PL
 |
EN
Nowa wersja platformy, zawierająca wyłącznie zasoby pełnotekstowe, jest już dostępna.
Przejdź na https://bibliotekanauki.pl
Preferencje help
Polish version English version Język
Widoczny [Schowaj] Abstrakt
Liczba wyników

Znaleziono wyników: 8

Liczba wyników na stronie
first rewind previous Strona / 1 next fast forward last
Wyniki wyszukiwania
Wyszukiwano:
w słowach kluczowych:  thermal-hydraulic analysis
help Sortuj według:

help Ogranicz wyniki do:
first rewind previous Strona / 1 next fast forward last
1
Content available Thermal-hydraulic analysis of LTS cables for the DEMO TF coil using simplifi ed models
100%
Lewandowska M. , Dembkowska A.
Nukleonika
|
2017
|
tom Vol. 62, No. 1
23--28
EN
The conceptual design activities for the DEMOnstration reactor (DEMO) – the prototype fusion power plant – are conducted in Europe by the EUROfusion Consortium. In 2015, three design concepts of the DEMO toroidal field (TF) coil were proposed by Swiss Plasma Center (EPFL-SPC, PSI Villigen), Italian National Agency for New Technologies (ENEA Frascati), and Atomic Energy and Alternative Energies Commission (CEA Cadarache). The proposed conductor designs were subjected to complete mechanical, electromagnetic, and thermal-hydraulic analyses. The present study is focused on the thermal-hydraulic analysis of the candidate conductor designs using simplifi ed models. It includes (a) hydraulic analysis, (b) heat removal analysis, and (c) assessment of the maximum temperature and the maximum pressure in each conductor during quench. The performed analysis, aimed at verification whether the proposed design concepts fulfil the established acceptance criteria, provides the information for further improvements of the coil and conductors design.
2
Content available Validation of the heat and mass transfer models within a pressurized water reactor containment using the International Standard Problem No. 47 data
86%
Bury T. , Składzień J. , Fic A.
Archives of Thermodynamics
|
2012
|
tom Vol. 33, no. 2
23-46
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.
3
Content available remote Modeling of liquid film flow in annuli
72%
Anglart H.
Journal of Power Technologies
|
2014
|
tom Vol. 94, nr 5
8--15
EN
One of the challenges in thermal-hydraulic analyses of BWRs is correct prediction of dryout occurrence in fuel assemblies. In practical applications the critical powers in fuel assemblies are found from correlations that are based on experimental data. The drawback of this approach is that correlations are valid only for these fuel assemblies on which the experiments have been conducted. Other restrictive factors are the limited ranges of experimental working conditions including pressure, mass flux and axial power distributions. To overcome the above-mentioned limitations, several different approaches have been proposed to predict the dryout occurrence. One of them is to employ a phenomenological model of annular flow, in which the mass transfer between the liquid film and the gas core is based on entrainment and deposition correlations. Most of these correlations are derived from water-air flows in vertical tubes and their applicability to other geometries in general, and rod-bundles in particular, should be analysed. This paper presents an analysis of the entrainment rate in vertical annuli. Using the standard approach to calculate the entrainment rate, one can demonstrate that the results deviate from measurements. It has been shown that modifying the entrainment correlation based on data obtained in the annulus geometry leads to an essential improvement in the predictive capability of the phenomenological model of annular two-phase flow.
4
Content available remote Coupling of CFD and lumped parameter codes for thermal-hydraulic simulations of reactor containment
72%
Bury T.
Computer Assisted Methods in Engineering and Science
|
2013
|
tom Vol. 20, no. 3
195--206
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.
5
Content available remote Thermal-hydraulic analysis of the improved LTS conductor design concepts for the DEMO TF coil
72%
Lewandowska M. , Sedlak K.
Przegląd Elektrotechniczny
|
2016
|
tom R. 92, nr 4
179--182
EN
The improved design concepts for the LTS TF coil system of DEMO have been proposed in 2013 by EPFL-CRPP PSI Villigen and ENEA Frascati. The present study is focused on the thermal-hydraulic analysis of the conductor designs, which includes: hydraulic analysis, heat removal analysis and assessment of the maximum temperature and the maximum pressure in each conductor during quench.
PL
W 2013 r. zespoły z EPFL-CRPP i ENEA Frascati opracowały udoskonalone koncepcje projektowe kabli dla poszczególnych warstw cewki TF tokamaka DEMO. W pracy przedstawiono analizę cieplno – przepływową projektów kabli obejmującą: analizę hydrauliczną, oszacowanie zdolności usuwania ciepła oraz oszacowanie maksymalnej temperatury oraz maksymalnego ciśnienia podczas utraty stanu nadprzewodzenia.
6
Content available remote Efficiency of selected methods of hydrogen removal from a nuclear reactor's containment building
72%
Bury T.
Journal of Power Technologies
|
2014
|
tom Vol. 94, nr 5
96--104
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.
7
Content available remote Conceptual design and analysis of a cryogenic system for a new test facility for high temperature superconductor current leads (HTS CLs)
72%
Lewandowska M. , Bagnasco M.
Przegląd Elektrotechniczny
|
2012
|
tom R. 88, nr 7a
140-143
EN
In view of the R&D program on HTS CL at EPFL-CRPP PSI Villigen it is foreseen to extend the capabilities of an existing facility (JORDI) in order to be able to perform tests of HTS CL in different conditions. The present work is focused on the conceptual design of the cryogenics needed for the facility upgrade considering the requirement of an as extended as possible range of operation. Two cooling options for the copper part of a CL are considered. The use of heat exchangers for the optimization of the cryogenic system is discussed and their size is assessed.
PL
W związku z trwającymi w EPFL-CRPP PSI Villigen pracami B&R nad nadprzewodnikowymi krioprzepustami prądowymi (HTS CLs) przewiduje się uruchomienie tam stanowiska eksperymentalnego do prowadzenia testów HTS CLs w różnych warunkach. W pracy zaprezentowano projekt koncepcyjny układu kriogenicznego dostarczającego chłodziwo do tego stanowiska. Rozważono możliwość dwóch opcji chłodzenia części miedzianej HTS CLs. Przedyskutowano konieczność zastosowania wymienników ciepła w układzie kriogenicznym i oszacowano ich rozmiar.
8
Content available remote Computer codes in the safety analysis for nuclear power plants. Computational capabilities of thermal-hydraulic tools, using the example of the RELAP5 code
58%
Skrzypek E. , Skrzypek M.
Journal of Power Technologies
|
2014
|
tom Vol. 94, nr 5
41--50
EN
Safety is a paramount concern of the Nuclear Power Program in Poland. To this end there is a need to investigate the design of the proposed reactor and its operation principles and perform multiple analyses both before the reactor start-up (The Pre-Construction Safety Report (PCSR) and during its operational life. In the worldwide nuclear community hundreds of people are involved in this complicated and complex process. Due to the sophistication of the phenomena occurring during operation and accidents, the number of analyses is increasing rapidly. Currently, much interest in this field is focused on the use of computer codes and high computational power.
first rewind previous Strona / 1 next fast forward last
JavaScript jest wyłączony w Twojej przeglądarce internetowej. Włącz go, a następnie odśwież stronę, aby móc w pełni z niej korzystać.