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Fusion Engineering, 1993., 15th IEEE/NPSS Symposium on

Date 11-15 Oct. 1993

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  • 15th IEEE/NPS Symposium Fusion Engineering Volume Two

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  • Conference Author Index

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  • Tokamak Physics Experiment safety analyses and enviromental safety, and health compliance activities

    Page(s): 962 - 965 vol.2
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    The Tokamak Physics Experiment (TPX) is a new fusion machine proposed to be built at the Princeton Plasma Physics Laboratory (PPPL). This paper describes results of the on-going safety analyses and Environmental, Safety, and Health (ES&H) activities in support of this project. The TPX deuterium and tritium operation perspectives, radiological design objectives, results of dose calculations for normal and postulated accident scenarios, and nonradiological impacts are all addressed. As part of the ES&H considerations, this paper provides an overview of the TPX Environmental Assessment (EA), the EA approval process, and a brief discussion of other environmental issues being addressed for TPX. Results show that: nonradiological impacts are minor; operational and accidental releases of tritium, activated gases, or activated solids are within design objectives and regulatory limits; and TPX can be designed, constructed, and operated to meet all regulatory requirements View full abstract»

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  • High heat flux experiments of plasma facing components for next fusion devices

    Page(s): 830 - 833 vol.2
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    To develop plasma facing components (PFC) for the next fusion devices, JAERI has been carrying out high heat flux and high particle flux experiments on the divertor modules and candidate materials in JAERI Electron Beam Irradiation System (JEBIS). (1) To investigate the feasibility and the advantage of a saddle type divertor modules, which has unidirectional (1-D) carbon fiber reinforced carbon composites (CFCs) armour tile brazed on OFHC-copper heat sink, high heat flux experiments have been carried out under a cyclic heat load of 24.5 MW/m 2 at a duration of 30 s. After 1000 cycles, no degradation of thermal response and no defect in the module was found. (2) To reduce the residual stress around the brazed interface, we have developed small specimens with the new materials combination of W-30Cu composites heat sink and 1-D CFC armour tile and carried out the high heat flux experiments under a cyclic heat load of 15 MW/m2 at a duration of 20 s. After 1,000 cycles, no cracks have been observed at the interface. (3) To evaluate the erosion of armour tiles by high heat flux, we have measured the erosion of CFCs and isotropic graphite up to 1100°C under a heat flux of 1800 MW/m2 for the duration of 1.5~2 ms. It is clear that the erosion of carbon based materials increases with the bulk temperature and decreases with the thermal conductivity. (4) To evaluate the erosion by high particle flux, we have developed a new irradiation device, which have a high hydrogen particle flux of 1021/m2/s at 50~100 eV View full abstract»

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  • Research and development assessments for Prometheus heavy ion and laser driven inertial fusion energy reactor designs

    Page(s): 989 - 992 vol.2
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    Research and development assessments for two inertial fusion energy (IFE) reactor design concepts developed in the Prometheus study are presented in this paper. The assessment here focuses on resolving the technical issues identified for the critical components unique to IFE: target, driver, and cavity/first wall. The two designs considered are based on heavy-ion and laser drivers View full abstract»

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  • Development of high strength electrical connections using copper electrodeposition

    Page(s): 1125 - 1127 vol.2
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    The poloidal field coils on the Alcator C-Mod Tokamak are manufactured from C107 or C110 half hard copper strip. A copper C10100 flag is joined to the ends of the coil turns in order to transfer the current from the coaxial leads. The coil design requires that any joint made between the flag and the coil turn be as strong, statically and in fatigue, as the coil strip. Standard methods such as welding, brazing and hard soldering would reduce the tensile strength by annealing the copper in the area of the joint. A soft solder connection with a full lap joint was not considered due to a premature failure in an initial test. In investigating alternative joining methods, tests were performed on samples of electrodeposited copper provided by the A.J. Tuck Co., of Brookfield, Connecticut. The tests showed that the joint met C-Mod structural specifications at 77K. Encouraged by these results, M.I.T., embarked on a testing program involving electroformed butt joints. Through an extensive engineering-program, reliable electroformed joints were obtained in production. The production engineering involved the construction of a special facility to control the process parameters. Details are presented here for duplicating this facility View full abstract»

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  • Graphical user interface of RFX control and data acquisition system

    Page(s): 577 - 580 vol.2
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    The paper presents the user interface aspects of the control and data acquisition system (SIGMA) of the RFX nuclear fusion experiment. SIGMA uses programmable controllers (PLCs) for slow signals and VAX based CAMAC for the fast data acquisition and timing tasks. The paper briefly summarises the technical aspects of the user interface: PLC based systems use operating consoles which employ IBM PC/AT compatibles running a commercial plant supervision package. The VAX based part employs VAXstations as user consoles with an extensive graphical user interface (initially DECwindows, now Motif). The use of these tools is illustrated by presenting the entire operation sequence for an RFX shot View full abstract»

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  • Design of the TPX limiter and armor components

    Page(s): 1193 - 1196 vol.2
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    The TPX limiter and armor systems are designed for steady-state operation from day one operation, at 18 MW plasma input power, to a possible upgrade to 45 MW. All components are designed for remote handling. Carbon-carbon (C-C) composites are the baseline plasma facing material for all limiter and armor systems. Where applicable, all components are made from low activation materials. The TPX limiter system consists of the inboard toroidal limiter, the outboard toroidal limiter, and three discrete poloidal limiters. These limiters are used for plasma startup and to protect the vessel, passive plates, and equipment in the ports from the energetic particle fluxes during steady-state operation. In addition, the inboard limiter protects the vacuum vessel from steady-state neutral beam shine-though and from neutral beam faults. The TPX armor components consist of two major systems: the neutral beam armor that protects the outer vessel wall and equipment in the ports, and the ripple armor that intercepts the trapped energetic particles that are drifting vertically in the ripple region. Different design concepts are employed for these plasma facing components (PFC) depending on their expected heat loads. Inboard and outboard limiters are designed with mechanically restrained C-C composite tiles mounted on cooled support plates. Components which must withstand higher heat loads, such as neutral beam and ripple armor, are made of C-C composite tiles brazed to actively-cooled copper View full abstract»

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  • Plasma vertical stability and feedback control for TPX

    Page(s): 638 - 641 vol.2
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    The n=0 axisymmetric vertical stability and vertical position control have been examined for the Tokamak Physics Experiment. The passive stabilization is accomplished by using stabilizer plates close to the plasma. The present configuration is found to provide robust stability over a wide range of plasma parameters. The active feedback control of the plasma vertical position is done using coils located inside the vacuum vessel. These are required to control random disturbances leading to ⩽1.0 cm RMS displacements from the midplane, and acceptable coil currents and voltages are found View full abstract»

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  • Electromagnetically plugged gaseous divertors [fusion reactor]

    Page(s): 846 - 849 vol.2
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    The novel gas box divertor concept is based on the presence of an electromagnetic force on the neutral gas-plasma boundary that confines the neutral gas and impurities away from the main plasma chamber. The electromagnetic force is generated by J×B forces on a high density plasma with a low degree of ionization. The force is transmitted from the plasma to the neutral gas through ion-neutral drag. The plasma volumetric force balances the pressure gradient in the neutral gas. The performance of argon and hydrogen target plasmas has been compared and the divertor region modelled. Since the plasma and the neutral gas in the divertor chamber are highly collisional, a fluid analysis over most of the region is appropriate View full abstract»

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  • Nuclear analysis of the first wall/blanket/shield of the Prometheus inertial fusion energy reactors

    Page(s): 765 - 768 vol.2
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    The design of the cavity and shielding system of the Prometheus reactors was undertaken by UCLA in conjunction with an IFE conceptual reactor study led by McDonnell Douglas. A wetted wall by a thin lead film was used as part of the first wall system (FWS) designated to protect the cavity against the microexplosions. The FWS is attached to the blanket/reflector/plena (B/R/P) system. For low activation consideration, SiC was used as the structural material and low pressure helium (1.5 MPa) was chosen as the coolant in the Li2O B/R/P system. In this paper, the results of the neutronics parametric study that led to the baseline design of the FWS and B/R/P system are given. The attenuation characteristics of two types of shielding materials were studied, namely, reinforced concrete and low-activation Pb/B4C composite. The latter was chosen as the bulk and beamlines shield due to its low activation View full abstract»

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  • Tritium outgassing from solid waste

    Page(s): 979 - 982 vol.2
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    Waste disposal costs and volumes of waste produced in large scale tritium facilities point towards the need for categorizing the waste by level of activity and for reducing the quantity of waste sent to the facility. Reliable, labour minimized techniques for measuring the total activity contained within each waste parcel destined for a waste storage facility underpin the viability of this goal. A program examining the feasibility of classifying and discriminating the various types of tritiated waste through outgassing is underway. The types of waste examined are metal and non-metal. The metallic waste consists of stainless steel, copper and aluminium which have been used in tritium systems. The non-metallic waste consists of vinyl gloves and paper products generated on a daily basis in our laboratory. An effort is made to correlate the global outgassing rates with the tritium inventory in the waste. The tritium inventory is estimated by soaking selected waste components and counting the activity of the leachant. The outgassed tritiated species are tracked by pulling air through the waste bag into a water bubbler with a pump. Water soluble species are then trapped within the bubbler. Outgassing rates of >0.1 nCi/day are detectable in our experiment. Outgassing rates ranging from ~1 to 2000 nCi/kg.day are measured. Waste inventories varied from ~1 to 1000 μCi View full abstract»

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  • TdeV control system upgrade

    Page(s): 573 - 576 vol.2
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    The continuous enhancement of the TdeV tokamak has exhausted the capacity of the original control system (two PDP 11/73 computers with CAMAC interfaces). Recently, new subsystems using Siemens PLCs and VME-based rtVAX had to be installed in parallel with the original control system. It was then decided to upgrade the TdeV control system while preserving the existing CAMAC interfaces. This system is a hierarchical, decentralized control system with a distributed database. The Vsystem software package, from Vista Control Systems Inc., is used to implement the supervisory control function, the man-machine interface (X-Windows), which run on VAXstations 4060, and the control database, which is distributed on the workstations and VME-based rtVAX processors running VAXeln. The CIMPlus PDS VMS driver, from Data Concepts Inc., has been used in Vsystem to implement the communication with Siemens PLCs. All the processors, including the PLCs, communicate via Ethernet View full abstract»

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  • Manufacturing and testing of a relevant scale mockup based on monoblock concept

    Page(s): 871 - 874 vol.2
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    The results obtained from small-scale mockups manufactured on the monoblock design concept have proven that the solution appears promising for a conventional divertor operating with heat fluxes in the range 10 to 15 MW/m2 with a thermal fatigue cycle exceeding 1000 cycles at full power. The concept developed in laboratory-scale brazing tests and the results of component design optimization have been implemented in a larger multi-tube mockup with the following goals: (a) to demonstrate the possibility of manufacturing defect-free industrial components on a standard basis; (b) to establish the feasibility and performance of a number of design features, including supporting concepts, tile shaping, etc.; (c) to check the capability of NDE techniques to predict defects relevant to the thermal performance of the component. The divertor mockup consists of six half-meter-long armored tubes obtained by brazing CFC to TZM molybdenum alloy. Two types of CFC were used to investigate the advantages of 3-d CFCs with respect to more conventional and cheaper 2-d CFC. The brazing process utilizes three variants of a process developed in laboratory trials and based on selected combinations of active braze filler/CFC surface conditioning procedures. The supporting structure is based on the sliding support concept intended to assure a compromise between the requested thermal stability of the component and the buildup of secondary stresses deriving from mechanical constraints. The FE thermal and thermal mechanical analysis of the divertor mockup structure is reported and the critical areas of sliding support are highlighted for comparison with experimental results. The main results of NDE and experimental high heat flux tests are reported and discussed View full abstract»

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  • The TPX superconducting magnet fabrication process

    Page(s): 798 - 801 vol.2
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    The TPX Toroidal Field (TF) and Poloidal Field (PF) Coils use both Nb3Sn and Nb-Ti internally cooled, cabled superconductor (ICCS). A prime consideration in the design of the coils was the minimization of joints in the conductor cable to save space, reduce heating losses sand increase reliability. Therefore, the traditional double pancake approach was discarded in favor of the continuously wound coil. The recently developed use of the roll bender in winding the coils permits the backwinding, which is required in this geometry. The Nb3Sn TF coils, the central solenoid coils and inner-most PF coils require the wind-react-insulate sequence. The Nb-Ti outer PF Coils are also insulated after the coils are wound because the insulation system used in all the coils requires the insulation of rigid polyimide sheet between turns and pancakes before Vacuum Pressure Impregnation (VPI) of the glass epoxy system. This paper proposes and discusses a manufacturing plan based upon the use of the roll bender in the winding of the coil, and a special fixture to spread apart the turns of the coil and permit the application of the insulation materials prior to VPI View full abstract»

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  • Chemical and Waste Management Tracking and Report-Generating Systems

    Page(s): 927 - 929 vol.2
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    The Chemical and Waste Management and Report-Generating System are software packages for personal computers to provide tracking and inventory controls, and report preparation capabilities for compliance with local, state and federal regulations. Designed by Princeton Plasma Physics Laboratory, these packages will: streamline the inventory and tracking of chemicals and waste products; and provide cost-effective production of required reports in standardized formats View full abstract»

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  • Anomalous behaviour in ICCS

    Page(s): 1128 - 1133 vol.2
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    ICCS conductors behave anomalously. Several anomalies have been documented: quenches driven by pulsing fields (referred to as ramp-rate limitations), and energy margins that show a transition between a “well-cooled” and an “ill-cooled” regime (marking a transition between high energy margin at low transport current and low energy margin at higher currents). In this paper, a novel model will be described that attempts to interpret experimental data. In the experiments, 3×3×3 ICCS Nb3Sn sample coils have been tested for stability to ramping fields and transport currents. Three such coils have been tested, one made of triplets made of all SC strands, a second with one copper and two SC strands and a third one with one steel and two SC strands. The analysis deals mainly with unpressurized cases, but the model also applies to supercritical He. The hypothesis is that non-uniform helium and superconductor distribution in the conductor is the cause of the “anomalous” behaviour. The energy margin as a function of the transport current has also been calculated, and it shows a transition. The limiting current is determined by the highest local fraction of bunching (averaged over a cross section). The model, however, has difficulty explaining multiple ramps View full abstract»

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  • Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    Page(s): 558 - 561 vol.2
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    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D's efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well as a summation waveform indicative of the total injected power act a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor running at 20 MHz. Data acquisition is accomplished using a National instrument's 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments' LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator's logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers end added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system's function and capabilities View full abstract»

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  • Drastic improvement of Ic of Nb3Sn CIC conductor by prestraining at room temperature

    Page(s): 1166 - 1169 vol.2
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    The strain sensitivity of Nb3Sn cable is well known. However the practical process to compensate for this effect when 316 LN is used for the jacket has never been considered. In this paper different proposals are analysed in order to prevent the 316 LN jacket contracting more than the Nb3Sn cable. A first experiment performed in the FBI test facility of KfK has shown that a prestrain of 0.3% carried out at 275 K on a short straight sample of cable in conduit conductor (3×3×4 Nb3Sn strands of 0.73 mm in a 316 L conduit) produced an improvement of the critical current. The improvement in this condition is about 80%. Different designs of tooling usable for the CS and TF coils of ITER are described View full abstract»

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  • Dynamic tritium accountancy for ITER

    Page(s): 969 - 974 vol.2
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    For the time being fusion technology development involves relatively small quantities of tritium. Consequently, it is sufficient to apply so-called “conventional” accountancy tools. However, it is foreseeable that tritium operations-and thus the amount of tritium-will increase substantially. An advanced accountancy methodology will satisfy the resulting new requirements. In this study the advanced accountancy methodology is developed and applied to the situation envisaged with an ITER-type fuel cycle. Firstly, this task comprises modeling of fuel cycle operations, providing the “true” data of the in-process inventories. As both the fuel cycle subsystems and networking them are susceptible to changes, flexible tools of simulation are necessary. Secondly, a measurement model takes care of the true data, handles data reduction, and applies mathematical methods to confirm the final inventories on a statistical basis. Then, in a third step, the test statistics might verify whether or not tritium has been lost View full abstract»

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  • 4 MW upgrade to the DIII-D fast wave current drive system

    Page(s): 1073 - 1076 vol.2
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    The DIII-D fast wave current drive (FWCD) system is being upgraded by an additional 4 MW in the 30 to 120 MHz frequency range. This capability adds to the existing 2 MW 30 to 60 MHz system. Two new ABB transmitters of the type that are in use on the ASDEX-Upgrade tokamak in Garching will be used to drive two new water-cooled four-strap antennas to be installed in DIII-D in early 1994. The transmission and tuning system for each antenna will be similar to that now in use for the first 2 MW system on DIII-D, but with some significant improvements. One improvement consists of adding a decoupler element to counter the mutual coupling between the antenna straps which results in large imbalances in the power to a strap for the usual current drive intrastrap phasing of 90°. Another improvement is to utilize pressurized, ceramic-insulated transmission lines. The intrastrap phasing will again be controlled in pairs, with a pair of straps coupled in a resonant loop configuration, locking their phase difference at either 0 or 180°, depending upon the length of line installed. These resonant loops will incorporate a phase shifter so that they will be able to be tuned to resonance at several frequencies in the operating band of the transmitter. With the frequency change capability of the ABB generators, the FWCD frequency will thus be selectable on a shot-to-shot basis, from this preselected set of frequencies. The schedule is for experiments to begin with this added 4 MW capability in mid-1994. The details of the system are described View full abstract»

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  • Structural analysis of the TPX plasma facing components

    Page(s): 1197 - 1201 vol.2
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    Structural design calculations and finite element analyses of the conceptual designs of the Tokamak Physics Experiment (TPX) divertor and inboard limiter were performed for thermal and electromagnetic induced mechanical loads. Finite element analyses of the divertor and inboard limiter support structures were performed for loads due to halo currents and eddy currents due to a plasma disruption. The results show the conceptual designs satisfy primary stress allowables. A number of scoping studies were performed to evaluate the thermal and structural response of various tile materials and designs for the TPX divertor. The purpose of these studies was to investigate what possible gains would occur if the present design requirements for the heat flux surfaces were eased. The studies were performed for beryllium and various carbon-carbon materials brazed to a dispersion strengthened copper tube. The studies included the effects of a soft copper compliant layer of varying thicknesses interfacing the copper tube and the tile. Elastic-plastic thermal stress analyses were performed of 1D, 2D, and 4D carbon-carbon and beryllium monoblock designs and for a saddleblock design with 1D carbon-carbon. The residual stresses and amount of plastic straining in the copper tube during the braze cycle are accounted for in computing the stress state after the brazing process and during steady state operating conditions View full abstract»

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  • Quench detection & instrumentation for the Tokamak Physics Experiment magnets

    Page(s): 802 - 805 vol.2
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    The design of the Local Instrumentation & Control (I&C) System for the Tokamak Physics Experiment (TPX) superconducting PF & TF magnets is presented. The local I&C system monitors the status of the magnet systems and initiates the proper control sequences to protect the magnets from any foreseeable fault. Local I&C also stores magnet-system data for analysis and archiving. Quench Detection for the TPX magnets must use a minimum of two independent sensing methods and is allowed a detection time of one second. Proposed detection methods include the measurement of; (1) normal-zone resistive voltage, (2) cooling-path helium flow, (3) local temperature in the winding pack, (4) local pressure in the winding pack. Fiber-optic based isolation systems are used to remove high common-mode magnet voltages and eliminate ground loops. The data acquisition and fault-detection systems are computer based. The design of the local I&C system incorporates redundant, fault-tolerant, and/or fail-safe features at all component levels. As part of a quench detection R&D plan, a Quench Detection Model Coil has been proposed to test all detection methods. Initial cost estimates and schedule for the local I&C system are presented View full abstract»

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  • Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    Page(s): 917 - 920 vol.2
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    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The closed cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed View full abstract»

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  • Design methods and actual performances of conductors for the superconducting coils of tokamaks

    Page(s): 667 - 670 vol.2
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    Conductors presently retained for the coils of large fusion machines are cable-in-conduit conductors made of about one thousand of superconducting strands cabled inside a jacket and cooled by a forced flow of helium. The current density is the key parameter for die machine design, as it reacts strongly on the size and on the cost. An optimized conductor design has been pursued to maximize the current density while fulfilling several criteria regarding protection, stability and safety margin. The influence of some parameters such as: field, strain, effective diameter of filaments and heat transfer coefficient is analyzed. A Nb3Sn conductor developed by CEA relevant for NET/ITER has been tested successfully at 6.2 K up to 50 kA and 12 T. Following the same concept a complete design of a conductor for the central solenoid of ITER is proposed View full abstract»

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