Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

详细信息    查看全文

• 作者：V. M. Alipchenkov ; A. M. Anfimov ; D. A. Afremov ; V. S. Gorbunov…
• 关键词：system thermal ; hydraulic code ; liquid–metal coolant ; fast reactor system ; lead ; sodium ; heat–mass exchange ; closing equations ; verification ; validation
• 刊名：Thermal Engineering
• 出版年：2016
• 出版时间：February 2016
• 年：2016
• 卷：63
• 期：2
• 页码：130-139
• 全文大小：255 KB
• 参考文献：1. White Book of Nuclear Thermal Engineering, Ed. by E. O. Adamov (GUP NIKIET, Moscow, 2001) [in Russian].
  2.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).
  3.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].
  4.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).
  5. Electronic Resource. http://​www.​inl.​gov/​relap5/​default.​htm
  6. Electronic Resource, http://​www-cathare.​cea.​fr/​scripts/​home/​publigen/​content/​templates/​show.​asp?​L=​EN&​P=​134
  7.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).
  8.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).
  9.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).
  10.A. V. Beznosov, Yu. G. Dragunov, and V. I. Rachkov, Heavy Heat Carriers in Atomic Engineering, (IzdAt, Moscow, 2007) [in Russian].
  11.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).
  12.J. K. Fink and L. Leibowitz, Thermodynamic and Transport Properties of Sodium Liquid and Vapor, (ANL/RE-95, Argonne Nat. Lab., 1995).CrossRef
  13.O. D. Kuznetsova and A. M. Semenov, “New reference data on the thermodynamic properties of sodium vapor,” High Temp. 38, 26–32 (2000).CrossRef
  14. 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].
  15.Yu. A. Zeigarnik and V. D. Litvinov, Boiling of Alkaline Metals in Channels (Nauka, Moscow, 1983) [in Russian].
  16.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).
  17.H. M. Kottowski and C. Savatteri, “Fundamentals of liquid metal boiling thermohydraulics,” Nucl. Eng. Design 82, 281–304 (1984).CrossRef
  18.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).CrossRef
  19.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].
  20.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.​]
  21.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).CrossRef
  22.A. I. Leonov and V. F. Prisnyakov, “Sodium extreme overheating at boiling up,” Teplofiz. Vys. Temp. 10, 149–152 (1972).
  23.F. E. Dunn, The SAS4A/SASSYS-1 LMR Accident and Systems Analysis Codes, 1996, vol. 4.
  24.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).CrossRef
  25.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).
  26.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.
  27.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].
  28.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).
  29.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].
  30.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].
  31.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).
  32.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].
  33.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].
  34.A. A. Molodtsov, Extended Abstract of Candidate’s Dissertation in Technical Science (Nizhny Novgorod, 2007).
  35.O. O. Novozhilova, Extended Abstract of Candidate’s Dissertation in Technical Science (Nizhny Novgorod, 2007).
  36.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]
  37.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).
  38.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/S0040363613050020CrossRef
  39.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).CrossRef
  
• 作者单位：V. M. Alipchenkov (1) (2)
  A. M. Anfimov (3)
  D. A. Afremov (4)
  V. S. Gorbunov (3)
  Yu. A. Zeigarnik (1) (2)
  A. V. Kudryavtsev (4)
  S. L. Osipov (3)
  N. A. Mosunova (1)
  V. F. Strizhov (1)
  E. V. Usov (1)
  
  1. Nuclear Safety Institute (IBRAE), Russian Academy of Sciences, ul. Bolshaya Tulskaya 52, Moscow, 115191, Russia
  2. Joint Institute for High Temperatures (OIVT), Russian Academy of Sciences, ul. Izhorskaya 13/2, Moscow, 125412, Russia
  3. Afrikantov Experimental Design Office of Mechanical Engineering (AO OKBM Africantov), pr. Burnakovskii 15, Nizhny Novgorod, 603074, Russia
  4. Dollezhal Research and Development Institute of Power Engineering (NIKIET), ul. Malaya Krasnoselskaya 2/8, Moscow, 107140, Russia
  
• 刊物类别：Engineering
• 刊物主题：Engineering Thermodynamics and Transport Phenomena
  Russian Library of Science
  
• 出版者：MAIK Nauka/Interperiodica distributed exclusively by Springer Science+Business Media LLC.
• ISSN：1555-6301

文摘

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid–metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid–metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat–mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level. Keywords system thermal-hydraulic code liquid–metal coolant fast reactor system lead sodium heat–mass exchange closing equations verification validation