Fusion Engineering

The ARIES-AT power core was evolved with the overall objective of achieving high performance while maintaining attractive safety features, credible maintenance and fabrication processes, and reasonable design margins as a rough indication of reliability. The blanket and divertor designs are based on Pb-17Li as coolant and breeder, and low-activation SiC f /SiC as structural material. Flowing Pb-17Li in series through the divertor and blanket is appealing since it simplifies the coolant routing and minimize the number of cooling systems. However, Pb-17Li provides marginal heat transfer performance in particular in the presence of MHD effects and the divertor design had to be adapted to accommodate the peak design heat flux of 5 MW/m 2 . The blanket flow scheme enables operating Pb-17Li at a high outlet temperature (about 1100 • C) for high power cycle efficiency while maintaining SiC f /SiC at a substantially lower temperature consistent with allowable limits. Waste minimization and additional cost savings are achieved by radially subdividing the blanket into two zones: a replaceable first zone and a life of plant second zone. Maintenance methods have been investigated which allow for end-of-life replacement of individual components. This paper summarizes the results of the design study of the ARIES-AT power core focusing on the blanket and divertor and including a discussion of the key parameters influencing the design, such as the SiC f /SiC properties and the MHD effects, and a description of the design configuration, analysis results and reference operating parameters.

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.

The company JEMA has delivered to the Joint European Torus (JET facility in Culham) two high voltage switching mode power supplies (HVSMPS) each rated 130 kVdc and 130 A. One HVSMPS feeds the grids of two PINI loads. This paper describes the main control issues and the algorithms developed for the project. The most demanding requirements from the control point of view is an absolute accuracy of ±1300 V and the possibility of performing up to 255 re-applications of the high voltage during a 20 s pulse. Keeping the output voltage ripple to the specified tolerance has been a major achievement of the control system. Since the output stage is formed of several modules (120) connected in series, their stray capacitance to ground significantly influences the individual contribution of each single module to the global output voltage. Two complementary techniques have been used to balance the effects of the stray capacities. The fast re-applications requirement has a significant impact on the intermediate dc link. This section is composed of a capacity of 0.83 F, which feeds the 120 invertor modules. The dc link is fed by a 12 pulse SCR rectifier, whose matching transformers are connected to the 36 kV grid. Every re-application and every voltage shutdown supposes a quasi-instantaneous power step of 17 MW. Fast open loop algorithms have been implemented in order to keep the dc link voltage within acceptable margins. Moreover, the HVSMPS output characteristics have to be maintained during the rapid and important voltage fluctuations of the 36 kV mains (28-37 kV). The general control system is based on a Simatic S7 PLC, and a SCADA user interface. Up to 1000 signals are acquired. The control system has shown to be also a useful tool to allow for a rapid and accurate identification of faults and their origin.

Data logging and data distribution will be two main tasks connected with data handling in ITER. Data logging refers to the recovery and ultimate storage of all data, independent of the data source. Data distribution is related, on the one hand, to the on-line data broadcasting for immediate data availability and, on the other hand, to the off-line data access. Due to the large data volume expected, data compression is a useful candidate to prevent the waste of resources in communication and storage systems. On-line data distribution in a long-pulse environment requires the use of a deterministic approach to be able to ensure a proper response time for data availability. However, an essential feature for all the above purposes is to apply lossless compression techniques. This article reviews different lossless data compression techniques based on delta compression. In addition, the concept of cyclic delta transformation is introduced. Furthermore, comparative results concerning compression rates on different databases (TJ-II and JET) and computation times for compression/decompression are shown. Finally, the validity and implementation of these techniques for long-pulse operation and real-time requirements is also discussed.

The ITER toroidal field model coil ( A. Ulbricht et al. / Fusion Engineering and Design 73 (2005) the leadership of European Fusion Development Activity/Close Support Unit (EFDA/CSU), Garching, in collaboration with the European superconductor laboratories and the European industry. The TFMC was developed and constructed in collaboration with the European industry consortium (AGAN) and Europa Metalli LMI supplied the conductor. The TFMC was tested in the test phase I as single coil and in phase II in the background field of the EURATOM LCT coil in the TOSKA facility of the Forschungszentrum Karlsruhe. In phase I, the TFMC achieved an ITER TF coil relevant current of about 80 kA and further representative test results before the end of the EDA. In the more complex test phase II, the coil was exposed to ITER TF coil relevant mechanical stresses in the winding pack and case. The tests confirmed that engineering design principles and manufacturing procedures are sound and suitable for the ITER TF full size coils. The electromagnetic, thermo hydraulic and mechanical operation parameters agree well with predictions. The achieved Lorentz force on the conductor was about 800 kN/m. That has been equivalent to the Lorentz forces in ITER TF coils.

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&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.

The ITER instrumentation and control (I&C) system is the term encompassing all hardware and software required to operate ITER. It has two levels of hierarchy: the central I&C systems and the plant systems I&C. The central I&C systems comprise CODAC (Control, Data Access and Communication), the central interlock system (CIS) and the central safety systems (CSS). The central I&C systems are "in-fund", i.e. procured by ITER Organization (IO), while plant systems I&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&C, poses a major challenge for the realization and integration of the ITER I&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&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&C system is to provide a fully integrated and automated control system for ITER. Standardization of plant systems I&C is of primary importance and has been the highest priority task during the last year. The target of associated R&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&C.

The reversed shear plasma configuration is examined as the basis for a tokamak fusion power plant. Analysis of plasma equilibrium and ideal MHD stability, bootstrap current and current drive, plasma vertical stability and position control, divertor physics and plasma power balance are used to determine the operating point parameters that maximize fusion power density and minimize the recirculating power fraction. The final plasma configuration for the ARIES-RS power plant obtains i of 4.96%, plasma driven current fraction of 91%, plasma current of 11.3 MA, toroidal field of 8.0 T and major and minor radius of 5.5 and 1.4 m. The current drive system utilizes fast wave, lower hybrid and high frequency fast wave current drive to obtain maximum current profile flexibility, requiring 580 MW of power. A divertor solution is found which employs neon impurity injection to enhance the radiation in the scrape-off layer (SOL) and divertor and results in a combined particle and heat load in the divertor of 5 6 MW m − 2 . The plasma is driven with a Q of 25 and is at a thermally stable operating point. The plasma is assumed to be in an ELMy H-mode, with low amplitude and high frequency ELMs. © 1997 Elsevier Science S.A. All rights reserved.

The RF vacuum window design proposed for ITER is similar to the one in operation on JET. It uses titanium alloys for the outer and inner conductors and double conical ceramic insulators. However, unlike the JET window, beryllia is the dielectric material. The design is complicated considerably by the need for continuous RF operation, which necessitates active cooling. Residual stress after manufacture has been optimised with particular attention given to subcritical crack growth in the ceramic. In addition, finite element methods were used to model the steady state temperature gradient from the combined effects of volumetric heating of neutron irradiation, RF surface heating and dielectric loss heating. A scoping study is included in the modelling methodology of bonded joints and on the hydraulic requirements of the window cooling system.

The EU will be providing the largest contribution to the ITER electron cyclotron (EC) heating and current drive (H&CD) system (20 MW, CW at 170 GHz). The contribution includes one third of the H&CD gyrotrons, their associated power supplies and four upper port launcher antennas. In all areas of participation, the EU EC partnership (coordinated by the European Fusion Development Agreement) aims toward advancing the technology, while staying within a specified cost envelope. This is portrayed in the co-axial gyrotron development that offers the potential to double the output power per source (2.0 MW), increasing the delivered power for a fixed number of auxiliary systems. The EU partnerships also attempt to increase performance for the entire EC system, in particular the launching antennas. The proposed front steering launcher design offers greater control of MHD activity than the previous remote steering design and opens up the possibility of an enhanced performance UL. The EC physics M. A. Henderson et al. / Fusion Engineering and Design 82 (2007) 454-462 455 requirements are repartitioned between the upper and equatorial launchers for a synergetic balance, which increases the EC physics capabilities while relaxing some of the engineering requirements.

In the present study a complete neutronic analysis has been performed for the current design of the JT-60SA toroidal field coil (TFC) system. The MCNP5 Monte Carlo code has been used to calculate the nuclear heating, neutron spectra and absorbed dose in the TFC components, assuming a DD neutron emission rate of 1.5 × 10 17 n/s (and 1% DT). Nuclear heating of the winding pack is lower than 0.3 mW/cm 3 and the maximum nuclear heating of the TFC case is 0.4 mW/cm 3 . The overall nuclear heating, including the safety margin, is less than 8 kW. Spatial distribution of the nuclear heating has been provided along poloidal, radial and toroidal directions as to be used for thermo-hydraulic analysis and the design of TFC system. The absorbed dose to insulator is as low as to avoid the replacement during the whole life of the machine. Neutron fluxes have been used as input for a preliminary activation analysis performed with FISPACT inventory code. Activity and contact dose rates have been calculated at different cooling times, after 10 years of operations in some representative zone of the winding pack and the case. All the TFC materials can be easily recycled within the first day after shutdown and the hands-on recycling is possible within less than 30 years.

Fusion nuclear technology (FNT) research in the United States encompasses many activities and requires expertise and capabilities in many different disciplines. The US Enabling Technology program is divided into several task areas, with aspects of magnet fusion energy (MFE) fusion nuclear technology being addressed mainly in the Plasma Chamber, Neutronics, Safety, Materials, Tritium and Plasma Facing Component Programs. These various programs work together to address key FNT topics, including support for the ITER basic machine and the ITER Test Blanket Module, support for domestic plasma experiments, and development of DEMO relevant material and technological systems for blankets, shields, and plasma facing components. In addition, two inertial fusion energy (IFE) research programs conducting FNT-related research for IFE are also described. While it is difficult to describe all these activities in adequate detail, this paper gives an overview of critical FNT activities. (N.B. Morley). ponents, systems and technologies of the plasma chamber that are required to contain, shield, extract energy from, and breed tritium fuel for the thermonuclear fusion plasma. FNT advances will be needed both for near-term magnetic (MFE) and inertial (IFE) fusion energy experiments, and ultimately for MFE and IFE energy-producing power reactors. An incomplete list of FNT components and systems includes: 0920-3796/$ -see front matter

Neutronics and nuclear data have an essential role in the development programme leading to the realization of a fusion power plant. A well-qualified nuclear database and validated computational tools are required in the nuclear design of fusion devices for reliable neutronics and activation calculations with the associated uncertainties. The EU is conducting a large effort on neutronics and nuclear data

Fusion Advanced Studies Torus (FAST) aims to contribute to the exploitation of ITER and to explore innovative DEMO technology. FAST has been designed to study, in an integrated scenario: (a) relevant plasma-wall interaction problems, with a large power load (P/R ∼ 22 MW/m; P/R 2 ∼ 12 MW/m 2 ) and with a full metallic wall; (b) to tackle operational problems in regimes with relevant fusion parameters; (c) to investigate the non-linear dynamics of fast particles (alpha like) in burning plasmas. FAST will operate on a wide parameters range, namely in high performance H-mode (BT ∼ 8.5 T; IP ∼ 8 MA) as well as in advanced Tokamak operation up to full non-inductive current scenario (IP ∼ 2 MA). The main heating is based on 30 MW ICRH, but the ports have been designed to allocate up to 20 MW of 1 MeV NNBI. Helium gas at 30 K is used for cooling of the full machine, a preliminary analysis shows the possibility of realizing FAST with a complete superconductor set of coils. An innovative active system is under development to reduce and to control the magnetic ripple. Tungsten (W) or liquid lithium (L-Li) has been chosen for the divertor material plates and the code EDGE2D has been used to optimize the divertor geometry.

Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840 MW of fusion power of which 767 MW is in the form of energetic particles (alpha power) and 3073 MW is in the form of neutrons. The alpha plus auxiliary power total 909 MW of which 430 MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

The availability specifications in IFMIF project are very high to reach it. A RAM planning has been proposed in order to develop this task. As a first step, a critical review of the previous RAM activities developed during the last 10 years in the framework of the IFMIF program is made. In this first step the software tools available for RAM studies are analyzed and it is proposed that RiskSpectrum should be used in the future in the IFMIF project. It is an interactive tool for reliability and safety analysis and it allows a complete organization analysis and presentation of risk and reliability information. As a first validation example, the target facility has been modelled using this code in order to benchmark the previous RAM results calculated by Markov chains. The results obtained are according to the previous RAM analysis carried out in conceptual design activity. Also, the need both of contrasted failure rate data and of developing an updated RAM analysis including last changes in the IFMIF design project is showed as compulsory.

Development of effective and reliable coatings is a key to the viability of most, if not all, fusion blanket systems. The specific purpose and requirements of the coatings vary widely, depending on the blanket concept. The efforts on coating development to date have focused primarily on electrically insulating coatings for the self-cooled lithium concepts with a vanadium alloy structure, and the tritium barrier coatings for the water-cooled, Pb Á/Li (WCLL) breeder concepts with ferritic steel structures. Although other coating materials are under consideration, most of the effort on the electrically insulating coatings has focused on CaO and AlN coatings on V Á/4Cr Á/4Ti alloy structure. Most of the effort on the tritium barrier coating development is focused on Al 2 O 3 coatings formed on aluminized ferritic steels. This paper presents an overview of the status of coating development for the various fusion concepts with emphasis on the materials interaction and chemistry control issues associated with the formation, stability and performance of the coatings. # (D.L. Smith). Fusion Engineering and Design 61 Á/62 (2002) 629 Á/641

All the damages that are not reflected in the market price are called external costs. The external costs of fusion were elaborated with the ExternE methodology. The external costs are in the range of a few mEuro/kWh depending on the plant model. The external costs are in general not dominated by the impacts due to radioactive emissions and releases. All stages of the life cycle contribute significantly to the external cost value. The external costs of fusion are in the same range as the external costs of photovoltaics and wind energy.

NSTX is a proof-of-principle experiment aimed at exploring the physics of the 'spherical torus' (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, among other advantages. The low aspect ratio (R/a, typically 1.2 -2 in ST designs compared to 4 -5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called 'center stack' in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.

The water-cooled lithium-lead (WCLL) blanket is based on reduced-activation ferritic-martensitic steel as the structural material, the liquid alloy Pb-17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept.

Future fusion experiments, aiming to demonstrate steady state reactor operation, require physics driven plasma control based on increasingly complex plasma models. A precondition for establishing such control systems is widely automated data analysis, which can provide integration of multiple diagnostic on a large scale. Even high quality online data evaluation, which is essential for the scientific documentation of the experiment, has to be performed automatically due to the huge data sets being recorded in long discharge runs. An automated system that can handle these requirements will have to be built on reusable software components that can be maintained by the domain experts: diagnosticians, theorists, engineers and others. For Wendelstein 7-X a service oriented architecture seems to be appropriate, in which software components can be exposed as services with well defined interface contracts. Although grid computing has up to now been mainly used for remote job execution, a more promising service oriented middleware has emerged from the recent grid specification, the open grid service architecture (OGSA). It is based on stateful web services defined by the web service resource framework (WSRF) standard. In particular, the statefulness of services allows to setup complex models without unnecessary performance losses by frequent transmission of large and complex data sets. At present, the usability of this technology in the W7-X CoDaC context is under evaluation by first service implementations.

The power plant conceptual study (PPCS) carried out in Europe between 2001 and 2004, aimed at the demonstration of the credibility, the safety and environmental advantages and the potential economic viability of future fusion power reactors. A set of requirements was issued by industry and utilities concerning safety, operational and economic aspects. In this framework, the helium-cooled lithium-lead (HCLL) reactor concept has been studied, which was based on limited extrapolation in both present physics and technology. The HCLL blanket is considered in Europe as a possible concept for DEMO. The net electric power was set at about 1500 MW e. The maintenance scheme considered is based on the removal of blanket modules through the equatorial ports and the removal of divertor cassettes through the lower ports of the vacuum vessel. This scheme shows the potential for good overall plant availability (75%). The balance of plant has been studied, taking into account the following systems: the primary heat transport system, the lithium lead circulation system and the secondary heat transfer system. A safety analysis has been carried out. A set of envelope accident sequences was identified using a functional failure mode and effects analysis (FFMEA) methodology and the deterministic assessment of the scenarios was carried out using best estimate computer codes. It is concluded that the HCLL reactor concept meets the overall objectives of the PPCS.

This paper presents the non-destructive testing of brazed joints between the cooling tube and the heat sink by infrared (IR) thermography, which is becoming a useful and recognised technique to evaluate the quality of joints. A sensitive IR camera generates an image of the surface based on the temperature of each point on the surface and sufficient to evaluate the overall quality of the thermal contact of brazed joints. In steady state superconducting Tokamak (SST-1), the plasma facing components (PFC) are actively cooled during plasma operation. The cooling tubes are brazed on the heat sink. The IR image is used to visualize the heat transfer between the cooling tube and the heat sink, which enables to identify the significant faults in thermal contact. Three-dimensional (3D) thermal finite element (FE) analyses have been performed to simulate the brazing defects. The experimental observations obtained from IR thermography have confirmed the FE simulations. Additional non-destructive investigations such as X-ray images of the test section also have confirmed the experimental observations.

The TJ-II remote participation system (RPS) is focused on providing remote access to elements that depend exclusively on characteristics of the TJ-II environment: data acquisition, diagnostics control systems and TJ-II operation tracking. Four key points were taken into account prior to starting the software design: access security, software execution platforms, software maintenance and distribution and delivery of operation events. The first, access security, was addressed by means of a distributed authentication and authorization system, PAPI. Regarding the other points, the development was based on the use of web servers (due to their standard character, flexibility and scalability) and Java technologies (due to their open nature, security properties and technological maturity). Software deployment was prepared to make use of the Java Network Launching Protocol (JNLP). On-line message distribution was planned according to a message oriented middleware. At present, the TJ-II RPS manages over 1000 digitization channels and 20 diagnostic control systems. The TJ-II RPS architecture is flexible, scalable and powerful enough to be applied to distributed environments and, in particular, it could be used in the ITER environment.

Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials and components have provided validation of nuclear data used in shielding and activation calculations for ITER, as well as an assessment of the related uncertainties. Further validation activity is still needed which could be performed in ITER, especially for activation cross sections, for dose rate calculations and for tritium production rates. In particular, concerning the nuclear performance of breeder blankets to be tested in ITER, the paper discusses the extensive preparatory work, to be carried out in a coordinated way among the participating Parties, required to finalize the design of neutronics Test Blanket Modules (TBMs), so that these tests can provide a complete and comparable, therefore useful, picture of the different concepts. Neutronics experiments being carried out at neutron generators on TBM mock-ups are providing important results on the quality of relevant neutron cross sections and for the preparation of the measurement techniques.

Within the European power plant conceptual study (PPCS) four fusion power plant "models", all based on the tokamak concept, have been developed. Two of these models, A and B, were developed considering limited extrapolations both in physics and in technology. PPCS model A is based on the WCLL blanket [M. Gasparotto et al., DEMO blanket technology R&D results in the EU, Fusion Eng. Des. 61-62 (2002) 263-271] and on a water-cooled divertor concept. PPCS model B is based on the HCPB blanket [M. Gasparotto et al., DEMO blanket technology R&D results in the EU, Fusion Eng. Des. 61-62 263-271] and on a helium-cooled divertor concept. For each of the more advanced concepts, models C and D, an advanced physics scenario has been identified and combined with advanced blanket concepts that allow higher thermodynamic efficiencies of the power conversion systems. Two key innovative developments are worthy of especial note: the development of a scheme for the scheduled replacement of the internal components and the conceptual design for a helium-cooled divertor.

For JET to fulfil its mission in preparing ITER operation, the installation of an electron cyclotron resonance heating system on JET would be desirable. The study described in this paper has investigated the feasibility of installing such a system on JET. The principal goals of such a system are: current drive over a range of radii for NTM stabilization, sawtooth control and current profile tailoring and central electron heating to equilibrate electron and ion temperatures in high performance discharges. The study concluded that a 12 gyrotron, 10 MW, system at the ITER frequency (170 GHz) adapted for fields of 2.7-3.3 T would be appropriate for the operation planned in JET. An antenna allowing toroidal and poloidal steering over a wide range is being designed, using the ITER upper launcher steering mechanism. The use of ITER diamond windows and transmission line technology is suggested while power supply solutions partially reusing existing JET power supplies are proposed. Detailed planning shows that such a system can be operational in about 5 years from the time that the decision to proceed is taken. The cost and required manpower associated with implementing such a system on JET has also been estimated.

Tokamaks and Stellarators generate heat loads that are pulsating in nature. The helium refrigerators employed for cooling these fusion devices are, however, designed for a steady heat load. Unless they are appropriately adapted and controlled, process parameters under pulsed load may vary to such an extent which may lead of plant instability. In this paper, performance of a J-T stage as a part of modified Claude cycle based helium refrigerator has been analyzed subjecting it to a pulsed heat load for understanding of the behavior of the components. Pulsed heat load results in the fluctuation of mass flow rate in the return stream that needs to be mitigated. The analyses focus both on the equipment and their interactions in the cycle and a concept of parallel heat exchangers has been applied. There is a steady decrease in liquid level in Dewar when the plant is designed with a capacity equal to the time-averaged heat load. Introduction of parallel heat exchanger mitigates mass flow rate fluctuation by 15% with only one additional heat exchanger at the last stage. There is promise for a higher mitigation effect when more heat exchangers are used in parallel in more number of stages. The study may be extended to entire helium plant used in fusion devices.

The vacuum vessel (VV) design is being finalized including interface components, such as the support rails and feedthroughs of coils for mitigation of edge localized modes (ELM) and vertical stabilization (VS) of the plasma (ELM/VS coils). It was necessary to make adjustments in the locations of the blanket supports and manifolds to accommodate the design modifications in the ELM/VS coils. The lower port gussets were reinforced to keep a sufficient margin under the increased VV load conditions. The VV support design is being finalized as well, with an emphasis on structure simplification. The design of the inwall shielding (IWS) has progressed, considering assembly and required tolerances. The layout of ferritic steel plates and borated steel plates will be optimized based on on-going toroidal field ripple analysis. The VV instrumentation was defined in detail. Strain gauges, thermocouples, displacement meters and accelerometers shall be installed to monitor the status of the VV in normal and off-normal conditions to confirm all safety functions are performed correctly. The ITER VV design was preliminarily approved, and the VV materials including 316L(N) IG were already qualified by the Agreed Notified Body (ANB) according to the procedure of Nuclear Pressure Equipment Order.

Within the European Union, the two major breeding blanket concepts presently being developed are the helium cooled pebble bed (HCPB), and the helium cooled lithium lead (HCLL) blankets. For both concepts, different conceptual designs are being discussed with temperature windows in the range 250-550 • C for conservative approaches based on reduced activation ferritic-martensitic (RAFM) steels, and in the range 250-650 • C for more advanced versions, taking into account oxide dispersion strengthened (ODS) steels. As a final result of a systematic development of RAFM-steels in Europe, the 9% CrWVTa alloy EUROFER was specified and produced in an industrial scale with a variety of product forms. A large characterisation program is being performed including irradiation in materials test reactors between 60 and 450 • C (≤15 dpa), and in a fast breeder reactor at 330 • C up to 30 dpa. EUROFER is resistant to high temperature ageing, and the existing creep-rupture data (∼30,000 h, 450-600 • C) indicate long-term stability and predictability. R. Lindau et al. / Fusion Engineering and Design 75-79 (2005) [989][990][991][992][993][994][995][996] The ODS variant of EUROFER shows superior tensile and creep properties compared to EUROFER. Applying a new production route has diminished the problem of lower ductility and inferior impact properties. A reliable joining technique for ODS and RAFM steels employing diffusion welding was successfully developed.

The largest pulsed superconducting coils ever built, the Central Solenoid (CS) Model Coil and Central Solenoid Insert Coil were successfully developed and tested by international collaboration under the R&D activity of the International Thermonuclear Experimental Reactor (ITER), demonstrating and validating the engineering design criteria of the ITER Central Solenoid coil. The typical achievement is to charge the coil up to the operation current of 46 kA, and the maximum magnetic field to 13 T with a swift rump rate of 0.6 T/s without quench. The typical stored energy of the coil reached during the tests was 640 MJ that is 21 times larger than any other superconducting pulsed coils ever built. The test have shown that the high current cable in conduit conductor technology is indeed : S 0 9 2 0 -3 7 9 6 ( 0 1 ) 0 0 2 3 5 -6

A Lower Hybrid Current Drive (LHCD) system would be the prime method to drive current in SST-1. The LHCD system is based upon two TH2103D klystrons of M/s Thomson CSF at 3.7 GHz capable of delivering 500 kW CW microwave power. The klystrons have been commissioned and deliver the microwave power through two standard WR-284 waveguides. It is electrically isolated from the main system using DC breaks. Two high power circulators protect the tube from high reflection. Directional couplers measure forward and reflected power in the system. The output power is deposited on two matched dummy loads during testing. In the event of arcing or any other fault, the high voltage from the klystron is removed in less than 5 ms, with the help of a crowbar system, based on a pressurized rail gap switch, for the protection of the klystron. High power (up to 200 kW), 1000-s tests of two klystrons have been conducted and detailed test results are described in this paper. It also describes the high power microwave system at 3.7 GHz that is being used to test the high power microwave components, being designed and fabricated for use in the LHCD system on SST-1.

NSTX is a proof-of-principle experiment aimed at exploring the physics of the 'spherical torus' (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, among other advantages. The low aspect ratio (R/a, typically 1.2 -2 in ST designs compared to 4 -5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called 'center stack' in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.

The main elements present in the explosives are hydrogen (H), nitrogen (N), oxygen (O) and carbon (C). The elemental density ratios (C/O, N/O, H/N, etc.) of explosives are different from the other substances. The explosives are characterized by low H and C concentration and high N and O concentration. The elemental composition of explosive materials can be found by measuring the ␥-rays produced due to the activation of explosive material by 14 MeV neutrons. A pulsed fast neutron system is used for the identification of H, N and O. Ammonium nitrate sample has been irradiated by pulsed 14 MeV neutrons of 10 8 neutrons/pulse and the pulse width less than a microsecond. The material has been irradiated with 5-7 neutron pulse in order to increase the ␥-ray count. The resulting ␥-ray spectrum due to the activation of ammonium nitrate is recorded and presented in this work. The ␥-ray spectrum has been recorded using 3 × 3 Bismuth Germanium Orthosilicate (BGO) and fast digitizer with sampling rate of 2 GS/s. The signal is transferred through fiber optic cable to the personal computer for further processing and analyzing. Pulse-height spectrum is constructed through a dedicate MATLAB routine in which the number of energy bins can be kept as per requirement.

Wendelstein 7-X (W7-X) is a fully optimized low-shear stellarator and shall demonstrate the reactor potential of this fusion plant. It is presently under construction at the Greifswald Branch Institute of IPP. The superconducting magnet system will allow continuous operation, limited only by the plasma exhaust system whose capacity is designed for 30 min full power operation. The Wendelstein 7-X (W7-X) coils and structures are part of the largest superconducting fusion device being constructed at present. They represent a technical challenge at industrial level and the need for proven techniques and manufacturing processes in accordance to the highest quality standards. The production of these components requires a management of monitoring for quality and tests. The coil system consists of 20 planar and 50 non-planar coils. They are supported by a pentagonal 10 m diameter, 2.5 m high coil support structure (CSS). The CSS is divided into five modules. Each module consists of two equal half modules. The manufacturing status of the CSS and the main project management and technical challenges will be presented. The lessons learned in the large scale production of this difficult kind of support structure will be presented as relevant experience for the realization of similar systems for future fusion devices, such as ITER.

A new transient recorder module architecture was developed to fulfil many of today's requirements of data acquisition for plasma diagnostics on fusion experiments. This architecture is supported by the availability of new high-density devices in the fields of digital signal processors and programmable logic devices, which can provide features such as multi-channel data readout in real-time, real-time digital signal processing and a large quantity of onboard memory. This paper describes the design and implementation of a transient recorder module in compliance with this new architecture, which, along with the developed software, can be efficiently used either as a stand-alone or integrated in a multi-unit data acquisition system. The module encloses all aforementioned capabilities in an eight-channel peripheral component interconnect (PCI) unit. All channels are differential, galvanic isolated at 1 kV and over-voltage protected. Acquisition rate is 2 M samples per second with 14-bit resolution. Local data storage capacity is 256 M samples.

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value ␦B T = 0.08% up to ␦B T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular,ˇN values above the 'no-wall limit' (ˇN ∼ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.

Packed fluidization is a novel technique, in which small particles are allowed to fluidize in the interstices of relatively large and stationary packing to enhance heat transfer rates of a unary packed bed of same size pebbles. In the present study, heat transfer in unary packed bed and binary packed fluidized bed were investigated and compared in terms of the effective thermal conductivity. In the experimental works, large pebbles (size: 3– 10 mm) of two different materials viz. lithium titanate and alumina and small particles (size: 231–780 μm), also of two different materials viz., lithium titanate and silica were used. It was found that the heat transfer in unary packed bed is enhanced due to the packed fluidization and in terms of the effective thermal conductivity; the enhancement was up to 260%. It was found that the volume fraction of small particles, operating gas velocity, particle to pebble size ratio and the type of materials have a significant effect in enhancing the effective thermal conductivity and heat transfer rates. It was also found that 60% (v/v) of small particles in the interstitial voids of packing pebbles can give much improved effective thermal conductivity keeping all other operating parameters same. Possible mechanism for heat transfer enhancement due to packed fluidization has been proposed. Based on the results for different particle and pebble sizes, materials and process variables viz. particles to pebble size ratios, operating gas velocity ratios, volume fraction of small particles in the interstices and bed wall temperatures, we could arrive at optimum conditions for determining effective thermal conductivity and a correlation to estimate the same.

ABSTRACT：The development strategy, related R&D activities based on the mapload of Chinese Fusion Power Plant (FPP) program are presented briefly. A conceptual design study of fusion DEMO plant based on the Helium-cooled/Solid Breeder/LAFMs (HCSB) concept has being carried out recently. The design progress, the basic structure outline, as well as the key issue of HCSB-DEMO fusion plant are introduced.

The ITER Neutral Beam Test Facility (NBTF) will be located in the area of the RFX Consortium, (Council of National Researches -CNR), in Padua, Italy. The ITER NB accelerates negative deuterium ions with maximum energy of 1 MeV and maximum beam current of 40 A. The production of tritium from this facility, mainly due to the D-D nuclear reaction on the injector calorimeter, might be a concern from the radiological safety point of view. In the present paper, tritium release in atmosphere is evaluated. Analysis was performed using three different computer codes. Different release scenarios were taken into account, and doses to reference groups (population and workers) were calculated. Our results show that the dose limit of 1 mSv/year stated by Italian law for the population is not exceeded around the facility and away from it. Even considering the most conservative scenario, the dose rate per year to the reference group is about three order of magnitude lower than the dose limit.

The tritium confinement strategy adopted during the past years in the ITER hot cell building is compared to the safety requirements given by the standard ISO-17873 "Nuclear facilities-criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors". In fact, this is the reference safety guideline recommended by French licensing authorities. Several features of the considered design of the hot cell building are not in agreement with these guidelines. Main discrepancies concern the zoning of the hot cell complex, the flow rates of ventilation, and the possibility to recycle the room atmosphere and to detritiate the effluent air. These aspects are discussed together with some proposed modifications of the design.