Search

Abstract: In Europe the work on the specification and design of a demonstration power plant (DEMO) is being carried out by EFDA in the power plant physics and technology (PPP&T) programme. DEMO will take fusion from experimental research into showing the potential for commercial power generation. During the fusion reaction, components in the tokamak become highly activated. The estimated dose rate levels after shutdown (zero decay time) due to 60dpa accumulation in steel (blanket) and 30dpa (divertor) are 13.1–17.4kGy/h (blanket); 8.8–11.6kGy/h (divertor) [1], much higher than those to be encountered at ITER. Upon removal from the tokamak, components would be transported to the hot cell facility with attention to minimizing the spread of activated dust and tritium contamination. It is proposed to use a sealed cask of ∼20tonnes, running on air castors with 50% lifting capability redundancy. Due to the number and complexity of the routes taken by this transporter it would have to be an un-tethered semi-autonomous system. This poses some technical challenges, including providing sufficient battery capacity, reliable guidance and a fail safe un-tethered control system. The mass of the components being moved is assumed here to range from a few tonnes to in excess of one hundred tonnes. Before the removed in-vessel components can be processed in the hot cell, they would require a period of cooling, approximately 2.5 years, to allow dose rate and decay heating to reduce. This reduces the decay heating level to ∼1–1.5kW/m3 and a contact dose rate to ∼250Sv/h, which is more suited to dexterous man-in-the-loop remote handling (RH). During maintenance, many components would require replacement or refurbishment. The hot cell facility would have to provide all the associated RH functions and operate fully remotely. With no human intervention, an accordingly robust RH recovery system would be required. This paper describes the first steps being taken towards the design of the DEMO hot cell. It will show a comparison of the current DEMO in-vessel maintenance concepts from a hot cell perspective, describe a proposed ex-vessel transport system, and summarize the facilities that have been identified as required within the hot cell, examine current RH technology and discuss the identified critical development issues. [Copyright &y& Elsevier]

Abstract: The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. It is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I p =5.5MA. In late 2007 the BA Parties, prompted by cost concerns, asked the JT-60SA Team to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility. Subsequently the Integrated Project Team has undertaken a machine re-optimization followed by engineering design activities aimed to reduce costs while maintaining the machine radius and plasma current. This effort led the Parties to the approval of the new design in late 2008 and hence final design and procurement activities have commenced. The paper will describe the process leading to the re-baselining, the resulting final design and technical solutions and the present status of procurement activities. [Copyright &y& Elsevier]

Reducing plasma flow clearly decreases the stability of tearing modes in multiple regimes (sawtooth, hybrid) in both high- and low-aspect-ratio tokamaks (DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)], Joint European Torus [M. Keilhacker and the JET Team, Plasma Phys. Controlled Fusion 41, 301 (1999)], National Spherical Torus Experiment [M. Ono, S. M. Kaye, Y.-K. M. Peng, Nucl. Fusion 40, 557 (2000)], each with distinct means of lessening rotation). Further, reducing flow makes pre-existing “saturated” islands larger at the same beta (β). Thus lower plasma flow impairs high-beta operation owing both to the destabilization and to the impact of tearing-mode islands. Experimental results suggest that flow shear (not flow) at the tearing rational surface is classically stabilizing, making the effective tearing stability index Δ′ of the total current density profile more negative (more stable). In this picture, with profiles and all else the same, the minimum metastable beta at which neoclassical tearing modes (NTMs) can destabilize is proportional to -Δ′ and hence lower flow and flow shear lead to possible destabilization (depending on seeding) at lower beta. Similarly, if destabilized, the saturated NTM island width is proportional to -β/Δ′ and thus increases as flow and flow shear are reduced. A working model gives a significant level of stabilizing shear if the plasma toroidal angular flow shear -dΩ[lowercase_phi_synonym]/dr at a given rational surface is of order of the inverse of the product of the local values of the parallel magnetic shear length Ls and the Alfvén time τA. Experimental data are fitted for the effect of this normalization of flow shear in a simple empirical model for both onset and saturation of tearing modes. Most theoretical literature is on the consequence of flow shear on tearing stability at zero beta; tokamaks at high beta have large magnetic Prandtl number (an issue for the sign of the flow effect) and very large Lundquist number. It is in this regime that theory will be compared with experimentally based empirical models. The consequence for future tokamaks with low rotation may be lower tearing stability than now expected. [ABSTRACT FROM AUTHOR]

Abstract: The exploitation of ITER tokamak will require diagnostics for machine protection, inputs to plasma control systems, evaluation and analysis of plasma parameters and performances. The equatorial visible/infrared wide angle viewing system and the radial neutron camera are the two main diagnostics of Procurement Package 11 (PP11), one of the diagnostic procurements under the responsibility of Europe, which also contains Equatorial Port Plug 1 and seven other diagnostics supplied by Europe or other ITER partners and integrated in the same Port Plug. Significant progress has recently been made in the development of the equatorial visible/infrared wide angle viewing system and the radial neutron camera. This paper gives an overview of the major technical achievements on these two diagnostics and points out the urgent needed R&D activities. [Copyright &y& Elsevier]

Abstract: In the framework of the Broader Approach Agreement, Europe is involved in the design activities for the Japanese Tokamak JT-60SA, investigating, among several issues, the operation of the superconducting TF magnets and their subsystems, aimed at the reactor conceptual design definition. In particular, one of the main critical aspects to study is the heating of the conductor due to both direct component of energy deposited by neutrons and by secondary gamma generated during plasma operation, and heat generated by the radiation on casing and transferred to the winding pack. Indeed, the operating temperature and the relevant temperature margin (i.e. the operating safety margin) of the magnet will depend strongly on the heat loads and on the capability of the coolant to remove it. Furthermore, the heat power to the conductor will depend on several aspects, namely the thickness of insulating material, the mass flow rate of helium flowing in the conductors and its thermodynamic properties at operating conditions, and the layout of the superconductors constituting the winding. Moreover, a crucial aspect in the final design will be the presence and position of the casing cooling channels. In this paper a 2D sensitivity analysis of heat transfer from casing to winding pack with respect to cooling channels number and position is presented, based on the reference layout of the magnet. As a result, we evaluated the optimum number and positioning of cooling channels needed, as a trade-off between magnet operating limits and available cryogenic power and if, at limit, they could be even neglected in normal operation, keeping dwell-time within reasonable values. [Copyright &y& Elsevier]

Abstract: The ITER neutral beam (NB) system presents several challenges and a robust program is necessary in order to achieve the requirements within the tight constraints due to the ITER construction plan. The establishment of full scale NB test facilities (NBTF) has therefore become a centre piece of the international NB development strategy. This paper describes the status of the design activities that have been undertaken in Europe to develop the components for the heating NB injectors along with the main plans and results of the R&D activities. A description of the programme towards the establishment of the test facilities and the planned activities is also reported. [Copyright &y& Elsevier]

Abstract: 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 ∼10KGy/h for IVVS down to ∼1Gy/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). [Copyright &y& Elsevier]

Abstract: The scope of this paper is a preliminary assessment of the maintenance scheme in support of the European study for the next generation of fusion reactor: DEMO. Despite other fusion machine requiring remote handling maintenance operations, DEMO is supposed to work under steady state operational conditions. Therefore, requirement on the maintenance scheme is stronger. To target a good availability of the machine along machine operation plan, it is necessary to draw an adequate maintenance scheme. Indeed, due to the high fluxes generated by the plasma in the vacuum vessel, the first wall lifetime is limited, so the frequent replacement is necessary. On current fusion experimental machine, as first wall load conditions are less severe, DEMO condition implies high level of requirement on maintenance time. During DEMO lifetime, several full first wall replacements are expected. To provide access to the vacuum vessel machine for first wall removal, preparatory work is required to set the machine to adequate maintenance conditions and to open the machine properly, the same situation at the end of the maintenance period. Shutdown duration for first wall replacement should be as short as possible to reach the availability target of the machine. From this statement, the maintenance duration should not exceed 20% of the total lifetime of the reactor operation. First wall segmentation (i.e. total number of component to replace) has a high impact onto the replacement time. Considering the number of feasible designs for the first wall segmentation, we concentrate remote handling concept assessments one type of segmentation, the one minimizing the numbers of modules to replace . Assumption on Divertor segmentation for these DEMO studies have similarities with Divertor ITER design; therefore ITER design output is relevant . We assume divertor removal performed in shadow time, while removing the other first wall modules. [Copyright &y& Elsevier]

Abstract: High heat flux loaded components which will be installed in the ITER Divertor require a heat flux removal capability in the range 5–10MW/m2 at steady-state and up to 20MW/m2 in transients. Within the ITER plasma facing components procurement context, each party should demonstrate its technical capability to carry out the manufacturing with the required quality. This is achieved through the successful manufacturing and testing of medium-size qualification prototypes. Each Qualification Prototype consists of three high heat flux units mounted onto an actively cooled supporting structure. Currently, the SATIR method has been identified by the ITER Organization as the basic test to decide upon the final acceptance of the ITER Divertor components. SATIR testing was performed on each CFC part of European HHF units prior to the insertion of the twisted tape and prior to assembling the units onto the steel support structure. The paper deals with SATIR results of all qualification prototypes manufactured by European industry. [Copyright &y& Elsevier]

Abstract: Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R&D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER. [Copyright &y& Elsevier]

Abstract: The European Joint Undertaking for ITER and the Development of Fusion Energy or (“Fusion for Energy” of F4E for short) is a new organisation that has been established with the main objective of providing Europe''s contribution to the ITER International Organisation (IO) as its Domestic Agency. Fusion for Energy is also the Implementing Agency for the Broader Approach projects being carried out with Japan and, in the longer term, will prepare a programme for the construction of demonstration fusion reactors (DEMO). The threefold mission of Fusion for Energy is consistent with the fast track strategy for the realisation of fusion energy. This paper aims to provide an overview of the current status of Fusion for Energy and highlight some of the opportunities available for research organisations and industry to participate. [Copyright &y& Elsevier]

Abstract: Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper. [Copyright &y& Elsevier]

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised. [ABSTRACT FROM AUTHOR]

Perturbations induced by edge density fluctuations on electron cyclotron wave beams propagating in fusion plasmas are studied by means of a quasi-optical code. The effects of such fluctuations are illustrated here by showing the beam propagation in the case of single harmonic perturbations to the wave phase front. [ABSTRACT FROM AUTHOR]