scholarly journals Computational Fluid Dynamics Modeling of Steam Condensation on Nuclear Containment Wall Surfaces Based on Semiempirical Generalized Correlations

2012 ◽  
Vol 2012 ◽  
pp. 1-7 ◽  
Author(s):  
Pavan K. Sharma ◽  
B. Gera ◽  
R. K. Singh ◽  
K. K. Vaze
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Pressure Rise ◽  
Dynamics Modeling ◽  
Steam Condensation ◽  
Noncondensable Gases ◽  
Nuclear Containment ◽  
Design Basis ◽  
Lumped Parameter ◽  
Transport Calculation

In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.

Influence of Sump on Containment Thermal Hydraulics: Synthesis of the TOSQAN Tests

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Emmanuel Porcheron ◽  
Pascal Lemaitre ◽  
Amandine Nuboer
Keyword(s):  
Nuclear Power ◽  
Severe Accident ◽  
Water Evaporation ◽  
Plant Water ◽  
Pressurized Water Reactor ◽  
Steam Condensation ◽  
Pressurized Water ◽  
Noncondensable Gases ◽  
Nuclear Core ◽  
Hydraulic Conditions

During the course of a severe accident in a nuclear power plant, water can be collected in the sump containment through steam condensation on walls, cooling circuit leak, and by spray systems activation. Therefore, the sump can become a place of heat and mass exchanges through water evaporation and steam condensation, which influences the distribution of hydrogen released in containment during nuclear core degradation. The objective of this paper is to present the analysis of semi-analytical experiments on sump interaction between containment atmosphere for typical accidental thermal hydraulic conditions in a pressurized water reactor (PWR). Tests are conducted in the TOSQAN facility developed by the Institut de Radioprotection et de Sûreté Nucléaire in Saclay. The TOSQAN facility is particularly well adapted to characterize the distribution of gases in a containment vessel. A tests’ grid was defined to investigate the coupled effect of the sump evaporation with wall condensation, for air steam conditions, with noncondensable gases (He, SF6), and for steady and transient states (two depressurization tests).

scholarly journals Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

2011 ◽  
Vol 32 (4) ◽  
pp. 67-79
Author(s):  
Tomasz Bury
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Hydrogen Combustion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Lumped Parameter ◽  
Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

A Review of Safety Analysis Philosophies for Nuclear Reactors

2017 ◽  
Vol 3 (3) ◽  
Author(s):  
A. S. Schneider ◽  
N. Yair
Keyword(s):  
Decision Making ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Safety Analysis ◽  
Historical Background ◽  
Natural Approach ◽  
Design Basis ◽  
Design Basis Accident ◽  
Defense In Depth

Various questions can be examined when discussing safety in general. Among these, some key issues are the attitude toward risk and its acceptance, the ways of identifying, analyzing, and quantifying risks, and societal factors and public opinion toward risks. The identification and quantification of risks are central in the regulatory framework and decision making and will be the focus of this article. Various approaches have been used for safety analysis over the years. This paper will survey some of the central attitudes in the nuclear reactor regulation philosophy and discuss the historical background surrounding them. Among these, we mention the “defense-in-depth” approach, the design basis accident (DBA), and beyond design basis accident (BDBA) analyses and discuss the rather subjective nature of their associated decision making. We maintain that it has long been recognized that the natural approach that comes out of the scientific method of inquiry is the probabilistic one, which in today's tools is conducted through the probabilistic safety analysis (PSA) method. This approach unlike the deterministic one, which produced concepts like DBA and defense-in-depth, enables us to put risks into context and to compare different risks such as those posed by different activities in particular or by other industries in general. It has consequently been gaining wide acceptance in regulatory bodies around the world as an effective tool in the inspection and regulation of nuclear reactors. Yet, it is also recognized that despite significant development over the past few decades, PSA still suffers from some well-known deficiencies. Its main benefit at this point is its contribution to identification and prioritization of design features, maintenance, management, and quality assurance (QA) important to safety. PSA has mostly been used in the nuclear power industry, but in recent years it has also started to be incorporated in research reactor (RR) safety analysis, and we therefore cover the subject of PSA usage for this purpose as well.

Surge in a Low-Speed Radial Compressor

1998 ◽  
Author(s):  
Corine Meuleman ◽  
Frank Willems ◽  
Rick de Lange ◽  
Bram de Jager
Keyword(s):  
Flow Rate ◽  
Pressure Rise ◽  
Flow Coefficient ◽  
Lumped Parameter Model ◽  
Pressure Oscillations ◽  
Vaned Diffuser ◽  
Flow Angle ◽  
Radial Compressor ◽  
Low Speed ◽  
Lumped Parameter

Surge is measured in a low-speed radial compressor with a vaned diffuser. For this system, the flow coefficient at surge is determined. This coefficient is a measure for the inducer inlet flow angle and is found to increase with increasing rotational speed. Moreover, the frequency and amplitude of the pressure oscillations during fully-developed surge are compared with results obtained with the Greitzer lumped parameter model. The measured surge frequency increases when the compressor mass flow is throttled to a smaller flow rate. Simulations show that the Greitzer model describes this relation reasonably well except for low rotational speeds. The predicted amplitude of the pressure rise oscillations is approximately two times too small when deep surge is met in the simulations. For classic surge, the agreement is worse. The amplitude is found to depend strongly on the shape of the compressor and throttle characteristic, which are not accurately known.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

Current status of design basis earthquake level for nuclear power plant sites in several countries

2021 ◽  
Author(s):  
Hoseon Choi ◽  
Seung Gyu Hyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
South Korea ◽  
Seismic Activity ◽  
Nuclear Power ◽  
Current Status ◽  
Seismic Safety ◽  
Design Basis ◽  
Design Earthquake ◽  
Design Construction

<p>According to strict criteria step by step for site selection, design, construction and operation, the seismic safety of nuclear power plant (NPP) sites in South Korea are secured by considering design basis earthquake (DBE) level capable of withstanding the maximum ground motions that can occur on the site. Therefore, it is intended to summarize DBE level and its evaluation details for NPP sites in several countries.</p><p>Similar but different terms are used for DBE from country to country, i.e. safe shutdown earthquake (SSE), design earthquake (DE), SL2, Ss, and maximum calculated earthquake (MCE). They may differ when applied to actual seismic design process, and only refer to approximate comparisons. This script used DBE as a representative term, and DBE level was based on horizontal values.</p><p>The DBE level of NPP sites depends on seismic activity of the area. Japan and Western United States, where earthquakes occur more frequently than South Korea, have high DBE values. The DBE level of NPP sites in South Korea has been confirmed to be similar or higher compared to that of Central and Eastern Unites Sates and Europe, which have similar seismic activity.</p>

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

2021 ◽  
Author(s):  
Haiying Chen ◽  
Shaowei Wang ◽  
Xinlu Tian ◽  
Fudong Liu
Keyword(s):  
Radiation Dose ◽  
Effective Dose ◽  
Half Life ◽  
Nuclear Power ◽  
Outer Boundary ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Typical Design ◽  
Total Effective Dose

Abstract The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

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