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1
Content available remote Reaktory prędkie IV generacji chłodzone ciekłymi metalami
100%
Darnowski P.
Spektrum
|
2012
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tom nr 11-12
XXII-XXIX
PL
Omówiono podstawowe zagadnienia związane z projektowaniem reaktorów chłodzonych ciekłymi metalami, które najprawdopodobniej w przyszłości będą trzonem całej technologii reaktorów prędkich. Przedstawiono rozważania na temat technologii, jaka będzie wykorzystana w pierwszym reaktorze IV generacji. Nakreślono niektóre z najważniejszych zagadnień dotyczących reaktorów prędkich powielających chłodzonych ciekłym metalem - LMFBR (Liquid Metal Fast Breeder Reactor).
EN
Discussed are basic problems connected with designing of liquid metal cooled reactors which most probably in the future will be the main core of the whole fast reactors technology. Presented are speculations on technology which can be applied in the first of 4th generation fast reactors. Outlined are some of the most important problems concerning fast breeder reactors cooled by liquid metal - LMFBR (Liquid Metal Fast Breeder Reactor).
2
Content available remote Transmutation: reducing the storage time of spent fuel
63%
Mazgaj P. , Darnowski P.
Journal of Power Technologies
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2014
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tom Vol. 94, nr 5
27--34
EN
Transmutation can reduce the storage time of spent fuel. Efficient transmutation requires a high flux of neutrons and can therefore be done only in nuclear reactors. The article shows the concepts of different solutions of transmutation in nuclear reactors. Knowledge of transmutation is supplemented by information on spent fuel and its radiotoxicity.
3
Content available Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor
63%
Darnowski P. , Uzunow N.
Nukleonika
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2015
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tom Vol. 60, No. 1
171--179
EN
An analysis of the infl uence of addition of minor actinides (MA) to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modifi ed BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-oflife (BOL) state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fl uxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW) decreases with increasing MA inventory in the fuel and confi rms that the addition of MA has a negative effect on the delayed neutron fraction.
4
Content available remote MOX and UOX Fuel Melt Margin for European Pressurized Reactor
51%
Stręciwilk P. O. , Darnowski P. , Dominiak A. , Domański R.
Journal of Power Technologies
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2013
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tom Vol. 93, nr 3
169--177
EN
Safety of Nuclear Power Plants (NPP) is the most important issue during its design and maintenance. Crucial area is nuclear isle where irradiated elements occur. During severe accidents in nuclear reactor core very dangerous is possibility of fuel melt which can lead to release of enormous amounts of radioactive elements. Nowadays Uranium Oxide fuels (UOX) as well as Mixed Oxides fuels (MOX) is under consideration for operating existing and planned NPPs. In this paper prepared Thermal-Hydraulics (TH) model and reliable thermal conductivity of UOX and MOX fuels relations are used for the margin to melt for UOX and MOX fuels calculations. This evaluation is performed for European Pressurized Reactor (EPR) geometry and thermophysical parameters.
5
Content available Coordination number for random distribution of parallel fibres
51%
Darnowski P. , Furmański P. , Domanski R.
Archives of Thermodynamics
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2017
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tom Vol. 38, no. 1
3--26
EN
This paper presents the results of computer simulations carried out to determine coordination numbers for a system of parallel cylindrical fibres distributed at random in a circular matrix according to twodimensional pattern created by random sequential addition scheme. Two different methods to calculate coordination number were utilized and compared. The first method was based on integration of pair distribution function. The second method was the modified sequential analysis. The calculations following from ensemble average approach revealed that these two methods give very close results for the same neighbourhood area irrespective of the wide range of radii used for calculation.
6
Content available remote Analiza skutków powstawania niekondensujących gazów i pary wodnej podczas ciężkiej awarii wrzącego reaktora jądrowego
51%
Darnowski P. , Niewiński G. , Pawluczyk M. , Pilichowska J.
Przemysł Chemiczny
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2017
|
tom T. 96, nr 3
521--524
PL
Przedstawiono wyniki symulacji przebiegu ciężkiej awarii ze stopieniem się rdzenia reaktora jądrowego typu BWR (boiling water reactor). Scenariuszem jest awaria z całkowitą utratą zasilania obiektu jądrowego SBO (station blackout). Przeanalizowano procesy zachodzące w układzie reaktora oraz w obudowie bezpieczeństwa. Szczególną uwagę skupiono na zagadnieniach powstawania gazów niekondensujących (H₂, CO, CO₂, CH₄) oraz pary wodnej, które odgrywają fundamentalną rolę w utrzymaniu szczelności obudowy. Wysunięto wnioski dotyczące przebiegu symulowanej awarii i ryzyka, jakie stwarzają gazy niekondensujące i para wodna.
EN
The effects of a total loss of power with station blackout and subsequent core meltdown for the model of boiling water reactor were simulated by using specialized computer program. The simulation included the processes running inside the reactor core as well as inside cooling systems and containment. The anal. was focused on the mechanisms of non-condensable gas generation (H₂, CO, CO₂, MeH) and steam prodn., which created the highest threat of damage of the reactor containment.
7
Content available Monte Carlo analysis of the battery-type high temperature gas cooled reactor
51%
Grodzki M. , Darnowski P. , Niewiński G.
Archives of Thermodynamics
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2017
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tom Vol. 38, no. 4
209–-227
EN
The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.
8
Content available remote Comparison of simple design of sodium and lead cooled fast reactor cores
45%
Mazgaj P. , Darnowski P. , Gurgacz S. , Lipka M. , Dziubanii K.
Journal of Power Technologies
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2014
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tom Vol. 94, nr 5
16--25
EN
This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid metal cooled fast reactor core, combined with simple neutron population computing for an infinite pin cell lattice. Two types of coolant were studied: liquid sodium and liquid lead, with all requirements regarding safety conditions observed. Temperature distributions along the cooling channel and distributions in the radial direction were prepared, then criticality calculations were performed for MOX fuel using MCNP Monte Carlo code.
9
Content available Simulations of the AP1000-based reactor core with SERPENT computer code
45%
Darnowski P. , Ignaczak P. , Orębski P. , Stępień M. , Niewiński G.
Archive of Mechanical Engineering
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2018
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tom Vol. LXV, nr 3
295--325
EN
The paper presents the core design, model development and results of the neutron transport simulations of the large Pressurized Water Reactor based on the AP1000 design.The SERPENT 2.1.29 Monte Carlo reactor physics computer code with ENDF/BVII and JEFF3.1.1 nuclear data libraries was applied. The full-core 3D models were developed according to the available Design Control Documentation and the literature. Criticality simulations were performed for the core at the Beginning of Life state for Cold Shutdown, Hot Zero Power and Full Power conditions. Selected core parameters were investigated and compared with the design data: effective multiplication factors, boron concentrations, control rod worth, reactivity coefficients and radial power distributions. Acceptable agreement between design data and simulations was obtained, confirming the validity of the model and applied methodology.
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