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    Elimination of the coil shielding for MFE-reactors through a liquid-protected first-wall
    (King Fahd Univ Petroleum Minerals, 2002) Sahin, S; Sahinaslan, A; Sahin, HM
    The idea of a protective, flowing, liquid zone to protect the first wall of a magnetic fusion energy (MFE) reactor from the direct exposure of the fusion reaction products is not new. This could extend the lifetime of the first wall to the lifetime of the fusion power plant, namely to 30 years. The present work discusses the possibility that such a liquid zone could lead also to the elimination of the magnetic coil shielding for WE reactors. Contrary to a related previous work, the liquid wall is now placed at the outermost periphery of the plasma chamber, in order to leave a greater space for the fusion plasma volume and consequently to lead to higher fusion power with the same plasma parameters. In this work, SS-304 type steel, SiC, and graphite are selected as structural materials. Different types of liquid coolant with tritium breeding capabilities (Flibe, Li17Pb83, natural lithium, all with natural lithium component) are investigated to protect the first wall from neutron- and bremsstrahlung-radiation and fusion reaction debris. The calculations are conducted for a power generation of 1GW(el) over 30 years of reactor operation with a thermodynamically conversion efficiency of 35 % leading to 2.857 GW(th) by a capacity factor of 100 %, The most important improvements through the placement of the protective liquid wall at the outer periphery in the new blanket can be cited as follows. Such a blanket: would in practice not necessitate extra shielding for superconducting coils around the fusion plasma chamber; would open the possibility of utilization of conventional stainless steel for fusion reactors due to the sufficiently low residual radioactivity in the structural materials after decommissioning of the plant. Research efforts and costs, involved in searching new alternative ceramic structural materials, such as SiC and graphite, based on unproven technology can be saved; and would make it possible to produce higher fusion power with a greater plasma volume.
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    Reduced shielding mass for the VISTA spacecraft
    (Springer Heidelberg, 2002) Sahin, S; Sahin, HM; Sahinaslan, A
    An innovative concept for the direct utilization of fusion energy with laser ignited (D,T) capsules for propulsion is presented with the so called VISTA (Vehicle for Interplanetary Space Transport Applications) concept. VISTA's overall geometry is that of a 50degrees-half-angle cone to avoid massive radioactive shielding. The 50degrees-half-angle maximizes the jet efficiency, and is determined by selecting the optimum pellet firing position along the axis of the cone with respect to the plane of the magnet coil. The pellet firing position is in the vacuum. Assuming a total fusion power production of 17 500 MW with a repetition rate of 5 Hz and 3500 MJ per shot, the propulsion power in form of charged particles has been calculated as similar to7000 MW, making similar to40% of the total fusion power. About 60% of the fusion energy is carried by the leaking neutrons out of the pellet. Most of them (96%) escape into vacuum without striking the space ship. Only 4% enter the frozen hydrogen expellant in the conical shape (about 50 gr.). Two design limits are discussed, 5 and 1 mW/cm(3). Total peak nuclear heat generation in the coils is calculated as 4.7 mW/cm(3). The peak neutron heating is 1.9 mW/cm(3) and the peak gamma-ray heating density is 2.8 mW/cm(3). However, volume averaged nuclear heat generation in the coils is much lower. It is calculated as 0.18, 0.48, and 0.66 mW/cm(3) for neutron, gamma-ray, and total nuclear heating, respectively. With higher design limits for nuclear heat generation in the coils and using natural lithium in the shielding, it was possible to reduce the net shielding mass from 595 tonne down to 170 tonne, making <3% of the vehicle mass, by a total vehicle mass of VISTA similar to 6 000 tonne.