Các cơ sở lý thuyết, tình trạng hiện tại của sự phát triển và triển vọng cải tiến tiếp theo của mã tính toán thủy nhiệt thế hệ mới HYDRA-IBRAE/LM để mô phỏng các hệ thống lò phản ứng nhanh

White Book of Nuclear Thermal Engineering, Ed. by E. O. Adamov (GUP NIKIET, Moscow, 2001) [in Russian]. P. N. Alekseev, V. G. Asmolov, A. Yu. Gagarinskii, N. E. Kukharkin, Yu. M. Semchenkov, V. A. Sidorenko, S. A. Subbotin, V. F. Tsibul’skii, and Ya. I. Shtrombakh, “On strategy of nuclear energetic development in Russia to 2050,” At. Energ. 111, 183–196 (2011). P. L. Kirillov, A Handbook on Thermohydraulic Calculations in Atomic Energetic. Vol. 3. Thermohydraulic Processes at Transfer and Non-Standard Regimes. Severe Accidents. Protection Coating. Codes, Their Possibilities, Uncertanities, (IzdAt, Moscow, 2014) [in Russian]. B. I. Nigmatulin, A. V. Vasilenko, and S. L. Solov’ev, “The creation of computer codes of the new generation: An important problem in Russian nuclear power engineering,” Therm. Eng. 49, 879–887 (2002). Electronic Resource. http://www.inl.gov/relap5/default.htm Electronic Resource, http://www-cathare.cea.fr/scripts/home/publigen/content/templates/show.asp?L=EN&P=134 J. W. Spore, J. S. Elson, S. J. Jolly-Woodruff, T. D. Knight, J.-C. Lin, R. A. Nelson, K. O. Pasamehmetoglu, R. G. Steinke, and C. Unal, TRAC-M/FORTRAN 90 (Version 3.0) LA-UR-00-910. Theory manual, Los Alamos Nat. Lab. (2000). S. Z. Rouhani, R. W. Shumway, W. L. Weaver, and C. L. Kullberg, TRAC-BF1/MOD1 Models and Correlations, Idaho Nat. Eng. Lab., Contract No. DE-AC0776ID01570, FIN No. L2031, UREG/CR-4391 (1992). M. M. Giles, G. A. Jayne, S. Z. Rouhani, et al., TRACBF1/MOD1: An advanced best estimate computer program for boiling water reactor accident analysis. V. 1: Model description, (US Nat. Reactor Lab, Merylend, 1992). A. V. Beznosov, Yu. G. Dragunov, and V. I. Rachkov, Heavy Heat Carriers in Atomic Engineering, (IzdAt, Moscow, 2007) [in Russian]. Yu. V. Yudov, “A two-fluid model of unsteady circuit thermohydraulics and its numerical realization in the KORSAR computer code,” Therm. Eng. 49, 895–900 (2002). J. K. Fink and L. Leibowitz, Thermodynamic and Transport Properties of Sodium Liquid and Vapor, (ANL/RE-95, Argonne Nat. Lab., 1995). O. D. Kuznetsova and A. M. Semenov, “New reference data on the thermodynamic properties of sodium vapor,” High Temp. 38, 26–32 (2000). Liquid-Metal Heat Carriers, Ed. by V. M. Borishanskii, S. S. Kutateladze, I. I. Novikov, and O. S. Fedynskii, (Atomizdat, Moscow, 1967), 2nd ed. [in Russian]. Yu. A. Zeigarnik and V. D. Litvinov, Boiling of Alkaline Metals in Channels (Nauka, Moscow, 1983) [in Russian]. A. D. Efanov, A. I. Sorokin, E. F. Ivanov, G. P. Bogoslovskaya, V. P. Kolesnik, S. S. Martsinyuk, V. L. Mal’kov, G. A. Sorokin, and K. S. Rymkevich, “An investigation of the heat transfer and stability of liquid-metal coolant boiling in a natural circulation circuit,” Therm. Eng. 50, 194–201 (2003). H. M. Kottowski and C. Savatteri, “Fundamentals of liquid metal boiling thermohydraulics,” Nucl. Eng. Design 82, 281–304 (1984). Y. Kikuchi, K. Haga, and T. Takahashi, “Experimental study of steady-state boiling of sodium flowing in a single-pin annular channel,” J. Nucl. Sci. Technol. 12, 83–91 (1975). A. V. Zhukov, P. L. Kirillov, and N. M. Matyukhin, Thermohydraulic Calculation of Fuel Assemby of Fast Reactors with Liquid Metal Cooling (Energoatomizdat, Moscow, 1985) [in Russian]. M. E. Kuznetsova, A. A. Butov, I. S. Vozhakov, I. G. Kudashov, E. V. Usov, S. I. Lezhnin, and N. A. Pribaturin, “System of relations used for closing of equations for two-liquid model used for analysis of damage development in the reactors with liquid metal cooling,” Proc. 18th All-Russ. Sci.-Tech. Conf. “Power Engineering, Effectiveness, Reliability, Safety”, Tomsk, 2012. [http://www.lib.tpu.ru/fulltext/c/2012/C15/092.pdf.] K. Takahashi, Y. Fujii-e, and T. Suita, “Incipient boiling phenomena of sodium under forced convection by direct heating,” J. Nucl. Sci. Technol. 9, 603–612 (1972). A. I. Leonov and V. F. Prisnyakov, “Sodium extreme overheating at boiling up,” Teplofiz. Vys. Temp. 10, 149–152 (1972). F. E. Dunn, The SAS4A/SASSYS-1 LMR Accident and Systems Analysis Codes, 1996, vol. 4. A. W. Cronenberg, H. K. Fauske, S. G. Bankoff, and D. T. Eggen, “A single bubble model for sodium expulsion from a heated channel,” Nuclear Eng. Design 16, 285–293 (1971). M. N. Vlasov, A. S. Korsun, Yu. A. Maslov, I. G. Merinov, and V. S. Kharitonov, “Calculation studies of core structure flow for determination of the basic parameters of integral turbulent model. Longitudinal flow,” Vestn. Nats. Issl. Yadern. Univ. Mos. Inzh. Fiz. Inst. Matem. Kompt. Model. 2, 314–318 (2013). H. Tanabe, “Test and analysis on steam generator tube failure propagation,” in Proc. IAEA Specialists’ Meeting on Steam Generator Failure and Failure Propagation Experience, Aix-en-Provence, France, 1990. I. A. Kuznetsov and V. M. Poplavskii, Safety of Atomic Power Stations with Reactors on Fast Neutrones, Ed. by V. I. Rachkov, (IzdAt, Moscow, 2012) [in Russian]. V. M. Poplavskii, M. S. Pinkhasik, Yu. E. Bagdasarov, and N. P. Aristarkhov, “Experimental study of sodium interaction with water,” Teploenergetika, No. 6, 70–74 (1966). Yu. E. Bagdasarov, M. S. Pinkhasik, I. A. Kuznetsov, and F. A. Kozlov, Technical Problems of Reactors on Fast Neutrons (Atomizdat, Moscow, 1969) [in Russian]. A. V. Zhukov, P. L. Kirillov, and N. M. Matyukhin, Thermohydraulic Calculation of Fuel Asembly of Fast Reactors with Liquid Metal Cooling, (Energoatomizdat, Moscow, 1985) [in Russian]. A. V. Zhukov, Yu. A. Kuzina, A. P. Sorokin, V. N. Leonov, V. P. Smirnov, and A. G. Sila-Novitskii, “An experimental study of heat transfer in the core of a BREST-OD-300 reactor with lead cooling on model, Therm. Eng. 49, 175–184 (2002). V. I. Subbotin, M. Kh. Ibragimov, P. A. Ushakov, V. P. Bobkov, A. V. Zhukov, and Yu. S. Yur’ev, Hydrodynamics and Heat Transfer in Atomic Energetic Plants (Foundations of Calculation) (Atomizdat, Moscow, 1975). Vol. 1 [in Russian]. A. V. Beznosov and T. A. Bokova, Equipment of Energetic Contours with Heavy Liquid Metal Heat Carriers in Atomic Power Engineering. A Tutorial, (Nizhegorod. Gos. Tekhn. Univ., Nizhny Novgorod, 2012) [in Russian]. A. A. Molodtsov, Extended Abstract of Candidate’s Dissertation in Technical Science (Nizhny Novgorod, 2007). O. O. Novozhilova, Extended Abstract of Candidate’s Dissertation in Technical Science (Nizhny Novgorod, 2007). P. V. Kolobaeva, “HYDRA-IBRAE/LM/V1 calculation code verification on experiments with lead-bismuth heat carrier, executed in Central ConstructionTechnological Institute,” in Proc. XV Young Scientist Sci. School of Institute of Safety Problems of Atomic Power Engineering Development of Russ. Acad. Sci.,; Preprint No. IBRAE-2014-02. (Moscow, Institute of Safety Problems of Atomic Power Engineering Development of Russ. Acad. Sci., 2014), pp. 112–115. [http://www.ibrae.ac.ru/docs/109/2014i02s.pdf] Yu. V. Yudov, S. N. Volkova, and Yu. A. Migrov, “The closing relationships of the thermohydraulic model of the KORSAR computer code,” Therm. Eng. 49, 901–908 (2002). V. M. Alipchenkov, V. V. Belikov, A. V. Davydov, D. A. Emel’yanov, and N. A. Mosunova, “Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors,” Therm. Eng. 60, 331–337 (2013). DOI: 10.1134/S0040363613050020 W. Ma, E. Bubelis, A. Karbojian, B. R. Sehgal, and P. Coddington, “Transient experiments from the thermal-hydraulic ADS lead bismuth loop (TALL) and comparative TRAC/AAA analysis,” Nucl. Eng. Design, No. 236, 1422–1444 (2006).