原子力工学

Research and development activities for light water rector power plant have been reviewed focusing on those of multiphase technology in last ten years. In various fields such as reactor safety research, development of thermal-hydraulic analysis codes, improvement of reactor components, design of advanced, nextgeneration, or innovative light water reactors, researches related to multiphase technology have been performed. In future, demands for cost reduction and prevention for severe accident will be increased in development of future light water reactors. There will be increasing interest in replacing the current constitutive relation models with more mechanically or micro-mechanically-based models to make developmental work more efficient.

Current status of research on geological disposal of radioactive wastes is described briefly, mainly focused on fluid flow and mass transport phenomena such as nuclide migration in geological media, colloid transportation, hydro-thermal-mechanical coupled process, moisture migration through clay, bentonite extrusion and erosion, long-term rock creep deformation.

Radioactive aerosols produced by cutting components and structures may cause internal exposures of workers and secondary contamination in dismantling nuclear facilities.It must be important for safety to understand the diffusion behaviors of aerosols in dismantling activities under various working conditions. Therefore diffusion behaviors of aerosols have been studied by experiments such as flow visualization and by numerical analyses based on the fluid dynamics.The results were compared with those obtained in the analyses.As a result, it was confirmed that the diffusion behaviors of aerosols in in-air plasma arc cutting can be simulated by numerical analyses.

Although nuclear power is one of the indispensable energy sources to support modern life styles in developed countries, it becomes harder and harder to increase its capacity.Newspaper reported that there are numbers of evidences showing the suppression effect on cancer by the low level of radiation.It is expected for public people that the fear for radiation induced harm on health will mitigate through the explanation based on scientific evidences.Safe management of radioactive waste is one of the most serious issues to be solved.The neutron at fast reactors can eat more effectively the long lived several nuclear waste materials from light water reactor system.The key issue is to develop the fast reactor fuel cycle system technologies that are more economical, more proliferation resistant and higher breeding ratio.The Metallic Fuel Cycle is one of the options for the future fast breeder reactor and its related fuel cycle that enable to give the answer for the radioactive waste issues.The attractiveness of the metallic fuel cycle concept is briefly described.

The following thermal-hydraulic performance is necessary to develop the new fuel assembly for Boiling Water Reactor.(1) critical power under steady state and unsteady conditions (2) pressure drop (3) void fraction (4) channel stability. The purpose of this report is to explain about the test section and measurement technique for thermal-hydraulic performance of BWR fuel.

High-power pulsed spallation neutron sources are being developed in the world. A mercury target is installed at MLF ( Materials and Life science experimental Facility ) in J-PARC (Japan Proton Accelerator Research Complex), which will promote innovative science. At the moment the proton beams bombard the mercury target, pressure waves will be generated in mercury by thermally shocked heat deposition. Cavitation will be induced through the pressure wave propagation in mercury and eroded the vessel inner surface contacting with mercury, i.e. pitting damage. The eroded vessel wall is damaged by cyclic fatigue because pulsed proton beams strike the target repeatedly. In the paper, the effects of pitting damage, cyclic fatigue damage and mercury environment on the lifetime of the mercury target, and mitigation techniques for pressure waves and cavitation damage are described.

Research and development activities for two-phase analysis codes for nuclear reactor design and safety analyses have been reviewed focusing in recent twenty five years. For reactor safety evaluation, large-scale tests were performed to confirm effectiveness of ECCS in 1980’s and 1990’s. These test results were succeeded to so-called best-estimate codes such as RELAP5 , TRAC codes. Severe accident researches were performed in 1980’s and 1990’s and accident management methods were studied. Detailed simulation methods such as subchannel analysis, multi-dimensional analyses have been developed based on test results and computational technology enhancement in 1990’s and 2000’s. Future scope is summarized briefly.

Reliable prediction of two-phase flow characteristics is important for safety and economy improvements of BWR plants. We have been developing two-phase flow measurement tools and techniques for BWR thermal hydraulic conditions, such as a 3D time-averaged X-ray CT system, an ultrasonic liquid film sensor and a wire-mesh sensor. We applied the developed items in experiments using the multi-purpose steam-water test facility known as HUSTLE, which can simulate two-phase thermal-hydraulic conditions in a BWR reactor pressure vessel, and we constructed a detailed instrumentation database. We validated a 3D two-phase flow simulator using the database and developed the reactor internal two-phase flow analysis system.

In this study, a numerical analysis method applicable to estimation of the boiling heat transfer has been developed. Currently, the experimental correlations or the empirical laws have been applied to evaluate the boiling heat transfer. Therefore, it is difficult to predict the effects of the change of the heated surface geometry, thermal-hydraulic conditions, the surface activation or modification, because out of the application range of these correlations. The purpose of this work is to construct the boiling two-phase analysis method for thermo-fluid phenomena, and to realize “Design-by-Analysis” independent on the experiments and empirical laws. For this purpose, it is important to predict steam-water interface structure characteristics of the two-phase flow directly. Until now, for evaluating the boiling phenomena, Diffusive Interface Model for the bubble interface tracking was applied. In this model, the steam-water interface is diffuse with a finite width, and values of the thermodynamic properties change between water and steam smoothly within the interface region. For evaluating the wettability of heated surface, the surface energy is estimated by using the phase-field model. The wetting phenomena during boiling are able to be analyzed directly with this model. We present the numerical results of nucleate pool boiling phenomenon by using the developed analysis method. We succeeded in simulating the boiling process, vapor bubbles nucleation, growth, and departure behavior on a heated surface. By present analysis method, it was confirmed that the boiling heat transfer coefficient could be evaluated quantitatively without the experimental correlations.

In order to mitigate severe accidents of nuclear power plants, a passive safety system is one of influential measures. As the passive safety system uses natural power such as gravity and/or phase changes instead of electrical power, multiphase flows appear in the system. In this paper, multiphase flows in the passive safety system are introduced from several developments. First, a PCCS uses condensation in heat transfer tubes to remove decay heat. Second, oscillated water-vapor multiphase flow can provide cooling water to the heated surface of core-catcher cooling channel. At last, three-phase flows comprising gases, solid aerosols and liquids in safety measures against severe accidents are mentioned.

So far, needs for nuclear safety have much encouraged multiphase flow research, which, in return, has contributed to enhance nuclear safety. In view of this, the author has been conducting his research on gas-liquid two-phase flow related to nuclear reactor safety. This paper reviews such research history and gives some future prospects.

Thinking over the series of Japan-US two-phase flow seminars, the necessity and meanings of the multi-phase flow research are mentioned over the research conducted concerning the accident at Fukushima Daiichi Nuclear Power Station. There could be found the heavy concerns on the difficulty of elucidating the complex phenomena at the accident; none-the-less, there surely exist not a few places in which the multiphase flow research could play the important and active roles, repeatedly conducting and accumulating research on the steady road to fulfill the elucidation of the accident and recover the trust, after the fact of which I would believe that we could draw a future image of new nuclear powers.

In view of the importance of one-dimensional thermal-hydraulic analysis for rod bundle geometry, extensive efforts to improve constitutive equations have been made in recent years. In the present review article, state-of-the-art of the rod-bundle constitutive equations for flow regime map, void fraction, covariance, and interfacial area concentration models is reviewed. Among them, the constitutive relations for covariance and interfacial area concentration models may have the potential to improve the robustness of the conventional analysis method. Some sensitivity analysis results using TRAC-BF1 code with the new constitutive equations are summarized and reviewed.