scholarly journals Full Scope Modeling and Analysis on the Secondary Circuit of Chinese Large-Capacity Advanced PWR Based on RELAP5 Code

2015 ◽  
Vol 2015 ◽  
pp. 1-7
Author(s):  
Dao-gang Lu ◽  
Fan Zhang ◽  
Dan-ting Sui ◽  
Xue-zhang Xi ◽  
Lei-bo Yu
Keyword(s):  
Nuclear Power ◽  
Peak Load ◽  
Severe Accident ◽  
Secondary Circuit ◽  
Two Phase ◽  
Large Capacity ◽  
Modeling And Analysis ◽  
Heat Transfer Efficiency ◽  
Design Values ◽  
Under Construction

Chinese large-capacity advanced PWR under construction in China is a new and indispensable reactor type in the developing process of NPP fields. At the same time of NPP construction, accident sequences prediction and operators training are in progress. Since there are some possible events such as feedwater pumps trip in secondary circuit may lead to severe accident in NPP, training simulators and engineering simulators of CI are necessary. And, with an increasing proportion of nuclear power in China, NPP will participate in regulating peak load in power network, which requires accuracy calculation and control of secondary circuit. In order to achieve real-time and full scope simulation in the power change transient and accident scenarios, RELAP5/MOD 3.4 code has been adopted to model the secondary circuit for its advantage of high calculation accuracy. This paper describes the model of steady state and turbine load transient from 100% to 40% of secondary circuit using RELAP5 and provides a reasonable equivalent method to solve the calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heat transfer efficiency. The validation of the parameters shows that all the errors between the calculation values and design values are reasonable and acceptable.

Influence of Gas Properties on Gas-Liquid Two-Phase Flow

2020 ◽  
Author(s):  
Miki Saito ◽  
Taizo Kanai ◽  
Satoshi Nishimura ◽  
Yoshihisa Nishi
Keyword(s):  
Removal Efficiency ◽  
Nuclear Power ◽  
Two Phase Flow ◽  
Wire Mesh ◽  
Severe Accident ◽  
Phase Flow ◽  
Stagnant Water ◽  
Two Phase ◽  
Bubble Characteristics ◽  
Gas Properties

Abstract Understanding the mechanism of fission product (FP) removal by pool scrubbing is essential for improving the prediction accuracy of FP emissions concerning severe accident (SA) in a nuclear power plant. Since FP migrates from a gas-phase to a liquid-phase via a gas-liquid interface, the FP removal efficiency by pool scrubbing is largely affected by the flow regime of gas-liquid two-phase flow. In order to gain a deeper understanding of the influence of gas properties on flow regimes, experiments were performed by injecting helium (He) and nitrogen (N2) gas mixtures of several volumetric ratios through a pool of stagnant water. The result suggests clear effects of gas compositions on gas-liquid two-phase flow, where both void and holdup fractions were found to increase with N2 fraction in the supplied gas. The results were compared with previous studies, and a detailed analysis of bubble characteristics for different compositions of gases was performed using a wire-mesh sensor (WMS). This paper also illustrates further research aspects needed to discuss the effect of its results on FP removal efficiency in a SA, and to acquire comprehensive physics behind such gas property influences on two-phase flow.

Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident

2016 ◽  
Author(s):  
Alexandre Zanchetti ◽  
Mickael Hassanaly ◽  
Hervé Cordier ◽  
Antonio Sanna ◽  
Namane Mechitoua ◽  
...  
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Cfd Simulation ◽  
Severe Accident ◽  
Water Spray ◽  
Hydrogen Distribution ◽  
Two Phase ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Spray System

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations. The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.

Experimental Investigation on Heat Transfer for Two-Phase Flow Under Natural Convection

2012 ◽  
Author(s):  
Milan Amižić ◽  
Estelle Guyez ◽  
Jean-Marie Seiler
Keyword(s):  
Heat Transfer ◽  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Liquid Pool ◽  
Severe Accident ◽  
Bottom Plate ◽  
Interfacial Region ◽  
Two Phase ◽  
Wide Range

In the frame of severe accident research for the second and the third generation of nuclear power plants, some aspects of the concrete cavity ablation during the molten corium–concrete interaction are still remaining issues. The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt-through. For the purpose of experimental investigation of thermal-hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched jointly by CEA, EDF, IRSN, GDF-Suez and SARNET. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the smallest pool configuration (50 cm × 25 cm × 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s. This paper presents some preliminary conclusions deduced from the experiments which involve a liquid pool with the gas injection only from the bottom plate. A comparison with existing models for the assessment of heat transfer has also been carried out.

Study on Flow Structure in a Supersonic Steam Injector

2018 ◽  
Author(s):  
Yuki Kamata ◽  
Masaya Fujishiro ◽  
Akiko Kaneko ◽  
Yutaka Abe
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Structure ◽  
Nuclear Power ◽  
Severe Accident ◽  
Analysis Model ◽  
Operating Characteristics ◽  
Two Phase ◽  
High Heat Transfer ◽  
Discharge Pressure

Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant. Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model. The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.

Architecture and Suppression Design of Alarm System in HTR-PM

2017 ◽  
Author(s):  
Chao Guo ◽  
Qianqian Jia ◽  
Xiaojin Huang ◽  
Shuqiao Zhou
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Critical Role ◽  
Visual Display ◽  
Severe Accident ◽  
Alarm System ◽  
Control Room ◽  
Efficient Operation ◽  
Set Up ◽  
Under Construction

Nuclear safety is one of the key issues for a nuclear power plant (NPP). The alarm system plays a critical role for the safe and efficient operation of an NPP which is a significant human-machine interface in the main control room. The multi-modular NPPs have multiple reactor modules coupled to one steam turbine. One critical problem for the multi-modular NPP is that the complexity of the alarm system is greatly increased, which threatens the human-factor safety and the operation reliability. On the other hand, the main control room usually suffers from too many alarms to be handled at the same time after the accident, which is difficult for the operator to find out specific initiating event and may cause severe accident. The High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) which is under construction in Shandong province of China, is a typical multi-modular NPP with two reactor modules coupled to one steam turbine. In this paper, the architecture of the alarm system of HTR-PM is introduced. Different from conventional full-digital alarm system in the NPPs, a set of alarm tiles which are set up at the top of the large display panel are adopted to improve the alarm identification, and the alarm tiles are classified to groups of reactor 1#, conventional island, and reactor 2#, respectively. These alarm tiles cooperates with the alarm indication on the visual display unit to help to locate the accident location as soon as possible. The suppression design in case of alarm overloading are also discussed in this paper. Techniques like the dead band and first alarm indication are adopted in the alarm system of HTR-PM. Two kinds of suppression logics on condition and priority are discussed in the end of this paper. The work showed in this paper can contribute to improve the design of alarm systems in other NPPs, especial the multi-modular NPPs.

Device for Measuring Hydrogen Concentration in Containment During Severe Accident

2013 ◽  
Author(s):  
Sha Luo ◽  
Shaobo Wang ◽  
Liang Qin ◽  
Feng Pang
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Measuring Method ◽  
Nuclear Safety ◽  
Concentration Measurement ◽  
Severe Accident ◽  
Under Construction ◽  
Hydrogen Concentration Measurement

The hydrogen explosion in Fukushima nuclear accident seriously challenged the safety of nuclear power plants in the world. Therefore, strengthening hydrogen concentration measurement in containment has become very important. To raise the nuclear safety level of in-service and under construction nuclear power plants in China, based on a document which was issued by the National Nuclear Safety Administration of China (NNSA, 2012), the function requirements, components, storage, and arrangement outlines of the hydrogen monitoring system after a severe accident was thoroughly analyzed in this article. Besides, two kinds of techniques for hydrogen concentration measurement in containment, respectively for direct measuring method and gas sampling measuring method, were also discussed and compared. Notably, based on the direct measuring method, the 718th Research Institute successfully developed the CH-15 type hydrogen concentration measuring device that suitable for used in severe accident. The hydrogen sensor, with independently intellectual property rights, is based on catalytic principles and installed inside the containment. It has many unique characteristics, such as wide measurement range, high measurement accuracy, and is capable of continuous measurement with multiple points. This device has simple structure and small size, with low energy consumption, so it is very suitable for being installed in the in-service and under construction nuclear power plants in China and abroad.

Analytical Evaluation of Debris Cooling and Spreading Behaviors at Molten Core in Severe Accident

2014 ◽  
Author(s):  
Akihiro Kobayashi ◽  
Shuichiro Miwa ◽  
Michitsugu Mori
Keyword(s):  
Molten Metal ◽  
Nuclear Power ◽  
Fluid Model ◽  
Severe Accident ◽  
Two Phase ◽  
Temperature Dependent Viscosity ◽  
Fluid Analysis ◽  
Volume Of Fluid Model ◽  
Dependent Viscosity ◽  
Thermal Fluid Analysis

On March 11, 2011, severe accident occurred at Fukushima Daiichi Nuclear Power Plant, and Units 1 to 3 of the plant have led to core melt. That is to say, melted fuel rods and core internals fell to the bottom of the Reactor Pressure Vessel (RPV). It is also believed that molten core has leaked into the reactor containment vessel. In order to plan for a safe molten core removal from the reactor, it is important to estimate the conditions of molten core by conducting analysis. Particular importance of the analysis is to understand the mechanisms of molten core spreading-cooling processes. However, sufficient understanding of this process has not been obtained yet. The main purpose of this study is to evaluate molten metal spreading-cooling phenomena and subsequently, estimate the conditions of the molten metal. In order to achieve the purpose, the Computational Fluid Dynamics (CFD) for thermal fluid analysis, STAR-CCM+ was utilized. In the simulation of the unsteady two-phase flow, the volume of fluid model was applied for the spreading and interfacial surface formation of molten metal with the surrounding air. The key parameter for the molten metal spreading is the temperature dependent viscosity of molten metal. To assess the validity of this model, the analysis of the VULCANO VE-U7, molten metal spreading experiment, has been compared with simulation results.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management
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