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PL
Zwrócono uwagę na zagrożenie zniszczenia betonowych obiektów elektrowni jądrowych wskutek odwapnienia faz zaczynu cementowego. Badania dekalcyfikacji w środowisku jonów chlorkowych prowadzono w warunkach laboratoryjnych.
EN
The attention on the risk of destruction of concrete objects in nuclear power plants due to decalcification of hardened cement paste phases. Study on decalcification process in water solution of chloride ions was conducted under laboratory conditions.
EN
Distributions of media streams flowing in a cross-flow tube and fin heat exchanger are usually non-uniform. This could be an effect of the heat exchanger construction, its installation method, design of a flowing channel or all those factors combined. The problem of the non-uniform media flow in heat exchangers of different types is not new, and it has been investigated by many researchers. Early results were sometimes ambiguous. More recent outcomes indicate that the effect of the non-uniform inflow of heat carriers to the heat exchanger could be significant it may adversely affect the device’s efficiency to a large extent. Investigations of tube and fin cross-flow heat exchangers carried out for almost twenty years at the Institute of Thermal Technology of the Silesian University of Technology, by way of experiments and numerical simulations, also confirm these latest conclusions. The reduction in overall heat exchanger capacity, comparing to the uniform inflow of media, may reach up to 18%. This work presents results of experimental and computational investigations of tube, fin, cross-flow, double row heat exchangers air-water. The heat exchangers under consideration are built in the form of two rows of elliptic tubes with rectangular fins. The ribbing structure of the first heat exchanger is uniform. This device was investigated primarily in order to determine its efficiency but also the range and the form of non-uniform inflow of air. The air flow distribution was tested on a special test station during a series of measurements. The results of the analysis of this heat exchanger were used to design a second heat exchanger with a non-uniform structure of fins on individual tubes. It was assumed that by changing the heat transfer surface (thickening the fins) in the region of high air speed, the efficiency of modified heat exchangers could be enhanced. Testing this hypothesis is the main aim of this work. The experimental results generally confirm the hypothesis, showing a rise in efficiency of up to 8%. However, it should be noted that the design of the modified ribbing structure is not optimal and changing this structure impacts the hydraulic resistance and distribution of air mass flow rate at the heat exchanger inflow. This effect should be considered when evaluating the results.
EN
This paper presents the results of thermodynamic analyses of a system using a horizontal ground heat exchanger to cool a residential building in summer and heat it in the autumn-winter period. The main heating device is a vapour compression heat pump with the ground as the lower heat source. The aim of the analyses is to examine the impact of heat supply to the ground in the summer period, when the building is cooled, on the operation of the heating system equipped with a heat pump in the next heating season, including electricity consumption. The processes occurring in cooling and heating systems have an unsteady nature. The main results of the calculations are among others the time-dependent values of heat fluxes extracted from or transferred to the ground heat exchanger, the fluxes of heat generated by the heat pump and supplied to the heated building by an additional heat source, the parameters in characteristic points of the systems, the temperature distributions in the ground and the driving electricity consumption in the period under analysis. The paper presents results of analysis of cumulative primary energy consumption of the analyzed systems and cumulative emissions of harmful substances.
EN
A cross-flow, tube and fin heat exchanger of the water – air type is the subject of the analysis. The analysis had experimental and computational form and was aimed for evaluation of radiative heat transfer impact on the heat exchanger performance. The main element of the test facility was an enlarged recurrent segment of the heat exchanger under consideration. The main results of measurements are heat transfer rates, as well as temperature distributions on the surface of the first fin obtained by using the infrared camera. The experimental results have been next compared to computational ones coming from a numerical model of the test station. The model has been elaborated using computational fluid dynamics software. The computations have been accomplished for two cases: without radiative heat transfer and taking this phenomenon into account. Evaluation of the radiative heat transfer impact in considered system has been done by comparing all the received results.
EN
The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal- -hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational effiiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.
6
Content available CFD modeling of passive autocatalytic recombiners
EN
This study deals with numerical modeling of passive autocatalytic hydrogen recombiners (PARs). Such devices are installed within containments of many nuclear reactors in order to remove hydrogen and convert it to steam. The main purpose of this work is to develop a numerical model of passive autocatalytic recombiner (PAR) using the commercial computational fluid dynamics (CFD) software ANSYS-FLUENT and tuning the model using experimental results. The REKO 3 experiment was used for this purpose. Experiment was made in the Institute for Safety Research and Reactor Technology in Julich (Germany). It has been performed for different hydrogen concentrations, different flow rates, the presence of steam, and different initial temperatures of the inlet mixture. The model of this experimental recombiner was elaborated within the framework of this work. The influence of mesh, gas thermal conductivity coefficient, mass diffusivity coefficients, and turbulence model was investigated. The best results with a good agreement with REKO 3 data were received for k-ε model of turbulence, gas thermal conductivity dependent on the temperature and mass diffusivity coefficients taken from CHEMKIN program. The validated model of the PAR was next implemented into simple two-dimensional simulations of hydrogen behavior within a subcompartment of a containment building.
EN
Large amounts of gaseous hydrogen may be released into the containment building during a severe accident in a water cooled nuclear reactor. The main methods of hydrogen removal from the containment are described in brief in this paper. HEPCAL - an in-house lumped parameter computer code - was used for simulation purposes and the results were used to evaluate the efficiency of various hydrogen removal systems.
8
Content available remote Numerical model of a cross-flow heat exchanger with non-uniform flow of media
EN
This paper presents thermal-hydraulic analyses of finned cross flow heat exchangers working in media flow maldistribution conditions. The authors postulate a possibility of performing such analyses through the use of CFD models of recurrent segments of the heat exchangers. Media inflow to each recurrent segment may be individually defined and thus the flow maldistribution in the whole heat exchanger could be considered. The methodology of creating these models, running calculations and results of very initial experimental validation is presented in the paper.
EN
The work deals with thermal-hydraulic analyses of a pressurized water reactor containment response to accidents caused by a rupture of primary circuit. The in-house system computer code HEPCAL-AD and CFD ANSYS Fluent have been coupled for these simulations. The aim of this work is verification of possible ways of the codes coupling. The assessment of each method has been done by comparing the computational results with experimental data obtained from testing rigs of the AP-600 reactor containment cooling system. Additional simulations of a loss-of-coolant accident (LOCA) have been carried out as well, and compared with outcomes of the AP-600 reactor simulator.
EN
Passive autocatalytic recombiners (PAR) is the only used method for hydrogen removal from the containment buildings in modern nuclear reactors. Numerical models of such devices, based on the CFD approach, are the subject of this paper. The models may be coupled with two types of computer codes: the lumped parameter codes, and the computational fluid dynamics codes. This work deals with 2D numerical model of PAR and its validation. Gaseous hydrogen may be generated in water nuclear reactor systems in a course of a severe accident with core overheating. Therefore, a risk of its uncontrolled combustion appears which may be destructive to the containment structure.
EN
Assuming that a maldistribution of media flow may have significant meaning for a cross-flow heat exchanger performance it is important to determine the form and scope of this non-uniformity. Experimental investigations of this problem are conducted for over ten years at the Institute of Thermal Technology of the Silesian University of Technology by using a computer controlled thermoanemometric probe. This is a very invasive technique that applies properly only in laboratory conditions. As the infra-red thermography was applied at some stage of realized studies, the authors assume a possibility of applying this technique for evaluation of media flow maldistribution, qualitative assessment at least. The experiments performed for the case of a typical fin and tube water cooler have been described in this paper, as well as the most important conclusions.
EN
A lumped parameter type code, called HEPCAL, has been worked out in the Institute of Thermal Technology of the Silesian University of Technology for simulations of a pressurized water reactor containment transient response to a loss-of-coolant accident. The HEPCAL code has been already verified and validated against available experimental data, which in fact have been taken from separate effect tests mainly. This work is devoted to validation of the latest version of the HEPCAL code against experimental data from more complex tests. These experiments have been performed on three different test rigs (called TOSQAN,MISTRA and ThAI) and a part of them became the basis of the International Standard Problem No. 47 (ISP-47) dedicated to containment thermal-hydraulics. Selected experiments realized within the framework of the ISP-47 project have been simulated using the HEPCAL-AD code. The obtained results allowed for drawing of some important conclusions concerning heat and mass transfer models (especially steam condensation), two-phase flow model and buoyancy effects.
EN
The paper deals with numerical thermodynamic analyses of cross-flow finned tube heat exchangers of the gas-liquid type. The authors postulate that some improvement of the heat exchanger performance may be achieved by applying a special ribbing structure - fitted to certain media flow conditions. First, the measurements have been carried out in order to determine the air inflow non-uniformity and next an own computer code HEWES has been used for numerical simulations in order to evaluate the impact of the measured non-uniformity on the exchangers' efficiency for the heat exchangers with unified ribbing structure. The numerical simulations have been repeated for heat exchangers with special ribbing structures - determined on the basis of experimental results. Results confirm the hypothesis - some increase in the total heat flow rate may be observed for the considered heat exchanger.
PL
Praca dotyczy termodynamicznej analizy krzyżowoprądowego, ożebrowanego wymiennika ciepła typu gaz-ciecz. Autorzy weryfikowali hipotezę zakładającą możliwość uzyskania poprawy efektywności działania urządzenia w wyniku zastosowania specjalnej, dostosowanej do warunków przepływu czynników, struktury ożebrowania. W pierwszej kolejności wykonane zostały pomiary określające zakres i postać niejednorodnego dopływu powietrza do analizowanego wymiennika ciepła, a następnie wykorzystano własny kod komputerowy HEWES do oceny wpływu tej niejednorodności na wydajność cieplną wymiennika. Obliczenia te dotyczyły wymiennika ciepła z jednorodnym ożebrowaniem. Symulacje numeryczne zostały następnie powtórzone dla przypadku wymiennika ze specjalną strukturą ożebrowania, która została dobrana dla wyznaczonych eksperymentalnie warunków dopływu powietrza. Uzyskane rezultaty potwierdzają postawioną na wstępie hipotezę - widoczny jest pewien wzrost wartości strumienia ciepła przekazywanego w wymienniku.
EN
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.
EN
The nuclear power share in the world's electricity production is about 16-17%. There are almost 440 nuclear reactors operating today and over 60 being constructed in the world. Most of them are pressurized water reactors. Two trends in safety systems development may be observed: an evolutionary approach and a revolutionary approach. The paper deals with the evaluation of these trends based on the results of simulations of loss-of-coolant accidents for two selected designs of the third generation pressurized water reactors: EPR and AP-1000.
PL
Udział energetyki jądrowej w światowej produkcji energii elektrycznej wynosi obecnie 16-17% Na świecie pracuje prawie 440 energetycznych reaktorów jądrowych, a ponad 60 jest budowanych. Większość z nich to reaktory wodne ciśnieniowe. Aktualnie można zaobserwować dwa trendy w rozwoju systemów bezpieczeństwa elektrowni jądrowych: podejście ewolucyjne oraz podejście rewolucyjne. W pracy podjęto próbę oceny tych dwóch trendów w oparciu o wyniki symulacji awarii rozszczelnieniowych dla dwóch wybranych rozwiązań konstrukcyjnych reaktorów wodnych ciśnieniowych trzeciej generacji: reaktora EPR oraz AP-1000.
EN
The work deals with experimental and numerical thermodynamic analyses of cross-flow finned tube heat exchangers of the gas-liquid type. The aim of the work is to determine an impact of the gas non-uniform inlet on the heat exchangers performance. The measurements have been carried out on a special testing rig and own numerical code has been used for numerical simulations. Analysis of the experimental and numerical results has shown that the range of the non-uniform air inlet to the considered heat exchangers may be significant and it can significantly affect the heat exchanger efficiency.
17
Content available remote Model numeryczny i analiza cieplno-przepływowa eksperymentu CASP-3
PL
W pracy przedstawione zostały wyniki analizy cieplno-przepływowej eksperymentu fizycznego, którego celem było odtworzenie warunków przepływu ciepła w obudowie bezpieczeństwa wodnego ciśnieniowego reaktora jądrowego po rozszczelnieniu pierwotnego obiegu chłodzenia. Do symulacji wykorzystano komercyjny pakiet CFD Fluent oraz dwuwymiarowe modele rozważanego obiektu.
EN
The paper presents results of thermal-hydraulic analysis of a physical experiment aimed in reconstruction of heat transfer conditions within containment of a pressurized water reactor after a rupture of the primary cooling circuit. The commercial CFD package Fluent has been used for simulations of the experiment. The simulations have been realized for two-dimensional models of the object under consideration.
EN
The paper deals with experimental and numerical thermodynamic analysis of a cross-flow heat exchanger with non-uniform flow of mediums through the exchanger. The main aim of the work is validation of a computer code for analyses of the considered type exchangers. The validation procedure was realized by means of comparison of the experimental and numerical results. Next, the sensitivity analysis of the code was also accomplished. Selected results of analyses carried out and the most important conclusions are presented in the paper.
EN
A loss-of-coolant accident (LOCA) is one of the most serious accident which may happen in the nuclear factor cooled and moderated by water under a high pressure. A threat of the core uncover during such accident causes that LOCA became a design basis accident (DBA) and have to be simulated to prove that engineered safety systems are able to manage the potential consequences of such accident. It is obvious that LOCA can not be investigated by means of full-scale physical experiments. Thus the mathematical modeling and numerical simulations are widely used for analyses of LOCA. Two types of computer codes are applied :or these purposes at the moment: a one-dimensional system codes (also referred as the lumped parameter codes) and a three-dimensional field codes (mostly based on CFD). The work presents an initial results of LOCA analyses performed by means of coupling a domestic system code called HEPCAL and a commercial CFD program FLUENT. The simulations have been realized for design data of advanced pressurized water reactors - AP-600 and EPR.
PL
Awaria rozszczelnieniowa pierwotnego obiegu chłodzenia połączona z wyciekiem chłodziwa jest jedną najpoważniejszych awarii, które mogą się wydarzyć w układach reaktorów jądrowych chłodzonych i moderowanych wodą pod ciśnieniem. Z oczywistych względów przebieg awarii rozszczelnieniowych nie może być badany na drodze eksperymentalnej w pełnej skali. Z tego powodu modelowanie matematyczne i symulacje numeryczne są powszechnie stosowaną metodą badań. Stosowane są obecnie do tych celów dwie grupy kodów komputerowych: jednowymiarowe kody systemowe (określane również jako kody o parametrach skupionych) oraz kody przestrzenne (modele trójwymiarowe, oparte zazwyczaj o numeryczną mechanikę płynów). W pracy przedstawiono wstępne rezultaty termodynamicznych analiz awarii rozszczelnieniowych zrealizowanych za pomocą sprzężenia własnego kody systemowego HEPCAL oraz komercyjnego pakietu CFD - programu FLUENT. Symulacje przeprowadzono dla układów zaawansowanych reaktorów wodny ciśnieniowych - reaktora AP-600 oraz reaktora EPR.
20
Content available remote Postęp w bezpieczeństwie jądrowym od czasów Czarnobyla
PL
Względy bezpieczeństwa stanowią nadrzędne kryterium warunkujące rozwój techniki reaktorowej. Ocena bezpieczeństwa elektrowni jądrowych jest procesem złożonym i prowadzi się ją zarówno na etapie projektowania jak i eksploatacji obiektu. Opisano filozofię bezpieczeństwa elektrowni jądrowych. Przedstawiono rozwój systemów bezpieczeństwa reaktorów wodnych. Wymieniono zagadnienia bezpieczeństwa w blokach z reaktorami jądrowymi.
EN
Safety reasons are the primary criterion conditioning reactor engineering development. Evaluation of NPPs safety is a complex process and it is carried through all the time both on object designing and exploitation stages. Described is safety philosophy of NPPs. Presented is water reactor safety systems development. Mentioned are safety problems concerning nuclear reactor units.
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