[en] OAK-B202: Temperature Profiles and (ell)=1 Nonuniformity within Cryogenic ICF Targets

No abstract available

[en] Scoping studies for the Next European Torus (NET) using the SUPERCOIL system code are described. Capital-cost-optimized devices satisfying constraints imposed on stresses/strains, fields, access, etc., are compared. The main objective is to determine the impact of design characteristics, performance objectives, and underlying plasma physics assumptions on the parameters and cost of NET. The background against which the main parameters of NET have been chosen is developed and illustrated by the NET study points used during the conceptual design phase. Supporting studies extrapolating NET design and physics assumptions to DEMO and power reactors are performed to allow the reactor relevance of the physics performance and testing program of NET to be justified

[en] This paper discusses tritium technology in the united States which has advanced considerably since the 1988 Tritium Conference in Toronto. This advance has come in facilities, processing, and safety related technologies and in an ever increasing commitment to compliance related issues. The major laboratories in the U.S. tritium programs continue to be (Westinghouse) Savannah River Site, EG and G Mound, Lawrence Livermore National Laboratory, Sandia National Laboratory and the Los Alamos National Laboratory. Each of the Laboratories have made some significant changes in their programs and/or facilities in the past four years

[en] In order to discuss the long-term reliability and safety of zirconium-cobalt alloy for tritium use, the release behavior of decay helium from ZrCo tritide has been investigated for one and a half years with a radio-gaschromatograph. In this paper, the results show that the release fractions of the total decay ³He in ZrCo tritide are less than 3% and has been almost constant for 18 months under the following conditions: the operating temperatures = 293 ∼ 523 K, the atom ratios (T/ZrCo) = 0.3 ∼ 1.4, and the number of hydrogenation-dehydrogenation cycles before tritiation = 1 ∼ 10. Moreover, residual decay ³He was not released even if ZrCo was heated to 873 K, though most of the tritium was released. It became clear that the decay ³He was quite immovable in ZrCo tritide under these experimental conditions

[en] The ITER Conceptual Design Activity (CDA) was a three-year, 400 professional-year effort to design a next step tokamak. The Activity was conducted under the auspices of the IAEA jointly by EURATOM, Japan, the USSR and USA. The main ITER parameters are summarized in the paper. An engineering design phase (EDA) lasting 5-6 years is planned to begin in 1992. Fuel Cycle design studies carried out as part of the CDA concluded that suitable options existed or could be developed to satisfy all tritium-handling requirements for the machine within the EDA time and resource framework. During the EDA, special emphasis will be required on design integration and optimization

[en] The R and D for tritium technology and studies of tritium science in Japan has been carried out in the national institutions (JAERI, PNC and others), in the universities, and in industry. In fusion fuel technology, the technological R and D of tritium processing and handling has been carried out by TPL-JAERI, including collaboration with TSTA (United States) and basic and innovative research conducted by the universities. PNC has conducted R and D and engineering studies in the heavy water quality control system for FUGEN reactor, a heavy water cooled reactor ATR, and in the tritium recovery system for the heavy water coolant. Since the last meeting in Toronto, there are many developments in the above researches which will be summarized in this paper with some results

[en] In this paper a nuclear fusion process in the dielectric medium of submicrometre metal particles is proposed that reduced the Coulomb repulsion forces. The process may allow nuclear fusion at a rather moderate lukewarm temperature of 10,000 K

[en] Analytical studies on a traveling wave direct energy converter (TWDEC) for D-³He fueled fusion are carried out. The energy of 15 MeV carried by fusion protons is too high to handle with an electrostatic device. The TWDEC controls these high energy particles on the base of the principle of a Linac. This traveling wave method is discussed and the details of proton dynamics and excitation mechanism of electric power are clarified. The TWEDC consists of a modulator and decelerator. The applied traveling wave potential to the modulator modulates the velocity of fusion proton beams. This modulation makes a form of bunched protons at a down stream of the modulator. The decelerator has a set of meshed grids, each of which is connected to a transmission circuit. The phase velocity of excited wave on the transmission circuit is controlled the same way as that of decelerated protons. The kinetic energy 15 MeV of proton beams changes into an oscillating electromagnetic energy on the transmission circuit. This highly efficient direct energy converter of fusion protons brings a fusion reactor with a high plant efficiency. 4 refs., 4 figs

[en] A rotational instability is observed in a field-reversed configuration plasma. Onset time, growth rate and modal frequency of the instability are measured in connection with the ion diamagnetic drift frequency over a wide rage of plasma parameters. When the plasma is stabilized by a quadrupole field, it shrinks axially with time and its particle confinement time becomes shorter than that of a nonstabilized plasma. A numerical calculation of the field profile reveals that the distortion of the confinement field by the quadrupole field at the ends of the separatrix is responsible for the degradation of particle confinement. However, the multipole field with a higher pole number than the quadrupole can stabilize the plasma without degradation of the particle confinement. 13 refs., 6 figs