Suppression Tank Research Articles

A Preliminary Study on Direct-Contact Condensation Between Water and Vapor in A Suppression Tank for Fusion Reactor Safety

A preliminary study on the direct-contact condensation in the ITER pressure suppression tank during the ICE event was performed experimentally and analytically. The water-vapor two phase flow behavior in the depressurized water tank was investigated by the flow visualization experiment and it was found that the present numerical approach is very effective to predict the condensation effect in the ITER suppression tank during the ICE event.

Analysis of Pressure Rise in an ITER-Like Fusion Reactor During In-Vessel LOCA by A Modified TRAC-PF1 Code

Damage of cooling tubes of plasma facing components (PFCs) results in water discharge into a vacuum vessel (W) of a fusion reactor. Flashing in vacuum, water pool boiling and impingement-jet on a surface of the PFC are the main heat transfer phenomena responsible for steam production that causes a rapid pressurization of the W. This is called an in-vessel loss-of-coolant accident (LOCA) event or ingress-of-coolant event (ICE). The ICE event is one of the most severe accidents in the fusion reactors.The integrated ICE test facility was constructed to demonstrate the safety design approach of International Thermonuclear Experimental Reactor (ITER) and obtain validation data for the ITER safety analysis codes. Then, an experimental study was performed using the integrated ICE test facility and at the same time the code validation study with the TRAC code was carried out. The pressure rise characteristics in the current ITER machine during the ICE event were analyzed numerically using the verified TRAC-PF1 code and the effects of the relief pipe diameter and suppression tank volume regarding to the pressure rise due to the ICE events were clarified quantitatively from the present analytical results.

減圧下での水蒸発と凝縮に関する実験と解析

When cooling tubes installed in plasma-facing components (PFCs) of a fusion reactor are damaged by some accidents, water may be discharged into a plasma chamber (PC) and vacuum vessel (VV) of the fusion reactor. In such a case, the discharged water will impinge on a surface of PFCs at high temperature and evaporate and then the VV may be broken due to pressurization by steam. This event is called an ingress-of-coolant event (ICE). The water-vapor two-phase flow behavior in the PC and VV during the ICE was investigated using an integrated ICE test facility. Furthermore, the condensation effect due to vapor inside a suppression tank which is installed to reduce the pressurization by the ICE was observed visually. In addition, the thermal-hydraulics during the ICE were validated numerically by the TRAC-PF1 code and analytical results on the pressure rise were in good agreement with the experimental results.

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1.

The semi three-dimensional numerical analysis model of the FDR-ITER to simulate the in-vessel LOCA or the ICE was constructed using the modified TRAC-BF1 code in which the vacuum vessel, the pressure suppression tank and the relief pipe header are modeled using one VESSEL component. Analytical results obtained from the modified TRAC-BF1 code were compared with those from the MELCOR code concerning pressure in the vacuum vessel when a cooling pipe of 0.6 m2 flow area breaks in it. The mass flow rate of the injected water into the vacuum vessel and the initial wall temperature in the modified TRAC-BF1 input data were the same as those calculated in the MELCOR analysis. The maximum pressure in the vacuum vessel obtained from the modified TRAC-BF1 code is 10% higher than that from the MELCOR code, but it is still below the design pressure 0.5 MPa of the FDR-ITER vacuum vessel. The semi three-dimensional analysis using the modified TRAC-BF1 obtained that the maximum delay in opening time among the four rupture-disks is within 3 ms. The maximum pressure in the vacuum vessel does not exceed the design pressure, even if one of rupture-disks is not opened.

Numerical study on pressure rise characteristics in simulated ITER structural components during ingress-of-coolant events

A numerical study was performed to predict the pressure rise characteristics during ingress-of-coolant events (ICE) in an integrated test facility which simulated structural components of International Thermonuclear Experimental Reactor (ITER). The integrated ICE tests were planned to clarify the two-phase flow characteristics from the plasma chamber through the divertor and vacuum vessel to the suppression tank during the ICE events and demonstrate adequacy of the ITER safety design approach, and obtain validation data for safety analysis codes for fusion experimental reactors. The TRAC code, which was originally developed for the accident analysis in light water reactors, was used in the present study. It was clarified quantitatively from the numerical predictions that the pressure inside the vacuum enclosure in ITER during the ICE events receives the injected water flow rate, orifice size and pitches, and wall and water temperatures, and then, the suppression tank is very effective to reduce the pressure rise.

303核融合炉事故事象時における熱流動挙動に関する数値解析

Numerical analyses on thermal-hydraulic characteristics in fusion reactors at accident events such as ICE (Ingress-of-Coolant Event) and LOVA (Loss-of-Vacuum Event) were carried out due to make sure adequacy of the safety design in ITER (International Thermonuclear Experimental Reactor) and the results were visualized numerically to enhance the understanding of physical phenomena during ICE and LOVA events. Relation between pressure rise and void fraction in ITER at the ICE event was obtained with a TRAC code. Air-ingress behavior into water was also clarified quantitatively with a CIP code. Furthermore, exchange flow and dust mobilization behavior through breaches on the fusion reactor during the LOVA event were analyzed based on equations of compressible fluid and dust momentum. It was found from the results of present numerical visualization that a suppression tank system which will be used in ITER is very effective and the ITER pressure suppression system during ICE and LOVA events is adequate.

Analysis of the Chemical Behavior of Iodine in the Suppression Tank of the LOFT Facility During Experiment LP-FP-2 with IODE and IMPAIR-2/M

The significance of iodine for source term quantification has been studied by investigating its chemical behavior under the prototypical conditions of a hypothetical severe accident within the containment. As a result, some computer codes were developed and their validation is currently under way. The loss-of-fluid test (LOFT) program was one of the most relevant research projects in the area of nuclear safety. Its last experiment, LP-FP-2, simulated a V-sequence. A great deal of information was recorded on the fission product release, transport, and deposition. A theoretical approach to the chemical behavior of iodine in the blowdown suppression tank (BST) of the LOFT facility was attempted with the IODE and IMPAIR-2/M codes. The comparison of the predictions with the existing experimental data led to the conclusion that the BST system behaved as a low-volatility system, with most of the iodine in the form of the soluble nonvolatile species iodide. Only a partial conversion to volatile molecular iodine was observed due to the presence of radiation. However, the intensity of the γ field was so weak that this transformation was not quantitatively meaningful.