL-TBM) nuclear behaviour under irradiation in ITER, carried out at the Department of Nuclear Engineering of the University of Palermo adopting a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of the HCLL-TBM was set-up and inserted into an ITER 3D semi-heterogeneous model that realistically simulates the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source was adopted for the calculations. The main features of the HCLL-TBM nuclear response were assessed, paying a particular attention to the neutronic and photonic deposited power, the tritium production rate and the spatial distribution of their volumetric densities. Structural material irradiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates. [Copyright &y& Elsevier]

Abstract: The ITER conductor is made of a superconductor cable with multi-stage twisted strands. The spatial structure of such a cable is essential for understanding the effect of load and strain on the performance of a superconducting cable. Unfortunately, there have no analytical models so far to describe accurately such twisted spatial structure. This paper proposes a twisting matrix model to accurately describe the spatial structure of a multi-stage twisted superconducting cable. Although the current model only simulates the first step of a cabling process, it still gives approximately neighboring relation between strands and sub-cables. This neighboring relation is essential for any cabling simulation. The 2nd step simulation, i.e., compaction, can be realized by finite element method based on the cable spatial structure constructed by the proposed one. The simulation results by this model agree well with the observations for pitch and orientation. [Copyright &y& Elsevier]

Abstract: 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. [Copyright &y& Elsevier]

Dust mobilization in a vacuum vessel is one of the key issues endangering the security of the International Thermonuclear Experimental Reactor (ITER) in case of loss of vacuum accidents. The turbulent behavior of particles in turbulent flows has to be modeled for successful numerical simulations about particle mobilization. In this study a Lagrangian approach is adopted to formulate the particle transport especially for dust-dilute flows mostly encountered in the vacuum vessel of ITER. Based on the logic frame of the approach and the used computational fluid dynamics (CFD) computer code in the study, a hybrid turbulent particle dispersion model is proposed. The hybrid model features both a deterministic separated flow model and a stochastic separated flow (SSF) model, which are two popular turbulent dispersion models applied in particle simulations, and takes the advantages of the both models. The proposed model is implemented into the particle model of the CFD code successfully and the simulation results are verified against the experimental data. The verifications manifest the validities of the proposed model. In this paper general information about the work of dust mobilization is introduced and the particle turbulent dispersion models are reviewed briefly at first. The hybrid model is then proposed based on the SSF model. An experiment about particle dispersions in an advective wind channel flow with decaying turbulence in the streamwise direction is reviewed in the third section. In the following section about model verification, the decaying turbulence parameters in the channel flow are verified against the experimental data as the first step, and the parameters about the particle dispersions in the verified flow field are then verified against the data. The work is concluded finally. [ABSTRACT FROM AUTHOR]

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

Three planned test blanket modules (TBMs) and 18 toroidal field coils break the axisymmetry of the ITER magnetic field. In this paper, the plasma response on these non-axisymmetric fields is studied quantitatively. For this purpose, self-consistent, three-dimensional, free-boundary equilibria of type ITER scenario 4 are computed. The resulting 3D equilibrium magnetic fields are then compared with the corresponding axisymmetric fields to which the vacuum perturbation fields are superimposed. The studies are performed for various normalized plasma pressures. [ABSTRACT FROM AUTHOR]

Impacts of pellets injected from the low-field side (LFS) on plasma in ITER are investigated using the 1.5D BALDUR integrated predictive modeling code. In these simulations, the pellet ablation is described using the neutral gas shielding (NGS) model. The pellet ablation model is coupled with the plasma core transport model, which is a combination of the MMM95 anomalous transport model and NCLASS neoclassical transport model. The boundary conditions are assumed to be at the top of the pedestal, in which the pedestal parameters are predicted using a pedestal model based on the theoretical-based pedestal width scaling (either magnetic and flow shear stabilization width scaling, or flow shear stabilization width scaling, or normalized poloidal pressure width scaling) and the infinite- n ballooning mode pressure gradient limit. These pedestal models depend sensitively on the density at the top of the pedestal, which can be strongly influenced by the injection of pellets. The combination of the MMM95 and NCLASS models, together with the pedestal and NGS models, is used to simulate the time evolution of the plasma current, ion and electron temperatures, and density profiles for ITER standard type-I ELMy H-mode discharges during the injection of LFS pellets. It is found that the injection of pellets results in a complicated plasma scenario, especially in the outer region of the plasma and the plasma conditions at the boundary in which the pellet has an impact on increasing the plasma edge density, but reducing the plasma edge temperature. The LFS pellet has a stronger impact on the edge as compared to the center. For fusion performance, the pellet can result in either enhancement or degradation, depending sensitively on the pellet parameters; such as the pellet size, pellet velocity, and pellet frequency. For example, when a series of deuterium pellets with a size of 0.5 cm, velocity of 1 km/s, and frequency of 2 Hz are injected into the ITER plasma from the LFS, the plasma performance, evaluated in terms of Q, can increase to 72% of that before the use of pellets. It is also found that the injection of pellets results in an increase in the ion and electron densities, but does not enhance the central plasma density. On the other hand, it results in the formation of another peak of the plasma density in the outer region near the plasma edge. The formation of the density peak results in the reduction of plasma transports near the edge by decreasing the contributions of ion-temperature-gradient and trapped electron modes, as well as kinetic ballooning modes. [ABSTRACT FROM AUTHOR]

Current status review of a tokamak reactor-based research is intended for the average readers. Using the project of International Thermonuclear Experimental Reactor (ITER) as an example, physical and engineering issues are discussed. Socio-economic aspects of a future fusion power plant (FPP) are considered. Draft of Russian Roadmap to FPP is presented. In conclusion possibilities of nuclear fuel production in a fusion reactor are outlined. [ABSTRACT FROM AUTHOR]

In the quest for new energy sources, the research on controlled thermonuclear fusion has been boosted by the start of the construction phase of the International Thermonuclear Experimental Reactor (ITER). ITER is based on the tokamak magnetic configuration, which is the best performing one in terms of energy confinement. Alternative concepts are however actively researched, which in the long term could be considered for a second generation of reactors. Here, we show results concerning one of these configurations, the reversed-field pinch (RFP). By increasing the plasma current, a spontaneous transition to a helical equilibrium occurs, with a change of magnetic topology. Partially conserved magnetic flux surfaces emerge within residual magnetic chaos, resulting in the onset of a transport barrier. This is a structural change and sheds new light on the potential of the RFP as the basis for a low-magnetic-field ohmic fusion reactor. [ABSTRACT FROM AUTHOR]

Numerous manufacturers and different strand processing techniques are involved with the production of the Nb3 Sn strand material required for ITER. The superconducting transport properties of brittle Nb3 Sn layers strongly depend on their strain state. Hence, the thermal compression and the substantial transverse load in combination with the key choice for the cabling pattern of the CICCs, will determine their performance. Knowledge of the influence of axial strain, periodic bending, and contact stress on the critical current (Ic) of the used Nb3 Sn strands is inevitable to gain sufficient confidence in an economic design and a stable operation of ITER. We have measured the Ic and n-value of Nb3 Sn strands from various manufacturers in the TARSIS facility, when subjected to spatial periodic bending and contact stress. The Ic and n-values have been determined for applied axial compressive and tensile strain varying from -0.8% up to +0.5%, between T = 4.2 K and 10 K and B = 6 T to 14 T. The strain sensitivity varies appreciably for different strand types. We present a selection of the results obtained so far. [ABSTRACT FROM AUTHOR]

Describes the design of the International Thermonuclear Experimental Reactor (ITER) magnet and model coils. Key features of the ITER Final Design Report magnet design; Magnet design drivers for Reduced Technical Objectives/Reduced Cost-ITER; Characteristics of the model coils.

Assesses the radial plate design of the Toroidal Field superconducting coil winding pack of the International Thermonuclear Experimental Reactor. Design options for the winding pack; Features of the radial plate design; Effective elastic characteristics of the winding pack; Stress analysis.

Focuses on the use of the pancake-wound technique in the development of the central solenoid (CS) coil of the Reduced Technical Objective/Reduced Cost for the International Thermonuclear Experimental Reactor. Coil design; Two alternatives for the CS conduit reinforcement; Assessment of the CS structure performance through an axisymmetric finite element method.

Evaluates the performance of the coils made of Cable-In-Conduit Nb3Sn conductors for the Central Solenoid of the International Thermonuclear Experimental Reactor. Performance of fast ramp tests; Distribution of the current inside the cable; Measurement of the critical current; Determination of the fastest possible current ramps.

Focuses on the development of an Nb3Sn conductor joint for the Central Solenoid (CS) Model Coil of the International Thermonuclear Experimental Reactor. Configuration of the butt joint for the CS Model Coil; Fabrication of hairpin cable; Joint resistance; Measurement of the quench current in various temperatures and fields.

Describes the design and fabrication of an Central Solenoid Model Coil developed by the International Thermonuclear Experimental Reactor. Overview of the inner module fabrication; Peak field produced by the model coil; Coil winding procedure; Insulation technology and materials.

Presents an abstract for a paper describing the preliminary operational plan and physics program for the International Thermonuclear Experimental Reactor (ITER) in La Jolla, California, and its implications for plasma measurements. Key objectives of the first ten years of ITER.

Summarizes detailed calculations of alpha ripple loss in the International Thermonuclear Experimental Reactor (ITER) in La Jolla, California, finding ripple losses similar to previous predictions for full current operation to the 24 toroidal field (TF) coil ITER; Simulation methods used including the development of an analytic fit to the TF ripple data field.

Provides information on the International Thermonuclear Experimental Reactor (ITER) in La Jolla, California, a multiphase project proceeding under the auspices of the International Atomic Energy Agency. Terms of operation under a four-party agreement; Objective of ITER; Major design requirements for ITER.

Reports the progress of the engineering design activities (EDA) for the International Thermonuclear Experimental Reactor (ITER). Aim of the international collaborative project, ITER; Details on the early developments of the ITER; Status of the ITER design.

Indicates the results of line brightness temperature calculations for parameters at the International Thermonuclear Experimental Reactor (ITER). Explanation of results on the Tokamak Fusion Test Reactor (TFTR); Feasibility of time-resolved measurements of the central ion temperature and central plasma parameters on ITER; Presence of high fluxes of neutrons.

The predictions of gyrokinetic and gyrofluid simulations of ion-temperature-gradient (ITG) instability and turbulence in tokamak plasmas as well as some tokamak plasma thermal transport models, which have been widely used for predicting the performance of the proposed International Thermonuclear Experimental Reactor (ITER) tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1996 (International Atomic Energy Agency, Vienna, 1997), Vol. 1, p. 3], are compared. These comparisons provide information on effects of differences in the physics content of the various models and on the fusion-relevant figures of merit of plasma performance predicted by the models. Many of the comparisons are undertaken for a simplified plasma model and geometry which is an idealization of the plasma conditions and geometry in a Doublet III-D [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] high confinement (H-mode) experiment. Most of the models show good agreements in their predictions and assumptions for the linear growth rates and frequencies. There are some differences associated with different equilibria. However, there are significant differences in the transport levels between the models. The causes of some of the differences are examined in some detail, with particular attention to numerical convergence in the turbulence simulations (with respect to simulation mesh size, system size and, for particle-based simulations, the particle number). The implications for predictions of fusion plasma performance are also discussed. © 2000 American Institute of Physics. [ABSTRACT FROM AUTHOR]