scholarly journals Analysis of in-vessel accident progression in VVER1000 NPP during SBO accident with external reactor vessel cooling method

2021 ◽  
Vol 10 (2) ◽  
pp. 1-14
Author(s):  
Long Doan Manh ◽  
Thai Nguyen Van ◽  
Thanh Tran Chi
Keyword(s):  
Molten Pool ◽  
Reactor Vessel ◽  
Fuel Cladding ◽  
Support Plate ◽  
Cooling Strategy ◽  
Cooling Method

In this study, the MELCOR v1.8.6 code was utilized to perform an analysis of the in-vessel accident progression in VVER1000 reactor during the Station Black-Out (SBO) accident with and without external reactor vessel cooling (ERVC) strategy. The analysis presented the predictions of the main phenomena during the accident such as failure of fuel cladding, collapse of lower core support plate, relocation of core debris to lower plenum and mass of debris components in lower plenum, and provided comparisons between two cases in term of main parameters such as integrity time of reactor and structure components of molten pool. These parameters are very important inputs for further research on the application of external vessel cooling strategy for VVER1000 reactor.

scholarly journals Research on Cooling Method in Surfacing Repair Process of Aero Compressor Blade

2020 ◽  
Vol 2020 ◽  
pp. 1-19
Author(s):  
Shijie Dai ◽  
Miao Gong ◽  
Liwen Wang ◽  
Tao Wang
Keyword(s):  
Heat Transfer ◽  
Forced Convection ◽  
Heat Sink ◽  
Molten Pool ◽  
Cooling Effect ◽  
Modeling And Simulations ◽  
Cooling Method ◽  
Jet Impact ◽  
Convection Cooling ◽  
Vertical Jet

For the cooling method in surfacing repairing, most of the research focuses on the method based on the fixture structure. However, due to the low thermal conductivity and ultrathin alloy blade, the heat transfer speed from the molten pool to fixture is slow. When the heat is transferred to the fixture, most of the molten pool has solidified and absorbed or segregated out some impurities. Therefore, how to cool the welding area directly is more critical. For this reason, the thermal cycle characteristics of typical points of the blade and the heat transfer process of the key area of the fixture are analyzed, the original cooling time is calculated, and two innovative cooling methods based on lateral forced convection cooling and vertical jet impact forced convection cooling are proposed. For lateral forced cooling, with “AF-field” lateral convection cooling modeling, the cooling effects of characteristic points and sections under different flow velocities are calculated. For vertical jet impact cooling, the pressure, flow rate, and convective heat flux distribution on the wall under different impact heights and nozzle diameter are calculated. The influence of different inlet flow rates on cooling performance is influenced, based on the analysis results of impact modeling, the moving heat sink model is established, and the cooling effect under different heat sink-source distances is calculated. The heat transfer rules of two methods are analyzed in detail through modeling and simulations. The results show that both methods can improve the cooling effect and the vertical jet impact cooling method has an effect that is more obvious. When the nozzle radius is 2 mm, the impact height is 4d, the inlet flow velocity is 35 m/s, and the distance is 7 mm, and the cooling time under the vertical jet impact method is shortened by 12.5%, which can achieve better cooling effect. The experiment further validates the effectiveness of the modeling and simulations.

Thermal Loading Failure Analysis of Molten Pool in the Reactor Vessel

2010 ◽  
Author(s):  
Xiao Yang ◽  
Yanhua Yang ◽  
Bo Kuang ◽  
Xiaoliang Fu
Keyword(s):  
Heat Flux ◽  
Molten Pool ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Loading ◽  
Success Probability ◽  
Local Heat ◽  
Reactor Vessel ◽  
Transfer Model ◽  
Severe Accidents

Uncertainties are addressed in the special context of assessing and managing risks from rare, severe-consequence hazards. Risk Oriented Accident Analysis Methodology (ROAAM) is used to analyze uncertainties during severe accidents analysis in nuclear power plants. In-vessel Retention (IVR) is one of the mitigations for severe accidents which will cause core damage. By external reactor vessel cooling (ERVC), the integrity of the reactor vessel is preserved. The success criterion for IVR is the local heat flux on the wall of lower head is less than the critical heat flux (CHF). This paper analyzes the uncertain parameters which decide the mitigation to be successful or fail. Two bounding structures and 4 molten pool steady states are defined. And the success probability of IVR is evaluated with a molten pool heat transfer model. Then the effectiveness of IVR-ERVC under the two bounding structures is evaluated.

Analysis and Optimization of Heat Treatment Process for Lower Core Support Plate and Core Barrel of Reactor Vessel Internals

2017 ◽  
Author(s):  
Dongan Liu ◽  
Shaoxuan Lin ◽  
Zonghua Ding
Keyword(s):  
Heat Treatment ◽  
Power Plant ◽  
Nuclear Power ◽  
Treatment Process ◽  
Reactor Vessel ◽  
Heat Treatment Process ◽  
Safe Operation ◽  
Support Plate ◽  
General Rules ◽  
Fuel Assemblies

Lower Core Support Plate (LCSP) and Core Barrel (CB) are key components of reactor vessel internals. Especially, since the fuel assemblies are installed on the LCSP, its flatness is critical for the safe operation of fuel assemblies. However, for SM1 and HY1 nuclear power plant (NPP), after heat treatment of the weld between LCSP and CB, the LCSP deforms seriously and its flatness exceeds the limitation, which results in a time-consuming and costly reprocessing. A numerical model of heat treatment process between LCSP and CB was developed first. The general rules of temperature and deformation distribution of LCSP and CB were obtained. Also, an experiment was conducted to validate the model. With the validated model, the deformation mechanism of LCSP due to heat treatment is studied. At last, the heat treatment process between LCSP and CB was optimized to avoid similar issues for the following NPPs.

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3D©

2002 ◽  
Author(s):  
D. L. Knudson ◽  
J. L. Rempe
Keyword(s):  
Heat Transfer ◽  
Molten Pool ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Complex Processes ◽  
Lower Head ◽  
Environmental Laboratory ◽  
Molten Materials ◽  
Core Materials

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D© has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D© relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D© are outlined.

A Method for Online Calibration of MAAP Simulations in a Severe Accident Management System Database

2017 ◽  
Author(s):  
Weifeng Xu ◽  
Fangqing Yang ◽  
Peng Chen ◽  
Yehong Liao
Keyword(s):  
Management System ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Outlet Temperature ◽  
Online Calibration ◽  
Support Plate ◽  
Initial Event ◽  
Accident Management ◽  
Accident Simulation ◽  
Plate Failure

During a nuclear plant accident, five accident events are usually considered, including core uncovery, core outlet temperature arrived at 650 °C, core support plate failure, reactor vessel failure and containment failure. In accident emergency aspect, when an accident happens, the initial event can be utilized in the severe accident management system which is based on MAAP to simulate the long process of the accident, so as to provide support for operators to take actions. However, in MAAP, many sensitivity parameters exist, which reflect phenomenological uncertainty or models uncertainty and will influence the happening time of the five accident events above. In this paper, based on MAAP5 and LOCAs, the CPR1000 is simulated to analyze the influences of MAAP5’s sensitivity parameters reflecting phenomenological uncertainty on the accident process, which is aimed to find out the sensitivity parameters associated to the five important accident events and build the database between these sensitivity parameters and five accident events’ happening time. Then, based on the research above, a preliminary approach to optimize the MAAP5’s accidents simulation is introduced, which is realized by adjusting sensitivity parameters. Finally, the application of this research will be showed in a severe accident management system developed by us. The research results offer great reference significance for the severe accident simulation and prediction in MAAP5.

Reactor Cavity Analysis Under Steam Explosive Conditions by TNT Model

2016 ◽  
Author(s):  
Seung-Huyn Kim ◽  
Yoon-Suk Chang ◽  
Yong-Jin Cho
Keyword(s):  
Shock Wave ◽  
Steam Explosion ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Vessel ◽  
Dynamic Effects ◽  
Element Mesh ◽  
Cooling Strategy ◽  
Hydraulic Conditions ◽  
Physical Phenomena

Steam explosion may occur in nuclear power plants by fuel-coolant interactions when the external reactor vessel cooling strategy is failed. This phenomenon can cause shock wave that endangers surrounding reactor cavity wall due to resulting dynamic effects. Even though extensive researches have been performed to predict influences of the steam explosion, due to complexity of physical phenomena and thermal-hydraulic conditions, it is remained as one of possible hazards. The object of this study is to examine characteristics of reactor cavity and nuclear components under representative steam explosion conditions. In this context, an assembled finite element mesh was generated and evaluated by the trinitrotoluene model. As a result, stresses, strains and displacements of the reactor cavity and nuclear components were calculated. Subsequently, crack evaluation of reinforced concrete was performed and their results were discussed.

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

2014 ◽  
Author(s):  
Pavlin P. Groudev ◽  
Antoaneta E. Stefanova ◽  
Petya I. Vryashkova
Keyword(s):  
Parametric Study ◽  
Cold Water ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fuel Cladding ◽  
Accident Management ◽  
Management Guidance ◽  
Fuel Behavior ◽  
Accident Conditions

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.

Sign in / Sign up

Export Citation Format

Share Document