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Artykuły w czasopismach na temat "U-Zr-Fe-O"
Tsurikov, D. F., V. N. Zagryazkin, V. Yu Vishnevskii, E. K. Diakov, A. Yu Kotov i V. M. Repnikov. "U–Zr–Fe–O Melt density". Atomic Energy 107, nr 4 (październik 2009): 247–54. http://dx.doi.org/10.1007/s10512-010-9222-2.
Pełny tekst źródłaAsmolov, V. G., V. N. Zagryazkin i D. F. Tsurikov. "The thermodynamics of U-Zr-Fe-O melts". High Temperature 45, nr 3 (czerwiec 2007): 305–12. http://dx.doi.org/10.1134/s0018151x07030042.
Pełny tekst źródłaAsmolov, V. G., V. N. Zagryazkin i D. F. Tsurikov. "Estimation of the density of U-Zr-Fe-O melts". High Temperature 46, nr 4 (30.07.2008): 579–82. http://dx.doi.org/10.1134/s0018151x08040202.
Pełny tekst źródłaOhgi, Hiroshi, Yuji Nagae i Masaki Kurata. "THERMODYNAMIC EVALUATION ON SOLIDIFICATION PATH FOR U-ZR-FE-O CORIUM". Proceedings of the International Topical Workshop on Fukushima Decommissioning Research 2022 (2022): 1066. http://dx.doi.org/10.1299/jsmefdr.2022.0_1066.
Pełny tekst źródłaBottomley, Paul David W., Mairead Murray-Farthing, Dario Manara, Thierry Wiss, Bert Cremer, Cos Boshoven, Patrick Lajarge i Vincenzo Rondinella. "Investigations of the melting behaviour of the U–Zr–Fe–O system". Journal of Nuclear Science and Technology 52, nr 10 (10.04.2015): 1217–25. http://dx.doi.org/10.1080/00223131.2015.1023381.
Pełny tekst źródłaFUKASAWA, Masanori, Shigeyuki TAMURA i Mitsuhiro HASEBE. "Development of Thermodynamic Database for U—Zr—Fe—O—B—C—FPs System". Journal of Nuclear Science and Technology 42, nr 8 (sierpień 2005): 706–16. http://dx.doi.org/10.1080/18811248.2004.9726440.
Pełny tekst źródłaPöml, Philipp, i Boris Burakov. "Study of the redistribution of U, Zr, Nb, Tc, Mo, Ru, Fe, Cr, and Ni between oxide and metallic phases in the matrix of a multiphase Chernobyl hot-particle extracted from a soil sample of the Western Plume". Radiochimica Acta 106, nr 12 (27.11.2018): 985–90. http://dx.doi.org/10.1515/ract-2018-2957.
Pełny tekst źródłaSUDO, Ayako, Fumiki MIZUSAKO, Kuniyoshi HOSHINO, Takumi SATO, Yuji NAGAE i Masaki KURATA. "Fundamental Study on Segregation Behavior in U–Zr–Fe–O System during Solidification Process". Transactions of the Atomic Energy Society of Japan 18, nr 3 (2019): 111–18. http://dx.doi.org/10.3327/taesj.j18.029.
Pełny tekst źródłaNandan, Shambhavi, Florian Fichot i Bruno Piar. "A simplified model for the quaternary U-Zr-Fe-O system in the miscibility gap". Nuclear Engineering and Design 364 (sierpień 2020): 110608. http://dx.doi.org/10.1016/j.nucengdes.2020.110608.
Pełny tekst źródłaKhabensky, V. B., V. I. Almjashev, E. B. Shuvaeva, E. V. Krushinov, S. A. Vitol, A. A. Sulatsky, S. Yu Kotova i V. V. Gusarov. "Experimental determination of spatial inversion pointof coexisting molten phases in the U-Zr-Fe-O system". Nuclear Propulsion Reactor Plants. Life Cycle Management Technologies., nr 3 (2021): 63–81. http://dx.doi.org/10.52069/2414-5726_2021_3_25_63.
Pełny tekst źródłaRozprawy doktorskie na temat "U-Zr-Fe-O"
Quaini, Andrea. "Étude thermodynamique du corium en cuve - Application à l'interaction corium/béton". Thesis, Université Grenoble Alpes (ComUE), 2015. http://www.theses.fr/2015GREAI061/document.
Pełny tekst źródłaDuring a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore, depending on the considered scenario, the corium can be formed by a liquid phase or by two liquids, one metallic the other oxide. The objective of this thesis is the investigation of the thermodynamics of the prototypic in-vessel corium U-Pu-Zr-Fe-O. The approach used during the thesis is based on the CALPHAD method, which allows to obtain a thermodynamic model for this complex system starting from phase diagram and thermodynamic data. Heat treatments performed on the O-U-Zr system allowed to measure two tie-lines in the miscibility gap in the liquid phase at 2567 K. Furthermore, the liquidus temperatures of three Zr-enriched samples have been obtained by laser heating in collaboration with ITU. With the same laser heating technique, solidus temperatures have been obtained on the UO2-PuO2-ZrO2 system. The influence of the reducing or oxidising on the melting behaviour of this system has been studied for the first time. The results show that the oxygen stoichiometry of these oxides strongly depends on the oxygen potential and on the metal composition of the samples. The miscibility gap in the liquid phase of the U-Zr-Fe-O system has been also observed. The whole set of experimental results with the literature data allowed to develop the thermodynamic model of the U-Pu-Zr-Fe-O system. Solidification path calculations have been performed for all the investigated samples to interpret the microstructures of the solidified samples. A good accordance has been obtained between calculation and experimental results. Heat treatments on two ex-vessel corium samples showed the influence of the concrete composition on the nature of the liquid phases formed at high temperature. The observed microstructures have been interpreted by means of calculation performed with the TAF-ID database. In parallel, a novel experimental setup named ATTILHA based on aerodynamic levitation and laser heating has been conceived and developed to obtain high temperature phase diagram data. This setup has been validated on well-known oxide systems. Furthermore, this technique allowed to observe in-situ, by using an infrared camera, the formation of a miscibility gap in the liquid phase of the O-Fe-Zr system by oxidation of a Fe-Zr sample. The next step of the development will be the nuclearization of the apparatus to investigate U-containing samples. The implementation of a very fast visible camera (5000 Hz) to investigate the thermo-physical properties of in-vessel and ex-vessel corium mixtures is also underway. The synergy between the development of experimental and calculation tools will allow to improve the thermodynamic description of the corium and the severe accident code using thermodynamic input data
MAURIZI, ANNE. "Reactivite chimique a haute temperature dans le systeme (u, zr, fe, o). Contribution a l'etude de la zircone comme recuperateur de corium". Paris 6, 1996. http://www.theses.fr/1996PA066617.
Pełny tekst źródłaBrunel, Alan. "Propriétés thermodynamiques et thermophysiques des liquides à haute température : applications aux combustibles nucléaires". Electronic Thesis or Diss., Sorbonne université, 2022. http://www.theses.fr/2022SORUS426.
Pełny tekst źródłaDuring a severe accident involving the meltdown of the core of a pressurized water nuclear reactor, the nuclear fuel will react with the zircalloy cladding around it and the structural materials of the core to make a high temperature magma called corium. Depending on its composition and its temperature, the corium can stratify because of two non-miscible metallic and oxidic liquids. For some stratification configurations, the heat flow can focus on the vessel’s wall, threatening its integrity with a corium flowing outside of it. The aim of this thesis is to collect thermodynamic and thermophysic data on a prototypical corium, the U-Zr-Fe-O system. The thermodynamic data collected in this thesis are related to the definition of the liquid miscibility gap and the compositions of the liquids in the U-Zr-Fe-O system and its sub-systems, depending on the composition and the temperature. Compositions of interest were selected after performing thermodynamic calculation by the CALPHAD method with the TAF-ID V13 database. The corresponding samples underwent heat treatments and post-treatment analyses to measure the compositions of the liquids and to compare them to thermodynamic calculations. An iron rich liquid miscibility gap and a zirconium rich one were highlighted in the Fe-Zr-O system. Although calculations were in agreement with data from the first miscibility gap at 1990 °C, measurements in the zirconium rich miscibility gap at 2420 °C and 2650 °C reveal an underestimation of the zirconium quantity in the metallic liquid and its overestimation in the oxidic liquid by the model. Studies on the UO2-Zr-Fe system at 2423 °C show that the liquid miscibility gap definition and the compositions of the liquids depend on the quantity of iron in the system, the U/Zr ratio and corium oxidation degree. Furthermore, the zirconium molar fraction is underestimated by the model in the metallic liquid to the benefit of iron, and is overestimated in the oxidic liquid. Finally, the oxygen solubility in the metallic liquid is underestimated by the model. Thermophysic data were collected thanks to the improvement of the ATTILHA experimental setup, allowing the study of oxygen sensitive or radioactive liquids at high temperature by using a laser heating. Experimental values on liquidus and eutectic transformation temperatures of the oxygen-rich domain of the Zr-O system were acquired with this setup. Furthermore, the development of the aerodynamic levitation allows us the investigation liquids’densities for the Zr-Fe2O3 and the Zr-UO2 systems between 1884 °C and 2268 °C for different zirconium molar fractions. Densities of liquids from the Zr-Fe2O3 system were used to refine surface tension values acquired on the VITI-MBP setup at CEA Cadarache. These values confirmed the surfacting properties of the oxygen on these liquids. The experimental data collected during this thesis will be used to feed the databases and to improve the forecast of the corium’s behavior during a severe accident
Części książek na temat "U-Zr-Fe-O"
McFarland, Ben. "Unfolding the Periodic Table". W A World From Dust. Oxford University Press, 2016. http://dx.doi.org/10.1093/oso/9780190275013.003.0007.
Pełny tekst źródłaStreszczenia konferencji na temat "U-Zr-Fe-O"
Carénini, L., i F. Fichot. "The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies". W 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60598.
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