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In physical theory models, the hysteretic response of concentric braces is simulated by modelling the bracings with two nonlinear fibre-section beam–column elements connected together with an initial out-ofstraightness imperfection (Δ0). The width of such imperfection is the parameter governing the numerical prediction of the brace buckling. In this paper, the accuracy of different formulations of Δ0 is investigated and validated against monotonic and cyclic experimental results from literature carried out on different bracing configurations. Correlations and measures of scatter between the predicted response and experimental performance are evaluated. A statistical analysis on both X-braced and inverted-V-braced frames based on Monte Carlo simulation is presented and discussed for the following reasons: (a) in order to account for the dispersion of the nonlinear dynamic-evaluated seismic performance due to epistemic uncertainties associated with examined formulations for camber width and (b) to provide modelling correction factors of simulated interstorey drift demand. The analysis shows that the formulations used for Δ0 affect the drift demand, the collapse mechanism and have a noticeably impact on the seismic response parameters at collapse. In particular, inverted-V-braced frames are more influenced than X-braced configurations, because the prediction of brace–beam interaction is very sensitive to Δ0.

This paper presents the main lessons that can be learned from two decades of Plasma Control development experience. It will present the general architectural choices and will provide examples from key systems, and, more importantly, will highlight the plasma operation requirements that have driven the developments. The aim is to present a meaningful set of functional and non- functional requirements that were derived from the JET experience and to discuss their potential applicability to future experimental devices, ITER in particular. ©2009 IEEE.

he plasma in a tokamak is magnetically confined through electromagnetic fields generated by a set of poloidal field coils, with feedforward nominal voltages or currents based on an approximate plasma model. Therefore, feedback control is necessary. A good magnetic control is essential for achieving good performance in present devices and expected to be of paramount importance in future tokamaks with a burning plasma due to the more stringent requirements in terms of plasma-wall clearance, stability margins, limited number of tolerable disruptions and so on. Magnetic fields are able to control plasma current, position and shape as well as three dimensional instabilities on the resistive time scales (resistive wall modes). They also have an influence on the control of the plasma current density profile. This paper focuses on the axisymmetric control of plasma current, position and shape, illustrating modelling assumptions and techniques, controller design procedures, assessment of controller performance and magnetic diagnostic requirements. © 2008 American Institute of Physics.

(#NP1.010) Bibliographic Code: 2006APS..DPPNP1010V

A modification of the FTU tokamak (toroidal field BT = 8T, plasma current IP = 1.6MA, minor radius a = 0.3m, major radius R = 0.93m) is proposed in order to extend the FTU operation to strongly shaped plasmas (FTU-D: R = 1m, a = 0.18 - 0.2m, elongation = 1.6, triangularity up to = 0.8). FTU has a circular vacuum vessel and was built to produce circular plasmas, however unbalancing the currents in the windings of the air core transformer a plasma shaping can be produced. Single Null (SN) and Double Null (DN) equilibria have been studied with a maximum current in the range 0.350-0.450 MA. The scientific aim of the project is the investigation of the advanced tokamak operation, characterised by the simultaneous achievement of high (normalised beta) and high bootstrap current fraction ( fB) in regimes with high-energy confinement obtained by current and pressure profile control. The main features of FTU-D, with respect to other existing tokamaks, are the high magnetic field ( BT = 5 - 2.5T), the high density and aspect ratio value (A = R/a = 5-6) and the possibility of investigating regimes with dominant electron heating.

This paper is focused on the plasma shape and current control in a large, next generation tokamak characterized by long-duration plasma discharges. The new features of these fusion devices are described and their impact on the feedback control requirements are investigated. Many of the controller design issues are addressed in the H∞/μ synthesis framework, including the model reduction, the power constraints, the voltage limits and plant uncertainties. The application to the case of the International Thermonuclear Experimental Reactor (ITER) shows the effectiveness of this approach for the design of plasma controllers.

Background and Objectives: Treating segmental tibial and ankle bone loss after radical surgery for chronic osteomyelitis is one of the most challenging problems encountered by orthopaedic surgeons. Open tibia and ankle fractures occur with an incidence of 3.4 per 100,000 and 1.6 per 100,000, respectively, and there is a high propensity of developing fracture-related infection with associated chronic osteomyelitis in patients. Segmental tibial and ankle bone loss have recently received new and improved treatments. Materials and Methods: Above all, 3D printing allows for the customization of implants based on the anatomy of each patient, using a personalized process through the layer-by-layer deposition of materials. Results: This article presents different cases from the authors’ experience. Specifically, four patients suffered tibia and ankle fractures and after radical surgery for chronic osteomyelitis combined with high-performance antibiotic therapy underwent ankle reconstruction/arthrodesis with custom-made tibial spacers. Conclusions: Thanks to 3D-printed patient-specific devices, it is possible to perform surgical procedures that, for anatomical reasons, would have been impossible otherwise. Moreover, an improvement in overall functionality and an important reduction in pain were shown in the last follow-up in all patients.

The plasma control system (PCS) in a tokamak device is in charge of controlling the evolution of plasma parameters according to prescribed models against uncertainties and disturbances. The PCS accomplishes its mission by receiving data from the diagnostics and by computing in real-time the commands to the actuators that affect the plasma behavior. The design of the ITER PCS includes many different aspects, which are not limited to the design of control algorithms, including the definition of the verification and validation tests for various components and commissioning procedures. Moreover, contributions come from different parties that adopt heterogeneous sources. To homogenize these contributions and to keep track of the PCS life-cycle throughout various design stages, a specific system-engineering approach has been adopted on top of the standard ITER life-cycle and the requirement management process. Such an approach relies on a database (DB) implemented using Enterprise Architect that allows modeling various aspects of the PCS design using SysML. This article gives an overview of the adopted approach and describes the structure and the current content of the PCS DB.

The present paper deals with concept, prototyping and application of a tensioning system for FRP ties into masonry structures. The proposed system, based on the use of FRP strands instead of traditional steel ties, has the aim to produce a compression stress state on masonry walls where it is applied. Given the objective difficulty in tensioning a FRP strand, it was necessary to both characterize and prototype a suitable connection system between the strand and the pulling system. The experimental phase concerned both the manufacturing of the pulling system and the study of used materials, as well as the characterization of the impregnation technology of FRP ties. The above described system has been produced and used in the framework of the structural retrofitting of the “Real Albergo dei Poveri” building in Naples.

The Ignitor Program maintains the objective of approaching D-T ignition conditions by incorporating systematical advances made with relevant high field magnet technology and with experiments on high density well confined plasmas in the present machine design. An additional objective is that of charting the development of the high field line of experiments that goes from the Alcator machine to the ignitor device. The rationale for this class of experiments, aimed at producing poloidal fields with the highest possible values (compatible with proven safety factors of known plasma instabilities) is given. On the basis of the favourable properties of high density plasmas produced systematically by this line of machines, the envisioned future for the line, based on novel high field superconducting magnets, includes the possibility of investigating more advanced fusion burn conditions than those of the D-T plasmas for which Ignitor is designed. Considering that a detailed machine design has been carried out (Coppi et al 2013 Nucl. Fusion 53 104013), the advances made in different areas of the physics and technology that are relevant to the Ignitor project are reported. These are included within the following sections of the present paper: main components issues, assembly and welding procedures; robotics criteria; non-linear feedback control; simulations with three-dimensional structures and disruption studies; ICRH and dedicated diagnostics systems; anomalous transport processes including self-organization for fusion burning regimes and the zero-dimensional model; tridimensional structures of the thermonuclear instability and control provisions; superconducting components of the present machine; envisioned experiments with high field superconducting magnets. © 2015 IAEA.

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-level objectives for future work are also discussed. © 2015 Elsevier B.V. All rights reserved.

The JET Shape Controller (SC) uses nine distinct circuits, powering the JET poloidal field coils, to control in real time the coil currents, and the plasma shape, current and position. The control scheme presently used [1] is based on a Multiple Input Multiple Output (MIMO) controller, which is designed to decouple the inductive coupling of the different coils. Achieving such a decoupling, the SC allows the user to tune independently the time response of each circuit. As a matter of fact the intended decoupling algorithm has been incorrectly coded in the JET SC system. This paper describes the modelling and experimental activities performed to correct the code error, and to improve the performance on a subset of the controlled parameters. (C) 2015 Elsevier B.V. All rights reserved.

We describe a software system known as the Plasma Control System Simulation Platform (PCSSP) that is being constructed to support development of the ITER plasma control system. When mature, PCSSP will provide support for industry standard practices such as model-based controller design, simulation testing of controllers, auto-generation of controller code, hardware-and software-in-The-loop testing, use of source code management tools, and open-source methods in development of the software and control algorithms for the ITER Plasma Control System (PCS). It will also contribute to fusion-specific objectives such as validation of ITER pulse schedules prior to their use in experimental operation. We also describe a more mature but less sophisticated software suite known as TokSys, which was developed at General Atomics to support plasma control development and is expected to eventually merge with PCSSP. Plasma control development activities that use or have used these software systems are also described. We will discuss the ongoing ITER PCS control algorithm development for initial plasma operation, which takes advantage of the existing PCSSP infrastructure and library of modules. Plasma control development and initial plasma operation of the EAST and KSTAR tokamaks relied heavily on the TokSys suite, which is now in routine use at EAST, KSTAR, DIII-D, and NSTX-U. © 2015 IEEE.

ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e. g., current profile regulation, tearing mode (TM) suppression), control mathematics (e. g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e. g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER. (C) 2015 AIP Publishing LLC.