Development of a Multiphase Particle Method for Melt-Jet Breakup Behavior of Molten Core in Severe Accident

2020 ◽  
Author(s):  
Zidi Wang ◽  
Yuzuru Iwasawa ◽  
Tomoyuki Sugiyama
Keyword(s):  
Nuclear Power ◽  
Particle Method ◽  
Transient Heat Conduction ◽  
Severe Accident ◽  
Lagrangian Description ◽  
Potential Threat ◽  
High Order Scheme ◽  
Containment Vessel ◽  
Jet Breakup ◽  
Solidification Processes

Abstract In a hypothetical severe accident in a light water reactor (LWR) nuclear power plant, there is a possibility that molten core released from the reactor vessel gets in contact with water in the containment vessel. In this so-called fuel-coolant interactions (FCIs) process, the melt jet will breakup into fragments, which is one of the important factors for a steam explosion, as a potential threat to the integrity of the containment vessel. The particle method could directly and easily capture the large deformed interfaces by particle motions, benefiting from its Lagrangian description and meshless framework. In order to investigate the melt-jet breakup with solidification processes, a multiphase particle method with arbitrary high order scheme is presented in this study. In addition, an interfacial particle shifting scheme is developed to suppress the unnatural particle penetration between different phases. The convergence rate with different order is firstly confirmed by a verification test in terms of both explicit and implicit calculations. Then, a transient heat conduction between two materials is carried out and quite good results are obtained. After that, a rising bubble benchmark is performed to show the feasibility of modelling for deformation and collapse. Improvements of clear interface are indicated compared with previous reported results. Two important multiphase instabilities, namely the Rayleigh-Taylor instability and the Kelvin-Helmholtz instability, are studied since they play important roles during the melt-jet breakup. The results achieved so far indicate that the developed particle method is capable to analyze the melt-jet breakup with heat transfer.

Numerical Investigation of Melt Jet Breakup With Different Shapes in Water Pool

2017 ◽  
Author(s):  
Hui Cheng ◽  
Jiyun Zhao
Keyword(s):  
Surface Area ◽  
Steam Explosion ◽  
Nuclear Power ◽  
3D Model ◽  
Severe Accident ◽  
Interface Surface ◽  
Jet Breakup ◽  
Different Shapes ◽  
Interface Surface Area ◽  
2D Model

During a severe accident in nuclear power plant, core damage may occur due to decay heat and molten fuel can pour into and interact with water resulting in steam explosion. The energetics of steam explosion strongly depends on the initial premixing stage during which the molten fuel undergoes a coarse fragmentation process, which determines the surface area for fuel-coolant contact and heat transfer. Extensive research has been done to understand the premixing stage, however, most of the studies are focused on the cylindrical jet interaction with water. In fact, during core melt, the molten fuel may pour near the edge of core, so the shapes and size of melt jet may differ significantly based on specific conditions. In this paper, numerically study on the melt jet breakup with different shapes in pool water are conducted, such as elliptical shape with VOF method. Firstly, the deformation of molten jet under the same conditions in 2D model is compared with 3D model and shows that the breakup of 3D model is quite different from 2D model, the integration of 3D model is maintained much better than 2D model. Then the characteristics of breakup of elliptic cylindrical melt jet are analyzed and compared with cylindrical melt jet. The results shows that the interface surface area of elliptic cylindrical jet is nearly twice the cylindrical jet.

scholarly journals Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

2014 ◽  
Vol 2014 ◽  
pp. 1-7 ◽  
Author(s):  
Min Yoo ◽  
Sung Min Shin ◽  
Hyun Gook Kang
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Transfer Model ◽  
Potential Threat ◽  
Severe Accidents ◽  
Development Of Instrument ◽  
Novel Concept

Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations), and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

Long-Term Cooling Strategy for the Primary Containment Vessel of the Kashiwazaki-Kariwa Nuclear Power Station in a Severe Accident

2017 ◽  
Author(s):  
Satoshi Kawaguchi ◽  
Satoshi Mizuno ◽  
Yoshihiro Oyama
Keyword(s):  
Heat Exchanger ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Power Station ◽  
Working Temperature ◽  
Containment Vessel ◽  
Portable Pump

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident. If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower. TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout). This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment. Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied. (1) RPV heat removal by restoring the RHR system (2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger (3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers (4) RPV heat removal using the clean up water system (CUW) By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.

Melting Simulation Using a Particle Method With Angular Momentum Conservation

2017 ◽  
Author(s):  
Masahiro Kondo ◽  
Shota Ueda ◽  
Koji Okamoto
Keyword(s):  
Angular Momentum ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Particle Method ◽  
Momentum Conservation ◽  
Severe Accident ◽  
Angular Momentum Conservation ◽  
The Core ◽  
Melting And Solidification

To analyze the core degradation and relocation behavior of melts in a severe accident of nuclear power plant, the melting and solidification in the complexed geometry is to be calculated. For the calculation of such complexed behavior, a new particle method conserving angular momentum is proposed and applied for the melting simulation. When solid melts, it may move like a rigid body. The angular momentum conservation is important to capture such kind of motion. The potential of the new particle method was confirmed with a calculation of the melting in dam break geometry and cantilever geometry.

Development of Numerical Simulation for Jet Breakup Behavior in Complicated Structure of BWR Lower Plenum: (3) Influence by Complicated Structure on Jet Breakup and Fragmentation Behavior

2014 ◽  
Author(s):  
Ryusuke Saito ◽  
Yutaka Abe ◽  
Akiko Kaneko ◽  
Takayuki Suzuki ◽  
Hiroyuki Yoshida ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Nuclear Power ◽  
Constant Term ◽  
Severe Accident ◽  
Shearing Stress ◽  
Complicated Structure ◽  
Simulation Code ◽  
Fragmentation Behavior ◽  
Jet Breakup ◽  
Image Velocimetry

To estimate the state of Reactor Pressure Vessel (RPV) of Fukushima Daiichi nuclear power plant, it is important to clarify the breakup and the fragmentation behavior of molten material jet in BWR lower plenum by a numerical simulation. To clarify the effects of complicated structures on jet breakup and fragmentation behavior experimentally and construct the benchmarks of the simulation code, we conduct the visualized experiments simulating the severe accident in the BWR. In this study, the jet breakup behavior, the fragmentation behavior and internal/external velocity profiles of the jet were observed by the backlight method and the particle image velocimetry (PIV). From experimental results, it is clarified that the complicated structures prolong the jet breakup length or make the fragments fallen together to the lower plenum similar to the bulk state. In addition, it is clarified that strong shearing stress occurs at the crest of interfacial waves at side of the jet when fragments are generated. Finally, the fragment diameters measured in the present study well agree with the theory suggested by Kataoka et al. (1983) by changing the coefficient term at each experimental condition. Thus, it is suggested that the fragmentation mechanism is mainly controlled by shearing stress and the fragment diameter can be estimated by adjusting the constant term.

Simulation of TOSQAN 101 Containment Spray Test With Combined Eulerian CFD and Droplet-Tracking Modelling

2008 ◽  
Author(s):  
Miroslav Babic´ ◽  
Ivo Kljenak ◽  
Matjazˇ Leskovar ◽  
Borut Mavko
Keyword(s):  
Boundary Condition ◽  
Gas Phase ◽  
Nuclear Power ◽  
Sensitivity Study ◽  
Severe Accident ◽  
Research Network ◽  
Lagrangian Description ◽  
Global Variables ◽  
Simulation Results ◽  
Droplet Tracking

The purposes of containment spray system operation during a severe accident in a light water reactor (LWR) nuclear power plant (NPP) are to depressurize the containment by steam condensation on spray droplets, to reduce the risk of hydrogen burning by mixing the containment atmosphere, and to collect radioactive aerosols from the containment atmosphere. While the depressurization may be predicted fairly well using lumped-parameter codes, the prediction of mixing and collection of aerosols requires a local description of transport phenomena. In the present work, modelling of sprays on local instantenous scale is presented and the Design of Experiment (DOE) method is used to assess the influence of boundary conditions on the simulation results. The TOSQAN 101 spray test, which was used for a benchmarking exercise within the EU Severe accident research network of excellence (SARNET), was simulated, and simulation results were compared to experimental data. The modelling approach is based on a Lagrangian description of the dispersed liquid phase (droplets), an Eulerian approach for the description of the continuous gas phase, and a two-way interaction between the phases. The simulations are performed using a combination of the computational fluid dynamics (CFD) code CFX4.4, which solves the gas transport equations, and of a newly proposed dedicated Lagrangian droplet-tracking code. The intent of the presented work is to assess the modeling of sprays and liquid on the wall in the presented approach with the emphasis on the heat and mass transfer between liquid and gas phase. The simulation-experiment comparison of available global and local variables demonstrates that the proposed approach is suitable for prediction of global variables evolution and of the non-homogenous structure of the atmosphere. The boundary condition steady-state sensitivity study is performed and shows that the global variables are mostly affected by the wall temperature boundary condition.

scholarly journals Important Severe Accident Research Issues After Accident at Fukushima Daiichi Nuclear Power Station

2013 ◽  
Author(s):  
Jun Sugimoto
Keyword(s):  
Nuclear Power Station ◽  
Working Group ◽  
Nuclear Power ◽  
Severe Accident ◽  
Power Station ◽  
Fukushima Accident ◽  
Research Issues ◽  
Containment Vessel ◽  
Accident Research ◽  
Fukushima Daiichi

After the accident at Fukushima Daiichi Nuclear Power Station several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, some recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started after the Fukushima accident in this working group. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. In January 2012, Research Expert Committee on Evaluation of Severe Accident was established in AESJ in order to investigate severe accident related issues for future LWR development and to propose action plans for future severe accident research, in collaboration with this working group. Based on these activities and also author’s personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Specifically those are investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations.

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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