Fusion Science

The water-cooled ceramic breeder (WCCB) blanket is one candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), and it is being developed at the Institute of Plasma Physics of the Chinese Academy of Sciences. This paper presents an updated WCCB blanket concept design that is expected to cover the CFETR operation modes of 200 MW, 500 MW, 1 GW, and 1.5 GW fusion power and achieve tritium self-sufficiency with the latest core parameters (major radius R = 7.2 m; minor radius a = 2.2 m) and mission. The feasibility of the new blanket design is evaluated considering neutronics and thermal-hydraulics aspects.

In this paper we propose to use tritiated liquid effluents from a CANDU type reactor as a tracer, to study hydrodynamics on Danube-Black Sea Channel. Tritiated water can be used to simulate the transport and dispersion of solutes in mentioned Channel, because it has the same physical characteristic as water. Measured tracer response curves produced from controlled evacuations provide an efficient method of obtaining necessary data. This paper presents the establishing of proper mixing length, and the base line of tritium concentration in studied area. These first steps were used to construct the unit-peak attenuation (UPA) curve for a sector of the Danube-Black Sea Channel.

The ETE is a spherical tokamak with aspect ratio A = 1.5 (major radius of 0.3m and minor radius of 0.2m) under development at LAP/INPE. The ETE incorporates some innovative features that resulted in a compact and light weighted device with good plasma accessibility. Since the first plasma obtained at the very end of 2000 (Ip = 12kA, duration of 2ms, B o = 0.1T), the machine is operational and improvements are being done in order to achieve the planned final parameter values for the first phase of operation (Ip = 220kA, duration 15ms, B o = 0.4T), which are limited by the available capacitors. The efforts are being focused on incrementing the energy of the capacitor banks, lessening the stray magnetic fields in the plasma region, conditioning the vacuum vessel wall, implementing diagnostics and optimizing the discharge parameters. Presently, plasma currents in the range of 40-60kA (duration of 6-12 ms) are routinely obtained. Electron temperatures up to 160eV and plasma densities up to 3.0x10 19 m -3 are being reached.

RF system for VEPP-5 booster ring has been created at Budker INP. It operates at 700 MHz and consists of the RF power source based on KU-393 klystron, waveguide sec-tion with wave-to-coax transitions, accelerating cavity, and control system. The cavity higher order modes are damped by resistive loads to eliminate the beam instability. Results of the cavity cold tests and RF system high power level tests are presented.

High power proton beams are available for neutron science, tritium target development, nuclear physics, material science, isotope manufacturing, and weapons research at the Los Alamos Neutron Science Centre (LANSCE) linac. The RF powerplant, particularly the 201.25 MHz power amplifiers (PA) powering four drift tube linac (DTL) tanks, is under a lot of strain due to the amount of concurrent beam users. These amplifiers were created about 30 years ago and have run at up to 3 megawatts with 12% duty factors. However, other recent modifications and upgrades to the 201.25 MHz RF powerplant have greatly improved the operation, and higher beam current for a proposed Long Pulse Spallation Source (LPSS) cannot be delivered simultaneously with other beams at high duty factor. Improved operating procedures through monitoring of tube vacuum and final power amplifier plate dissipation have raised RF system reliability above 95% for operation periods in 1993-95. With just a few power tubes, the system will provide high peak power and duty factor, enhanced DC to RF efficiency, and a streamlined cooling system. Plans are in progress to design a novel final power amplifier that can use low-level RF modulation for amplitude control.

Significant developments in the Ignitor program concerning both physics and technology are reported. These include i) the confirmation of the validity of the path to ignition based on the high density regimes (n 0 ~ 10 21 m -3 ) following relevant experiments by LHD (Japan); ii) the adoption of intermediate temperature superconducting material (MgB 2 ) for the largest poloidal field coils and the relevant R&D effort; iii) the identification and the analysis of the optimal double-X point configurations with plasma currents I p ~ 9-10 MA that can be adopted to achieve ignition in the H-regime; iv) the completion of the analysis of the ICRH strategy, optimizing the time necessary to reach ignition; v) the completion of the detailed engineering design of all the components of the machine core (e.g., toroidal and poloidal field magnets, supporting mechanical structures, plasma chamber, first wall system, remote handling system, cryogenic cooling system, tritium system); vi) the construction and initial tests of the first injector capable of launching pellets with almost 4 km/s as required by the high plasma densities and temperatures at which ignition can be achieved in Ignitor; vii) the development of new diagnostic systems that can complement the electromagnetic diagnostics to provide the input for the control system of both the position and the shape of the plasma column; viii) the development of prototype coils of the electromagnetic diagnostics with ceramic insulators that are resilient to neutron and gamma radiation; ix) the initial conceptual studies of power producing reactors based on the "reactor physics" that Ignitor is designed to investigate.

The results of recent experiments in the L-2M stellarator are presented. The experiments were carried out with the use of boronization of the stellarator vacuum chamber. This made it possible to change the conditions of the plasma experiment substantially, reducing the radiation loss power to 10 15% of the power input. Under these conditions, at a heating power above 120 kW, a jump (pedestal) by 100 150 eV appears in the electron temperature at the separatrix. When operating with a high heating power and plasma density above n > 10^19/m^3, we observed a spontaneous, sudden and very fast (about 0.2 ms) drop in the plasma energy by 510% of its value, measured by the diamagnetic diagnostics. After this fast drop, the plasma energy slowly restores or even increases somewhat. The probability of such events increases with increasing density and heating power. The fast change in the plasma parameters occurs only at the edge while not influencing the core plasma parameters. The fast drop...

The results of recent experiments in the L-2M stellarator are presented. The experiments were carried out with the use of boronization of the stellarator vacuum chamber. This made it possible to change the conditions of the plasma experiment substantially, reducing the radiation loss power to 10 - 15% of the power input. Under these conditions, at a heating power above 120 kW, a jump (pedestal) by 100 - 150 eV appears in the electron temperature at the separatrix. When operating with a high heating power and plasma density above n > 10^19/m^3, we observed a spontaneous, sudden and very fast (about 0.2 ms) drop in the plasma energy by 5- 10% of its value, measured by the diamagnetic diagnostics. After this fast drop, the plasma energy slowly restores or even increases somewhat. The probability of such events increases with increasing density and heating power. The fast change in the plasma parameters occurs only at the edge while not influencing the core plasma parameters. The fas...

In this paper, the multipactor RF breakdown in a parallel-plate waveguide partially filled with a ferrite slab magnetized normal to the metallic plates is studied. An external magnetic field is applied along the vertical direction between the plates in order to magnetize the ferrite. Numerical simulations using an in-house 3D code are carried out to obtain the multipactor RF voltage threshold in this kind of structures. The presented results show that the multipactor RF voltage threshold at certain frequencies becomes considerably lower than for the corresponding classical metallic parallel-plate waveguide with the same vacuum gap.

During the D-T operation of the International Thermonuclear Experimental Reactor (ITER), the dose inside ITER building is attributed to neutrons and gamma-rays streaming from various openings in the machine, whereas after shutdown, it is solely due to the decay gammas from activated components inside the building. In the present study, two cases of different geometrical arrangement are considered (Case A and Case C) in which the impact of the size of the opening of the NBI horizontal ports and presence of the NBI structure on the dose profiles everywhere in ITER building was studied and compared to the case (Case B) where all the horizontal ports are plugged. Contours and mapping of the dose rate profiles during operation, at shutdown, and at one-day, one-week, and one-month after shutdown, were generated for the three cases. Zone classification requirements for each hall/room and gallery of the building in terms of the maximum allowed exposure dose during operation and after shutdown were examined to identify where these requirements are met during 'hands-on' maintenance activities. It is shown that the requirements are satisfied at all times after shutdown in the halls above the ground, even during operation. For the conservative Case A, the dose rates are excessively high inside the equatorial NBI hall and inside the access halls above and below (divertor hall). In Case C, all the halls/vaults and galleries are accessible 1-week after shutdown, except inside the NBI hall whose dose rate (6.2 × 10 5 mSv h − 1) exceeds the required zone classification (3-10 mSv h − 1). All areas are accessible a week after shutdown in Case B which represents the building side away from the NBI ports. However, in all these cases, the accessibility criterion inside the rooms beneath the cryostat is not met at all times after shutdown due to the rather thin bottom bioshield (50 cm) which may necessitate some design changes.

An extensive benchmark exercise has been conducted on ITER with the objective to test and validate different approaches for the use of CAD generated geometry data for Monte Carlo transport calculations with the MCNP code. The exercise encompassed the generation of a dedicated neutronics CATIA model based on available engineering CAD design data, the conversion into MCNP geometry, the verification of the converted models, and a number of calculations to compare the different approaches with regard to the performance and the validity of the results obtained. The paper briefly reviews the different approaches and provides a detailed description of the ITER benchmark effort, its results and conclusions showing that the approaches have reached the maturity level to allow their application to real ITER design analyses. This is considered an essential step forward for neutronics analysis tools to satisfy ITER quality assurance rules.

We have analyzed the availability and reliability of the flashlamp-pumped, Nd:glass amplifiers that, as a part of a laser now being designed for future experiments, in inertial confinement fusion (ICF), will be used in the National Ignition Facility (NIF). Clearly , in order for large ICF systems such as the NIF to operate effectively as a whole, all components must meet demanding availability and reliability requirements. Accordingly, the NIF amplifiers can achieve high reliability and availability by using reliable parts, and by using a cassette-based maintenance design that allows most key amplifier parts to be replaced within a few hours. In this way, parts that degrade slowly --as the laser slabs, silver reflectors, and blastshields can be expected to do, based on previous experience -can be replaced either between shots or during scheduled maintenance periods, with no effect on availability or reliability. In contrast, parts that fail rapidly --such as the flashlamps --can and do cause unavailability or unreliability. Our analysis demonstrates that the amplifiers for the NIF will meet availability and reliability goals, respectively, of 99.8% and 99.4%, provided that the 7680 NIF flashlamps in NIF have failure rates of less than, or equal to, those experienced on Nova, a 5000-lamp laser at Lawrence Livermore National Laboratory (LLNL).

Fast wave current drive experiments have been performed in JET plasmas with electron internal transport barriers produced with LHCD. The central plasma current was difficult to affect, even though the calculated current drive efficiency was fairly high, 0.07A per W absorbed by the electrons. The main reasons are: the strongly inductive nature of the plasma current; the interplay between the fast wave driven current and the bootstrap current, which, due to the dependence of the bootstrap current on the poloidal magnetic field, decreases the bootstrap current as the driven current increases; and parasitic absorption of the waves that decreased the power absorbed by the electrons. The measured difference in the central current density for co and counter current drive is larger than the response expected from current diffusion calculations, suggesting a faster current diffusion than that given by neoclassical resistivity. Effective direct electron heating, comparable to the indirect electron heating with H-minority heating, is found for the dipole phasing of the antennas without producing a significant fast ion pressure and with low impurity content in the divertor plasma even though the single pass damping is only a few percent. For the ±90 o phasings producing current drive, with a similar single pass damping, strong degradation of the heating is observed with strong increases in the BeII and CIV line intensities in the divertor. The degradation depends on the phasing of the antennas and increases with reduced single pass damping, consistent with RF-power being lost by dissipation of rectified RF-sheath potentials at the antennas and walls. Asymmetries in direct electron heating, lost power, production of impurities, fast ions and gamma-rays are seen for co and counter current drive that are consistent with differences in the absorption on residual 3 He ions due to the RF-induced pinch.

Thisdocumentwasprepared= an accountof worksponsoredby sn agencyof the UnitedStatesGovernmen~NeithertheUnitedStatesGovernment nor theUniversity ofCalifornianor anyofthar employ-makesanywarranty,expressor implied,or assumes anykgalkbilityorr aspomibfity forthe -racy, completenesqorusefuirsess of any informatio~apparat~produ~or process disdoosd, or represents that its use wouidnot infringe privately owned rights Reference hesein to any specific commercial products, process, or service by trade namq trademar& manufacturer, or otherwise, does not necessarily constitute orimplyits endomemen~recommendation or favoring by the United States Govemmcnt or the University of California. The views and opinions of authois expressed herein do not necesswily state or reflect those of the United States Govenunemt or the University of CaIiforniq and shall not be used for advetiing or product endorsement purposes.

In recent three-ion species (majority D and H plus a trace level of 3 He) ICRF heating experiments on Alcator C-Mod, double mode conversion on both sides of the 3 He cyclotron resonance has been observed using the phase contrast imaging (PCI) system. The MC locations are used to estimate the species concentrations in the plasma. Simulation using TORIC shows that with the 3 He level <1%, most RF power is absorbed by the 3 He ions and the process can generate energetic 3 He ions. In mode conversion (MC) flow drive experiment in D(3 He) plasma at 8 T, MC waves were also monitored by PCI. The MC ion cyclotron wave (ICW) amplitude and wavenumber kR have been found to correlate with the flow drive force. The MC efficiency, wave-number k of the MC ICW and their dependence on plasma parameters like Te0 have been studied. Based on the experimental observation and numerical study of the dispersion solutions, a hypothesis of the flow drive mechanism has been proposed.

For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt (t&lt;τE) response of the scrape-off-layer (SOL) plasma to Lower Hybrid Radiofrequency (LHRF) power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current dri...

The preliminary conceptual design for the Electron Cyclotron (EC) system of the future European DEMOnstration fusion power plant is ongoing in the EUROfusion Consortium. This represents one of the key aspects in order to assess the performances and the integration capabilities of such a system in EU DEMO as well as in the alternative reactor configuration Flexi-DEMO. Different options for the antenna, namely remote steering antenna (RSA), open ended waveguides (OEWG) and the front steering antenna (FSA) later on renamed in mid steering antenna (MSA) to notify that the MSA is protected behind the breeding blanket (BB) in DEMO, are investigated, analyzing their performance for several injection angles and launch points. This activity considers the constraints given by physics and engineering requirements, as for example the maximum power per port and the necessary local current drive to stabilize neo-classical tearing modes (NTMs) with a proper deposition width. The beam tracing calculations have been performed on different scenarios, providing information on plasma accessibility and deposited power. The microwave design and initial ideas about the ex-vessel EC transmission lines routing will be shown.

I present a short survey on some mathematical results concerning to the existence of solutions and some numerical experiments on a non-local two dimensional free boundary problem modeling the magnetic confinement of a plasma in a Stellarator device. One of the main difficulties of the magnetically controlled plasma fusion (in axisymmetric geometric devices as Tokamaks or non axisymmetric geometric ones as Stellarators), is to determinate the conditions of the magnetic field and of the current density in order to maintain the plasma far from the camera walls of the nuclear reactors. I will prove the existence of a solution (Díaz, Padial, Rakotoson 1998) by means of a Galerkin argument type for a new family of elliptic problems associated to an equivalent formulation to the original problem but in a non-local form. Finally, by finite elements, I will show some numerical results (Díaz, Galán, Padial 2012).

Submitted for the DPP17 Meeting of The American Physical Society Nonsymmetric 3D MHD equilibrium and radial localization of trapped particles 1 WRICK SENGUPTA, HAROLD WEITZNER, Courant Institute of Mathematical Sciences, NYU-Quasisymmetry and omnigeneity are key ideas proposed to ensure radial confinement of trapped particles in a stellarator. These constraints have stringent restrictions on magnetic geometry, some aspects of which are yet to be fully explored. In this work we obtain a local 3D MHD equilibrium expansion by analytically solving the MHD equilibrium equations. This expansion, although local, is sufficient to explore the deeply trapped particle physics, since it is carried out around a region of local minima of the magnitude of the magnetic field. Based on this analytical 3D equilibrium solution, we obtain the aforementioned constraints. We then extend this local analysis to a global one by expanding around the magnetic axis of the stellarator. Effects of curvature and torsion of the axis are treated self-consistently. We demonstrate that due to toroidal mode coupling, the expansion in flux coordinate near the axis is logarithmic and not purely algebraic. These non analytic terms can not be in general neglected. Our results show that it is far easier to satisfy the omnigeneity condition than the quasisymmetry requirement. This implies, that there exists a large class of equilibrium close to quasisymmetry, which are still omnigeneous and allow inclusion of symmetry breaking error fields.

The ITER tokamak will be operating with 5 microwave diagnostic systems. While they rely on different physics, they share a common need: transmitting low and high power microwave in the range of 12 GHz to 1000 GHz(different bandwidths for different diagnostics) between the plasma and a diagnostic area tens of meters away. The designs proposed for vacuum windows, in-vessel waveguides and antennas are presented together with the development activities needed to finalise this work.

An ion cyclotron resonance heating (ICRH) system has been designed, fabricated indigenously and commissioned on Tokamak Aditya. The system has been commissioned to operate between 20•0 and 47•0 MHz at a maximum power of 200 kW continuous wave (CW). Duration of 500 ms is sufficient for operation on Aditya, however, the same system feeds the final stage of the 1•5 MW ICRH system being prepared for the steady-state superconducting tokamak (SST-1) for a duration of 1000 s. Radio frequency (RF) power (225 kW) has been generated and successfully tested on a dummy load for 100 s at 30•0 MHz. Lower powers have been coupled to Aditya in a breakdown experiment. We describe the system in detail in this work.

A 139 GHz microwave scattering diagnostic is used to study the penetration of externally launched 0.8 GHz lower hybrid waves in current drive (fie $6 x1012cm-3) experiments, on MIT&#x27;s Versator II tokamaK. A 7 W Extended Interact~ on Oscillator is used ...

During normal ITER operation, Electron Cyclotron (EC) heating and current-drive system will generate stray microwave radiations. Three cases have been identified with their associated load localisation and power. The effect on first-wall diagnostics, mineral insulated cables and vacuum windows are described. The most affected diagnostic port plugs are the #11, #12 and #17. In these areas, mitigation strategies will have to be found to protect the vacuum windows from the cross-polarised stray radiation.

The ITER tokamak will be operating with 5 microwave diagnostic systems. While they rely on different physics, they share a common need: transmitting low and high power microwave in the range of 12 GHz to 1000 GHz(different bandwidths for different diagnostics) between the plasma and a diagnostic area tens of meters away. The designs proposed for vacuum windows, in-vessel waveguides and antennas are presented together with the development activities needed to finalise this work.

The efficiency of the energetic ion confinement is reduced in a tokamak plasma by the non-axisymmetric field, namely the ripple field. The ripple field is produced by the finite number of the toroidal field coils. It is affected by the non-axisymmetric finite beta effect. The three-dimensional MHD equilibrium calculation code VMEC is used to analyze the non-axisymmetric finite beta effect in a ripple tokamak. In the VMEC code, the flux coordinates are used, so the calculation region is limited to the area of plasma. To calculate the orbit of them outside the plasma, we develop the field calculation code, which is based on the Biot-Savart law. The details of the method and results are described in this study.

We are reporting preliminary experimental results on the injection and trapping in the NRL Modified Betatron. In internal injection, the beam is produced by a diode located within the chamber. Using one half of the 1 m toroidal vacuum chamber, the propagated beam current and displacement are monitored as functions of the applied magnetic fields (B , and the toroidal field, B). Our preliminary results are in agreement with theoretical predictions.

Current state-of-the-art and next generation laser systems-such as those used in the NIF and LIFE experiments at LLNL-depend on ever larger optical elements. The need for wide aperture optics that are tolerant of high power has placed many demands on material growers for such diverse materials as crystalline sapphire, quartz, and laser host materials. For such materials, it is either prohibitively expensive or even physically impossible to fabricate monolithic pieces with the required size. In these cases, it is preferable to optically bond two or more elements together with a technique such as Chemically Activated Direct Bonding (CADB ©). CADB is an epoxy-free bonding method that produces bulk-strength bonded samples with negligible optical loss and excellent environmental robustness. The authors have demonstrated CADB for a variety of different laser glasses and crystals. For this project, we will bond quartz samples together to determine the suitability of the resulting assemblies for large aperture high power laser optics. The assemblies will be evaluated in terms of their transmitted wavefront error, and other optical properties.

We apply the code package SCENIC to a two field-period quasi-axisymmetric stellarator. Ion cyclotron resonance heating (ICRH) is applied both on the high-and low-field side to a 1% 3 He minority in a deuterium plasma. It is shown that due to toroidal variations, the results are considerably different from similar tokamak studies. In particular, toroidal variations in power deposition and pressure are created and accentuated during radio frequency heating, such that modifications to the magnetic equilibrium depend on toroidal angle. We demonstrate that due to enhanced particle loss, low-field side heating is significantly less efficient than high-field side heating, and that toroidally trapped particles impose upper power limits for efficient radio frequency injection.

The development of a 4.6 GHz reflectometer system to detect LH waves in Alcator C-Mod plasmas is presented together with preliminary experimental results. Hard X-ray data of the 2008 Alcator C-Mod campaign indicated an anomalous decrease of photon generation in high density plasmas during LHCD experiments, suggesting LH waves did not propagate inside the plasma separatrix. Calculations and simulations show that the nonlinear interaction between a reflectometer probe beam and LH waves in the region where wave matching conditions are satisfied generates back-scattered, frequency shifted signals. To measure these signals, the high frequency stages of existing 60 GHz and 75 GHz O-mode reflectometer channels were modified, and intermediate frequency and power detecting stages have been added. The ray tracing code, GENRAY, and full wave simulations, e.g., TORIC and LHEAF, will be used to optimize plasma conditions for maximum response of frequency shifted signals. Both 1D and 2D transient...

This document is intended for publication in the open literature. It is made available on the clear understanding that it may not be further circulated and extracts or references may not be published prior to publication of the original when applicable, or without the consent of the Publications Officer, EUROfusion Programme Management Unit,

One of the main challenges in the development of a plasma diagnostic and control system for DEMO is the need to cope with unprecedented radiation levels in a tokamak during long operation periods. A list of diagnostics required for plasma control has been developed during the pre-conceptual design phase. Different approaches are proposed for the integration of these diagnostics in DEMO: in equatorial and upper ports, in the divertor cassette, on the inner and outer surfaces of the vacuum vessel and in diagnostic slim cassettes, a modular approach developed for diagnostics requiring access to the plasma from several poloidal positions. According to each integration approach, diagnostics will be exposed to different radiation levels, with a considerable impact on their design. This paper provides a broad overview of the radiation environment that diagnostics in DEMO are expected to face. Using the water-cooled lithium lead blanket configuration as a reference, neutronics simulations w...

In the frame of the EFDA Task HCD-08-03-01 (EU contribution to the ITER LHCD Development Plan) the preliminary design of the LHCD system, described in the Detailed Design Description DDD 5.4 LH (2001), has been revised to take into account progresses on technology and on knowledge. In particular a revised layout of the ITER machine and the successful development of a prototype klystron at 5 GHz with a target RF power of 500 kW CW have determined significant changes in the design of the main transmission lines and on the modularity of the launcher.

This paper describes the design and performance of flashlamp-pumped, Nd:glass. Brewster-angle slab amplifiers intended to be deployed in the National Ignition Facility (NIF). To verify performance, we tested a full-size, three-slab-long, NIF prototype amplifier, which we believe to be the largest flashlamp-pumped Nd:glass amplifier ever assembled. Like the NIF amplifier design, this prototype amplifier had eight 40-cm-square apertures combined in a four-aperture-high by two-aperture-wide matrix. Specially-shaped reflectors, anti-reflective coatings on the blastshields, and preionized flashlamps were used to increase storage efficiency. Cooling gas was flowed over the flashlamps to remove waste pump heat and to accelerate thermal wavefront recovery. The prototype gain results are consistent with model predictions and provide high confidence in the final engineering design of the NIF amplifiers. Although the dimensions, internal positions, and shapes of the components in the NIF amplifiers will be slightly different from the prototype, these differences are small and should produce only slight differences in amplifier performance.

The latest design of the Water Cooled Lithium Lead blanket for the European DEMO reactor is based on a Single Module Segment (SMS) concept for the blanket segments: the removal of the gaps between the modules (present in the former Multi Module structure design), with the consequent reduction of the neutron streaming, improves the tritium production and the shielding capabilities. A nuclear analysis has been carried out with the Monte Carlo N-Particle transport code version 5 (MCNP5) and the Joint Evaluated Fission and Fusion nuclear data libraries version 3.2 (JEFF 3.2). A detailed three-dimensional MCNP DEMO model with the Single Module layout has been generated. Three-dimensional neutron and gamma transport simulations have been carried out in order to assess the tritium production, nuclear power deposition in each subcomponent of the blanket and nuclear heating density, neutron flux, neutron damage (dpa) and Helium production, relevant for the design of the system. The results confirm the fulfilment of tritium self-sufficiency and shielding requirements of DEMO. Furthermore, a parametric analysis has been carried out by varying the thickness of the first wall tungsten layer to evaluate how the blanket performances are influenced by the tungsten layer thickness.

The Princeton Beta Experiment-Modified (PBX-M) has a close-fitting, conducting, passive plate, stabilizing shell which nearly surrounds highly indented, bean-shaped plasmas. The proximity of this electrically isolated shell to a large fraction of the plasma surface allows measurements similar to previous work on other tokamaks using floating probes and limiters. Measurements were performed to characterize the plasma-induced voltages on the PBX-M passive plate stabilizing shell during high-p" plasmas. Voltage differences were measured between the respective passive plate toroidal and poloidal gaps, the respective passive plates and the vessel, and an outer poloidal graphite limiter and its passive plate. The calibration and qualification testing procedures are discussed. The initial measurements found that the largest voltages were observed at plasma start-up and at the plasma current disruption and exhibited characteristics depending on operating conditions. The highest voltages observed have been at disruption and were less than 2 kV DISCLAIMER This report was prepared as a» account of worlr sponsored by an agency of (he United Stales Government.

The paper reports on latest developments for the ITER Ion Cyclotron Heating and Current Drive system: imminent acceptance tests of a prototype power supply at full power; successful factory acceptance of candidate RF amplifier tubes which will be tested on dedicated facilities; further design integration and experimental validation of transmission line components under 6MW hour-long pulses. The antenna Faraday shield thermal design has been validated above requirements by cyclic high heat flux tests. R&D on ceramic brazing is under way for the RF vacuum windows. The antenna port plug RF design is stable but major evolution of the mechanical design is in preparation to achieve compliance with the load specification, warrant manufacturability and incorporate late interface change requests. The antenna power coupling capability predictions have been strengthened by showing that, if the plasma scrape-off layer turns out to be steep and the edge density low, the reference burning plasma can realistically be displaced to improve the coupling.

A dedicated optical system is designed for the breakdown of the first ITER plasma, when the machine will not be equipped with the equatorial ECRH-launcher. Seven beams from the upper launcher will be transmitted via several in-vessel mirrors through the resonance layer to a beam-dump. The whole system was simulated with the PROFUSION package to obtain the beam parameters, estimate spill-over effects and determine the required polarization in the launcher in order to have X-mode polarization in the resonance.

experiments involve the use of a variety of materials that generate a number of radiological issues. Along with the use of tritium and depleted uranium, shots generating neutrons create prompt radiation fields as well as fission and activation products. In order to assure readiness for these hazards, a series of readiness reviews was conducted as the hazards were introduced. Each step was built upon the previous steps, as well as the basic infrastructure and operating capability of the laser facility. A detailed preparation plan for the introduction of these hazards was developed. This included ensuring required equipment was in place and ready, all plans and procedures were developed, and personnel were trained and qualified to perform work in the environment. The approach for preparing the facility for operations under the new set of conditions, the preparations for the readiness reviews, the review process, as well as the approach to initial operations are discussed.

This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor the University of California nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or the University of California. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or the University of California, and shall not be used for advertising or product endorsement purposes. This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint is made available with the understanding that it will not be cited or reproduced without the permission of the author. This report has been reproduced directly from the best available copy.