Accessible Requires Authentication Published by De Gruyter October 4, 2019

Heat transfer to water near the critical point: evaluation of the ATHLET thermal-hydraulic system code

Wärmeübergang an Wasser in der Nähe des kritischen Punkts: Evaluierung des Thermohydraulik-Rechenprogramms ATHLET
T. Gschnaidtner, I. Aymerich Rodrigáñez, G. Lerchl, C. Wieland and H. Spliethoff
From the journal Kerntechnik

Abstract

The heat transfer coefficient is an essential measure in the predesign of supercritical water-cooled reactors (SCWRs). At supercritical pressures, three distinct heat transfer modes exist: normal, improved, and deteriorated. The heat transfer behavior of supercritical water in the pseudo-critical range is different from that of single-phase fluids in the subcritical range. These heat transfer modes differ from those of single-phase flow at subcritical pressures, resulting in an unusual behavior of the heat transfer coefficients. Moreover, during accidental scenarios, when the operating pressure is reduced from supercritical to subcritical conditions, a boiling crisis may occur. During pressure reduction, temporary phenomena such as superheating of the cladding temperature can endanger the safe operation of SCWRs. In order to analyze operational and accidental scenarios of SCWRs, thermal-hydraulic system codes such as ATHLET are applied. However, the prediction capabilities of thermal-hydraulic system codes rely on a comprehensive validation work based on experimental data. This study presents an extensive analysis of the applicability of ATHLET at the near-critical pressure range. ATHLET is assessed against the LESHP-database and two trans-critical transient experiments. At supercritical pressures, the heat transfer coefficient correlations are evaluated with regard to their prediction accuracy and numerical problems including the “multiple solutions problems”. The trans-critical transient experiments are used to test the prediction capability of ATHLET with respect to transient heat transfer phenomena including critical heat flux, film boiling and return to nucleate boiling.

Kurzfassung

Der Wärmeübergangskoeffizient ist essentiell bei der Auslegung von überkritischen wassergekühlten Reaktoren (SCWR) im Hinblick auf den Wärmeübergang im Reaktorkern. Das Wärmeübergangsverhalten von überkritischem Wasser weicht insbesondere im pseudo-kritischen Bereich vom typischen Wärmeübergangsverhalten einphasiger Fluide im unterkritischen Bereich ab. Im überkritischen Druckbereich untergliedert man das Wärmeübergangsverhalten deshalb in drei verschiedene Wärmeübertragungsbereiche: normal, verbessert und verschlechtert. Zudem kann bei Störfällen in SCWRs eine Druckabsenkung vom überkritischen in den unterkritischen Druckbereich erfolgen, die zu einer Siedekrise führt. Dabei können zeitlich begrenzte Phänomene wie eine Überhitzung der Rohrwand mit anschließender Wiederbenetzung auftreten. Um Betriebs- und Unfallszenarien von SCWRs zu analysieren, werden im Allgemeinen Thermohydraulik-Systemrechenprogramme wie ATHLET eingesetzt. Zur genauen Vorhersage sowie Validierung der Simulationsergebnisse stützen sich Thermohydraulik-Systemrechenprogramme auf die Ergebnisse experimenteller Untersuchungen. Entsprechend ist es Ziel dieser Studie, ATHLET im Hinblick auf den nah-kritischen Druckbereich zu testen. Dazu werden die Simulationsergebnisse mit der LESHP-Datenbank und zwei transienten Experimenten im transkritischen Druckbereich verglichen. Die überkritischen Wärmeübergangskorrelationen werden hinsichtlich ihrer Vorhersagegenauigkeit und der auftretenden numerischen Probleme einschließlich des Problems von Mehrfachlösungen untersucht. Die transkritischen transienten Experimente dienen dazu, die Anwendbarkeit der in ATHLET implementierten Modelle im Hinblick auf den kritischen Wärmestrom, Filmsieden und Rückkehr zum Blasensieden zu testen.


E-mail:

References

1 Pioro, I.; Duffey, R.: Nuclear Power as a Basis for Future Electricity Generation. Journal of Nuclear Engineering and Radiation Science1 (2015) 11001, 10.1115/1.4029420 10.1115/1.4029420 Search in Google Scholar

2 Pioro, I.; (Ed.): Handbook of generation IV nuclear reactors. Elsevier-Woodhead Publishing, Duxford, UK, 2016 Search in Google Scholar

3 Leung, L.: Overview of Global Development of SCWR Concepts. in: Joint ICTP-IAEA Course on Science and Technology of Supercritical Water Cooled Reactors, Trieste, Italy, 27 June–01 July, 2011. Search in Google Scholar

4 International Atomic Energy Agency (IAEA) (Ed.): Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs). International Atomic Energy Agency, Vienna, Austria, 2014 Search in Google Scholar

5 Duffey, R.: The Development and Future of the Supercritical Water Reactor. CNL Nuclear Review5 (2016) 181188 Search in Google Scholar

6 Schulenberg, T.; Raqué, M.: Transient heat transfer during depressurization from supercritical pressure. International Journal of Heat and Mass Transfer79 (2014) 23324010.1016/j.ijheatmasstransfer.2014.07.084 Search in Google Scholar

7 Wang, F.; Yang, J.; Gu, H.-Y.; Zhao, M.; Li, H.-B.; Lu, D.-H.: Experimental research on heat transfer performance of supercritical water in vertical tube. Yuanzineng Kexue Jishu/Atomic Energy Science and Technology47 (2013) 933939, 10.7538/yzk.2013.47.06.0933 Search in Google Scholar

8 Shen, Z.; Yang, D.; Xie, H.; Nie, X.; Liu, W.; Wang, S.: Flow and heat transfer characteristics of high-pressure water flowing in a vertical upward smooth tube at low mass flux conditions. Applied Thermal Engineering102 (2016) 39140110.1016/j.applthermaleng.2016.03.150 Search in Google Scholar

9 Mokry, S.; Pioro, I.: Improvement of Supercritical Water Heat-Transfer Correlations for Vertical Bare Tubes. in: Proceedings of the 22nd International Conference on Nuclear Engineering, Prague, Czech Republic, 07 July, 2014, V005T17A019, 10.1115/ICONE22-30137 Search in Google Scholar

10 Lei, X.; Li, H.; Zhang, W.; Dinh, N. T.; Guo, Y.; Yu, S.: Experimental study on the difference of heat transfer characteristics between vertical and horizontal flows of supercritical pressure water. Applied Thermal Engineering113 (2017) 60962010.1016/j.applthermaleng.2016.11.051 Search in Google Scholar

11 Gang, W.; Pan, J.; Bi, Q.; Yang, Z.; Wang, H.: Heat transfer characteristics of supercritical pressure water in vertical upward annuli. Nuclear Engineering and Design273 (2014) 44945810.1016/j.nucengdes.2014.03.038 Search in Google Scholar

12 Zhao, M.; Gu, H. Y.; Li, H. B.; Cheng, X.: Heat transfer of water flowing upward in vertical annuli with spacers at high pressure conditions. Annals of Nuclear Energy87 (2016) 20921610.1016/j.anucene.2015.08.011 Search in Google Scholar

13 Wang, H.; Bi, Q.; Yang, Z.; Wang, L.: Experimental and numerical investigation of heat transfer from a narrow annulus to supercritical pressure water. Annals of Nuclear Energy80 (2015) 41642810.1016/j.anucene.2015.02.029 Search in Google Scholar

14 Licht, J.; Anderson, M.; Corradini, M.: Heat transfer to water at supercritical pressures in a circular and square annular flow geometry. International Journal of Heat and Fluid Flow29 (2008) 15616610.1016/j.ijheatfluidflow.2007.09.007 Search in Google Scholar

15 Wang, H.; Bi, Q.; Leung, L. K. H.: Heat transfer from a 2 × 2 wire-wrapped rod bundle to supercritical pressure water. International Journal of Heat and Mass Transfer97 (2016) 48650110.1016/j.ijheatmasstransfer.2016.02.036 Search in Google Scholar

16 Gu, H.-Y.; Hu, Z.-X.; Liu, D.; Li, H.-B.; Zhao, M.; Cheng, X.: Experimental study on heat transfer to supercritical water in 2 × 2 rod bundle with wire wraps. Experimental Thermal and Fluid Science70 (2016) 172810.1016/j.expthermflusci.2015.08.015 Search in Google Scholar

17 Gu, H. Y.; Hu, Z. X.; Liu, D.; Xiao, Y.; Cheng, X.: Experimental studies on heat transfer to supercritical water in 2 × 2 rod bundle with two channels. Nuclear Engineering and Design291 (2015) 21222310.1016/j.nucengdes.2015.05.028 Search in Google Scholar

18 Li, H.; Hu, Z.; Zhao, M.; Gu, H.; Lu, D.: Experimental investigation on transient heat transfer in 2 × 2 bundle during depressurization from supercritical pressure. Annals of Nuclear Energy109 (2017) 23724810.1016/j.anucene.2017.05.023 Search in Google Scholar

19 Kohlhepp, A.; Schatte, G. A.; Gschnaidtner, T.; Wieland, C.; Spliethoff, H.: Experimental Investigation of Temperature Profiles Along a Heated Tube at Depressurization from Supercritical Pressures at Medium Heat and Mass Fluxes. in: Proceedings of the 15th UK Heat Transfer Conference (UKHTC2017), London, UK, 04-05 September, 2017 Search in Google Scholar

20 Köhler, W.; Hein, D.: Influence of the wetting state of a heated surface on heat transfer and pressure loss in an evaporator tube. NUREG/IA-0003, Nuclear Regulatory Commission, Washington, DC (USA). Office of Nuclear Regulatory Research Kraftwerk Union AG, Erlangen (Germany, FR), 198610.2172/5236497 Search in Google Scholar

21 Hein, D.; Kastner, W.; Köhler, W.; Krätzer, W.: Der Wärmeübergang im Post-Dryout-Gebiet während transienter Vorgänge. Wärme- und Stoffübertragung16 (1982) 25125710.1007/BF01375650 Search in Google Scholar

22 Chen, W.; Fang, X.; Xu, Y.; Su, X.: An assessment of correlations of forced convection heat transfer to water at supercritical pressure. Annals of Nuclear Energy76 (2015) 45146010.1016/j.anucene.2014.10.027 Search in Google Scholar

23 Schatte, G. A.; Kohlhepp, A.; Wieland, C.; Spliethoff, H.: Development of a new empiri-cal correlation for the prediction of the onset of the deterioration of heat transfer to supercritical water in vertical tubes. International Journal of Heat and Mass Transfer113 (2017) 1333134110.1016/j.ijheatmasstransfer.2017.04.037 Search in Google Scholar

24 Zhang, Q.; Li, H.; Lei, X.; Zhang, J.; Kong, X.: Study on identification method of heat transfer deterioration of supercritical fluids in vertically heated tubes. International Journal of Heat and Mass Transfer127 (2018) 67468610.1016/j.ijheatmasstransfer.2018.07.058 Search in Google Scholar

25 Zahlan, H. A. M.; Leung, L. K. H.; Huang, Y.-P.; Liu, G.-X.: Assessment of Convective Heat Transfer Correlations Against an Expanded Database for Different Fluids at Supercritical Pressures. Journal of Nuclear Engineering and Radiation Science4 (2018) 1100410.1115/1.4037720 Search in Google Scholar

26 Gschnaidtner, T.; Schatte, G. A.; Kohlhepp, A.; Wang, Y.; Wieland, C.; Spliethoff, H.: A new assessment method for the evaluation of supercritical heat transfer correlations, par-ticularly with regard to the “multiple/no solutions” problem. Thermal Science and Engineering Progress7 (2018) 26727810.1016/j.tsep.2018.07.006 Search in Google Scholar

27 Austregesilo, H.; Bals, C.; Hora, A.; Lerchl, G.; Romstedt, P.; Schöffel, P.; von der Cron, D.; Weyermann, F.: ATHLET 3.1A: Models and Methods. 4th ed. 3, 2016 Search in Google Scholar

28 Lerchl, G.; Austregesilo, H.; Ceuca, S.; Glaeser, H.; Luther, W.; Schöffel, P.: ATHLET 3.1A: Validation. 4th ed. 3, 2016 Search in Google Scholar

29 Fu, S. W.; Liu, X. J.; Zhou, C.; Xu, Z. H.; Yang, Y. H.; Cheng, X.: Modification and application of the system analysis code ATHLET to trans-critical simulations. Annals of Nuclear Energy44 (2012) 404910.1016/j.anucene.2012.02.005 Search in Google Scholar

30 Zhou, C.; Yang, Y.; Cheng, X.: Feasibility analysis of the modified ATHLET code for supercritical water cooled systems. Nuclear Engineering and Design250 (2012) 60061210.1016/j.nucengdes.2012.06.021 Search in Google Scholar

31 Samuel, J.; Lerchl, G.; Harvel, G. D.; Pioro, I.: Investigation of ATHLET System Code for Supercritical Water Applications. in: Proceedings of the 22nd International Conference on Nuclear Engineering, Prague, Czech Republic, 07 July, 2014, V005T17A018, 10.1115/ICONE22-30136 Search in Google Scholar

32 Hegyi, G.; Keresztúrí, A.; Trosztel, I.; György, H.: Evaluation of Different Heat Transfer Correlations of the Supercritical Water by the ATHLET Code. in: Proceedings of the 22nd International Conference Nuclear Energy for New Europe (NENE), Bled, Slovenia, 09–12 September, 2013 Search in Google Scholar

33 Watts, M. J.; Chou, C. T.: Mixed convection heat transfer to supercritical pressure water. in: Heat Transfer 1982: Proceedings of the 7th International Heat Transfer Conference, 3rd ed., Munich, Germany, 06–10 September, 1982, pp. 49550010.1615/IHTC7.2970 Search in Google Scholar

34 Schatte, G. A.; Kohlhepp, A.; Gschnaidtner, T.; Wieland, C.; Spliethoff, H.: Heat Transfer to Supercritical Water in Advanced Power Engineering Applications: An Industrial Scale Test Rig. J. Energy Resour. Technol140 (2018) 6200210.1115/1.4039610 Search in Google Scholar

35 Cheng, X.; Yang, Y. H.; Huang, S. F.: A simplified method for heat transfer prediction of supercritical fluids in circular tubes. Annals of Nuclear Energy36 (2009) 1120112810.1016/j.anucene.2009.04.016 Search in Google Scholar

36 Zhao, M.; Gu, H. Y.; Cheng, X.: Experimental study on heat transfer of supercritical water flowing downward in circular tubes. Annals of Nuclear Energy63 (2014) 33934910.1016/j.anucene.2013.07.003 Search in Google Scholar

37 Gupta, S.; Farah, A.; King, K.; Mokry, S.; Pioro, I.: Developing New Heat-Transfer Correlation for SuperCritical-Water Flow in Vertical Bare Tubes. in: Proceedings of the 18th International Conference on Nuclear Engineering (ICONE18), 17–21 May, 2010, pp. 80981710.1115/ICONE18-30024 Search in Google Scholar

38 Mokry, S.; Pioro, I.; Farah, A.; King, K.; Gupta, S.; Peiman, W.; Kirillov, P.: Development of supercritical water heat-transfer correlation for vertical bare tubes. Nuclear Engineering and Design241 (2011) 1126113610.1016/j.nucengdes.2010.06.012 Search in Google Scholar

39 Bromley, L. A.; LeRoy, N. R.; Robbers, J. A.: Heat Transfer in Forced Convection Film Boiling. Ind. Eng. Chem.45 (1953) 2639264610.1021/ie50528a027 Search in Google Scholar

40 Berenson, P. J.: Film-Boiling Heat Transfer From a Horizontal Surface. Journal of Heat Transfer83 (1961) 35110.1115/1.3682280 Search in Google Scholar

41 Biasi, L.; Clerici, G. C.; Garribba, S.; Sala, R.; Tozzi, A.: Studies on burnout. Part 3. A new correlation for round ducts and uniform heating and its comparison with world data. Energ. Nucl.14 (1967) Search in Google Scholar

42 Groeneveld, D. C.; Shan, J. Q.; Vasić, A. Z.; Leung, L. K. H.; Durmayaz, A.; Yang, J.; Cheng, S. C.; Tanase, A.: The 2006 CHF look-up table. Nuclear Engineering and Design237 (2007) 1909192210.1016/j.nucengdes.2007.02.014 Search in Google Scholar

43 Groeneveld, D. C.; Moeck, E. O.: An Investigation of heat transfer in the liquid deficient regime. Atomic Energy of Canada Ltd., Chalk River (Ontario). Chalk River Nuclear Labs, 1969 Search in Google Scholar

44 Jackson, J. D.; Hall, W. B.: Forced convection heat transfer to fluids at supercritical pressure. in: Kakaç, S.; Spalding, D.B. (Eds.): Turbulent Forced Convection in Channels and Bundles. Hemisphere Publishing Corp, Washington, DC, USA, 1979, pp. 563611 Search in Google Scholar

Received: 2019-03-29
Published Online: 2019-10-04
Published in Print: 2019-10-14

© 2019, Carl Hanser Verlag, München