Mitsuhiko Shibata

Temperature distributions in a Tokamak vacuum vessel of fusion reactor after the loss-of-vacuum events occurred

Abstract If a loss-of-vacuum event (LOVA) occurred in a fusion reactor, buoyancy-driven exchange flows would take place at breaches of a vacuum vessel (VV) due to the temperature difference between the inside and outside of the VV. The exchange flows may bring mixtures of activated materials and tritium in the VV to the outside through the breaches, and remove decay heat from the plasma-facing components of the VV. Therefore, the LOVA experiments were carried out under the conditions that one or two breaches were opened and that the VV was heated to a maximum 200°C, using a small-scaled LOVA experimental apparatus. Air and helium gas were provided as working fluids. Fluid and wall temperature distributions in the VV were measured and the flow patterns in the VV were estimated from these temperature distributions. It was found that: (1) the exchange mass in the VV depended on the breach positions; (2) the exchange flow at the single breach case became a counter-current flow when the breach was at the roof of the VV and a stratified flow when it was at the side wall; (3) and that at the double breach case, a one-way flow between two breaches was formed.

Analysis and experimental results on ingress of coolant event in vacuum vessel

Experiments on the ingress of coolant event (ICE) in the vacuum vessel of a fusion reactor have been carried out in the Japan Atomic Energy Research Institute (JAERI) as one of the research and development tasks for the International Thermonuclear Experimental Reactor (ITER) to obtain the thermofluid data for validation of safety analysis codes. The ICE experiment is numerically analyzed using the transient reactor analysis code (TRAC) which is one of the codes preparing for the safety analysis of the ITER. The TRAC has been modified so as to analyze the ICE phenomena in the vacuum vessel of a fusion reactor. Several ICE experiments have been carried out as benchmark tests for the safety analysis codes. We have analyzed those experiments by using the TRAC, and considered the difference between the analysis and experimental results. Analysis results of the temperature in the vacuum vessel show a tendency completely different from the experimental result. It is clarified that the present TRAC has not been verified on the scattering behavior of water droplets.

Experimental results of functional performance of a vacuum vessel pressure suppression system in ITER

Abstract An integrated Ingress-of-Coolant Event (ICE) test facility was constructed to investigate quantitatively effectiveness and functional performance of a vacuum vessel pressure suppression system (VVPSS) during the ICE in International Thermonuclear Experimental Reactor (ITER). The integrated ICE test facility simulated structural components of the ITER VVPSS with a scaling factor of 1/1600. From the present study it was clarified experimentally that the ITER VVPSS is very effective to reduce the pressurization during the ICE because the condensation of vapor within a suppression tank is enhanced and the pressure rise characteristics strongly depend on the cross-sectional area of the relief piping, and then the pressure rise rate in ITER is expected between 100 and 200 kPa s−1.

Safety scenario and integrated thermofluid test

Abstract The largest mobilizable radioactive material inventory in the form of tritium and activated dust in a fusion reactor is estimated to be located in the vacuum vessel. The accident scenarios of postulated thermofluid transients such as ingress of coolant inside the vacuum vessel and the loss of vacuum boundary leading to the release of radioactive material are introduced. The accuracy of the present analysis method and database for evaluating the radioactive material release in such accident scenarios is assessed. The areas where the data and methods seem to be most uncertain are identified, such as the condensation of steam under vacuum condition, the activated dust mobilization and transport in and out of the vacuum vessel in the event of the transients. An approach to experimentally reduce such uncertainties in the evaluation of radioactive material release are presented. A combination of a number of specific test devices to reduce uncertainties in such areas as dust mobilization and transport, and an integrated thermofluid test facility to establish the evaluation methodology are proposed.

Mobilization Characteristics of Dust Conveyed by Buoyancy Flows during Loss-of-Vacuum Accident

Dust mobilization in a vacuum vessel (VV) of a fusion reactor during a loss-of-vacuum accident (LOVA) was studied numerically and experimentally from a viewpoint of the fusion safety research. After the LOVA occurred, air flows from the outside of the VV through a breach into the inside and the activated dust is blown up from the walls inside the VV, and then it can be considered that the dust is conveyed due to a buoyancy flow to the outside of the VV through the breach. The velocity, pressure and temperature distributions were obtained by the present numerical analyses and the dust mobilization behavior was predicted quantitatively. In addition, the dust transport characteristics from the inside of the VV through the breach to the outside due to the buoyancy effect was clarified by the visualization experiments.