Possibility of Simulating Natural Circulation in Fast Neutron Reactors Using a Light Water Test Facility
The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. The sodium-cooled BN-1200 reactor was selected as the reactor installation to be modeled. The development of novel reactor designs must be based on the results of experimental studies. Some problems of modeling thermohydraulic processes in BN type reactors are studied by using sodium test facilities. Experimental studies of natural convection processes using light water test facilities can be considered as a good alternative to those using sodium test facilities. To validate the model, the similarity theory and the «black box» method were used and their principles and applicability were analyzed. Using the «black box» method makes it possible to avoid detailed modeling of such components as the reactor core and heat exchangers, replacing them by a simplified representation of these components to simulate the integral characteristics of the existing real life equipment. The paper considers the basic criteria which determine the similarity of the thermohydraulic processes under study. The governing criteria of similarity were estimated based on the fundamental differential equations of natural convection heat transfer. Based on these criteria, a set of dimensionless values was obtained which show the correlation between the model parameters and the characteristics of the reactor facility. Besides, generalized relationships were derived which can be used to estimate the scaling factors for calculating the key values of the reactor facility based on the model parameters. These relationships depend on the thermal-physics parameters of the working fluids, the geometrical scale value and the ratio of the thermal power of the model to that of the reactor facility, i.e., model-to-reactor thermal power ratio. The conditions under which it is possible to model sodium coolant by light water with adequate accuracy were analyzed. An example is given of the numerical values of the scaling factors for one of the reference light water test facilities. The paper uses the experience of a number of foreign researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. According to the available estimates, the assumptions used do not result in considerable losses in accuracy. Thus, the natural circulation of the sodium coolant in the upper plenum of the fast-neutron reactor can be simulated with adequate accuracy by using light water test facilities.
- Takeda H., Koga T. Study on similarity rule for natural circulation water test of LMFBR. Specialists’ Meeting of IAEA: Evaluation of Decay Heat Removal by Natural Convection. O-arai Engineering Center, PNC, Japan. Feb. 22-25, 1993.
- Eguchi Y., Takeda H., Koga T., Tanaka N., Yamamoto K. Quantitative prediction of natural circulation in an LMFR with a similarity law and a water test. Nuclear Engineering and Design. 1997, v. 178, no. 3, pp. 295-307.
- Eguchi Y., Takeda H. Experimental and computational Study of prediction of Natural Circulation in top-entry Loop-type FBR. Specialists’ Meeting of IAEA: Evaluation of Decay Heat Removal by Natural Convection. O-arai Engineering Center, PNC, Japan. Feb. 22-25, 1993.
- Ieda Y., Kamide H., Ohshima H., Sugawara S., Ninokata H. Strategy of Experimental Studies in PNC on Natural Convection Decay Heat Removal. Specialists’ Meeting of IAEA: Evaluation of Decay Heat Removal by Natural Convection. O-arai Engineering Center, PNC, Japan. Feb. 22-25, 1993.
- Ishii M., Kataoka I. Scaling Criteria for LWR’s under Single-Phase and Two-Phase Natural Circulation. Joint NRC/ANS Meeting on Basic Thermal Hydraulic Mechanisms in LWR Analysis. Bethesda, MD, USA. Sept. 14-15, 1982.
- Ishii M., Kataoka I. Scaling laws for thermal-hydraulic system under single-phase and two-phase natural circulation. Nuclear Engineering and Design. 1984, v. 81, no. 3, p. 411-425.
- Bomelburg H.J. An Evaluation of the Applicability of Water Model Testing to Liquid Metal Engineering Problems. General, Miscellaneous and Progress Reports, LMEC-68-4. Liquid Metal Engineering Center, Atomics International. Feb. 26, 1968.
- Ushakov P.A., Sorokin A.P. About the Possibilities of Modeling the Fuel and the Coolant Temperature Distributions in Liquid Metal Cooled Reactors using Water and Air. IPPE Preprint -2346. Obninsk. FEI Publ., 1993. 34 p. (in Russian).
- Ushakov P.A., Sorokin A.P. Modeling Problems of Emergency Natural Convection Heat Removal in the Upper Plenum of Fast Reactors Using Water. Proc. of the IX-th Intern. Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-9). USA, October 3-8, 1999. 10.Ashirmetov M.R., Ershov G.A. The general design features of a BN-1200 unit. Available at: http://www.proatom.ru/modules.php?name=News&file =article&sid=4279. 25.01.2013 (accessed Jun. 20, 2021) (in Russian).
- Mitenkov F.M. The opportunities of developing the fast/neutron breeder reactors. Available at: http://elib.biblioatom.ru/text/mitenkov_razmyshleniya-o-perezhitom_2004/go,170/ (accessed Jun. 20, 2021) (in Russian).
- Ushakov P.A., Sorokin A.P. Role of the Reynolds number in modeling natural convection in liquid metals. Atomic Energy, 1998, v. 84, no. 5, pp. 309-313.
- Polezhaev V.I. The numerical calculation of the Navier-Stokes equations for the flowing and heat transfer in a closed two-dimensional region. Cand. Sci. (Engineering) Diss. Moscow. MEI Publ., 1967, 196 p. (in Russian).
- Mihatulin D.S, Chirkov A.Yu. The Compendium of Lectures in the Heat and Mass Transfer. Moscow. MGTU Publ., 2009, 148 p. (in Russian).
- Isachenko V.P., Osipova V.A., Sukomel A.S. Heat Transfer. Moscow. Еnergiya Publ., 1975, 416 p. (in Russian).
- Isaev S.I., Kozhinov I.A., Kofanov V.I., Leontiev A.I., Mironov B.M., Nikitin V.M., Petrazhitsky G.B., Hvostov V.I., Chukaev A.G., Shishov E.V., Shkola V.V. The Heat and Mass Transfer Theory. Moscow. Vysshaya Shkola Publ., 1979, 495 p. (in Russian).
- Kirillov P.L., Bogoslovskaya G.P. The Heat and Mass Transfer in Nuclear Power Systems. Moscow. Energoatomizdat Publ., 2000, 458 p. (in Russian).
- Guhman A.A. Introduction to the Similarity Theory. Textbook. 2-nd ed. Moscow. Vysshaja Shkola Publ., 1973, 296 p. (in Russian).
- Kirpichev M.V. The Similarity Theory. Moscow. AN SSSR Publ., 1953, 96 p. (in Russian).
- Kirillov P.L., Yur’ev Yu.S., Bobkov V.P. The Thermohydraulic Calculations Handbook (Nuclear Reactors, Heat Exchangers, Steam Generators). 2-nd ed. Moscow. Energoatomizdat Publ., 1990, 360 p. (in Russian).
- Suh K.Y., Todres N.E., Rohsenow W.M. Mixed Convective Low Flow Pressure Drop in Vertical Rod Assemblies: II-Experimental Validation. Trans. of the ASME. 1989, v.111, no. 4, pp. 966-973.
Link for citing the article: Slobodchuk V.I., Uralov D.A., Avramova E.A. Possibility of Simulating Natural Circulation in Fast Neutron Reactors Using a Light Water Test Facility. Izvestiya vuzov. Yadernaya Energetika. 2021, no. 3, pp. 146-157; DOI: https://doi.org/10.26583/npe.2021.3.12 (in Russian).