Modeling of the heat transfer process in steam generation in a vertical pipe
Authors: Satin A.A., Savelyev R.S.
Published in issue: #3(75)/2018
DOI: 10.18698/2308-6033-2018-3-1746
Category: Power, Metallurgic and Chemical Engineering | Chapter: Nuclear Reactor Engineering, Machines, Assemblies, and Nuclear Materials Technology
At present, different software complexes of computational fluid dynamics (ANSYS CFX, FLUENT, STAR CCM +, etc.) allow the calculation of the boiling process. However, the issues of application of boiling process models, their verification as well as their comparison with current engineering calculation methods used for solving practical problems have not been adequately studied. Engineering methods for calculating of boiling in nuclear power engineering allow estimating the vapor content for a heat exchange surface of simple geometry. There are practically no ways for determination of the vapor content, the structure wall temperature, and other parameters in complex three-dimensional geometry. In this paper, the hydro-gas-dynamic ANSYS CFX package was used to simulate the generation of steam in a vertical pipe. The results of calculation performed in the ANSYS CFX software environment are compared with calculations based on the criterial dependencies and on the experimental data. To substantiate the three-dimensional calculation, the Rensselaer Polytechnic Institute (RPI) Wall Boiling model is used. This model is used in all software complexes of computational fluid dynamics for simulating the boiling process. Verification of this model will make it possible to obtain a qualitative picture of the coolant parameter distribution during boiling for the heat exchange surface of a complex three-dimensional geometry.
References
[1] Solonin V.I. Teplogidravlicheskie protsessy v aktivnykh zonakh vodookhlazhdaemykh reaktorov [Thermohydraulic processes in the active zones of water-cooled reactors]. Moscow, BMSTU Publ., 2012, 140 p.
[2] Yemelyanov I.Ya., Mikhan V.I., Solonin V.I. Konstruirovanie yadernykh reaktorov [Nuclear reactor design]. N.A. Dollezhal, ed. Moscow, Energoizdat Publ., 1982, 400 p.
[3] Smolin V.N. Issledovanie teplogidravliki aktivnykh zon yadernykh reaktorov na modelnykh stendakh [Investigation of active zone thermohydraulics of nuclear reactors on model stands]. Moscow, FGUP NIKIET Publ., 2005, 150 p.
[4] Kirillov P.L., Yuryev Yu.S., Bobkov V.P. Spravochnik po teplogidravlicheskim raschetam: yadernye reaktory, teploobmenniki, parogeneratory [Handbook of thermohydraulic calculations: nuclear reactors, heat exchangers, steam generators]. Moscow, Energoatomizdat Publ., 1990, 360 p.
[5] Butterworth D., Hewitt G.F. Two-phase flow and heat transfer. New York, Oxford University Press Publ., 1977 [In Russ.: Butterworth D., Hewitt G. Teploperedacha v dvukhfaznom potoke. Moscow, Energiya Publ., 1980, 328 p.].
[6] Petukhov B.S., Genin L.G., Kovalev S.A., Solovyev S.L. Teploobmen v yadernykh energeticheskikh ustanovkakh [Heat transfer in nuclear power plants]. Moscow, MEI Publ., 2003, 528 p.
[7] Bartolomei G.G., Brantov V.G., Molochnikov Yu.S. Teploenergetika — Thermal Engineering, 1982, no. 3, pp. 20–22.
[8] ANSYS CFX-Solver Theory Guide. SAS IP, Inc ., 2011, 402 p.
[9] IAPWS-IF97. Release on the IAPWS Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam. The International Association for Properties of Water and Steam, Erlangen, September 1997. 48 p.