Experimental Investigations on the Prototypic Steam Explosion

2005 ◽  
Author(s):  
J. H. Song ◽  
J. H. Kim ◽  
B. T. Min ◽  
S. W. Hong
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Test Section ◽  
Steam Explosion ◽  
Nuclear Power ◽  
Weight Percent ◽  
Experimental Investigations ◽  
Dynamic Force ◽  
Reference Case

This paper discusses results of a series of steam experiments using a prototypic material representing the molten core of nuclear power plant. Five experiments are discussed in detail in addition to a brief review of the previous experiments, where the focuses were on the effect of corium composition and external trigger on the strength of the steam explosion. A mixture of UO2:ZrO2 is used for the experiment, where the weight percent of each component is changed. One experiment was performed with corium at a composition of 70:30 without an external trigger as a reference case. Three tests were performed by using corium at the same composition with an external trigger. The last experiment was performed using corium at a composition of 80:20 with an external trigger. Various parameters are measured including the dynamic pressures on the wall of the test section and the dynamic force at the bottom of the test section. From the experimental data, the strength of the steam explosion was evaluated. It is shown that the strength of the steam explosion highly depends on the composition. A comparison between the cases with an external trigger and the cases without external trigger indicates that there is no substantial escalation of the strength of the explosion due to an external trigger.

scholarly journals The Use of System Codes in Scaling Studies: Relevant Techniques for Qualifying NPP Nodalizations for Particular Scenarios

2014 ◽  
Vol 2014 ◽  
pp. 1-13 ◽  
Author(s):  
V. Martinez-Quiroga ◽  
F. Reventos
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Small Scale ◽  
Plant Operation ◽  
Essential Step ◽  
Hydraulic Safety ◽  
And Control

System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations. This paper describes the methodology developed at the Technical University of Catalonia in order to contribute to the qualification of Nuclear Power Plant nodalizations by means of scale disquisitions. The techniques that are presented include the so-calledKv-scaled calculation approach as well as the use of “hybrid nodalizations” and “scaled-up nodalizations.” These methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The paper explains both the concepts and the general guidelines of the method, while an accompanying paper will complete the presentation of the methodology as well as showing the results of the analysis of scaling discrepancies that appeared during the posttest simulations of PKL-LSTF counterpart tests performed on the PKL-III and ROSA-2 OECD/NEA Projects. Both articles together produce the complete description of the methodology that has been developed in the framework of the use of NPP nodalizations in the support to plant operation and control.

Operating Condition of Steam Injector as a Passive Cooling System at Severe Accident of Nuclear Power Plant

2012 ◽  
Author(s):  
Yutaka Abe ◽  
Shunsuke Shibayama ◽  
Akiko Kaneko ◽  
Chikako Iwaki ◽  
Tadashi Narabayashi ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Structure ◽  
Test Section ◽  
Nuclear Power ◽  
Water Jet ◽  
Back Pressure ◽  
Severe Accident ◽  
Opening Ratio ◽  
Discharge Pressure

Steam injector (SI) is a passive jet pump which is driven by high-performance steam condensation onto water jet and it is expected to be active at severe accident of nuclear power plant with no electricity. SI is mainly consists of convergent-divergent nozzle. Supersonic steam flow condenses onto water jet in the mixing nozzle and mass, momentum, and energy of steam is transferred to water in the mixing nozzle. Condensed water jet is accelerated at the throat and kinetic energy is converted into pressure in the diffuser, which produces higher pressure than inlet steam pressure. It is easy to apply the SI to nuclear power plant since SI has quite simple and compact structures. The objectives of the present study are to clarify the mechanism of heat and momentum transfer in the mixing nozzle and to determine operating range of SI for practical use. A transparent test section is adopted to conduct visualization of the flow structure with a high-speed video camera as well as measurement of pressure distribution in mixing nozzle, throat, and diffuser with changing back pressure. Fundamental parameters change between operative and inoperative state of the injector were evaluated by measuring pressure and temperature distribution along axial direction of the test section. Discharge pressure as one of operating characteristics of the injector was also measured in changing back pressure by decreasing the opening ratio of the back pressure valve attached downstream of the test section. It was confirmed that discharge pressure increased and the injector became inoperative unsteadily with decreasing opening ratio of the back pressure valve just after it produced the maximum discharge pressure. In the present investigation, this maximum discharge pressure is evaluated as the operation limit of the injector. Furthermore, discharge pressure from diffuser, which is one of the indicators of operating performance as well as operating limit is predicted from inlet condition adopting one-dimensional analysis model proposed previously. By comparing analytical result with experimental data, as well as visualization of flow structure in throat and diffuser, physics model including two-phase flow structure with shock wave which was observed at throat and diffuser are discussed in order to predict injector’s operation with high accuracy.

Possibilities of generalization of experimental data on lifetime change of nuclear-power-plant parameters

Atomic Energy ◽  
1992 ◽  
Vol 73 (3) ◽  
pp. 762-764 ◽  
Author(s):  
V. S. Emel'yanov ◽  
O. G. Kamyshnikov ◽  
V. I. Morozkin ◽  
Yu. I. Raevskii
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power

The Impacts of Scale Effect on Flow Patterns in a Supersonic Steam Injector

2017 ◽  
Author(s):  
Masaya Fujishiro ◽  
Yutaka Abe ◽  
Akiko Kaneko
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Scale Effect ◽  
Flow Rate ◽  
Test Section ◽  
Nuclear Power ◽  
Flow Patterns ◽  
Water Jet ◽  
Steam Flow ◽  
And Performance

From the viewpoint of an importance of safety, the nuclear power plant should be managed to prepare severe accidents. The performance of safety dropped by an accident is strongly to be minimized during the situation of station blackout. The installation of a steam injector (SI) into the nuclear power plant has long been expected. In the SI, the steam condenses due to the direct contact at the surface of water jet, resulting in the force attracting water. The force drives the circulation of an amount of coolant water. SI also works as a reactor condenser thanks to its high efficient performance during the condensation. Because any external forces to circulate water and steam are not required, SI can be operated without the electric powers. The structure of SI is similar to a convergent-divergent nozzle. After the flow acceleration at a throat, the discharged pressure is expected to exceed the inlet pressure. Owing to its quite simple structure, the reduced cost of installation and maintenance is also expected. The following previous studies for four cases of throat diameter clarified two-phase flow structures and heat transfer characteristics in water jet and performance of SI: (i) Narabayashi et al. (2000) examined for 5.5 and 6.5 mm in diameter; (ii) Osakabe et al. (2004) for 3.4 mm; (iii) Koizumi et al. (2006) for 4 mm; (iv) Abe et al. (2014) for 4, 6.5, and 8 mm. Although these clarified the operative state which formed a water jet, operative condition was not elucidated. Furthermore, the scale effect for various diameters of SI has not been discussed in detail. The aim of this study is to clarify scale effect of a test section on operating criteria and performance. Experiment was performed to clarify the scale effect by using three types of throat diameters: 4, 6.5, and 8 mm. As a result, three formations of a water jet were observed: (i) formation, (ii) incomplete formation, and (iii) no formation. We proposed a classification which enables us to categorize complex flow patterns into five regimes. We clarified the operating criteria of them by comparing water flow rate with steam flow rate. SI did not form a water jet on the condition with low steam flow rate. The suppling water was stopped, and only steam was supplied to the test section for the condition that steam latent heat was larger than subcooled water enthalpy.

TRACE Validation Against LOCA Transients Performed on FIX-II Facility

2013 ◽  
Author(s):  
Xiao Hu ◽  
Mian Xing ◽  
Weimin Ma
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation Models ◽  
Primary System ◽  
Choked Flow ◽  
Loss Of Coolant Accident ◽  
Trace Model ◽  
Transient Thermal

As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. The FIX-II experiments were performed to produce experimental data for understanding the initial stage of LOCA and so as to verify the computational codes. In the present study, based on FIX-II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 which correspond to different LOCA cases were developed to validate the TRACE code (version 5.0 patch 2). The predictions of the TRACE code including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core were compared with the experimental data. The results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model and choked flow model on the results, which show that both the models have significant influence on the outcome of the model.

Estimation of Ex-Vessel Steam Explosion Pressure Loads for AP1000 Nuclear Power Plant

2014 ◽  
Author(s):  
Zhigang Li ◽  
Jinghao Li ◽  
Meng Lin ◽  
Yanhua Yang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Explosion ◽  
Nuclear Power ◽  
Geometrical Model ◽  
Radioactive Material ◽  
Coolant Temperature ◽  
Reactor Accident ◽  
Melt Temperature ◽  
Explosion Pressure

An ex-vessel steam explosion is a fuel coolant interaction process which may occur when the reactor vessel fails and the molten core pours into the water in the reactor cavity during a severe reactor accident. A strong enough steam explosion in a nuclear power plant could endanger the containment integrity and lead to a direct release of radioactive material to the environment. In this article, a nuclear island geometrical model of AP1000 nuclear power plant was established and different scenarios of ex-vessel steam explosions in AP1000 NPP were simulated by MC3D code. Since the initial parameters with large quantity of uncertainties under accident condition may have important effects on the steam explosion, some initial parameters study were performed by varying the location of the melt release (75°,45°,30°,0°), the cavity water subcooling, the triggering time for explosion calculations, the melt temperature and the break size. Results indicate that the higher the melt temperature, the longer the triggered time and the lower the coolant temperature would lead to the more severe steam explosion. Besides, when the angle of break reaches 45 degree and the diameter of the break is 0.5m, the steam explosion causes the largest damage.

Life Extension of Bolted Flanged Connections From a Design Peculiarities Point of View

2004 ◽  
Author(s):  
Anatoly I. Efremov
Keyword(s):  
Experimental Data ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Point Of View ◽  
Life Extension ◽  
Design Improvement ◽  
Power Plant Equipment ◽  
Bolt Preload ◽  
Plant Equipment

Life extension of Bolted Flanged Connections (BFC) depends directly on early leakage that is a major cause of bolt degradation and failure. Experiments with standard BFC typically used in Nuclear Power Plant equipment and subjected to bolt preload and subsequent internal pressure revealed an influence of the BFC design peculiarities on early leakage. The experimental data became a basis for BFC design improvement.

Experimental Investigations on Temperature Distribution and Heat Removal Capability of Residual Heat Exchanger

2013 ◽  
Author(s):  
Jiangbo Wu ◽  
Qincheng Bi ◽  
Chengsi Zhou
Keyword(s):  
Heat Exchanger ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Rate ◽  
Nuclear Power ◽  
Heat Removal ◽  
Experimental Investigations ◽  
Outlet Temperature ◽  
Forced Circulation ◽  
Residual Heat

The residual heat exchanger in nuclear power plant is the key component of secondary side passive residual heat removal system, where the performance of removing decay heat by condensation and pool boiling to the secondary water storage tanks in the residual heat exchanger is crucial to the safety of the nuclear power plant. In the present paper, an experimental facility is built to evaluate the heat removal capability of the residual heat exchanger in both steady state natural circulation and forced circulation at different pressures. The high pressure steam is forced to flow and enter into heat exchanger with a slight inclined tube, which is installed in a large water pool. Experiments are carried out to study the characteristics of the steam condensation in the residual heat exchanger at different parameters. A calculation code for modeling this process and predicting the outlet temperature is also developed. The results show that the temperature difference between inlet and outlet increases with the increase of the inlet steam pressure due to the variation of latent heat. Meanwhile, the outlet temperature also increases with increasing flow rate. The calculation results accord with the experimental data at low mass flow rate. It is also found that the two calculation models proposed by J.R. Thome predict the flow pattern well, and the Shah’s equation is more suitable to estimate the heat transfer characteristic.

scholarly journals A simulation study of processes for mixing non-isothermal flows under dynamic effects

2020 ◽  
Author(s):  
Aleksandr Sataev ◽  
Vyacheslav Andreev ◽  
Denis Novikov ◽  
Julia Perevezentseva
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Vertical Component ◽  
Dynamic Force ◽  
Mixing Processes ◽  
Periodic Effect ◽  
Temperature Pulsations ◽  
Isothermal Flows ◽  
Isothermal Water

The processes for mixing of non-isothermal streams essentially define the parameters of the heat-carrier on an input in a core in modes with incomplete structure of the working equipment and, as a consequence, - a heat engineering condition of a core. Besides, the task of researching the temperature pulsations accompanying practically all modes of currents for non-isothermal streams is extremely relevant, as these pulsations lead to additional thermocyclic loadings on elements of the equipment and in many cases define its resource. The paper describes the research of mixing processes for non-isothermal water coolant flows in hydraulic model of ship nuclear power plant. In several experiments, attention was paid to the mixing processes when feeding non-isothermal flows through the circulation loops located opposite of each other. To simulate the effect of external dynamic force in the form of periodic effect on the spatial orientation of the model, the ship was tested on a stand "Swinging platform". These vibrations affected the mixing processes occurring within the model. The main impact they had on the transition time, temperature gradient, vertical component of the velocity projection. In the future, these parameters will be clarified and the influence of other factors on the mixing of non-isothermal flows in the ship's nuclear power plant will be studied in more detail.

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