Takeharu Misawa

Development of Analytical Procedures of Two-Phase Flow in Tight-Lattice Fuel Bundles for Innovative Water Reactor for Flexible Fuel Cycle

Abstract A research and development project to investigate thermal-hydraulic performance in the tight-lattice rod bundles of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been in progress at Japan Atomic Energy Agency in collaboration with power companies, reactor vendors, and universities since 2002. The FLWR can realize favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burnup, and long operation cycle, based on matured light water reactor technologies. Mixed-oxide fuel assemblies with tight lattice arrangement are used because they increase the conversion ratio by reducing the moderation of neutrons. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. Information about the effects of the gap width and grid spacer configuration on the flow characteristics in the FLWR core is still insufficient. Thus, we are developing procedures for qualitative analysis of thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this study, an advanced two-fluid model is developed to economize on the computing resources. In the model, interface structures larger than computational cells (such as liquid film) are simulated by the interface tracking method, and small bubbles and droplets are estimated by the two-fluid model. In this paper, we describe the outline of this model and the numerical simulations we performed to validate the model performance qualitatively.

Development of an Analytical Method on Water-Vapor Boiling Two-Phase Flow Characteristics in BWR Fuel Assemblies Under Earthquake Condition

Safe operation of nuclear reactors under earthquake conditions cannot be guaranteed because the behavior of thermal fluids under such conditions is not yet known. For instance, the behavior of gas-liquid two-phase flow during earthquakes is unknown. In particular, fluctuation in the void fraction is an important consideration for the safe operation of a nuclear reactor, especially for a boiling water reactor (BWR). The void fraction in the coolant is one of the physical parameters important in determining the thermal power of the reactor core, and fluctuations in the void fraction are expected to affect the power of the plant.

Assessment of Applicability of Two-Fluid Model Code ACE-3D to Heat Transfer Test of Supercritical Water Flowing in an Annular Channel

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code is applicable to the prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of the SCWR core.

Numerical Analysis on Heat Transfer-Characteristics of Supercritical Pressure Water in a Heated Tube Based on Three Dimensional Two-Fluid Model

In Japan Atomic Energy Agency (JAEA), the three-dimensional two-fluid model analysis code considering the supercritical pressure water based on ACE-3D (Advanced Code for Evaluation of 3-Dimensional two-phase flow) has been developed to establish the thermal-hydraulics design by numerical analytical approach for the Super Critical Water Reactor (SCWR). In this paper, in order to evaluate the prediction accuracy of ACE-3D for the heat transfer-characteristics at the pseudo critical point, a numerical analysis of the supercritical water using ACE-3D based on the three dimensional two-fluid model has been conducted for simulating the experiments in a heated tube with both upward and downward flow. For the turbulence model in this analysis, both the standard k-e model and the low-Reynolds number type k-e model which uses the Launder-Sharma model were examined to investigate the influence of the turbulence model on the heat transfer-characteristics near the heated wall near the pseudo critical point. As a result, it was found that the numerical results of wall temperature using the low-Reynolds number k-e model for upward flow in a heated tube were in good agreement with experimental data compared with that of using the standard k-e model.Copyright

Analytical Procedure on Fluid Mixing Phenomena in Boiling Water Reactor Core With Advanced Interface Tracking Method

Thermal-hydraulic design of a boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, when the reactor of a new design is developed, an actual size test that simulates its design is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for BWRs using an innovative two-phase flow simulation technology. For this design method, we are developed an advanced interface tracking method that improves fluid volume conservation, to enable high accuracy prediction of two-phase flow fluid mixing phenomena in the fuel bundles. It was incorporated in the detailed two-phase flow simulation code: TPFIT. And the vectorization and parallelization of TPFIT code was conducted to analyze enormous amounts of data. In this study, to verify the TPFIT code performance, the TPFIT code was applied to the air-water and steam-water bubbly two-phase flow in various flow channels and the numerical results were compared with experimental results. Furthermore, the numerical results applied to the fluid mixing phenomena in boiling water reactor rod bundles are shown, and the existing correlations for the fluid mixing phenomena are evaluated by use of these results.Copyright