Publications
Nuclear energy related journal papers and conference papers affiliated by IHE/ITC
Papers published after 2010
(1)
Mazgaj, P. E.; Darnowski, P.; Trewin, R.; Kral, P.; Puustinen, M. Impact of Selected Long-Term Operation Improvements Relevant to the Pressurized Thermal Shock in PWR. In HND2022 Conference Proceedings; Zadar, Croatia, 2022; p 9.
(2)
Mazgaj, P.; Darnowski, P.; Kaszko, A.; Hortal, J.; Dusic, M.; Mendizábal, R.; Pelayo, F. Demonstration of the E-BEPU Methodology for SL-LOCA in a Gen-III PWR Reactor. Reliability Engineering & System Safety 2022, 226, 108707. https://doi.org/10.1016/j.ress.2022.108707.
(3)
Kubiński, W.; Darnowski, P.; Palmi, K. Prediction of Nuclear Reactor Core Parameters Using Artificial Neural Network. In PHYSOR2022 Proceedings; ANS: Pittsburgh, PA, 2022; p 8.
(4)
Darnowski, P.; Mazgaj, P. MelSUA – An Open-Source MATLAB Toolbox for Sensitivity and Uncertainty Analysis with MELCOR Code. In HND2022 Conference Proceedings; Zadar, Croatia, 2022; p 10.
(5)
Darnowski, P.; Kubiński, W. Development of the MATLAB Tool for Optimization of the Nuclear Hybrid Energy System Dedicated for Hydrogen Production. In NENE2022 Conference Proceedings; Portoroz, Slovenia, 2022; p 8.
(6)
Żurkowski, W.; Sawicki, P.; Kubiński, W.; Darnowski, P. Application of Genetic Algorithms in Optimization of SFR Nuclear Reactor Design. Nukleonika 2021, 66 (4), 139–145. https://doi.org/10.2478/nuka-2021-0021.
(7)
Włostowski, M.; Domitr, P.; Darnowski, P. A Sensitivity Study of Critical Flow Modeling with MELCOR 2.2 Code Based on the Marviken CFT-21 Experiment. Energies 2021, 14 (16), 4985. https://doi.org/10.3390/en14164985.
(8)
Spirzewski, M.; Domitr, P.; Darnowski, P. Global Uncertainty and Sensitivity Analysis of MELCOR and TRACE Critical Flow Models against MARVIKEN Tests. Nuclear Engineering and Design 2021, 378, 111150. https://doi.org/10.1016/j.nucengdes.2021.111150.
(9)
Milewski, J.; Kupecki, J.; Szczęśniak, A.; Uzunow, N. Hydrogen Production in Solid Oxide Electrolyzers Coupled with Nuclear Reactors. International Journal of Hydrogen Energy 2021, 46 (72), 35765–35776. https://doi.org/10.1016/j.ijhydene.2020.11.217.
(10)
Lo Frano, R.; Cancemi, S. A.; Darnowski, P.; Ciolini, R.; Paci, S. Preliminary Analysis of an Aged RPV Subjected to Station Blackout. Energies 2021, 14 (15), 4394. https://doi.org/10.3390/en14154394.
(11)
Laskowski, R.; Smyk, A.; Uzunow, N. “Influence of Cooling Water Temperature on Performance of EPR Nuclear Power Plant.” Rynek Energii 2021, 152 (1).
(12)
Kubiński, W.; Darnowski, P.; Chęć, K. The Development of a Novel Adaptive Genetic Algorithm for the Optimization of Fuel Cycle Length. Annals of Nuclear Energy 2021, 155, 108153. https://doi.org/10.1016/j.anucene.2021.108153.
(13)
Kubiński, W.; Darnowski, P.; Chęć, K. Optimization of the Loading Pattern of the PWR Core Using Genetic Algorithms and Multi-Purpose Fitness Function. Nukleonika 2021, 66 (4), 147–151. https://doi.org/10.2478/nuka-2021-0022.
(14)
Kubiński, W.; Bojarski, P.; Darnowski, P. Application of Artificial Neural Network and Particle Swarm Optimization in Determining Selected Parameters of the Nuclear Reactor Core. In Proceedings of the European Nuclear Young Generation Forum ENYGF’21; September 27-30, 2021, Tarragona, Spain, 2021; p 5.
(15)
Domitr, P.; Włostowski, M. The Use of Machine Learning for Inverse Uncertainty Quantification in TRACE Code Based on Marviken Experiment. Nuclear Engineering and Design 2021, 384, 111498. https://doi.org/10.1016/j.nucengdes.2021.111498.
(16)
Stępień, M.; Niewiński, G.; Kaszko, A. Overview of the Safety Systems Used in Generation III and III+ of Reactors. Modern Engineering 2020, No. 3, 92–99.
(17)
Raaj Saasthaa Arumuga Kumar, E.; Pancholi, M. K.; Darnowski, P.; Dzido, A. Neutronic Performance of a Thorium Based Mixed Oxide Fuel in a Burner Sodium-Cooled Fast Reactor. Energy 2020, 212, 118744. https://doi.org/10.1016/j.energy.2020.118744.
(18)
Muszynski, D.; Andrzejewski, K. Modelling the Tapered Geometry of Real 3D Unit Cell of Polish MARIA Reactor in a Rectangular Prism Diffusion Code. Annals of Nuclear Energy 2020, 137, 107101. https://doi.org/10.1016/j.anucene.2019.107101.
(19)
Mazgaj, P.; Darnowski, P.; Niewiński, G. Analiza niepewności i wrażliwości dla wytwarzania się wodoru podczas eksperymentu PHEBUS FPT-1. In Współczesne zagadnienia TERMODYNAMIKI – XXIV Zjazd Termodynamików; Białobrzegi, 2020; p 8.
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Mazgaj, P.; Darnowski, P.; Kaszko, A.; Hortal, J.; Dusic, M.; Mendizábal, R.; Pelayo, F. Demonstration of the E-BEPU Methodology for LB-LOCA in NPP with PWR Reactor. In NENE2020 Conference Proceedings; 2020; p 8.
(21)
Lipka, M.; Rajewski, A. Regress in Nuclear District Heating. The Need for Rethinking Cogeneration. Progress in Nuclear Energy 2020, 130, 103518. https://doi.org/10.1016/j.pnucene.2020.103518.
(22)
Laskowski, R.; Smyk, A.; Uzunow, N. Wpływ Temperatury Wody Chłodzącej Na Osiągi Elektrowni Jądrowej z Reaktorem EPR. In Współczesne zagadnienia TERMODYNAMIKI – XXIV Zjazd Termodynamików; null: null, 2020.
(23)
Kubiński, W.; Darnowski, P.; Chęć, K. Optimization of BEAVRS PWR Loading Pattern Using a Novel Genetic Algorithm Based on Population Variance Control. In NENE2020 Conference Proceedings; 2020; p 8.
(24)
Kaszko, A.; Niewiński, G.; Stępień, M. Probabilistic Safety Assessment of ESBWR Gravity Driven Cooling System. Modern Engineering 2020, No. 1, 18–25.
(25)
Kaszko, A.; Kowal, K.; Potempski, S. Quantification of Initiating Events Probability Based on Fragility Functions and Bayesian Network Applied for Multi-Hazard. In EGU General Assembly 2020; Copernicus Meetings, 2020. https://doi.org/10.5194/egusphere-egu2020-21900.
(26)
Fan, W.; Li, H.; Anglart, H. A Study of Rewetting and Conjugate Heat Transfer Influence on Dryout and Post-Dryout Phenomena with a Multi-Domain Coupled CFD Approach. International Journal of Heat and Mass Transfer 2020, 163, 120503. https://doi.org/10.1016/j.ijheatmasstransfer.2020.120503.
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Fan, W.; Cherdantsev, A. V.; Anglart, H. Experimental and Numerical Study of Formation and Development of Disturbance Waves in Annular Gas-Liquid Flow. Energy 2020, 207, 118309. https://doi.org/10.1016/j.energy.2020.118309.
(28)
Fan, W.; Anglart, H. VarRhoTurbVOF: A New Set of Volume of Fluid Solvers for Turbulent Isothermal Multiphase Flows in OpenFOAM. Computer Physics Communications 2020, 247, 106876. https://doi.org/10.1016/j.cpc.2019.106876.
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Fan, W.; Anglart, H. VarRhoTurbVOF 2: Modified OpenFOAM Volume of Fluid Solvers with Advanced Turbulence Modeling Capability. Computer Physics Communications 2020, 256, 107467. https://doi.org/10.1016/j.cpc.2020.107467.
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Darnowski, P.; Włostowski, M.; Stępień, M.; Niewiński, G. Study of the Material Release during PHÉBUS FPT-1 Bundle Phase with MELCOR 2.2. Annals of Nuclear Energy 2020, 148, 107700. https://doi.org/10.1016/j.anucene.2020.107700.
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Darnowski, P.; Mazgaj, P.; Bašić, I.; Vrbanić, I.; Skrzypek, M.; Malesa, J.; Silde, A.; Hiittenkivi, J.; Štrubelj, L. Severe Accident Simulations Dedicated to the SAMG Decision- Making Tool Demonstration. In NENE2020 Conference Proceedings; 2020; p 8.
(32)
Cancemi, S. A.; Frano, R. L.; Ciolini, R.; Darnowski, P. Preliminary Analysis of Creep and Ageing Influence During SBO Accident. In NENE2020 Conference Proceedings; 2020; p 10.
(33)
Bryk, R.; Schollenberger, S. P.; Dennhardt, L.; Świrski, K. Parameter Study at the PKL Test Facility on Heat Transfer Mechanisms in the Steam Generator in Presence of Nitrogen, Steam and Liquid. Energy 2020, 211, 118612. https://doi.org/10.1016/j.energy.2020.118612.
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Bergagio, M.; Fan, W.; Thiele, R.; Anglart, H. Large Eddy Simulation of Thermal Mixing with Conjugate Heat Transfer at BWR Operating Conditions. Nuclear Engineering and Design 2020, 356, 110361. https://doi.org/10.1016/j.nucengdes.2019.110361.
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Spirzewski, M.; Anglart, H.; Stano, P. M. Uncertainty and Sensitivity Analysis of a Phenomenological Dryout Model Implemented in DARIA System Code. Nuclear Engineering and Design 2019, 355, 110281. https://doi.org/10.1016/j.nucengdes.2019.110281.
(36)
Raaj Saasthaa Arumuga Kumar, E.; Darnowski, P.; Kiritbhai Pancholi, M.; Dzido, A. Thorium Application in the Medium-Sized Sodium-Cooled Fast Reactor. E3S Web Conf. 2019, 137, 01030. https://doi.org/10.1051/e3sconf/201913701030.
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Niewiński, G.; Stępień, M. Energetyka jądrowa. Bezpieczna technologia czy zagrożenie dla ludzkości? Nierówności Społeczne a Wzrost Gospodarczy 2019, 57 (1). https://doi.org/10.15584/nsawg.2019.1.20.
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Malik, K.; Żbikowski, M.; Teodorczyk, A. Detonation Cell Size Model Based on Deep Neural Network for Hydrogen, Methane and Propane Mixtures with Air and Oxygen. Nuclear Engineering and Technology 2019, 51 (2), 424–431. https://doi.org/10.1016/j.net.2018.11.004.
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Kopka, P.; Potempski, S.; Kaszko, A.; Korycki, M. Urban Dispersion Modelling Capabilities Related to the UDINEE Intensive Operating Period 4. Boundary-Layer Meteorol 2019, 171 (3), 465–489. https://doi.org/10.1007/s10546-018-0399-6.
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Fan, W.; Li, H.; Anglart, H. Numerical Investigation of Spatial and Temporal Structure of Annular Flow with Disturbance Waves. International Journal of Multiphase Flow 2019, 110, 256–272. https://doi.org/10.1016/j.ijmultiphaseflow.2018.10.003.
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Domitr, P.; Darnowski, P.; Spirzewski, M. The Assessment of Critical Flow Models of MELCOR2.2 and TRACE V5.0 Patch 5 Against Marviken Critical Flow Tests. In NURETH-18 Conference Proceedings; ANS: Portland, OR, 2019.
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Darnowski, P.; Włostowski, M. Zintegrowane Analizy Awarii Ciężkich Na Przykładzie Eksperymentu Phebus FPT-1 z Wykorzystaniem Kodu Obliczeniowego MELCOR 2.2 – Część 2: Symulacje. Bezpieczeństwo Jądrowe i Ochrona Radiologiczna 2019, 2019 (3/4), 12–21.
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Darnowski, P.; Włostowski, M. Zintegrowane Analizy Awarii Ciężkich Na Przykładzie Eksperymentu Phebus FPT-1 z Wykorzystaniem Kodu Obliczeniowego MELCOR 2.2 – Część 1: Opis Instalacji, Modelu i Kwalifikacji …. Bezpieczeństwo Jądrowe i Ochrona Radiologiczna 2019, 2019 (1/2), 26–35.
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Darnowski, P.; Stępień, M.; Włostowski, M.; Świrski, K. Zachowanie Substancji Promieniotwórczych w Obiegu Pierwotnym Reaktora Jądrowego Typu PWR Podczas Ciężkiej Awarii. Bezpieczeństwo Jądrowe i Ochrona Radiologiczna 2019, 2019 (3/4), 22–32.
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Darnowski, P.; Pawluczyk, M. Analysis of the BEAVRS PWR Benchmark Using SCALE and PARCS. Nukleonika 2019, 64 (3), 87–96. https://doi.org/10.2478/nuka-2019-0011.
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Darnowski, P.; Mazgaj, P.; Skrzypek, E. MELCOR Simulations of the SBO in Gen III PWR with EVMR. In NENE2019 Conference Proceedings; 2019; p 8.
(47)
Bryk, R.; Schmidt, H.; Mull, T. Modeling of Emergency Condenser System Response to Loss of Coolant Accident in a BWR III+ Generation. Ekspolatacja i Niezawodnosc – Maintenance and Reliability 2019, 21, 468–475. https://doi.org/10.17531/ein.2019.3.13.
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Bryk, R.; Mull, T.; Schmidt, H. Experimental Investigation of LWR Passive Safety Systems Performance at the INKA Test Facility. E3S Web Conf. 2019, 137, 01035. https://doi.org/10.1051/e3sconf/201913701035.
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Bryk, R.; Dennhardt, L.; Schollenberger, S. Experimental Investigation of PWR Accident Scenarios at the PKL Test Facility. E3S Web Conf. 2019, 137, 01016. https://doi.org/10.1051/e3sconf/201913701016.
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Stępień, M.; Gurgacz, S.; Niewiński, G. Tor a bezpieczeństwo energetyczne. Reaktory torowe. Rynek Energii 2018, 137 (4).
(51)
Spirzewski, M.; Anglart, H. An Improved Phenomenological Model of Annular Two-Phase Flow with High-Accuracy Dryout Prediction Capability. Nuclear Engineering and Design 2018, 331, 176–185. https://doi.org/10.1016/j.nucengdes.2018.02.032.
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Darnowski, P.; Ignaczak, P.; Obrębski, P.; Stępień, M.; Niewiński, G. Archive of Mechanical Engineering. Archive of Mechanical Engineering 2018. https://doi.org/10.24425/124484.
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Bryk, R.; Schmidt, H.; Mull, T.; Ganzmann, I.; Herbst, O. Modeling of the Water Level Swell during Depressurization of the Reactor Pressure Vessel of the Boiling Water Reactor in Accidental Conditions. EiN 2018, 21 (1), 28–36. https://doi.org/10.17531/ein.2019.1.4.
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Bryk, R.; Schmidt, H.; Mull, T.; Herbst, O.; Ganzmann, I. A Model of Water Thermal-Hydraulics during Depressurization of a Vessel Filled with Water under Saturation Conditions. In 2018 International Interdisciplinary PhD Workshop (IIPhDW); 2018; pp 8–12. https://doi.org/10.1109/IIPHDW.2018.8388234.
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Bergagio, M.; Li, H.; Anglart, H. An Iterative Finite-Element Algorithm for Solving Two-Dimensional Nonlinear Inverse Heat Conduction Problems. International Journal of Heat and Mass Transfer 2018, 126, 281–292. https://doi.org/10.1016/j.ijheatmasstransfer.2018.04.104.
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Anglart, H.; Li, H.; Niewinski, G. Mechanistic Modelling of Dryout and Post-Dryout Heat Transfer. Energy 2018, 161, 352–360. https://doi.org/10.1016/j.energy.2018.07.011.
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Skrzypek, E.; Skrzypek, M.; Saas, L.; LeTellier, R. In Vessel Corium Propagation Sensitivity Study of Reactor Pressure Vessel Rupture Time with PROCOR Platform. Journal of Power Technologies 2017, 97 (2), 110–116.
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Niewiński, G.; Stępień, M.; Góral, K. Analysis of AP1000 Radioactive Material Release Accidents with MELCOR Accident Consequence Code System (MACCS). Journal of Power Technologies 2017, 97 (5), 446–454.
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Le Tellier, R.; Skrzypek, E.; Saas, L. On the Treatment of Plane Fusion Front in Lumped Parameter Thermal Models with Convection. Applied Thermal Engineering 2017, 120, 314–326. https://doi.org/10.1016/j.applthermaleng.2017.03.108.
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Grodzki, M.; Darnowski, P.; Niewiński, G. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor. Archives of Thermodynamics 2017, 38 (4), 209–227. https://doi.org/10.1515/aoter-2017-0032.
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Darnowski, P.; Potapczyk, K.; Świrski, K. Investigation of the Recriticality Potential during Reflooding Phase of Fukushima Daiichi Unit-3 Accident. Annals of Nuclear Energy 2017, 99, 495–509. https://doi.org/10.1016/j.anucene.2016.10.004.
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Darnowski, P.; Pawluczyk, M. Analysis of the BEAVRS PWR Benchmark with SCALE and PARCS. In NUTECH2017; 2017; p 13.
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Darnowski, P.; Niewiński, G.; Pawluczyk, M.; Pilichowska, J. Analysis of effects of formation of non-condensable gases and water vapor during a severe accident in a boiling water nuclear reactor. Przemysł Chemiczny 2017, 3 (96), 521–524. https://doi.org/10.15199/62.2017.2.5.
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Darnowski, P.; Potpaczyk, K.; Świrski, K. Studies on the Recriticality Potential during Fukushima Unit-3 Core Reflooding. In ERMSAR2017; SARNET: Warsaw, 2017; p 7.
(65)
Bryk, R.; Schmidt, H.; Mull, T.; Wagner, T.; Herbst, O.; Ganzmann, I. The Modelling of Condensation in Horizontal Tubes and the Comparison with Experimental Data. ITM Web Conf. 2017, 15, 03007. https://doi.org/10.1051/itmconf/20171503007.
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Bryk, R.; Schmidt, H.; Mull, T.; Wagner, T.; Ganzmann, I.; Herbst, O. Modeling of Kerena Emergency Condenser. Archives of Thermodynamics 2017, 38 (4), 29–51. https://doi.org/10.1515/aoter-2017-0023.
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Bergagio, M.; Thiele, R.; Anglart, H. Analysis of Temperature Fluctuations Caused by Mixing of Non-Isothermal Water Streams at Elevated Pressure. International Journal of Heat and Mass Transfer 2017, 104, 979–992. https://doi.org/10.1016/j.ijheatmasstransfer.2016.08.082.
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Bergagio, M.; Anglart, H. Experimental Investigation of Mixing of Non-Isothermal Water Streams at BWR Operating Conditions. Nuclear Engineering and Design 2017, 317, 158–176. https://doi.org/10.1016/j.nucengdes.2017.03.034.
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Pawluczyk, M.; Mazgaj, P.; Gurgacz, S.; Gatkowski, M.; Darnowski, P. Loss of Coolant Accident in Pressurized Water Reactor. Prediction of a 6-Inch Cold Leg Break with Relap5 and Cathare 2. Procedia Engineering 2016, 157, 333–340. https://doi.org/10.1016/j.proeng.2016.08.374.
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Odukoya, A.; Naterer, G. F.; Roeb, M.; Mansilla, C.; Mougin, J.; Yu, B.; Kupecki, J.; Iordache, I.; Milewski, J. Progress of the IAHE Nuclear Hydrogen Division on International Hydrogen Production Programs. International Journal of Hydrogen Energy 2016, 41 (19), 7878–7891. https://doi.org/10.1016/j.ijhydene.2015.09.126.
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Mazgaj, P.; Vacher, J.-L.; Carnevali, S. Comparison of CATHARE Results with the Experimental Results of Cold Leg Intermediate Break LOCA Obtained during ROSA-2/LSTF Test 7. EPJ Nuclear Sci. Technol. 2016, 2, 1. https://doi.org/10.1051/epjn/e2015-50020-7.
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Luks, A.; Pytel, K.; Tarchalski, M.; Uzunow, N.; Krok, T. Modelling of Thermal Hydraulics in a KAROLINA Calorimeter for Its Calibration Methodology Validation. Nukleonika 2016, 61 (4), 453–460. https://doi.org/10.1515/nuka-2016-0074.
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Gurgacz, S.; Bieder, U.; Gorsse, Y.; Swirski, K. CFD Simulations of Selected Steady-State and Transient Experiments in the PLANDTL Test Facility. J. Phys.: Conf. Ser. 2016, 745, 032042. https://doi.org/10.1088/1742-6596/745/3/032042.
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Gradecka, M.; Thiele, R.; Anglart, H. Computational Fluid Dynamics Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle With Grid Spacers. Journal of Nuclear Engineering and Radiation Science 2016, 2 (3). https://doi.org/10.1115/1.4032635.
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Darnowski, P.; Potapczyk, K.; Niewiński, G.; Gatkowski, M. Analysis of the Re-Criticality Potential during the Early In-Vessel Phase of a Station Blackout in a BWR Reactor. In ENS2016 Conference Proceedings; European Nuclear Society, 2016; Vol. ENC2016-A0026.
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Darnowski, P.; Potapczyk, K.; Gatkowski, M.; Niewiński, G. Development of One-Way-Coupling Methodology between Severe Accident Integral Code MELCOR and Monte Carlo Neutron Transport Code SERPENT. Procedia Engineering 2016, 147 (ICCHMT2016), 207–213. https://doi.org/10.1016/j.proeng.2016.08.358.
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Skrzypek, M.; Laskowski, R. Thermal-Hydraulic Calculations for a Fuel Assembly in a European Pressurized Reactor Using the RELAP5 Code. Nukleonika 2015, 60 (3), 537–544. https://doi.org/10.1515/nuka-2015-0110.
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Gutowska, I.; Furmanski, P.; Woods, B. Depressurized Loss of Coolant Accident Mitigation Method Framework at Oregon State University High Temperature Test Facility – 15427. In ICAPP 2015 Proceedings; Societe Francaise d&apos: France, 2015.
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Gurgacz, S.; Pawluczyk, M.; Mazgaj, P.; Darnowski, P.; Samul, K.; Skrzypek, M. EPR Medium Break LOCA Benchmarking Exercise Using RELAP5 and CATHARE; NUREG/IA; NUREG/IA NUREG/IA; US NRC, 2015; p 57.
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Gradecka, M.; Thiele, R.; Anglart, H. CFD Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle with Grid Spacers. In Proceedings of the 7th International Symposium on Supercritical Water-Cooled Reactors ISSCWR-7; Helsinki, 2015.
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Darnowski, P.; Uzunow, N. Minor Actinides Impact on Basic Safety Parameters of Medium-Sized Sodium-Cooled Fast Reactor. Nukleonika 2015, 60 (1), 171–179. https://doi.org/10.1515/nuka-2015-0034.
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Darnowski, P.; Skrzypek, E.; Mazgaj, P.; Świrski, K.; Gandrille, P. Total Loss of AC Power Analysis for EPR Reactor. Nuclear Engineering and Design 2015, 289, 8–18. https://doi.org/10.1016/j.nucengdes.2015.03.020.
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Darnowski, P.; Skrzypek, E.; Mazgaj, P.; Gatkowski, M. Simulations of Large Break Loss of Coolant Accident without Safety Injection for EPR Reactor Using MELCOR Computer Code. In ERMSAR2015; SARNET: Marsille, 2015; p 12.
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Badyda, K.; Kuźniewski, M. Analiza opłacalności budowy elektrowni jądrowej w Polsce. Energetyka 2015, 7.
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Skrzypek, E.; Skrzypek, M. Computer Codes in the Safety Analysis for Nuclear Power Plants. Computational Capabilities of Thermal-Hydraulic Tools, Using the Example of the RELAP5 Code. Journal of Power Technologies 2014, 94 (Nuclear Issue), 41–50.
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Migdal, M.; Niewinski, G. Inclusion of Specific Geometry of the Beryllium Blocks in the Computational Model of the MARIA Reactor. Journal of Power Technologies 2014, 94 (Nuclear Issue), 69–76.
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Mazgaj, P.; Darnowski, P.; Gurgacz, S.; Lipka, M.; Dziubanii, K. Comparison of Simple Design of Sodium and Lead Cooled Fast Reactor Cores. Journal of Power Technologies 2014, 94 (Nuclear Issue), 16–26.
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Mazgaj, P.; Darnowski, P. Transmutation: Reducing the Storage Time of Spent Fuel. Journal of Power Technologies 2014, 94 (Nuclear Issue), 27–34.
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Mazgaj, P.; Jimenez, D.; Lopez, D. Design of a Sub-Critical Reactor for Transmutation of Higher Actinides. Journal of Power Technologies 2014, 94 (Nuclear Issue), 35–40.
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Gutowska, I.; Woods, B. G. Examination of Rapid Depressurization Phenomena Modeling Problems in VHTR Following Sudden DLOFC Event. In Proceedings of the HTR 2014; 2014; p 8.
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Gurgacz, S.; Pawluczyk, M.; Niewinski, G.; Mazgaj, P. Simulation and Analysis of Pipe and Vessel Blowdown Phenomena Using RELAP5 and TRACE. Journal of Power Technologies 94 2014, 94 (Nuclear Issue), 61–68.
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Gatkowski, M.; Buchner, T.; Niewinski, G.; Mazgaj, P. Development of a Measurement and Reconstruction System for Determining the Phase Distribution in a Two-Phase FLow Vertical Tube Using Electrical Impedance Tomography. Journal of Power Technologies 2014, 94 (Nuclear Issue), 77–84.
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Darnowski, P. Reaktory Podkrytyczne Sterowane Akceleratorami – Realne Rozwiązanie Problemu Odpadów Nuklearnych? Ekoatom. 2014, pp 53–59.
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Suchcicki, S. J.; Uzunow, N.; Pytel, K.; Mieleszczenko, W.; Mołdysz, A. An Investigation of Oxide Layer Impact on Heat Transfer in a Fuel Element of the MARIA Reactor. Journal of Power Technologies 2013, 93 (4), 247–256.
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Stręciwilk, P. O.; Darnowski, P.; Dominiak, A.; Domański, R. MOX and UOX Fuel Melt Margin for European Pressurized Reactor. Journal of Power Technologies 2013, 93 (3), 169–177.
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Lewandowski, J.; Laskowski, R. A Simplified Mathematical Model of a U-Tube Steam Generator under Variable Load Conditions. Archives of Thermodynamics; 2013; No 3 September; 75-88 2013.
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Prusinski, P. A.; Potempski, S.; Borysiewicz, M.; Kowal, K.; Prusinski, A. M. CFD Analysis of the Safety Related Thermal Hydraulic Parameters Describing a FLow Domain of an Experimental Medical Installation (BNCT Converter) inside of the Research Reactor MARIA. Journal of Power Technologies 2012, 92 (4), 227–240.
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Gutowska, I.; Woods, B. Scaling Analysis of Depressurized Conduction Cooldown Event at the Oregon State University High Temperature Test Facility. In 2nd International Nuclear Energy Congress; 2012.
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Darnowski, P. Reaktory Prędkie IV Generacji Chłodzone Ciekłymi Metalami. Ekoatom. 2012.
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Kupecki, J.; Badyda, K.; Anglart, H. Sensitivity Analysis of Reacting Two-Phase FLow in Nuclear Heat-Based Gasification Process. Journal of Power Technologies 2011, 91 (2), 54–62.
(101)
Gutowska, I. Gadolinium Burnable Poison Concept for a MASLWR Core. In 1st International Nuclear Energy Congress Procedings; Warsaw, 2011.
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Gutowska, I. Analysis of the Phenomenon of Rapid Pressure Drop Due to LOCA Accident at the Oregon State University High Temperature Test Facility. In Proceedings X Conference on Research & Development in Power Engineering,; 2011.
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Uzunow, N. Concept of Incorporating a Peak-Load Hydrogen Turbine into the Second Loop of Pressurised-Water Reactor Power Unit. Archiwum Energetyki 2010, XL (1–2).
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Papers published before 2010
(1)
Pluta, Z.; Lahssuny, Y. M. The Effect of Material Properties on the Temperature Distribution in a Nuclear Fuel Rod. In Prace Naukowe PW Mechanika 2003z. 202; 2003; pp 107–118.
(2)
Adamski, J.; Miller, A. The New Construction of Nuclear Power Reactors the Source of Energy Saving. In Proceedings of the ENERGY SAVING conference; Lviv, Ukraine, 2003.
(3)
Adamski, J.; Miller, A. Elektrownie z Reaktorem Jądrowymi Najnowszych Generacji Źródłem Taniej Energii w XXI Wieku. In Zeszyty Naukowe Politechniki Warszawskiej s. Konferencje; 2003; pp 15–25.
(4)
Adamski, J.; Miller, A. Jądrowe Reaktory Prędkie Mnożące – Źródła Czystej Energii Dla XXI Wieku. In Seria Elektryka Nr kol. 280/2002Z51; 2002; pp 729–746.
(5)
Adamski, J. Jądrowe Reaktory Lekkowodne – Ekologiczne Urządzenia Do Wytwarzania Energii”. In Zeszyty Naukowe Politechniki Warszawskiej s. Konferencje z. 21; 2001; pp 5–26.
(6)
Lahsumy, Y. M.; Jędral, W. A Model for Predicting Nuclear Reactor Coolant Pump Behaviour During Normal and Abnormal Operating. In International Thermal Energy on Environment Congress Conditions Proceedings; Maroko, 1997.
(7)
Janczak, R. Znaczenie Bezpieczeństwa Jądrowegow Eksploatacji Elektrowni Jądrowych. In Spektrum Biuletyn Informacyjny, Nr 5, 1997; 1997.
(8)
Butuirat, F.; Kiełkiewicz, M. On Additivity of Coagulation Kernels. Annals of Nuclear Energy 1996, 23 (13), 1091–1096. https://doi.org/10.1016/0306-4549(95)00116-6.
(9)
Adamski, J. THE NUCLEAR REACTOR PRESSURE VESSEL AND THERMAL SHIELD ACTIVITY. Journal of Power Technologies 1996, 83.
(10)
Portacha, J.; Smyk, A.; Feituri, I. A. El. THE INFLUENCE OF THE HEAT TRANSFER COEFFICIENT VARIATIONS IN THERMAL SYSTEM ELEMENTS ON POWER AND DISTRIBUTION OF EXERGETIC LOSSES IN THE PWR NUCLEAR POWER PLANT. Journal of Power Technologies 1995, 79.
(11)
Lewandowski, J.; Mościcki, M. COMPARISON OF LIMITING MODELS OF STEAM PROPAGATION IN REACTOR CONTAINMENT DURING LOCA ACCIDENT – PORÓWNANIE GRANICZNYCH MODELI ROZPRZESTRZENIANIA SIĘ PARY W BUDYNKU REAKTORA W TRAKCIE AWARII TYPU LOCA. Journal of Power Technologies 1995, 81, 10.
(12)
Kiełkiewicz, M. INFLUENCE OF VAPOUR CONDENSATION ON REMOVAL OF AEROSOLS IN A NUCLEAR REACTOR CONTAINMENT – WPŁYW KONDENSACJI PARY NA USUWANIE AEROZOLI W OBUDOWIE BEZPIECZEŃSTWA REAKTORA JĄDROWEGO. Journal of Power Technologies 1995, 81, 9.
(13)
Kiełkiewicz, M. CONDENSATIONAL GROWTH OF DROPLETS. Journal of Power Technologies 1995, 79.
(14)
Butuirat, F.; Kiełkiewicz, M. Comparison of Two Approximate Methods of Aerosol Dynamics. Annals of Nuclear Energy 1995, 22 (7), 497–501. https://doi.org/10.1016/0306-4549(94)00068-P.
(15)
Kiełkiewicz, M. Accuracy of the Moments Method. Annals of Nuclear Energy 1994, 21 (3), 189–193. https://doi.org/10.1016/0306-4549(94)90061-2.
(16)
Mościcki, M. CONDENSATIONAL GROWTH OF AEROSOL PARTICLES – KONDENSACYJNY WZROST CZĄSTEK AEROZOLI. Journal of Power Technologies 1993, 78.
(17)
Kiełkiewicz, M. Condensational Growth and Removal of Droplets. Journal of Aerosol Science 1993, 24 (2), 227–235. https://doi.org/10.1016/0021-8502(93)90060-M.
(18)
Kiełkiewicz, M.; Mościcki, M. ZMIANY WIDMA MASOWEGO AEROZOLI WYWOŁANE KONDENSACJĄ PARY. Journal of Power Technologies 1992, 76.
(19)
Bader, P. Thermal Explosion under Film Boiling Conditions. International Journal of Heat and Mass Transfer 1992, 35 (9), 2271–2276. https://doi.org/10.1016/0017-9310(92)90069-5.
(20)
Adamski, J.; Archutowski, M.; Kiełkiewicz, M. ANALYSIS OF TRANSPORT AND RETENTION OF AEROSOLS IN CONTAINMENT’ OF WATER-COOLED NUCLEAR REACTOR – ANALIZA TRANSPORTU I RETENCJI AEROZOLI W OBUDOWIE BEZPIECZEŃSTWA REAKTORA JĄDROWEGO CHŁODZONEGO WODĄ. Journal of Power Technologies 1990, 74.
(21)
Archutowski, M.; Kiełkiewicz, M. CALCULATIONS OF SHUTDOWN POWER IN THERMAL NUCLEAR REACTORS – OBLICZENIA MOCY POWYŁĄCZENIOWEJ W TERMICZNYCH REAKTORACH JĄDROWYCH. Journal of Power Technologies 1989, 73.
(22)
Adamski, J. THE TRANSURANIUM ISOTOPES PRODUCTION IN THE NUCLEAR REACTOR WITH THE DUFERANT ISOTOPES COMPOSITION OP THE REACTOR CORE – WYZNACZANIE KONCENTRACJI TRANSURANOWCÓW W REAKTORZE JĄDROWYM Z PALIWEM URANOWYM. Journal of Power Technologies 1987, 70.
(23)
Grunwald, B.; Lewandowski, J.; Miller, A.; Plewa, J. THE MATHEMATICAL MODEL OF A SATURATED STEAM TURBINE FOR DYNAMICS ANALYSE OF TURBOSET FOR NUCLEAR POWER STATION – MODEL MATEMATYCZNY TURBINY NA PARĘ NASYCONĄ DO BADANIA DYNAMIKI TURBOZESPOŁU ELEKTROWNI JĄDROWEJ. Journal of Power Technologies 1980, 57, 16.
(24)
Matuła, J. THE SIMPLIFIED METHOD OF ESTIMATION OF FISSION PRODUCTS RELEASE FROM FUEL OF NUCLEAR TOWER REACTORS – UPROSZCZONA METODA OCENY WYDZIELANIA PRODUKTÓW ROZSZCZEPIENIA Z PALIWA JĄDROWEGO REAKTORA ENERGETYCZNEGO. Journal of Power Technologies 1979, 56, 11.
(25)
Brodowicz, K.; Jędrzejewska, W.; Kalicki, A.; Kasprzak, S.; Zieliński, M. RECTIFICATION OF BORIC ACID IN NUCLEAR TOWER PLANTS – DESTYLACJA KWASU BOROWEGO DLA POTRZEB ENERGETYKI JĄDROWEJ. Journal of Power Technologies 1979, 56, 12.
(26)
Matuła, J. THE EVALUATION OF FISSION PRODUCTS INVENTORY IN THE FUEL OF NUCLEAR POWER REACTORS – OCENA ZASOBÓW PRODUKTÓW ROZSZCZEPIENIA W PALIWIE JĄDROWYCH REAKTORÓW ENERGETYCZNYCH. Journal of Power Technologies 1977, 49.
(27)
Lewandowski, J. LINEAR MODEL OF DYNAMICS OF A SATURATED STEAM TURBINE SET FOR THE NUCLEAR POWER STATION – LINIOWY MODEL DYNAMIKI TURBOZESPOŁU NA PARĘ NASYCONĄ DLA ELEKTROWNI JĄDROWEJ. Journal of Power Technologies 1977, 48.
(28)
Podowski, M. ANALYSIS OF A CERTAIN OPERATOR EQUATION AS APPLIED TO THE EXAMINATION OF NUCLEAR REACTOR STABILITY – ANALIZA PEWNEGO RÓWNANIA OPERATOROWEGO I JEJ ZASTOSOWANIE DO BADANIA STABILNOŚCI REAKTORÓW JĄDROWYCH. Journal of Power Technologies 1972, 37.
(29)
Podowski, M. A FINITE DIFFERENCE SOLUTION OF PARTIAL DIFFERENTIAL EQUATIONS OF NEUTRON DIFFUSION – METODA RÓŻNICOWA ROZWIĄZYWANIA UKŁADU RÓWNAŃ RÓŻNICZKOWYCH CZĄSTKOWYCH TYPU RÓWNAŃ DYFUZJI NEUTRONÓW. Journal of Power Technologies 1972, 35.
(30)
Masłowski, A. OPTIMAL ESTIMATION METHOD FOR CERTAIN PHYSICAL PROCESSES – ON THE EXAMPLE OF PHENOMENA OCCURRING IN A NUCLEAR REACTOR – METODA OPTYMALNEJ ESTYMACJI PEWNYCH PROCESÓW FIZYCZNYCH NA PRZYKŁADZIE MODELU ZJAWISK ZACHODZĄCYCH W REAKTORZE JĄDROWYM. Journal of Power Technologies 1972, 37.
(31)
Masłowski, A. OPTIMAL ESTIMATION OF SOME OF THE SPACE-TIME PROCESSES IN NUCLEAR REACTORS – O OPTYMALNEJ ESTYMACJI NIEKTÓRYCH PRZESTRZENNOCZASOWYCH ZJAWISK ZACHODZĄCYCH W REAKTORACH JĄDROWYCH. Journal of Power Technologies 1971, 32.