Simple Method for Evaluation of [lbar] Used for Kinetic Calculations of One Point Reactor Model

Graphite Material has been widely used as first wall material in present large tokamak devices. However, overall evaluation with respect to vacuum engineering properties, interactions with plasmas, and thermal and mechanical properties has not been systematically performed so far, though these properties much depend on the kind of the graphite material. For the overall evaluation of the graphite as fusion first wall material, the “Graphite Project” was organized in 1986 under the support of the Ministry of Education, Science and Culture. More than 20 institutions participated in this project and 27 graphite material (isotropic graphite, pyrolytic carbon, C/C composite) supplied from 8 graphite manufactureres of Japan were characterized as “common samples”.

It was found that the vacuum engineering properties such as outgassing, effective surface area and hydrogen permeation significantly depended on the pore structure of the graphite. Both the outgassing quantity and the surface area were observed to be small for the graphite with low density. The mechanism of hydrogen permeability was explained by the molecular flow through the pore structure. The chemical sputtering yield of metal deposited graphite was significantly lower than that of the graphite with clean surface. The hydrogen retention was considerably reduced by the iron or titanium deposition onto the graphite surface. The yield of the radiation enhanced sublimation for the isotropic graphite was measured and the result showed that the yield of the isotropic graphite was quite similar to that of the pyrolytic carbon. The heat load experiments showed that most of the isotropic graphite failed at roughly the same heat load and the fracture toughness was also approximately the same. The C/C composite materials, however, had the thermal shock resistance and the fracture toughness, which were several times larger than that of the isotropic graphite.     

Dynamic Response of Hot/Cold Liquid Interfaces to Pump Speed Perturbations in a Thermal-Hydraulic Loop Simulating a PIUS-type Reactor

The operation of a PIUS-type reactor requires controlling the reactor pump speed to keep stationary the hot/cold liquid interfaces between the reactor coolant and cold borated water. The dynamic response of the interface location to pump speed perturbations is analyzed for an experimental loop simulating a PIUS-type reactor. The transfer function between the pump speed and the interface location is obtained by perturbing and Laplace-transforming the one-dimensional fluid momentum equations. The analytical results agree well with experimental data taken from the same facility. It is shown that the magnitude of the phase lag in the response of the interface location, which needs to be considered in designing a pump speed controller, primarily depends on the fluid inertia in the loop, the density lock flow area, and the density difference between the simulated reactor coolant and borated water.