Farklı Çalışma Parametrelerinde Hesaplamalı Akışkanlar Dinamiği Simülasyonu ile Kritik Isı Akısının Tahmini

Çekirdekli kaynamadan ayrılma, ısı akısı değeri, kritik ısı akısı denilen sınır değere ulaştığında meydana gelir ve ısı transferinin azalması ile ısıtılan yüzeyin zarar görmesine neden olabilecek ani sıcaklık artışına neden olur. Bu çalışmada dikey bir borudaki aşırı soğutulmuş akışta kritik ısı akısının meydana geldiği çekirdekli kaynamadan ayrılma koşulları hesaplamalı akışkanlar dinamiği simülasyonu tahmin edilmiştir. Gelişmiş duvar kaynama modeli ANSYS Fluent 2019 R3 yazılımı ile uygulanmıştır. Sabit bir basınçta değişen kütle akısı ve giriş sıcaklığı koşullarında simülasyonlar gerçekleştirilmiştir. Modeli doğrulamak için mevcut literatürden deneysel veriler toplanmış ve simülasyon sonuçları ile karşılaştırılmıştır. Elde edilen HAD sonuçlarına göre kritik ısı akısı değerinin aşırı soğutma seviyesi ve kütle akısı ile arttığı belirlenmiştir. Hesaplanan kritik ısı akısı değerleri ile deneysel veriler arasındaki ortalama sapma %16’dır. Elde edilen sonuçlar uygulanan modelin kritik ısı akısını tahmin etmede başarılı olduğunu göstermiştir.

Estimation of Critical Heat Flux in Different Operating Parameters by Computational Fluid Dynamics Simulation

Departure from nucleate boiling occurs when the heat flux value reaches the limit value called the critical heat flux, and it causes a sudden increase in temperature that can cause damage to the heated surface with the decrease of heat transfer. In this study, a computational fluid dynamics simulation of departure from nucleate boiling where critical heat flux occurs in a subcooled flow in a vertical pipe has been predicted. An improved wall boiling model has been applied using ANSYS Fluent 2019 R3 software. Simulations are carried out under varying mass flux and inlet temperature conditions at a constant pressure. Experimental data were collected from the literature to validate the model and the simulation results were compared with the experimental data. According to the simulation results, it was determined that the critical heat flux value increased with the subcooling level and mass flux. The average deviation between the calculated critical heat flux values and the experimental data is 16%. The results proved that applied model was qualified to predict critical heat flux accurately.

PDF

___

Azhar, M., 2019. Numerical Study of Nucleate Boiling Flows Using ANSYS Fluent. Proceedings of the 4th World Congress on Momentum, Heat and Mass Transfer. Paper No ICMFHT 114.
Bartolemei, G.G. and Chanturiya, V.M., 1967. Experimental study of true void fraction when boiling subcooled water in vertical tubes. Thermal Engineering, 14, 123-128.
Bartolemei, G.G. and Gorburov, V.I., 1969. Experimental study of vapour phase condensation in liquid subcooled below saturation temperature. Heat Production, 12, 58-62.
Becker, K.M., Ling, C.H., Hedberg S. and Strand G., 1983. An experimental investigation of post dryout heat transfer. KTH-NEL-33.
Burns, A. D., Frank, T., Hamill, I. and Shi, J. M., 2004. The Favre averaged drag model for turbulent dispersion in Eulerian multi-phase flows. 5th International Conference on Multiphase Flow. Paper No 392.
Celata, G. P., Cumo, M. and Mariani, A., 1993. Burnout in highly subcooled water flow boiling in small diameter tubes. International Journal of Heat and Mass Transfer, 36(5), 1269–1285.
Chen, Y., 2012. An Overview of Heat Transfer Phenomena. Kazi, S.N., InTechOpen, 193-218.
Cole, R., 1960. A photographic study of pool boiling in the region of the critical heat flux. AIChE Journal, 6(4), 533–538.
Del Valle, V. H. and Kenning, D. B. R., 1985. Subcooled flow boiling at high heat flux. International Journal of Heat and Mass Transfer, 28, 1907–1920.
Dong, X., Zhang, Z., Liu, D., Tian, Z. and Chen, G., 2018. Numerical Investigation of the Effect of Grids and Turbulence Models on Critical Heat Flux in a Vertical Pipe. Frontiers in Energy Research, 6(58), 1-11.
Filho, F. A. B., Ribeiro, G. B. and Caldeira A. D., 2015. A verification and validation of the new implementation of subcooled flow boılıng in a CFD code. International Nuclear Atlantic Conference.
Guerrero, E., Muñoz, F. and Ratkovich, N. (2017). Comparison between Eulerian and VOF models for two-phase flow assessment in vertical pipes. CT&F - Ciencia, Tecnología y Futuro, 7(1), 73 - 84.
Ishii, M., 1987. Two-fluid model for two phase flow. International Workshop on Two-phase Flow Fundamentals.
Kim, S. J., Rao, D. V., Okhuysen, B., Johns, R. and Baglietto, E., 2016. A CFD simulation effort on the Departure from Nucleate Boiling ( DNB ) in sub-cooled flow. OECD/NEA & IAEA Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety.
Kim, S. J., 2018. A status review on DNB prediction using CASL baseline boiling model and possible suggestions regarding wall boiling closures. United States.
Kurul, N. and Podowski, M. Z., 1991. On the modeling of multidimensional effects in boiling channels. Proceedings of the 27th National Heat Transfer Conference.
Lavieville, J., Quemerais, E., Mimouni, S., Boucker, M. and Mechitoua, N., 2006. NEPTUNE CFD V1.0 Theory Manual, EDF.
Lemmert, M. and Chawla, L. M., 1977. Heat Transfer Boiling. Hahne, E. and Grigull, U., Academic Press.
Moraga, F. J., Bonetto, F. J. and Lahey, R. T., 1999. Lateral forces on spheres in turbulent uniform shear flow. International Journal of Multiphase Flow, 25(6–7), 1321–1372.
Pope, S. B., 2000. Turbulent Flows. Cambridge: Cambridge University Press. 802. Ranz, W. E. and Marshall, W. R., 1952. Evaporation from drops. Chemical Engineering Progress, 48, 141–146.
Tentner, A., Lo, S., Loilev, A., Melnikov, V., Samigulin, M., Ustinenko, V. and Kozlov, V., 2006. Advances in computational fluid dynamics modeling of two-phase flow in a boiling water reactor fuel assembly. Proceedings of International Conference on Nuclear Engineering.
Tolubinski, V. I. and Kostanchuk, D. M., 1970. Vapor bubbles growth rate and heat transfer intensity at subcooled water boiling. Proceedings of the 4th International Heat Transfer Conference.
Tong, L. S., 1967. Prediction of departure from nucleate boiling for an axially non-uniform heat flux distribution. Journal of Nuclear Energy, 21(3), 241–248.
Tong, L. S. and Tang, Y.S., 1997. Boiling Heat Transfer and Two-Phase Flow, Second Edition. Hewitt, G.F. and Tien, C.L., Taylor and Francis, 1-5.
Vyskocil, L. and Macek, J., 2012. CFD simulation of critical heat flux in a tube. Computational Fluid Dynamics (CFD) for Nuclear Reactor Safety Applications - Workshop Proceedings.
Vyskocil, L. and Macek, J., 2015. CFD simulation of the departure from nucleate boiling. International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 3915–3927.
Waterhead, R. J., 1963. Nucleate Boiling Characteristics and the Critical Heat Flux Occurrence in Subcooled Axial-Flow Water Systems. United States.
Weisman, J. And Pei, B. S., 1983. Prediction of Critical Heat Flux in Flow Boiling at Low Qualities. International Journal of Heat and Mass Transfer, 26(10), 1463–1477.
Yakhot, V. and Orszag, S. A., 1986. Renormalization group analysis of turbulence I Basic theory. Journal of Scientific Computing, 1(1), 3–51.
1-https://ansyshelp.ansys.com/account/secured?returnurl=/Views/Secured/corp/v201/en/flu_th/flu_th.html, (14.03.2022).