Experimental Research on Pulsating Two-Phase Flow Pressure Drop Characteristics in Horizontal Rectangular Channel

We preliminarily designed a power tracking and control system using single-chip computers and industrial control computer in the electric heating simulated power loop. The system is an innovational design based on the proven simulated nuclear power loop, with increased techniques of step-less power regulation system and modeling nuclear feedback effect correctional programs. We promoted both hardware and software designs of this power tracking and control system in this paper. It used single-chip computers as the core control chips and an industrial control computer as the additional correctional program and record carrier. The process and implementation of the control software are presented, which is designed as a fuzzy theoretical nonlinear system. In order to ensure the subsequent updates, the access interface of the system is open for following correctional programs, including the correctional program of void fraction effect, temperature effect, hysteresis effect and heat power distribution effect. Taken hysteresis effect correctional program as an example, we use an offset tic-tac clock replacing the inherent tic-tac clock in different devices of the system in order to reduce the hysteresis effect of measuring and corresponding errors. We also put out a preliminary analysis of the accurate synchronization for the system at the end of the paper.

Hydrogen Distribution Assessment With CFD Tool in Help of Severe Accident Management

This paper presents the results of an analysis of the application of CFD tool to help hydrogen management. Some information pointed out the problem of hydrogen generation and distribution. Passive autocatalytic recombiners are the point of interest and mainly PAR units’ location. A severe accident is taken into account regarding the sources of hydrogen generation. The analysis of the severe accident progression is performed with MELCOR code. CFD tool Fluent (ANSYS) is applied to assess hydrogen and steam distribution in the atmosphere of the containment (confinement). The NPP unit of type WWER 440 (V230) is considered but as it is stressed this fact is irrelevant to phenomena and accident management targets.

HTR-PM Simulation Analysis of Accident Conditions Based on vPower Platform

HTR engineering simulator can be achieved by embedding THERMIX code into the vPower simulation environment. The engineering simulator consists of double-module reactors, two steam generators and entire secondary loop system for power generation with a water-steam Rankin cycle. The engineering simulator can be applied to simulate the steady-state operation, but also transient and accident state of HTR-PM. This paper analyzes the trends of reactor power, helium flow, steam generator inlet parameters, turbine inlet parameters and other key parameters under accident conditions, as well as the mutual influence between the two module reactors during accident process. Current simulation results are in good agreement with the design values and safety analysis results of HTR-PM.

Development of Numerical Simulation for Jet Breakup Behavior in Complicated Structure of BWR Lower Plenum: (4) Multi-Channel Experimental Analysis by Detailed Two-Phase Analysis Code TPFIT

In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred as can be seen at Fukushima Daiichi nuclear power plant accident. Especially, it is important to estimate the behavior of molten core jet in the lower part of the reactor pressure vessel at a severe accident. In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes, instrument guide tubes and core support plate. However, it is difficult to evaluate these effects on molten core jet experimentally. Therefore, we considered that multi-phase computational fluid dynamics approach is the best way to estimate the effects on molten core jet by complicated structure. The objective of this study is to develop the evaluation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum. So we are developing a simulation method to estimate the behavior of molten core jet falling down through the core support plate to the lower plenum of the BWR. The simulation method is based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Thus, the authors are carrying out experimental works by use of multi-phase flow visualization technique. In the experiments, time series of interface shapes are observed by high speed camera and velocity profiles in/out of the jet are measured by the PIV method. In this paper, we carried out analysis of the multi-channel experiment using the analytical method based on the TPFIT. Specifically, predicted results including interface shape and velocity profile in and out simulated molten material were compared with measured results. In the results, predicted results agreed with measured results qualitatively.

Takagi-Sugeno Fuzzy Load-Following Control of Nuclear Reactors Based on Reactor Point Kinetics Equations

A state differential feedback control system based Takagi-Sugeno (T-S) fuzzy model is designed for load-following operation of nonlinear nuclear reactor whose operating points vary within a wide range. Linear models are first derived from the original nonlinear model on several operating points. Next the fuzzy controller is designed via using the parallel distributed compensation (PDC) scheme with the relative neutron density at the equilibrium point as the premise variable. Last the stability analysis is given by means of linear matrix inequality (LMI) approach, thus the control system is guaranteed to be stable within a large range. The simulation results demonstrate that the control system works well over a wide region of operation.

Investigation of MHD RANS Modeling Base on DNS Database Under the Advanced Blanket Design Conditions Utilized Molten Salt

In this study, validations of Reynolds Averaged Navier-Stokes Simulation (RANS) based on Kenjeres & Hanjalic MHD turbulence model (Int. J. Heat & Fluid Flow, 21, 2000) coupled with the low-Reynolds number k-epsilon model have been conducted with the usage of Direct Numerical Simulation (DNS) database. DNS database of turbulent channel flow imposed wall-normal magnetic field on, are established in condition of bulk Reynolds number 40000, Hartmann number 24, and Prandtl number 5. As the results, the Nagano & Shimada model (Trans. JSME series B. 59, 1993) coupled with Kenjeres & Hanjalic MHD turbulence model has the better availability compared with Myong & Kasagi model (Int. Fluid Eng, 109, 1990) in estimation of the heat transfer degradation in MHD turbulent heat transfer.

Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily. Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.

Using the Reflectometric Technique for Monitoring the Boiling Flow Dynamics in a Once Through Steam Generator

In control systems design for nuclear facilities, the correct choice for sensors, transducers and actuators is not an easy task when different options must be evaluated. In particular, for a Once Through Steam Generator (OTSG), the control of its boiling flow dynamics is usually performed with sensors that may be slow in response, must be placed all through the zone of interest (or need heavy mechanical modifications) and may not distinguish correctly when the liquid/steam interface is not so well defined at high temperatures and pressures. In this paper, a theoretical study about a reflectometric technique applied to an OTSG is proposed. This technique can be used also for the liquid/steam levels monitoring of the boiling flow. The overall behaviour of the variables of interest and the first theoretical results show the benefits of such an innovative approach.

Experimental Study on Burning Process and Characteristic of Sodium Columnar Fire

A variety of sodium fire generated by the leakage of liquid sodium in the FBR is common. This paper focuses on the burning process and characteristics of sodium fire in a columnar flow. About 290°C liquid sodium was injected into a 7.9 m3 stainless steel cylindrical combustion space to shape the sodium columnar fire by 0.2 MPa high pressure nitrogen. The data of temperature field for the study of burning characteristic of sodium columnar fire have been collected by the temperature acquisition system located in the combustion space. The sodium flow maintains the columnar shape at first, and disperses by hydrodynamic effects on its way down. About 64s after the initiating time of sodium ejection for this experiment, the maximum temperature of the area close to the ejection center reaches over 1200°C. And the maximum temperature appears at the space of 1–1.5m from the plate. But the high temperature lasts for a short time and reduces rapidly. The radial temperature of the area far from the sodium flow is relatively low and generally about 200°C, and maximally about 350 °C. This study is helpful to evaluate the combustion characteristics and burning process of the sodium fire in the sodium-related facilities.

Modelling of Water Ingress in to Vacuum Vessel

An event of water coolant ingress into vacuum vessel is one of the most important events leading to severe consequences in nuclear fusion reactors. The ingress of coolant to the vacuum vessel could appear due to coolant pipe rupture of in-vessel components. Vacuum vessel could not withstand the high pressure inside. Pressure increase in vacuum vessel is due to water evaporation because of pressure difference and water contact with high temperature plasma facing components. If pressure in vacuum vessel is too high — safety valve opens and the steam containing activated dust will be transferred form the vacuum vessel to the environment. Therefore, it is important to understand the thermo hydraulic processes in vacuum vessel during the ingress of coolant event (ICE). There are few experimental investigations performed, modeling of ICE. In this article ingress of coolant event in vacuum vessel was modeled using RELAP5 code. RELAP5 is a “best estimate” system code suitable for the thermo-hydraulic analysis of all transients and postulated accidents in nuclear fission, light water reactor systems, including both large and small-break loss-of-coolant accidents as well as the full range of operational transients. The use of RELAP5 code for the accident analysis in nuclear fusion reactors allows to perform integral analysis of thermal-hydraulic processes in the cooling system and vacuum vessel. The comparisons of calculation results and experimental data showed that with some limitations the RELAP5 program could be used for the analysis of the thermal hydraulic processes in the vacuum vessel during ICE.