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.

Study on Flow Structure in a Supersonic Steam Injector

2018 ◽  
Author(s):  
Yuki Kamata ◽  
Masaya Fujishiro ◽  
Akiko Kaneko ◽  
Yutaka Abe
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Structure ◽  
Nuclear Power ◽  
Severe Accident ◽  
Analysis Model ◽  
Operating Characteristics ◽  
Two Phase ◽  
High Heat Transfer ◽  
Discharge Pressure

Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant. Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model. The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.

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.

Analysis of Inertization Strategies for the Filtered Containment Venting System in Cofrentes Nuclear Power Plant

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kevin Fernández-Cosials ◽  
Gonzalo Jiménez ◽  
César Serrano ◽  
Luisa Ibáñez ◽  
Ángel Peinado
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Computational Model ◽  
Nuclear Power ◽  
Inert Atmosphere ◽  
Severe Accident ◽  
Nitrogen Injection

During a severe accident (SA) in a nuclear power plant (NPP), there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system (FCVS) can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a SA, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components, or structures. For this reason, to protect the integrity of the FCVS against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes NPP is presented. It consists of a nitrogen injection located in three different points. A computational model of the FCVS as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the FCVS would be preserved.

Estimation of Radiological Consequences at a Proposed Nuclear Power Plant Site Using Atmospheric Dispersion Code

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Kwame Gyamfi ◽  
Sylvester Attakorah Birikorang ◽  
Emmanuel Ampomah-Amoako ◽  
John Justice Fletcher
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Atmospheric Dispersion ◽  
Atmospheric Stability ◽  
Assessment System ◽  
Severe Accident ◽  
Dispersion Modeling ◽  
Stability Class ◽  
Total Effective Dose Equivalent

Abstract Atmospheric dispersion modeling and radiation dose calculation have been performed for a generic 1000 MW water-water energy reactor (VVER-1000) assuming a hypothetical loss of coolant accident (LOCA). Atmospheric dispersion code, International Radiological Assessment System (InterRAS), was employed to estimate the radiological consequences of a severe accident at a proposed nuclear power plant (NPP) site. The total effective dose equivalent (TEDE) and the ground deposition were calculated for various atmospheric stability classes, A to F, with the site-specific averaged meteorological conditions. From the analysis, 3.7×10−1 Sv was estimated as the maximum TEDE corresponding to a downwind distance of 0.1 km within the dominating atmospheric stability class (class A) of the proposed site. The intervention distance for evacuation (50 mSv) and sheltering (10 mSv) were estimated for different stability classes at different distances. The intervention area for evacuation ended at 0.5 km and that for sheltering at 1.5 km. The results from the study show that designated area for public occupancy will not be affected since the estimated doses were below the annual regulatory limits of 1 mSv.

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

2006 ◽  
Author(s):  
Frank Kretzschmar
Keyword(s):  
Numerical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Material ◽  
Test Facility ◽  
Computer Codes ◽  
Direct Containment Heating ◽  
Heating Up

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant

2018 ◽  
Author(s):  
Wentao Zhu ◽  
Wenjing Li
Keyword(s):  
Decision Making ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Assessment Tool ◽  
Quantitative Risk Assessment ◽  
Severe Accident ◽  
Injection System ◽  
Retention Strategy ◽  
Level 2

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.

Design of Severe Accident Management Systems for Beyond Design Basis External Hazards at Paks NPP

2016 ◽  
Author(s):  
Tamás János Katona ◽  
András Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Management Systems ◽  
Basic Question ◽  
Severe Accidents ◽  
Design Basis ◽  
Accident Management ◽  
External Events

Paks Nuclear Power Plant identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of capacity / margins of existing severe accident management facilities, and construction of some mew systems and facilities. In all cases, the basic question was, what level of margin has to be ensured above design basis external hazard effects, and what level of or hazard has to be taken for the design. Paks Nuclear Power Plant developed certain an applicable in the practice concept for the qualification of already implemented and design the new post-Fukushima measures that is outlined in the paper. The concept and practice is presented on several examples.

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