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Fusion Engineering, 1995. SOFE '95. Seeking a New Energy Era., 16th IEEE/NPSS Symposium

Date Sept. 30 1995-Oct. 5 1995

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  • 16TH IEEE/NPSS SYMPOSIUM FUSION ENGINEERING

    Publication Year: 1995
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    Freely Available from IEEE
  • Keyword Index

    Publication Year: 1995
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    Freely Available from IEEE
  • Author index

    Publication Year: 1995
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    Freely Available from IEEE
  • Helium and tritium retention and migration in beryllium

    Publication Year: 1995 , Page(s): 948 - 951 vol.2
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    Efficiency of the application of beryllium for making such elements of ITER as neutron multiplicator, first wall coating and divertor plates depends on the helium and tritium mobility in this metal. Release parameters of helium and tritium from two grades of isotropic beryllium have been investigated after neutron irradiation at 550, 620, 780°C with fluence of 2.6·1021 cm-2 3.5-1021 cm-2 and 2.6·1021 cm-2 (E>0.1 MeV), correspondingly. Isothermic multi-stage annealings in the temperature range from 400 to 1300°C with the simultaneous monitoring of release gases was fulfilled by the mass-spectrometric method. It was shown that an intensive release of tritium took place at 500°C, while the first signs of helium release were revealed above 600°C. On the base of data obtained the diffusion parameters (D0, E) both gases in beryllium were calculated. Tritium mobility was found approximately three orders of magnitude as high as that of helium. The total amount of helium accumulated in the irradiated beryllium varied from 0.59 to 1.48 ntp cm3·g-1. Helium to tritium ratio was estimated as about 4. Beryllium swelling as a function of the quantity of retained helium is presented View full abstract»

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  • RF power diagnostics and control on the DIII-D, 4 MW 30-120 MHz Fast Wave Current Drive System (FWCD)

    Publication Year: 1995 , Page(s): 866 - 869 vol.2
    Cited by:  Papers (2)
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    The Fast Wave Current Drive System uses three 2 MW transmitters to drive three antennas inside the DIII-D vacuum vessel. This paper describes the diagnostics for this system View full abstract»

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  • Molecular dynamics simulation of self-sputtering of plasma facing material, Be

    Publication Year: 1995 , Page(s): 952 - 955 vol.2
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    Sputtering phenomenon can be evaluated by the molecular dynamics simulation approach as well as the experimental approach. Authors developed the evaluation system of sputtering using molecular dynamics and investigated the feasibility of this approach. This approach was applied to self-sputtering of beryllium because a selection of interatomic potential was comparatively easy. The energy of incident particles ranged from 30 to 300 electron-volts (eV). The incident angle was changed from 0 to 75 degrees normal to the surface of beryllium target. The sputtering yield and damage due to an incident particle was calculated using this approach. The results were compared with ones from references. The subjects was described when this approach was applied to the sputtering rate of plasma facing material View full abstract»

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  • Cyclotron radiation transport in D-3He fusion toroidal systems

    Publication Year: 1995 , Page(s): 1127 - 1130 vol.2
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    Heat transport by electron cyclotron radiation (ECR) in inhomogeneous high-temperature fusion plasmas in a strong magnetic field is analyzed both analytically and numerically within the framework of self-consistent description of ECR transport in plasmas in a strong magnetic field and kinetics of plasma electrons. Numerical results for the total ECR losses and spatial profile of ECR energy balance are presented for ITER-like parameters and the importance of profile effects in ECR energy balance-due to nonlocal (non-diffusive) character of ECR transport in fusion toroidal plasmas-is shown View full abstract»

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  • Engineering overview of the National Spherical Tokamak Experiment

    Publication Year: 1995 , Page(s): 1430 - 1433 vol.2
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    The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio (R/a=1.25; R=80 cm) device designed for a plasma current of 1 MA. It features auxiliary heating and current drive and a close-fitting conducting shell to maximize plasma pressure (43% beta). NSTX is designed for a 5 sec. experimental pulse to demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design takes maximum advantage of existing components and facilities from previous devices at PPPL to reduce the program costs. The device will be sited in the former Princeton Large Torus (PLT) test cell and will utilize the PLT radiation shielding, base structure, and cell utilities. NSTX will utilize the S-1 Spheromak vacuum vessel, poloidal field coils, and capacitor banks (for helicity injection). The Poloidal Beta Experiment-Modified (PBX-M) power supplies will be shared to power the PF and TF coil systems. Existing RF hardware and infrastructure will be used for heating systems. TFTR data acquisition and diagnostics resources are planned to be used. In total, NSTX will utilize site credits with a value of ~$50 M, reducing base construction cost of the device to $18.6 M. Twelve water-cooled copper demountable toroidal field (TF) coils provide the 5.4 kg (pulsed) and 3.5 kg (long pulse >5 sec) toroidal field at the plasma center. Poloidal fields are generated by windings contained in the center column and four pairs water-cooled copper coils supported directly on the vacuum vessel. One of the most critical components of the device is the center stack, which consists of the inner legs of the TF coils overwrapped with ohmic heating and poloidal field windings. The ohmic heating coil windings are designed to optimize the V-s and together with the PF coils, produce a flux swing of 1 V-s View full abstract»

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  • Cold versus hot fusion deuterium branching ratios

    Publication Year: 1995 , Page(s): 1622 - 1625 vol.2
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    A major source of misunderstanding of the nature of cold nuclear fusion has been the expectation that the deuterium branching ratios occurring within a palladium lattice would be consistent with the gas-plasma branching ratios. This misunderstanding has lead to the concept of the dead graduate student, the 1989's feverish but fruitless search for neutron emissions from cold fusion reactors, and the follow-on condemnation of the new science of cold fusion. The experimental facts are that in a properly loaded palladium lattice, the deuterium fusion produces neutrons at little above background, a greatly less-than-expected production of tritium (the tritium desert), and substantially more helium-4 than is observed in hot plasma physics. The experimental evidence is now compelling (800 reports of success from 30 countries) that cold nuclear fusion is a reality, that the branching ratios are unexpected, and that a new science is struggling to be recognized. Commercialization of some types of cold fusion devices has already begun View full abstract»

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  • Blanket-relevant liquid metal MHD channel flows: data base and optimization simulation development

    Publication Year: 1995 , Page(s): 956 - 959 vol.2
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    The problems of generalization and integration of test, theoretical and design data relevant to liquid metal (LM) blanket are discussed in present work. First results on MHD data base and LM blanket optimization codes are presented View full abstract»

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  • A rotating and radially sapping electrical double probe for TEXTOR 94

    Publication Year: 1995 , Page(s): 1051 - 1053 vol.2
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    A rotating and sapping (radially fast moving) electrical double probe has been developed for implementation at the tokamak TEXTOR 94. This new diagnostic system allows to measure a large number of parameters in the edge plasma. It consists of the TEXTOR interface, the highly flexible vacuum chamber, the rotating probe system and the linear drive unit. The motors to drive the rotation and the slow and fast displacements are computer controlled for operation during the discharges View full abstract»

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  • Development of a check valve for two-stage pellet injectors

    Publication Year: 1995 , Page(s): 1562 - 1565 vol.2
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    The two-stage light gas gun is presently the most suitable method for injecting solid deuterium pellets at velocities in excess of 2000 m/s. However, the use of this technique in plasma machines presents some drawbacks: (1) the relatively large mass of propellant, which is of the same order of the pellet mass (2) the shape of the pressure pulse, which does not provide an optimum acceleration, unless the pellet is forced in the barrel, providing a sufficient breakout pressure. These problems can be solved completely or in part, by interposing between the pump tube and the barrel a check valve with a suitable opening pressure. In this work a check valve developed at CNPM is described and experimental results are reported View full abstract»

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  • Investigations of radiation resistant beryllium grades for nuclear fusion applications

    Publication Year: 1995 , Page(s): 944 - 947 vol.2
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    R&D results on beryllium with high radiation resistance obtained recently are described in this report. The data are presented on nine different grades of isotropic beryllium manufactured by VNIINM and distinguished by both initial powder characteristics and properties of billets, made of these powders. The average grain size of the investigated beryllium grades varied from 8 to 26 μm, the content of beryllium oxide was 0.9-3.9 wt.%, the dispersity of beryllium oxide ~0.04-0.5 μm, tensile strength-250-650 MPa. All materials were irradiated in SM-2 reactor over the temperature range 550-780°C. The results of the investigation showed, that HIP beryllium grades are less susceptible to swelling at higher temperatures in comparison with hot pressed and extruded grades. Beryllium samples, having the smallest grain size, demonstrated minimal swelling, which was less than 0.8% at 750°C and Fs=3.7·1021 cm-2 (E>0.1 MeV). The mechanical properties, creep and microstructure parameters, measured before and after irradiation, are presented View full abstract»

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  • TPX heating and cooling system

    Publication Year: 1995 , Page(s): 1323 - 1326 vol.2
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    TPX, while having primarily super-conducting coils that do not require water cooling, still has very significant water cooling requirements for the plasma heating systems, vacuum vessel, plasma facing components, diagnostics, and ancillary equipment. This is accentuated by the 1000-second pulse requirement. Two major design changes, which have significantly affected the TPX heating and cooling system, have been made since the conceptual design review in March of 1993. These are (1) changing the vacuum vessel neutron shielding configuration, and (2) reducing the vacuum vessel operating temperature. This paper will discuss these changes and review the current status of the conceptual design. In all, six different heating and cooling supply requirements (temperature, pressure, water quality) for the various TPX components must be met. This paper will detail these requirements and provide an overview of the heating and cooling system design while focusing on the ramifications of the TPX changes View full abstract»

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  • What must DEMO do?

    Publication Year: 1995 , Page(s): 1157 - 1161 vol.2
    Cited by:  Papers (2)
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    The US fusion demonstration plant (DEMO) must satisfy certain top level requirements so that energy producers will confidently invest in a commercial fusion version for their next generation power plant. To instill that level of confidence to both the investor and the public, DEMO must achieve high standards in safety, low environmental impact, reliability, and economics. This is a most difficult set of goals to meet. The public is demanding ever more strict environmental rules and regulations. The hazards of radioactive and toxic waste and emissions are becoming better understood. The difficulties of establishing and maintaining long-lived repositories are enormous. Neighborhood action groups have an aversion to large power plants in their back yards. Utilities and independent power producers are reluctant to commit to a long-term financial arrangement for a new technology. To achieve these stringent goals, the competition is continuing to improve to meet these challenges. Only the best can adapt and survive. Can fusion meet the challenge? Does it have enough advantages to offset the difficulties ahead? Fusion has many inherent advantages. Fusion can be environmentally safe. By tailoring the materials used, it can achieve low level waste standards. It can meet demanding public standards for safety. It does not have a melt-down scenario. Abnormal events can be tolerated with no major hazard potential. Redundancy and robust engineering can be designed into the system and testing can verify demanding reliability standards. Plant economics can be achieved if rigid cost standards are established and maintained. The DEMO plant is not expected to achieve all requirements demanded of the commercial power plant, but it must demonstrate values close enough to the commercial machine so that extrapolation to the commercial carries minimal risk in all key areas. Specifically. DEMO must demonstrate all the major performance parameters in an integrated system similar to that of the commercial plant. It should be large enough so that all aspects of the DEMO can be confidently scaled to that of the commercial plant, including the economics, reliability, availability, and operability. The US Starlite DEMO project is establishing quantifiable top level requirements to assure that DEMO will satisfy the aforementioned needs for the commercial plant. At the same time, it must be determined that DEMO can be achieved by extrapolating today's current physics, engineering, and material databases View full abstract»

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  • Neutronics study of a spherical tokamak for component testing

    Publication Year: 1995 , Page(s): 1434 - 1437 vol.2
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    A conceptual design for the testing of components for future fusion devices has been analysed for its neutronics performance. Three-dimensional Monte Carlo modelling was used to obtain neutron fluxes in regions of interest, as well as reaction rates and atomic displacement rates for assessment of materials damage. The neutron flux spectrum in a typical blanket test module was computed. Coupled neutron/photon calculations provided heating rates at various locations. Results show that the concept is viable from a neutronics point of view, although some additional shielding of the poloidal field coils may be desirable View full abstract»

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  • Magnetic fusion reactor economics

    Publication Year: 1995 , Page(s): 1549 - 1554 vol.2
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    An almost primordial trend in the conversion and use of energy is an increased complexity and cost of conversion systems designed to utilize cheaper and more-abundant fuels; this trend is exemplified by the progression fossil→fission→fusion. The present projections of the latter indicate that capital costs of the fusion “burner” far exceed any commensurate savings associated with the cheapest and most-abundant of fuels. These projections suggest competitive fusion power only if internal costs with the use of fossil or fission fuels emerge to make them either uneconomic, unacceptable, or both with respect to expensive fusion systems. This “implementation-by-default” plan for fusion energy is re-examined by identifying in general terms fusion power-plant embodiments that might compete favorably under conditions where internal costs (both economic and environmental) of fossil and/or fission are not as great as is needed to justify the contemporary vision for fusion power. Competitive fusion power in this context will require a significant broadening of an overly focused program to explore the physics and symbiotic technologies leading to more compact, simplified, and efficient plasma-confinement configurations that reside at the heart of an attractive fusion power plant View full abstract»

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  • Co-wound voltage sensor R&D for TPX magnets

    Publication Year: 1995 , Page(s): 1534 - 1537 vol.2
    Cited by:  Papers (2)
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    The Tokamak Physics Experiment (TPX) will be the first tokamak to use superconducting cable-in-conduit-conductors (CICC) in all poloidal field (PF) and toroidal field (TF) magnets. Conventional quench detection, the measurement of small resistive normal-zone voltages (<1 V) in the magnets will be complicated by the presence of large inductive voltages (>4 kV). In the quench detection design for TPX, we have considered several different locations for internal co-wound voltage sensors in the cable cross-section as the primary mechanism to cancel this inductive noise. The Noise Rejection Experiment (NRE) at LLNL and the Noise Injection Experiment (NIE) at MIT have been designed to evaluate which internal locations will produce the best inductive-noise cancellation, and provide us with experimental data to calibrate analysis codes. The details of the experiments and resulting data are presented View full abstract»

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  • Current ripple in the coils of the TJ-II Spanish stellarator

    Publication Year: 1995 , Page(s): 1070 - 1073 vol.2
    Cited by:  Papers (1)
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    High precision coil current control, stability and ripple content are very important aspects for a stellarator design. The TJ-II coils will be supplied by network commutated current convertors and therefore the coil currents will contain harmonics which have to kept to a very low level. An analytical investigation as well as numerous simulations with EMTP, SABER(R) and other softwares, have been done in order to predict the harmonic currents and to verify the completion with the specified maximum levels. The calculations and the results are presented View full abstract»

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  • Use of a thermal analogy to find electrical resistances of the electrical breaks in the TPX passive stabilization systems

    Publication Year: 1995 , Page(s): 1299 - 1302 vol.2
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    The inner and outer passive stabilization systems for the Tokamak Physics Experiment (TPX) are similar in design in that they both utilize copper passive plates that form large toroidal rings. The rings are electrically continuous except at one toroidal location where a high resistance break must exist. Vertical conductors connect the rings together on either side of the electrical break forming a saddle coil. In order to prevent all the current during initial plasma start-up from flowing through the rings instead of the plasma, the resistances of the breaks for the inner and outer stabilizers must be greater than 70 and 300 μΩ respectively. A thermal-electrical analogy has been developed so that 2-D heat transfer finite element codes can be used to find the electrical resistances in the proposed designs of the high resistance breaks. This analogy is based on classical heat transfer theory using an electrical analogy for finding the equivalent conductances of materials that are in series or parallel. In these cases the conductivities of the materials are converted into conduction resistances. The conduction resistances are associated with actual electrical resistances, the heat transfer rate with current, and the temperature difference with potential drop. Therefore the basic heat transfer equation, q=KΔT, can be used to express the electrical equivalent equation, V=IR as ΔT=q(1/K). By imposing a temperature drop across the 2-D finite element thermal models of a break and having the code determine the total heat flow through the model, the resistance of the break, R=1/K, can then be calculated View full abstract»

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  • CaO insulator and Be intermetallic coatings on V-base alloys for liquid-lithium fusion blanket applications

    Publication Year: 1995 , Page(s): 960 - 963 vol.2
    Cited by:  Patents (1)
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    The objective of this study is to develop (a) stable CaO insulator coatings at the liquid-Li/structural-material interface, with emphasis on electrically insulating coatings that prevent adverse MHD-generated currents from passing through the V-alloy wall, and (b) stable Be-V intermetallic coatings for first-wall components that face the plasma. Electrically insulating and corrosion-resistant coatings are required at the liquid-Li/structural interface in fusion first-wall/blanket applications. The electrical resistance of CaO coatings produced on oxygen-enriched surface layers of V-5%Cr-5%Ti by exposing the alloy to liquid Li that contained 0.5-85 wt.% dissolved Ca was measured as a function of time at temperatures between 250 and 600°C. Crack-free Be2V intermetallic coatings were also produced by exposing V-alloys to liquid Li containing Be as a solute. These techniques can be applied to various shapes (e.g., inside/outside of tubes, complex geometrical shapes) because the coatings are formed by liquid-phase reactions View full abstract»

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  • Effect of large ion angular momentum spread and high current on inertial electrostatic confinement potential structures

    Publication Year: 1995 , Page(s): 1476 - 1481 vol.2
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    Prior inertial electrostatic confinement (IEC) studies have assumed that very low angular momentum (zero in the ideal case) is necessary to achieve a potential well structure capable of trapping energetic ions in the center of a spherical device. However, the present study shows that high-current ion beams having large-angular-momentum spread can also form deep potential well traps View full abstract»

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  • Improving plasma shaping accuracy through consolidation of control model maintenance, diagnostic calibration, and hardware change control

    Publication Year: 1995 , Page(s): 862 - 865 vol.2
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    With the advent of more sophisticated techniques for control of tokamak plasmas comes the requirement for increasingly more accurate models of plasma processes and tokamak systems. Development of accurate models for DIII-D power systems, vessel, and poloidal coils is already complete, while work continues in development of general plasma response modeling techniques. Increased accuracy in estimates of parameters to he controlled is also required. It is important to ensure that errors in supporting systems such as diagnostic and command circuits do not limit the accuracy of plasma parameter estimates or inhibit the ability to derive accurate plasma tokamak system models. To address this issue, we have developed more formal power systems change control and power system/magnetic diagnostics calibration procedures. This paper discusses our approach to consolidating the tasks in these closely related areas. This includes, for example, defining criteria for when diagnostics should be re-calibrated along with required calibration tolerances, and implementing methods for tracking power systems hardware modifications and the resultant changes to control models View full abstract»

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  • Magnetic aspects of the passive stabilizing structure in TPX

    Publication Year: 1995 , Page(s): 1338 - 1341 vol.2
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    The advanced operating modes of TPX require both passive stabilization and active control of critical magneto hydrodynamic (MHD) modes. TPX contains a inner passive stabilizer, an outer passive stabilizer, and a set of active fast position control coils inside the TPX vacuum vessel. The passive structure is placed close to the plasma surface and covers enough of the plasma surface area in appropriate locations to provide good inductive coupling to the unstable modes. Access requirements for heating systems, diagnostics, divertors, and planned maintenance operations constrain the area available for the conductor coverage. This paper describes the TPX passive stabilizing structure. This includes the special analysis techniques developed to (1) assess kink stabilizing performance, (2) approximate these structures as simple axisymmetric structures for startup, disruption, and position control analysis, and (3) estimate the field errors produced by the eddy currents in these structures View full abstract»

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  • Development of repetitive railgun pellet accelerator and steady-state pellet supply system

    Publication Year: 1995 , Page(s): 1566 - 1569 vol.2
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    A railgun system for repetitive high-speed pellet acceleration and a steady-state pellet supply system has been developed and investigated. Using a 2 m-long railgun system, the hydrogen pellet was accelerated to 2.6 km/sec by the supplied energy of 1.7 kJ. It is expected that the hydrogen pellet can be accelerated to 3 km/sec using the present pneumatic pellet accelerator and a 2 m-long augment railgun. A screw-driven hydrogen-isotope filament extruding system has been fabricated and will be tested to examine its applicability to steady-state extrusion of the solid hydrogen-isotope filament View full abstract»

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