Authors & Affiliations
Opanasenko A.N.1, Sorokin A.P.1, Trufanov A.A.1, Denisova N.A.1, Razuvanov N.G.3, Sviridov Eu.V.2, Belyaev I.A.3
1A.I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Russia
2Moscow Power Engineering Institute, Moscow, Russia
3Joint Institute of High Temperatures of Russian Academy of Science, Moscow, Russia
Sorokin A.P. – Dr. Sci. (Tech.), Deputy Director of Safety Department, A.I. Leypunsky Institute for Physics and Power Engineering.
Trufanov A.A. – Deputy Director General, Director of Safety Department, A.I. Leypunsky Institute for Physics and Power Engineering.
Denisova N.A. – Leading Engineer of Safety Department, A.I. Leypunsky Institute for Physics and Power Engineering.
Razuvanov N.G. – Dr. Sci. (Tech.), Leading Researcher, Joint Institute of High Temperatures of Russian Academy of Science.
Sviridov Eu.V. – Cand. Sci. (Tech.), Docent, Moscow Power Engineering Institute.
Belyaev I.A. – Cand. Sci. (Tech.), Head of the laboratory, Joint Institute of High Temperatures of Russian Academy of Science.
The results of experimental studies of the structure of the movement, non-isothermal temperature fields and coolant velocity in the water model in different regimes in relation to the various types of nuclear power. Research carried out on transparent models of reactor coolant system and for drop VVER channel sector model of the upper chamber of a fast reactor and water three-loop model of a fast reactor with integral equipment arrangement in the first circuit. Measurements of local velocities height and radius of the upper mixing chamber in a plane in the direction of the center of the core to the intermediate heat exchanger in the reactor on fast neutrons model for the first time possible to obtain three-dimensional picture of the velocity field in the volume of the upper chamber in a forced circulation of the coolant and reactor emergency cooling by natural convection. As a result, the measurement of temperature and velocity fields using specially designed measuring instruments identified areas with a stable temperature stratification with large gradients and fluctuations of temperature at the interface of stratified and recirculation zones. The results obtained enable us to judge about the amplitude and frequency characteristics of the temperature fluctuations in these potentially dangerous areas for different modes of operation of the reactors. The data obtained can be used to verify the thermal-hydraulic codes used to justify the design characteristics and safety of nuclear reactors.
experimental studies, a fast reactor, the reactor tank, experimental water model, the integrated arrangement, the upper chamber, thermal hydraulics, local characteristics, velocity, temperature stratification of the coolant, stationary processes, transients, emergency cooling, natural convection, the temperature non-uniformity, ripple, verification codes
1. Opanasenko A.N., Sorokin A.P., Zaryugin D.G., Markov M.V. Stratifikatsiya teplonositelya v yadernykh energeticheskikh ustanovkakh [Coolant stratification in nuclear power plants]. Atomnaya energiya - Atomic Energy, 2011, vol. 111, no. 3, pp. 131-163.
2. Opanasenko A.N., Sorokin A.P., Zaryugin D.G, Fedorov A.V. Eksperimental'nye issledovaniya poley temperatury i struktury dvizheniya teplonositelya na modeli bystrogo reaktora v elementakh pervogo kontura pri perekhode k raskholazhivaniyu estestvennoy tsirkulyatsiey [Experimental study of temperature fields and structure movement of coolant on fast reactor model in the first circuit elements in the transition to natural circulation cooldown]. Obninsk, IPPE Publ., pp. 102–111.
3. Opanasenko A.N., Sorokin A.P. Eksperimental'nye issledovaniya stratifikatsionnykh protsessov v elementakh kontura tsirkulyatsii YaEU razlichnogo tipa na vodyanykh modelyakh [Experimental studies of stratification processes in the element circuit circulation NPP different types of models on the water]. Trudy nauchno-tekhnicheskoy konferentsii "Teplofizika reaktorov novogo pokoleniya (Teplofizika–2016)" [Proc. Sci. and Tech. Conf. "Thermal physics of reactors of new generation (Thermal physics–2016)"]. Obninsk, 2016, pp. 104–106.
4. Shulz H. Experience with thermal fatigue in LWR piping caused by mixing and stratification. Specialists Meeting Proceedings. Paris, 1998, pp. 13–18.
5. Evaluation of Decay Heat Removel by Natural Convection. Japan, Oarai Engineering Center, PNC. IAEA, IWGFR/88. 1993. 158 p.
6. Ushakov P.A., Sorokin A.P. Problemy modelirovaniya na vode avariynogo osta teplovydeleniya estestvennoy konvektsiey v kamerakh bystrykh reaktorov [Problems of modeling by water of emergency decay heat removal by natural convection in chambers of fast reactors]. Preprint FEI-2585 – Preprint IPPE-2585. Obninsk, 1997.
7. Belyaev I.A., Razuvanov N.G., Zagorskiy V.S. Temperaturnyy datchik dlya izmereniya poley temperatury i komponent skorosti v magnitno-gidrodinamicheskom potoke zhidkogo metalla [Temperature sensor for the measurement of temperature and velocity fields in the magnetic component of the hydrodynamic flow of liquid metal]. Teplovye protsessy v tekhnike – Thermal processes in engineering, 2015, no. 12, pp. 556–572.
8. Opanasenko A.N. Teplogidravlika verkhney oblasti baka bystrogo reaktora v razlichnykh rezhimakh raboty [Thermal hydraulic of upper region of the fast reactor vessel in various operating modes]. Preprint FEI-2623 – Preprint IPPE-2623. Obninsk, 1997.