Thermal–Hydraulic Performance Analysis for AP1000 Passive Containment Cooling System

2013 ◽  
Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Fenglei Niu
Keyword(s):  
System Reliability ◽  
Nuclear Power ◽  
Failure Modes ◽  
Cooling System ◽  
Natural Circulation ◽  
Hydraulic Performance ◽  
Reactor Power ◽  
Physical Parameters ◽  
Cold Source ◽  
Process Failure

In order to improve the safety of new generation nuclear power plant, passive containment cooling system is innovatively used in AP1000 reactor design. However, since the system operation is based on natural circulation, physical process failure — natural circulation cannot establish or be maintained — becomes one of the important failure modes. Uncertainties in the physical parameters such as heat and cold source temperature and in the structure parameters have important effect on the system reliability. In this paper, thermal–hydraulic model is established for passive containment cooling system in AP1000 and the thermal–hydraulic performance is studied, the effect of factors such as air temperature and reactor power on the system reliability are analyzed.

Reliability of PCCS in AP1000 Based on Accident Development

2016 ◽  
Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Fenglei Niu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Physical Process ◽  
System Reliability ◽  
Nuclear Power ◽  
Cooling System ◽  
Natural Circulation ◽  
System Structure ◽  
System Failure ◽  
Process Failure

Passive containment cooling system (PCCS) is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink – atmosphere – based on natural circulation, independent of human response or the operation of outside equipments, so the reactor capacity of resisting external hazards (earthquake, flood, etc.) is improved. However since the system operation based on natural circulation, many uncertainty factors such as temperatures of cold and heat sources will affect the system reliability, and physical process failure becomes one of the important contributors to system failure, which is not considered in the active system reliability analysis. That is, the system will lose its function since the natural circulation cannot be established or kept even when the equipments in the system can work well. The function of PCCS in AP1000 is to transfer the heat produced in the containment to the environment and to keep the pressure in the containment below its threshold. After accidents the steam is injected to the containment and can be cooled and condensed when it arrives at the containment wall, then the heat is transferred to the atmosphere through the steel vessel. So the peak value of the pressure is influenced by the steam situation which is injected into the containment and the heat transfer and condensate processes under the accidents. In this paper the dynamic thermal-hydraulic (T-H) model simulating the fluid performance in the containment is established, based on which the system reliability model is built. Here the total pressure in the containment is used as the success criteria. Apparently the system physical process failure may be related to the system working state, the outside conditions, the system structure parameters and so on, and it’s a heavy work to analyze the influences of all the factors, so only the effects of important ones are included in the model. Monte Carlo (MC) simulation is used to evaluate the system reliability, in which the input parameters such as air temperature are sampled based on their probabilistic density distributions. The pressure curves along with the accident development are gained and the system reliabilities under different accidents are gotten as well as the main contributors. The results illustrate that the system physical process failure probabilities are varied under different climate conditions, which result in the system reliability and the main contributors to system failure changing, so the different methods can be taken to improve the system reliability according to the local condition of the nuclear power plant.

Reliability Analysis for Passive Containment Cooling System in Seismic Hazard

2012 ◽  
Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Xuefeng Lv ◽  
Fenglei Niu
Keyword(s):  
Seismic Hazard ◽  
Reliability Analysis ◽  
Failure Probability ◽  
System Reliability ◽  
Nuclear Power ◽  
Cooling System ◽  
Natural Circulation ◽  
System Operation ◽  
Ground Acceleration ◽  
Plant Safety

Passive containment cooling system is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink–atmosphere based on natural circulation. So it is important to the plant safety whether the system can work well or not in the seismic hazard. Since the system operation is independent of human interfere or the operation of outside equipments, the reliability of the system is improved, however, physical process failure become one of the important contributors to the system operation failure because natural circulation may not keep when the system configuration is different from the design. So it is necessary to analyze the system reliability in seismic situation. The equipment failure probability under earthquake is a function of the peak ground acceleration which is stochastic, and the fault tree method used in traditional probability safety assessment (PSA) for system reliability analysis is not power enough to deal with conditional probability. In this paper, a new analysis method for system reliability evaluate at seismic situation based on Monte Carlo (MC) simulation is put forward, and annual failure probability of passive containment cooling system in AP1000 in seismic hazard is calculated, the result is according with the AP1000 seismic margin evaluation.

Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Vanderley Vasconcelos ◽  
Wellington Antonio Soares ◽  
Raissa Oliveira Marques ◽  
Silvério Ferreira Silva Jr ◽  
Amanda Laureano Raso
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Human Factors ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
Cooling System ◽  
Human Reliability ◽  
Non Destructive ◽  
Plant Components

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

Passive System Reliability Modeling of the Combination of Failure Modes

2016 ◽  
Author(s):  
Luciano Burgazzi
Keyword(s):  
Reliability Analysis ◽  
Failure Mode ◽  
System Reliability ◽  
Probabilistic Models ◽  
Failure Modes ◽  
Natural Circulation ◽  
Heat Transfer Process ◽  
Passive System ◽  
Modes Of Failure ◽  
Degradation Failure

Innovative probabilistic models to extend the reliability analysis of passive systems under different modes of failure are proposed. The prevailing failure mode on the system can be predicted through the failure probability assessment on each specific mode. A realistic case is presented to analyze a passive system with two kinds of major failure modes — natural circulation stoppage due to e.g., isolation valve closure (a catastrophic failure) and heat transfer process degradation due to e.g., deposit thickness on component surfaces (a degradation failure). Modeling of each individual failure mode together with system reliability analysis is presented and results are discussed.

Research on Thermal Hydraulic Characteristics of Passive Containment Cooling System for Ship Nuclear Power Platform

2016 ◽  
Author(s):  
Chunhui Dai ◽  
Jun Wu ◽  
Sichao Tan ◽  
Zhenxing Zhao ◽  
Qi Xiao ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Sea Water ◽  
Cooling System ◽  
Natural Circulation ◽  
Three Dimensional ◽  
Electric Energy ◽  
Hydraulic Characteristics ◽  
Loss Of Coolant Accident ◽  
Numerical Computing

Ship nuclear power platform is a small and movable power plant on the sea, aiming at generating electric energy and producing fresh water, it provides support for the national energy strategy. Subsequent to a loss of coolant accident (LOCA), steam is vented in the reactor containment following vaporization of liquid and/or steam expansion. The temperature as well as pressure in the condensation rises synchronously. For removing heat and reducing pressure inside containment subsequent to a LOCA, the Passive containment cooling system of Ship nuclear power platform is designed. In order to establish and maintain the passive heat removing channel, steam condenses on the containment condenser tube surface, coupling natural convection of the seawater inside the tubes. The heat transfer mechanism of Passive containment cooling system is very complex. To solve this problem, a three dimensional heat exchanging/one dimensional natural circulation coupling numerical computing method is proposed to obtained the safety performance of the reactor containment. Models of heat exchanging process between steam which contains non-condensable gas inside the reactor containment and sea water outside are firstly established. Then the thermal-hydraulic characteristics of the steam and sea water beside the heat transfer tubes are obtained by a simulation which is carried out in a LOCA.

Optimization of Test Interval of Ignalina Nuclear Power Plant Auxiliary Feedwater Pumps

2009 ◽  
Author(s):  
Roman Voronov ◽  
Robertas Alzbutas
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Failure Modes ◽  
Cooling System ◽  
Model Parameters ◽  
Testing Interval ◽  
Safety Function ◽  
Core Cooling ◽  
Safety Systems

Some safety systems of the Ignalina Nuclear Power Plant (NPP) operate in standby mode. An equipment of such systems is periodically tested and that allows timely detect and repair equipment failures. The periodic testing is an important measure of ensuring systems’ operability and reliability. However, during the test and repair the equipment cannot perform it’s safety function, therefore too often testing decreases the availability of the system. This paper describes the mathematical model that represents how availability and reliability of the systems and their components depend on testing interval, taking into account different failure modes of the equipment. This model allows to find the optimal testing interval for the safety. As an example, the auxiliary feedwater pumps, that are a part of the Ignalina NPP Reactor Emergency Core Cooling System, are analysed. The model parameters calculation is based on Ignalina NPP data regarding pumps operation and failure as well as on general Nordic NPPs reliability data (T-Book) appling Bayesian approach for parameters updating. The analysed safety system is a redundant system that consists of six pumps and other equipment. Therefore a model for multiple components failure was developed. The model accounts for actual operational requirements of the system. The results of this model are compared with usually used binominal model.

URANUS: Korean Lead-Bismuth Cooled Small Modular Fast Reactor Activities

2011 ◽  
Author(s):  
Sungyeol Choi ◽  
Il Soon Hwang ◽  
Jae Hyun Cho ◽  
Chun Bo Shim
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Base Isolation ◽  
Cooling System ◽  
Natural Circulation ◽  
High Level Waste ◽  
Isolation System ◽  
Base Isolation System ◽  
Enriched Uranium ◽  
Seismic Base Isolation

Since 1994, Seoul National University (SNU) has developed an innovative future nuclear power based on LBE cooling advanced Partitioning and Transmutation (P&T) approach that leaves no high-level waste (HLW) behind with transmutation reactor named as Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual, and Economical Reactor (PEACER). A small modular lead-bismuth cooled reactor has been designated as Ubiquitous, Robust, Accident-forgiving, Nonproliferating and Ultra-lasting Sustainer (URANUS-40) with a nominal electric power rating of 40 MW (100 MW thermal) that is well suited to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. URANUS-40 is a pool type fast reactor with and an array of heterogeneous hexagonal core, fueled by proven low-enriched uranium dioxide fuels. The primary cooling system is designed to be operated by natural circulation. 3D seismic base isolation system is introduced underneath the entire reactor building allowing an earthquake of 0.5g zero period acceleration (ZPA) for the Safe Shutdown Earthquake (SSE). Also, the proliferation risk can be effectively managed by capsulized core design and a long refueling period (25yr).

Numerical Analysis of Containment Response Under Accident Conditions

2013 ◽  
Author(s):  
Shengzhi Yu ◽  
Jianjun Wang
Keyword(s):  
Geometric Modeling ◽  
Nuclear Power ◽  
Cooling System ◽  
Natural Circulation ◽  
Physical Models ◽  
One Dimensional ◽  
Modeling Process ◽  
The One ◽  
Accident Conditions ◽  
Control Volumes

On the basis of passive containment cooling system concept, the one-dimensional codes are developed for the analysis of containment thermal-hydraulic characteristics under accident conditions, which can be applied to deal with large dry concrete containment in nuclear power plant. In order to build up the hypothesis flowing path, the containment space is divided into a rising channel and downward ring during the geometric modeling process. In this paper, the physical models, the identification of solving methods and the verification of the codes are introduced. It is assumed that the control volumes take the adiabatic condition in addition to heat exchanger and breaks, and there is no exchange of mass, momentum and energy between each volume in the rising channel and corresponding volume in downward ring. In the analysis, we also assume that the heat in the containment can only be transferred through natural circulation by passive containment cooling system. Furthermore, the break is supposed in the center of bottom of the containment. In this paper, the responses of the containment are predicted with the codes under large LOCA scenario. Under the same conditions, the characteristics of the natural circulation are also analyzed through the codes for the passive containment cooling system. The results can provide some references for the design of the passive containment cooling system.

Thermal Hydraulic Performance Analysis for Passive Containment Cooling System of AP1000 in the Event of LOCA

2017 ◽  
Author(s):  
Talha Bin Mujahid ◽  
Yu Yu ◽  
Bin Wang ◽  
Muhammad Ali Shahzad ◽  
Fenglei Niu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Analysis ◽  
Hydraulic Performance ◽  
Steam Line ◽  
Loss Of Coolant ◽  
Main Steam Line ◽  
Main Steam

The design of a nuclear reactor containment building is of key importance in order to enhance the safety of a nuclear power plant. Owing to nuclear accidents such as TMI, Chernobyl and Fukushima, more and more attention is paid to the passive concept in nuclear power development. In order to improve the safety of new generation nuclear power plant, passive systems are widely used, passive containment cooling system in AP1000 is one of the typical example of such kinds of systems. It’s function is to transfer the heat produced in the containment to the atmosphere and keep the pressure in the vessel below the threshold under such accidents as Loss of coolant (LOCA), main steam line break (MSLB), etc. The system operates based on natural circulations inside the steel vessel and in the air baffle outside the containment, and the cooling water is sprayed to the steel surface to enhance the heat transfer process. A proper model simulating the system behavior is needed for system design and safety analysis, and a multivolume lumped parameter approach is employed in order to analyze the containment integrity and to study the long term response of postulated Loss of coolant (LOCA) accidents and Main steam line break (MSLB) accidents. However, the temperature and pressure distributions cannot be described detailed by such model, which is important to study the T-H characteristics in the containment. In this paper LOCA has been simulated on MATLAB using a given pipe break size and the response of containment is analyzed. Furthermore, the results are compared with the results in the Westinghouse Design Control Document 2002. Then the thermal hydraulic performance is studied, the factors such as the air temperature, containment pressure and mass flow rate of the coolant and their effects on the containment are analyzed. This research is done to get further insight on the safety analysis of reactor containment regarding maximum temperature and stress calculation inside the containment.

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