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We report on high efficiency multimode pumps that enable ultra-high efficiency high power ECO Fiber Lasers. We discuss chip and packaged pump design and performance. Peak out-of-fiber power efficiency of ECO Fiber Laser pumps was reported to be as high as 68% and was achieved with passive cooling. For applications that do not require Fiber Lasers with ultimate power efficiency, we have developed passively cooled pumps with out-of-fiber power efficiency greater than 50%, maintained at operating current up to 22A. We report on approaches to diode chip and packaged pump design that possess such performance.

The objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such information is essential to design and predict the performance of DEMO and future fusion reactors. It foresees to test six mock-ups of breeding blankets, called Test Blanket Module (TBM), in three dedicated ITER equatorial ports from the beginning of the ITER operation. The TBM and its associated ancillary systems, including cooling system and tritium extraction system, forms the Test Blanket System (TBS) that will be fully integrated in the ITER machine and buildings. This paper describes the main features of the six TBSs that are presently planned for installation and operation in ITER, the main interfaces with other ITER systems and the main aspects of the TBM Program management.

Tritium breeding blanket testing is an important element in the ITER mission. Up to six different concepts for tritium breeding blanket systems, referred to as Test Blanket Systems (TBS), will be tested in three equatorial ports of ITER. Successful TBS experiments in ITER represent an essential step on the path to DEMO for all the ITER Members’ fusion development plans. The ITER Members are in charge of the design, manufacturing and delivery of the TBSs to the ITER site. The IO has responsibility for preparing the necessary interfaces required for the installation of the TBSs. Moreover, the TBM Program has to be fully integrated in the ITER Research Plan and its testing objectives have to be synchronized with the planned ITER operations. The paper addresses the major implementation steps of the TBM Program in ITER, including the organizational aspects, its integration into the ITER Research Plan and the Operational Plan, the licensing procedure and also gives a short overview of the TBS/ITER interfaces issues.

Up to six mock-ups of different tritium breeding blanket systems, referred to as Test Blanket Systems (TBS), will be tested in three equatorial ports of ITER. The paper describes the recent studies performed by the IO for the preparation of the most urgent interfaces required for the integration of the TBSs into the ITER device and Tokamak complex. Main addressed items concern the impact of the TBM ferromagnetic structural material on plasma performance, the preliminary design of the TBM frame, the main dimensions and location for the TBM cooling systems and for the Tritium circuits, and preliminary requirement for measurement systems. A large effort has also been devoted to the proposed maintenance/refurbishment procedure and to the space and equipment requirements in the hot cell facility, including provisions for the shipping of TBM sub-components outside the ITER site for post-irradiation examinations.

A feasibility has been demonstrated for numerical reconstruction on the base of magnetic measurements for geometrical displacements or deformations occurred in the manufacture and assembly of magnet coils. For validation of the proposed approach the test results of reconstruction of possible misalignments and deviations of the ITER PF1 coil are presented.

This paper describes an important part of ITER Research Plan, on plasma–wall interaction (PWI). In order to maximize the flexibility of the machine during the initial operation (H and D phases), CFC will be used for the targets. Tungsten will be used for the other plasma-facing components of the divertor. In order to minimize the tritium retention, tungsten will fully cover the divertor targets before the DT phase. Extrapolation of heat loads on plasma-facing components (PFCs) during disruption and ELMs to ITER parameters indicates serious consequences of these phenomena. Therefore schemes for prediction and mitigation or avoidance of these phenomena need to be developed during construction and demonstrated in the early phase of ITER operation. T-retention and dust have important impacts on safety. Therefore the methods of measurement and removal of tritium and dust must be developed during construction and demonstrated in the early phase of ITER operation.

To guarantee acceptable quality of 1800 FW panels produced by six different parties (CN, EU, JA, KO, RF and US), a qualification program is essential. The qualification mock-up is 80 mm wide, 240 mm long and 81 mm thick with three beryllium tiles 10 mm thick. Three identical mock-ups will be fabricated by each of the six parties in 2006–2007 with the same method as for the ITER first wall panels. Heat flux tests will be performed on the qualification mock-ups in 2007–2008. The maximum design heat load on the ITER FW is 0.5 MW/m2 × 10,000 shots and 0.25 MW/m2 × 20,000 shots. The maximum heat flux due to MARFE (multifaceted axisymmetric radiation from the edge): 0.5–1.4 MW/m2 (up to 10 s duration) also needs to be taken into account in the heat load test conditions. Mechanical tests of joints are required using standardized methods. Only parties which have satisfied the acceptance criteria of the qualification tests can proceed to the procurement stage of the ITER FW. Semi-prototypes are also proposed before the ITER FW manufacturing.

Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (#1) has been nearly finalized. Design improvement of other sectors is in progress—in particular, of the VV Sectors #2 and #3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements.

Within the framework of the ITER Test Blanket Working Group, the ITER Parties have made several proposals for test blanket modules to be tested in ITER from the first day of H–H operation. This paper gives an overview of the proposed TBMs designs, of the ITER boundary conditions and of the expected TBM operating conditions. Operating conditions will vary throughout the various ITER phases, starting from the initial H–H phase where no neutrons and, therefore, no nuclear volume heating will be present, to the later D–T phase where pulses of up to 3000 s length may be expected. The paper is focused on the design requirements for the materials and subcomponents that will be used in the various TBMs, from the viewpoint of both the materials performance and the required R&D.

Testing of Breeding Blanket Modules is one of ITER's primary objectives. This paper discusses the major features and requirements of key tests to be performed in ITER for several DEMO-relevant Test Blanket Modules as proposed by the present six ITER Parties. The content is focused on the assessment work recently performed by the ITER Test Blanket Working Group mostly devoted to the establishment of a testing strategy, to the presentation of the Test Blanket Modules proposals and to the definition of the necessary interfaces with the ITER machine and buildings.

It is to be expected that, over time, an interaction of plasma with walls and the divertor will create in ITER hundreds of kilograms of beryllium, carbon or tungsten dust - loose particles with characteristic diameter

A divertor design, supported by R&D, capable of sustaining high heat loads and large electro-magnetic disturbances has been reported previously [1] , [2] . This paper reports on design improvements that, in response to reaction from researchers and industry, focus on cost reductions, holding to a minimum the number of component variants and pursuing the establishment of workable acceptance criteria for divertor armour joints (the latter reported in [3] ). In addition, comment from remote assembly experts has prompted improvements of the in-vessel handling and cassette to vessel attachments.

Recent progress of the ITER vacuum vessel (VV) design is presented. As construction approaches, the VV design has been improved, simplified and developed in more detail. The VV support system has been improved, and the design of the VV shells and the blanket supports has been simplified. The VV design simplifications have been driven by manufacturing requirements and recommendations resulting from cooperation with industry. To simplify the manufacture/maintenance of the port structures, a single wall concept is used for some ports. Structural analyses have been performed to validate all design modifications.

One of the objectives of ITER is to demonstrate fusion technology in an integrated system by performing testing of nuclear components, in particular to test design concepts of tritium breeding blanket relevant to a DEMO reactor. In the current ITER design three large equatorial ports have been allocated for blanket module testing. Typical testing conditions foreseen now include a surface heat flux of 0.1 MW/m 2 , a neutron wall load of 0.78 MW/m 2 , pulse length of 400 s with a duty cycle of 25%. After the first 10 years of operation one may expect to reach a total neutron fluence at the surface of test blanket modules ∼ 0.12 Mwy/m 2 . In the second 10 years of operation very long pulses and accumulation of neutron fluence ∼ 0.3 MWy/m 2 may be expected. Test modules must not compromise ITER safety and reliability. Water-cooled modules must have their own pressure suppression system. The mass of liquid lithium is strictly limited to avoid a hydrogen explosion. Breeding blanket testing in ITER is extremely important for DEMO breeding blanket development. The best effort has to be undertaken to coordinate the Parties' activities in this area and to achieve the best use of space and time available for blanket testing in ITER.

This paper presents the results of a modelling study conducted to estimate the deposition of hydrocarbon radicals in the regions of the ITER divertor hidden from the plasma (e.g., underneath the private flux region dome and in the pumping ducts). A Monte-Carlo code based on the approach described in Ref. [M. Mayer et al., J. Nucl. Mater. 313–316 (2003) 429] has been developed and applied for the actual ITER geometry. It allows for a full 3-dimensional treatment of particle trajectories and follows individual hydrocarbon particles until they either stick on the surfaces or leave the system. The cases analysed clarify the influence of several parameters such as the sticking coefficient of the hydrocarbon species, the pressure of the background gas, and the geometry. Consistent with experimental findings from tokamaks, this assessment shows that particles with relatively high sticking probability (>0.01) deposit primarily underneath the divertor structure, and only species with very small sticking probability (⩽10 −3 ) may reach and stick along the inner surfaces of the pumping ducts.

This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed.

We report on GaAlInAs/GaAs lasers manufactured by the industry’s biggest production MBE tool. This MBE reactor allows for growth on 23 three-inch diameter wafers at a time, at a cost that compares favorably with the MOCVD method. Data on chip-on-submount performance and uniformity across the entire MBE-growth area are presented and compared to the quality of material produced by smaller size production MBE tools. We also present data on performance characteristics of spatially combined fiber coupled passively cooled single emitter-based pumps. The data include performance characteristics of devices operating at ~805nm and ~975nm wavelengths when driven in CW, QCW and pulsed modes; both pumps use ~105μm core diameter fiber to launch power confined within NA

The focus of effort in ITER EDA since 1998 has been on the development of a new design to meet revised technical objectives and a cost reduction target of about 50% of the previously accepted cost estimate. Drawing on the design solutions already developed, and using the latest physics results and outputs from technology R&D projects, the Joint Central Team and Home Teams, working together, have been able to progress towards a new design which will allow the exploration of a range of burning plasma conditions, with a capacity to progress towards possible modes of steady state operation. The new ITER design, whilst having reduced technical objectives from those of its predecessor, will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy. The main features of the current design and of its projected performance are introduced and the outlook for construction and operation is summarized.

ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties.

International Thermonuclear Experimental Reactor (ITER) Engineering Design Activities (EDA) in safety and environment are approaching the point where conceptual safety design, topic studies and research will give way to project oriented engineering design activities. The Joint Central Team (JCT) is promoting safety design and analysis necessary for siting and regulatory approval. Scoping studies are underway at the general level, in terms of laying out the safety and environmental design framework for ITER. ITER must follow the nuclear regulations of the host country as the future construction site of ITER. That is, regulatory approval is required before construction of ITER. Thus, during the EDA, some preparations are necessary for the future application for regulatory approval. Notwithstanding the future host country's jurisdictional framework of nuclear regulations, the primary responsibility for safety and reliability of ITER rests with the legally responsible body which will operate ITER. Since scientific utilization of ITER and protection of the large investment depends on safe and reliable operation of ITER, we are highly motivated to achieve maximum levels of operability, maintainability, and safety. ITER will be the first fusion facility in which overall ‘nuclear safety’ provisions need to be integrated into the facility. For example, it will be the first fusion facility with significant decay heat and structural radiational damage. Since ITER is an experimental facility, it is also important that necessary experiments can be performed within some safety design limits without requiring extensive regulatory procedures. ITER will be designed with such a robust safety envelope compatible with the fusion power and the energy inventories. The basic approach to safety will be realized by ‘defense-in-depth’. The first priority will be in the prevention of accidents through the intrinsic features of the facility, quality assurance throughout, in design, construction, operation and maintenance, and appropriate provisions for human factors. Nevertheless, the plant will be designed to be ready for anomalous events. In addition, public will be protected with appropriate mitigative features, even for extremely unlikely and unforeseen hypothetical accidents to add safety margins as appropriate. Current safety design approaches are introduced in this paper, including a global methodology, off-normal plasma termination, decay heat removal, and containment and confinement strategies.

The majority of fiber laser volume applications are price sensitive. Therefore, the availability, quality and cost of singleemitter- based pumps will have decisive impact on the breadth of further fiber lasers' acceptance. Availability and cost should not come in expense of further improvement in pumps' performance and reliability. Here we report on optimized high-power and high-brightness wavelength stabilized CW devices. Performance of CW pumps rated for 100W and 50W power is discussed. Pumps launch over 98% output power into a spectral window of 975±0.5nm at driving currents ranging from 2A to 12A and the heatsink temperature variation from 20°C to 50°C. Such performance qualifies these wavelength-stabilized pumps for use in many air-cooled and special applications.

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L–H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v ∼ 60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH98/H98 were ∼3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

Advanced high volume applications require pumps with high power, high brightness, and high power efficiency. New generation devices meet all of these challenging requirements, while still maintaining the advantages of distributed pumping architecture including high reliability inherent to single emitter sources. Based on new-generation long-cavity diode chips, new pumps are capable of more than 60W CW power ex-fiber output (100 μm core diameter) into NA ~ 0.12. Peak power efficiency stays over 60%. All of the above is provided at room heatsink temperature, maintained by basic air- or water-cooling.

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation. © 2009 IAEA, Vienna.

Higher reliability and power efficiency achieved with low-demanding cooling make single emitter diodes a more effective pump source than monolithic laser diode arrays. Continuously improving performance and increasing brightness of single emitter pumps are accompanied with a steady reduction of cost of pumping (dollar-per-watt). Performance advantages do not compromise reliability of the pumps. These features ensure that single emitter diodes are the most effective solution even for multi-kWatt systems pumping. Here we report on a recent progress in single-mode and multi-mode edge-emitting diodes.

A low half-wave voltage V(pi) of 1.57 V was obtained with a 2-cm-long birefringent polymer waveguide modulator at a wavelength of 1.3 microm by use of a modulator design with a constant-bias electric field and a high-microbeta chromophore. The design allows the maximum achievable electro-optic coefficient of the material to be utilized. This electro-optic coefficient can be more than twice as high as the residue value that is used by conventional modulator designs, after fast partial relaxation following poling.

A polymer TM-pass polarizer is demonstrated by integrating photobleaching (PB) waveguides, which support only the TM mode, with mode-matched rib waveguides formed by reactive ion etching (RIE). The input and output rib guides support both TE and TM modes. The PB can be controlled to closely match the mode profile of the photobleached guide to the mode profile of the input and output rib waveguides in order to reduce the insertion loss. The polymer selected for photo-bleaching was PMMA-DRI. This electrooptic polymer has a PMMA backbone with the azo dye Disperse Red I side chain attached to the backbone.

We propose to use an extended electro-optic interaction between the two guided-wave IR beams in a nonlinear optical waveguide as the method for generation of polarization at the difference RF frequency. The electro-optic interaction is fast and the distributed waveguide interaction enables high efficiencies and higher total power. In addition, use of fiberoptics and integrated optics enables local generation of RF beams from the remotely located IR sources.

In this paper we report on trade study results relating to design features needed to accommodate seismic isolation in the ITER design and on selected features of a proposed seismic isolation system. ITER will implement the layout changes identified in this study, but seismic isolation will only be applied if detailed dynamic analysis based on the EPA design and selected site conditions indicate that it is essential to protect the viability of the project.

An assessment of ITER plasma parameters is carried out for the low activation phase that is required for commissioning the basic ITER systems including plasma control, heating and current drive. Such an operation is analysed for hydrogen, helium and deuterium plasmas for full field and current, as well as with magnetic field and plasma current reduced to half of their design values, B0 = 2.65 T, Ip = 7.5 MA. Both hydrogen and deuterium neutral beam injection (NBI) are considered. We assess the possible domain for safe operation, and the possible target plasmas for commissioning the NBI, electron cyclotron heating (ECH) and ion cyclotron heating (ICH) systems, taking into account the constraints imposed by NB shine-through loss, Greenwald limit and access to H-mode operation. Simulations with the Automated System for Transport Analysis (ASTRA) show that for 33 MW of NBI with 20 MW of ECH, H-mode access is marginal for hydrogen plasmas. Good H-mode confinement, expected at PNB + PEC + PIC > 1.5 PL–H, is more likely for the helium and deuterium cases. It is found that plasma parameters, such as normalized beta, plasma density and current flat-top duration, for full power/half field/half current operation can be similar to those required for the DT long pulse operation. Preliminary assessment is also made of the maximum of tritium and neutron yield achievable in a single shot at the deuterium phase of ITER operation.

A number of possible designs of external and in-vessel coils generating resonant magnetic perturbations (RMPs) for Type I edge localized modes (ELMs) control in ITER are analysed for the reference scenarios (H-mode, Hybrid and Steady-State) taking into account physical, technical and spatial constraints. The level of stochasticity (Chirikov parameter ∼1 at ψ1/2∼ 0.95) generated by the I-coils in the DIII-D experiments on ELMs suppression was taken as a reference. Designs with a toroidal symmetryn= 3 were considered to avoid lowernnumbers producing larger central islands, a potential trigger of MHD instabilities. The evaluation of RMP coils designs is done with respect to the RMPs spectrum that should produce enough edge ergodisation and minimum central perturbations at minimum current. The proposed designs include in-vessel, mid-ports and external coils. Changes in the equilibrium due to changes in the internal inductanceli, the poloidal beta βpand the edge magnetic shear in a reasonable range for ITER scenarios were demonstrated to have a small effect on the edge ergodisation. Present estimations were done without margins and for vacuum fields neglecting plasma response on RMPs. The validity of the vacuum approach was estimated analytically in thevisco-resistive linear responseregime using [1]. The typical radial magnetic field amplitudes produced by RMP coils in DIII-D and ITER are an order of magnitude or slightly above the critical values for the ‘downward’ bifurcation to the reconnected stage indicating the possibility of the islands formation in the pedestal region. Central islands (from the top of the pedestal) are expected to be screened.

A high-magnetic field tokamak is designed for plasma heating experiments by adiabatic compression of plasma over major and minor radii. Neutral beam injection heating is used as well. The plasma position and current control system includes a diagnostic facilities with preliminary signal processing, an automatic control system, a computer with interface, power supplies and ohmic heating coil, compression coil, equilibrium coils for the first and second discharge stage, vertical and horizontal field correction coils. While selecting the control system structure, the following tokamak features are taken into consideration: - low time constant ~ 0.4 as of the vacuum chamber; - strong magnetic coupling between poloidal field coils; - application of uncontrollable or weakly controllable power supplies for ohmic heating and compression coils; - application of a capacitor bank with a fast thyristor switch to supply the equilibrium coil. Combined adaptive program control and feedback loops is used. The program control is accomplished using numerical modelling of the control system together with the control object and the following correction (adaption) during dwell time according to the results of the preceding discharge.