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[en] Two advanced boiling water reactors (ABWRs) whose electric output power is 1356 MW have been commercially operated since 1996 and 1997 by the Tokyo Electric Power Company (TEPCO) in Japan. Features of an ABWR are reactor internal pumps (RIPs) placed in the lower plenum and downcomer, peripherally bottom-mounted on the reactor pressure vessel - which should require different modeling from the jet pumps and two recirculation pumps in the primary outer-loop recirculations of BWR-5.Efforts focused on modeling and simulating the ABWR with transient analyses by the point-kinetics model with the local reactivity modified by local importance weighting of the squared nodal power during start-up tests using the RETRAN-3D code, version MOD003 without three-dimensional kinetics. The core and reactor pressure vessel including ten ABWR RIPs and the steam lines were modeled, and simulations were carried out for the cases of the one-pump trip test, the changing-setpoint tests, the main-steam-isolation-valve-closure test, and the generator load rejection test with bypass.The analytical simulation with RETRAN-3D/MOD003 well reproduced the measured data of the ABWR in operation for the RIP trip and the transient tests and could demonstrate its validation for applying to the ABWR with modeling of RIPs
[en] The capability to predict the void fraction distribution and the interfacial area transport phenomenon in the subcooled boiling region is of considerable interest to BWR operation and safety. The existence of the thermodynamic non-equilibrium between the phases complicates analysis of the subcooled boiling flow. Few experiments have been performed to measure interfacial parameters under this condition. Most existing models or correlations are applicable only to limited experimental conditions. Consequently, it is desirable to establish a database of local interfacial parameters and boundary conditions. It is also desirable to develop reliable constitutive models for broad subcooled boiling conditions. This will significantly improve the numerical simulation of the performance of a nuclear reactor. (author)
[en] The importance of the cooling process from transient film boiling to quenching has been pointed out from the standpoint of nuclear reactor safety technology as the emergency cooling process for the fuel rods reached high temperature in a loss of coolant accident or a reactivity accident. In this study, as the first step to clarify the conditions to cause quenching and the mechanism of film boiling heat transfer till quenching arises, experiment was carried out with a horizontal fine platinum wire by laser measuring method. As the mechanism of causing quenching, solid-liquid direct contact model, axial heat conduction model and leading cooling model were proposed, but quenching phenomena have not been fully explained. In this experiment, using laser optical measurement, the behavior of steam film in the film boiling under subcooled and pressurized condition was clarified, and the effect of the variation of steam film on film boiling heat transfer was evaluated. The principle of measuring steam film thickness by laser method and the measuring method, the experimental setup, the calibration curve, film boiling curves, the spectral analysis and the mean value of the time variation of steam film thickness and other results are reported. (Kako, I.)
[en] A basic study has been performed experimentally to investigated the behavior of vapor film and pool film boiling on a horizontal platinum fine wire, and on a stainless steel sphere with diameter of 12.7 mm in subcooled water. A He-Ne gas laser beam is used for measuring the vapor film thickness and its time variation. The film boiling phenomena are observed by high speed photography, while at the same time measurements are taken of the temperature of the test specimen, and are also taken of the vapor film behavior with the laser beam for the sphere. (author)
[en] Steam injector (SI) is a simply designed and passive jet pump, in which external power sources and internal mechanical parts are inessential. The SI utilizes direct contact condensation between steam and water as an operational mechanism, and is capable of producing higher pressure water than the inlet fluid pressures. The accident in Fukushima Daiichi Nuclear Power Plant caused setback to the credibility and reliability of nuclear power. One way to regain its trust from the global community, it is suggested to develop and install passive coolant injection systems operable even during the station black out. The direct-contact condensation of steam in sub-cooling water is an important phenomenon in SI, since it relates to essentially-driven mechanism. In past years, the condensation of steam jet submerged in the quiescent sub-cooled water has studied by many researchers. In the present work, based on reviewing the previous research the important phenomena were simulated using the computational fluid dynamics (CFD) code. (author)
[en] Analytical and experimental studies have been conducted on large-scale steam injectors for a next-generation reactor. The steam injectors are simple, compact, passive steam jet pumps for a steam-injector-driven passive core injection system (SI-PCIS) or steam-injector-driven primary loop recirculation system (SI-PLR). In order to check the feasibility of such large-scale steam injectors we developed the separate-two-phase flow models installed in the PHOENICS Code, and scale-model tests were conducted for both SI-PCIS and SI-PLR. A 1/2 scale SI-PCIS model achieved a discharge pressure of almost 8 MPa with 7 MPa steam and 0.4 MPa water, and a 1/5 scale SI-PLR model attained a discharge pressure of 12.5 MPa with 3 MPa steam and 7 MPa water. Both results are in good agreement with the analysis, confirming the feasibility of both systems. The systems will help to simplify the next generation of BWRs.
[en] Highlights: • RETRAN-03 was utilized to analyze the effectiveness of the Isolation Condenser installed on ABWR. • Studies were conducted to observe the transient behavior of the reactor with IC. • Countermeasures to improve safety level in LWR operation features are proposed. - Abstract: The Great East Japan Earthquake occurred on March 11, 2011 fatally damaged the Fukushima Daiichi Nuclear Power Plant (NPP), caused prolonged station blackout (SBO). Following the SBO, the reactor water level gradually dropped due to the increase in steam discharge from safety relief valves and eventually led to nuclear fuel melt down. Almost four years have passed since the accident, and official reports by the Japanese regulatory have given the general description of causes and progressions of the fatal accident. Even after the Fukushima accident, more than 430 nuclear power plants are currently operating and over 80 units are under construction worldwide. From this fact alone, it is extremely important to learn from the Fukushima accident and enhance the safety culture of the reactor operation to completely eliminate the possibilities of catastrophic accidents seen in 2011. In this study, the best estimate transient thermal-hydraulics code, RETRAN-03/MOD04 was utilized to focus on the effectiveness of the Isolation Condenser (IC) installed on Advanced Boiling Water Reactors (ABWR). The ICs turned out to be one of the very few operable safety systems during the Fukushima accident, and this simple yet reliable safety system should be utilized to secure ABWR from possible reactor core damages. In the present paper, several case studies conducted utilizing the ICs are presented and methods of countermeasures to improve light water reactor safety level in design and operation features are proposed.
[en] The experimental work was performed to study the thermal behaviors with quenching and transient film boiling in an experiment where Zry-4 rods heated at about 1000 0C were submerged rapidly into distilled water. These thermal behaviors were affected strongly with coolant subcooling temperature, and dependent highly on large heat flow in a radial direction through the convective film boiling vapor film at the quench front rather than axial heat flow by conduction in a rod which was confirmed analytically to be small. (author)
[en] The Molten Core Concrete Interaction (MCCI) is one of the most severe incidents in nuclear power plants. This leads to the leak of radioactive material from the power plant. On March 11"t"h, 2011, it was very concerned that how depth is the concrete erosion formed by the accident in the Fukushima Daiichi Nuclear Power Plant. In this research, a performance of MCCI under the severe accident (Fukushima Daiichi Nuclear Power Plant accident) events was evaluated by numerical simulation and discusses the situation about the erosion in the containment vessel. (author)