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The cruising range of an electric vehicle (EV) can be greatly extended by a dynamic wireless charging system because the battery can be charged while the vehicle is moving. However, the investment in infrastructure can be enormous and has to be reduced to implement the system. This paper presents the design of a wireless power transfer system with road embedded transmitter coils to facilitate dynamic charging of EVs. A 30 cm by 1.6 m transmitter coil, which is relatively long compared with the 40-cmdiameter circular receiver coil, has been designed to reduce the cost of the transmitter circuit. A prototype transmitter coil has been embedded into a road for evaluation of its characteristics. The measured coupling coefficient of the transmitter and receiver coils is as small as 0.09 because of the asymmetry between the coils. Compensation circuits for the transmitter and receiver coils have been designed and optimized for such a small coupling coefficient condition. The designed dynamic charging system can transfer more than 1 kW of electric power with more than 90% efficiency. Five transmitter circuits have been embedded into a road surface for testing. Cement asphalt mortar was used to construct the road surface in this study to avoid damaging the transmitter coils by high temperature and high pressure during the construction work. The designed receiver circuit has been installed on a compact 2-seater EV. A demonstration was conducted to show that the retrofitted EV can successfully receive electric power transferred from the road.

Future experimental fusion devices and power reactors will continuously pump large amounts of helium, and deuterium and tritium (D-T). D-T to helium ratios greater than 10:1 are projected. Cryopumps are the logical choice to pump gases in the 10/sup 3/-m/sup 3//s range. A test program undertaken at Los Alamos that evaluated the capability of copumping the fusion exhaust gases on a liquid-helium cooled surface of activated carbon (coconut charcoal) is described. A compound cryopump installed in the VAC system of the Tritium Systems Test Assembly (TSTA) was used as the test article. In the test, copumping of helium and deuterium, and helium-3 and tritium was undertaken. Test results show that copumping is practical and that compound pumps may not be required for fusion applications. >

An all-metal apparatus has been constructed and installed in the main cell of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The apparatus is being used in an experiment that is part of a study to determine the feasibility of using metal tritides for storing tritium in fusion fuel processing. The structure will handle about 2600 Ci of tritium in the experiment. It allows measurement of equilibrium pressure on metal-hydrogen systems and was made to assure safe handling of tritium. The pressure-composition isotherms for the LaNi/sub 3/Mn/sub 2/-protium (H), deuterium (D), and tritium (T) systems were measured to study isotopic effects in the temperature range of 60 degrees C to 250 degrees C, and the pressure range below 120 kPa. >

Several experiments were performed with the Isotope Separation System (ISS) at the Tritium Systems Test Assembly (TSTA) from 1988 through 1989. These experiments included two-column experiments with D-T mixtures and four-column experiments with D-T mixtures and impurities (He and H/sub 2/). The objectives of these experiments were (1) to obtain fundamental data on cryogenic distillation columns under a wider range of operational condition than in previous work, (2) to observe the influence of He on the operation on cryogenic distillation columns, and (3) to obtain fundamental data on cryogenic distillation columns under the influenced of He. Previous ISS experiments have been performed to obtain fundamental data by using a single column or two columns of the ISS at the TSTA with D-T mixtures. In previous experiments, the HETP (height equivalent to a theoretical plate) value and the pressure drop in the packed section were measured in the vapor velocity range of 5-15 cm/s and in the reflux ratio of 5-40. The result showed that the HETP value was not influenced by the vapor velocity and there was no evidence of a flooding point or loading point. This year, the experiments were performed in the vapor velocity range of 2-15 cm/s and reflux ratio range of less than 230. Continuous separation of He from D-T mixtures was tested for the purpose of obtaining data on ISS operation under the influence of He. >

The Tritium Systems Test Assembly (TSTA) project at Los Alamos National Laboratory is charged with developing and demonstrating the fusion fuel processing requirement for the magnetic fusion energy program. The key fusion fuel streams are the plasma exhaust and the blanket stream. At present, the technology under development at TSTA includes only the plasma exhaust. The blanket is a logical upgrade to TSTA and the upgrade is the subject of an ongoing study. The program, called the Breeding Blanket Interface (BBI), is being investigated with the intention of installing such a processing system at TSTA in the 1992-1993 time frame. The BBI program was initiated in July, 1987. The conditions of the tritium product streams of various blankets were determined. Pre-conceptual designs of the BBI systems were developed for two blanket concepts: aqueous lithium salt solution blanket and solid breeder blanket. Preliminary estimates of costs have been performed and the schedule of the TSTA/BBI program has been developed. The schedule and planning of the TSTA/BBI program are discussed. The goal of the program is to establish the engineering data basis of the fuel cycle for an International Thermonuclear Engineering Reactor. >

Research and development of a fusion reactor fuel-processing system are being performed with a large amount of tritium at the Tritium Process Laboratory (TPL). Major experiments involve the following: the fuel cleanup system (FCU) for processing simulated plasma exhaust; the isotope separation system (ISS) with cryogenic distillation; the tritium permeation study with triton ion beam; isotope separation with the thermal diffusion columns; and analysis and measurement of tritium. Experiments have been performed at TPL since 1988. In recent experiments, some results are obtained with and without a large amount of tritium. >

The Tritium Systems Test Assembly (TSTA) at Los Alamos, New Mexico, is a full-scale facility dedicated to testing tritium processing for fusion reactors. Adding a breeder blanket interface (BBI) to the TSTA is being studied. The BBI is to test the processing required for the tritium output streams for the various fusion-reactor breeder blankets. In the current phase of the study, the characteristics of the output from the various blanket types are being evaluated. Defining the output stream with respect to H/T ratio, impurity content, and radionuclide content is emphasized. Assessments of the solid breeder blanket (ceramic, Li/sub 2/O) and the aqueous salt solution blanket are reported. For the aqueous lithium salt solution blanket, the chemical systems are very complex, involving a number of phenomena, including radiolysis, corrosion, and electrolysis. A system of about 0.01 ITER scale may be added to the TSTA. For the solid breeder blanket, no critical issues are identified. A system of about half-ITER scale, possibly even full scale, may be added to the TSTA. >

The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory has conducted its longest continuous integrated loop operation to data. This provided an opportunity to test some hitherto unproven capabilities of the TSTA fuel cleanup system (FCU). The purpose of the FCU is to remove impurities from a stream of hydrogen isotopes (Q/sub 2/) representative of torus exhaust gas. During this run, impurities loadings ranging from 60 to 179 sccm of 90% N/sub 2/ and 10% CH/sub 4/ were fed to the FCU. Each of the two FCU main flow molecular sieve beds (MSBs) were filled to breakthrough three times. The MSBs were regenerated during loop operations. The experiment yielded two principle findings. The first is that using the cryogenic molecular sieve is very robust method for removing impurities from fusion fuels. While loop flow rates, pressures, and impurities concentrations varied, the MSBs continued unaffected, completely removing impurities. The second is that the first step in the MSB regeneration process proved to be very effective. >

Tritium Systems Test Assembly (TSTA) in the Los Alamos National Laboratory is a reactor-scale fusion fuel processing loop and is operated by the Japan Atomic Energy Research Institute (JAERI) and the USDOE under the collaboration. An extended experimental campaign of the TSTA loop was performed with 100 grams of tritium in April-May 1992 in order to simulate a steady-state fusion fuel processing for 25 days. Continuous processing of simulated plasma exhaust (DT and H mixture, He, CH/sub 4/ and N/sub 2/) was successfully demonstrated. Impurity processing was successfully tested by two different techniques. US developed Fuel Cleanup System (FCU) purifies DT by cryogenic molecular sieve beds and decomposes tritiated water by Magnesium beds. Japanese developed J-FCU is based on purification by permeation through palladium alloy membrane and vapor electrolysis with a ceramic electrolyte cell. Isotope Separation System was stably operated with four interlinked columns and continuously produced pure tritium and deuterium stream while exhausting protium impurity. Real-time analysis in the ISS process was demonstrated with the laser Raman Spectroscopy. The result demonstrated the safe and stable operation of reactor-scale fusion fuel loop in a steady state.