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.

scholarly journals Mathematical modeling and numerical investigation of density wave instability of supercritical natural circulation loop

2021 ◽  
Vol 321 ◽  
pp. 04004
Author(s):  
Santosh Kumar Rai ◽  
Neha Ahlawat ◽  
Pardeep Kumar ◽  
Vinay Panwar
Keyword(s):  
Energy Conservation ◽  
Natural Circulation ◽  
Density Wave ◽  
Circulation Loop ◽  
Wave Instability ◽  
Conservation Equations ◽  
One Dimensional ◽  
Natural Circulation Loop ◽  
The One ◽  
Riser Height

In present paper, a mathematical model based on the one dimensional nonlinear mass, momentum and energy conservation equations has been developed to study the density wave instability (DWI) in horizontal heater and horizontal cooler supercritical water natural circulation loop (HHHC-SCWNCL). The one dimensional nonlinear mass, momentum and energy conservation equations are discretized by using finite difference method (FDM). The numerical model is validated with the benchmark results (NOLSTA model). Numerical simulations are performed to find the threshold stability zone (TSZ) and draw the stability map for natural circulation loop. Further, effect of change in diameter and riser height on the density wave instability of SCWNCL has been investigated.

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.

scholarly journals Neutron flux simulation of a one-dimensional reactor zone taking into account xenon generation

Energetika ◽  
2020 ◽  
Vol 65 (4) ◽  
Author(s):  
Zsófia Tóth ◽  
Dániel Péter Kis
Keyword(s):  
Numerical Methods ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Research Work ◽  
Reactor Core ◽  
Time And Space ◽  
One Dimensional ◽  
Safety And Reliability ◽  
The One

The energy in nuclear power plants is produced by thermal fission. It is extremely important to be able to monitor the processes in the reactor to ensure the safety and reliability of the power plant. One of the main traits of the reactor core is neutron flux. It changes in time and space therefore it is crucial to be able to simulate its changes with computer codes. In the research work a program code was established in the Matlab software with which the neutron flux of a one-dimensional zone can be simulated with homogenous and heterogenic zone parameters as well. The code is written using the one-group one-dimensional time- and space-dependent diffusion equation. The equation of an average delayed neutron group and xenon and iodine distributions was also included in the system to give a more precise look on the problem. The main innovation in the code is that numerical methods were used to solve the problem: the finite difference approach was applied for the place-dependent and for the time-dependent solution. The advantage of this code compared to other ones is that one-dimensional zones can be simulated in a really short time and it still gives a precise solution because of the complex numerical methods used.

Analysis of the Containment Thermal-Hydraulic Behavior of a Small Reactor During Loss of Coolant Accident

2017 ◽  
Author(s):  
Qian Lin ◽  
Weizhong Zhang
Keyword(s):  
Cooling System ◽  
Three Dimensional ◽  
Dimensional Model ◽  
Three Dimensional Model ◽  
One Dimensional ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
The One

The containment thermal hydraulics of a small reactor during loss of coolant accident (LOCA) is studied by a lumped parameter one-dimensional model and a three-dimensional model. The capability of a kind of heat exchanger type passive containment cooling system (PCCS) is analyzed by the one-dimensional model. The calculation results show that, the decay heat can be removed and the containment pressure can be decreased by the proposed PCCS. The steam and non-condensable gas (the air) distribution in the containment is investigated, the mixing and stratification behaviors are analyzed for several different cases, in which the PCCS and condenser are located at higher, base or lower position. The sensitivity analysis of the PCCS elevation shows that, in despite of the different gas stratification, the containment pressures are nearly the same. Similar conclusions can be obtained by the one-dimensional model and three-dimensional model. The preliminary results may indicate that, the designed PCCS and condenser can be located at a lower part, which will be benefit for the economy of the small reactor or meet other requirements.

Theoretical and Computational Analysis of an Auxiliary Power Unit (APU) Installation Cooling System

2012 ◽  
Author(s):  
Parthiv N. Shah ◽  
Tricia Waniewski Sur ◽  
R. Scott Miskovish ◽  
Albert Robinson
Keyword(s):  
Network Model ◽  
Cooling System ◽  
Dimensional Model ◽  
Ambient Conditions ◽  
Modeling Framework ◽  
Flow Network ◽  
One Dimensional ◽  
Dimensional Flow ◽  
System Pressure ◽  
The One

This paper presents a theoretical one-dimensional model and computational fluid dynamics (CFD) simulations of a tailcone-installed APU cooling system. The work is motivated by the need to deliver sufficient cooling airflow to critical components within an aircraft tailcone compartment. The cooling system considered herein utilizes (1) an eductor system at the APU exhaust and (2) a ram air scoop near an upstream inlet to the compartment to induce the necessary cooling flow during ground and in-flight APU operation. A one-dimensional flow network model provides a framework for the quantification and matching of eductor pumping and system pressure drop characteristics. Detailed CFD models that simulate internal tailcone compartment flows driven by ambient conditions external to the aircraft in ground or flight operation support the one-dimensional model and are used to characterize component performance and assess different scoop and eductor designs. The one-dimensional flow network model is calibrated to the CFD results to predict system cooling performance under known APU loads at points on the ground and in the flight envelope. The agreement between the models is encouraging and suggests the modeling framework and CFD techniques discussed will be applicable to future designs and improvements of eductor-driven aircraft compartment cooling systems.

IRSN R&D Actions on FP Behaviour for RCS, Containment and FCVS in Severe Accident Conditions

2016 ◽  
Author(s):  
Laurent Cantrel ◽  
Thierry Albiol ◽  
Loïc Bosland ◽  
Juliette Colombani ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Porous Materials ◽  
Nuclear Power ◽  
Source Term ◽  
Cooling System ◽  
Fission Products ◽  
Simulation Software ◽  
Severe Accident ◽  
Status Report ◽  
Diagnosis And Prognosis ◽  
Accident Conditions

This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behaviour in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. For the last topic, in link with the Fukushima post-accident management and possible improvement of mitigation actions for such SA, the FCVS topic is again on the agenda (see Status Report on Filtered Containment Venting, OECD/NEA/CSNI, Report NEA/CSNI/R(2014)7, 2014.) with a large interest at the international scale. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in ASTEC SA simulation software. After being initiated in the International Source Term Program (ISTP), researches devoted to the understanding of iodine transport through the RCS are still ongoing in the frame of a bilateral agreement between IRSN and EDF with promising results. In 2017, a synthesis report of the last 10 years of researches, which have combined experimental and fundamental works based on the use of theoretical chemistry tools, will be issued. For containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project operated by IRSN. The objective of the STEM project was to improve the evaluation of Source Term (ST) for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major fission products: iodine and ruthenium. More precisely, the STEM project provided additional knowledge and improvements for calculation tools in order to allow a more robust diagnosis and prognosis of radioactive releases in a SA. STEM data will be completed by a follow-up, called STEM2, to further the knowledge concerning some remaining issues and be closer to reactor conditions. Two additional programmes deal with FCVS issues: the MItigation of outside Releases in the Environment (MIRE) (2013–2019) French National Research Agency (NRA) programme and the Passive and Active Systems on Severe Accident source term Mitigation (PASSAM) (2013–2016) European project. For FCVS works, the efficiencies for trapping iodine with various FCVS, covering scrubbers and dry filters, are examined to get a clear view of their abilities in SA conditions. Another part, performed in collaboration with French universities (Lorraine and Lille 1), is focused on the enhancement of the performance of these filters with specific porous materials able to trap volatile iodine. For that, influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio …) will be tested. New kind of porous materials constituted by Metal organic Frameworks (MOF) will also be looked at because they can constitute a promising way of trapping.

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.

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.

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