Journal of computational fluids engineering (한국전산유체공학회지)

Korean Society of Computational Fluids Engineering (한국전산유체공학회)
권호 표지
DOI QR코드

DOI QR Code

THE DEVELOPMENT AND ASSESSMENT STRATEGY OF A THERMAL HYDRAULICS COMPONENT ANALYSIS CODE

열수력 기기해석용 CUPID 코드 개발 및 평가 전략

  • Received : 2011.04.27
  • Accepted : 2011.05.18
  • Published : 2011.06.30
Download PDF
⟨ Previous Next ⟩

Abstract

A three-dimensional thermal-hydraulic code, CUPID, has been developed for the analysis of transient two-phase flows at component scale. The CUPID code adopts a two-fluid three-field model for two-phase flows. A semi-implicit two-step numerical method was developed to obtain numerical solutions on unstructured grids. This paper presents an overview of the CUPID code development and assessment strategy. The governing equations, physical models, numerical methods and their improvements, and the systematic verification and validation processes are discussed. The code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER, are also presented.

Keywords

References

  1. 1996, Proc. OECD/CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements, Annapolis, MD, USA, 5-8 November 1996, NUREG/CP-0159, NEA/CSNI/R(97)4.
  2. 2003, Yadigaroglu, G., Andreani, M., Dreier, J. and Coddinggton, P., "Trends and needs in experimentation and numerical simulation for LWR safety," Nuclear Engineering and Design, Vol.221, pp.205-223. https://doi.org/10.1016/S0029-5493(02)00339-4
  3. 2010, Bestion, D., "Extension of CFD codes application to two-phase flow safety problems," Nuclear Engineering and Technology, Vol.42, pp.365-376. https://doi.org/10.5516/NET.2010.42.4.365
  4. 2005, Yadigaroglu, G., "Computational fluid dynamics for nuclear applications: From CFD to multi-scale CMFD," Nuclear Engineering and Design, Vol.235(2-4), pp.153-164. https://doi.org/10.1016/j.nucengdes.2004.08.044
  5. 2005, Staetke, H., et al, "Advanced three-dimensional two-phase flow simulation tools for application to reactor safety (ASTAR)," Nuclear Design and Design, Vol.235, pp.379-400. https://doi.org/10.1016/j.nucengdes.2004.08.052
  6. 2007, Guelfi, A., et al., "NEPTUNE: a new software platform for advanced nuclear thermal hydraulics," Nuclear Science and Engineering, Vol.156(3), pp.281-324. https://doi.org/10.13182/NSE05-98
  7. 2005, Bestion, D., "Status and perspective of two-phase flow modeling in the NEPTUNE multiscale thermal-hydraulic platform for nuclear reactor simulation," Nuclear Engineering and Technology, Vol.37,511-524.
  8. 2010, Niceno, B., Sato, Y., Badillo, A., M. Andreani, M., "Multi-scale modeling and analysis of convective boiling: Towards the prediction of CHF in rod bundles," to be published in the December issue of Nuclear Engineering and Technology.
  9. 2010, Hosoi, H., Yoshida, H., "Numerical simulation of air-water two-phase flow in 38 mm diameter pipe by advanced two-fluid model including effects of turbulent diffusion on bubbles," The7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety(NTHAS7), November14-17, Chuncheon, Korea.
  10. 2008, Jeong, J.J., Yoon, H.Y., Park, I.K., Cho, H.K., and Kim, J., "A Semi-implicit numerical scheme for transient two-phase flows on unstructured grids," Nuclear Engineering and Design, Vol.238, pp.3403-3412. https://doi.org/10.1016/j.nucengdes.2008.08.017
  11. 2010, Jeong, J.J., Yoon, H.Y., Park, I.K., Cho, H.K., Lee, H., "Development and preliminary assessment of a three-dimensional thermal hydraulics code, CUPID," Nuclear Engineering and Technology, Vol.42, pp.279-296. https://doi.org/10.5516/NET.2010.42.3.279
  12. 2006, Yun, B.J., Euh, D.J., and Song, C.H., "Investigation of the downcomer boiling phenomena during the reflood phase of a postulated large-break LOCA in the APR1400," Nuclear Technology, Vol.156, pp.56-68.
  13. 2006, Ishii, M. and Hibiki, T., Thermo-Fluid Dynamics of Two-Phase Flow, Springer.
  14. 2001, The RELAP5-3D Code Development Team, RELAP5-3D code manual volume I: Code structure, system models and solution methods, Idaho National Engineering and Environmental Laboratory.
  15. 1999, Jeong, J.J., Ha, K.S., Chung, B.D., and Lee, W.J., "Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1," Annals of Nuclear Energy, Vol.26(18), pp.1611-1642. https://doi.org/10.1016/S0306-4549(99)00039-0
  16. 1990, Barre, F., Bernard, M., "The CATHARE code strategy and assessment," Nuclear Engineering and Design, Vol.124, pp.257-284. https://doi.org/10.1016/0029-5493(90)90296-A
  17. 2003, Frepoli, C., Mahaffy, J.H., and Ohkawa, K., "Notes on the implementation of a fully-implicit numerical scheme for a two-phase three-field flow model," Nuclear Engineering and Design, Vol.225, pp.191-217. https://doi.org/10.1016/S0029-5493(03)00159-6
  18. 2005, Tentner, A., et al., "Computational fluid dynamics modeling of two-phase flow in a boiling water reactor fuel assembly," Proc.of Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications, Avignon, France,.
  19. 2006, CD-adapco, "STAR-CD Version 4.0 Methodology Manual", Chapter 13, CD-adapco, UK.
  20. 2006, Tentner, A., et al., "Advances in computational fluid dynamics modeling of two-phase flow in a boiling water reactor fuel assembly," Proc. of ICONE14, Miami, USA.
  21. 2008, Tentner, A., et al., "Computational fluid dynamics modeling of two-phase flow topologies in a boiling water reactor fuel assembly," Proc. of ICONE16, Orlando, USA.
  22. 2007, Ioilev, A., et al., "Advances in the modeling of cladding heat transfer and critical heat flux in boiling water reactor fuel assemblies," Proc. of NURETH-12, Pittsburgh, USA.
  23. 2008, Ustinenko, V., et al., "Validation of CFD-BWR, a new two-phase computational fluid dynamics model for boiling water reactor analysis," Nuclear Engineering and Design, Vol.238, pp.660-670. https://doi.org/10.1016/j.nucengdes.2007.02.046
  24. 2002, Yoneda, K., et al., "Flow structure and bubble characteristics of steam-water two-phase flow in a large-diameter pipe," Nuclear Engineering and Design, Vol.217, pp.267-281. https://doi.org/10.1016/S0029-5493(02)00157-7
  25. 1983, Kataoka, I., Ishii, M., and Mishima, K., "Generation and size distribution of droplet in annular two-phase flow," Trans. ASME, J. Fluid Engineering, Vol.105, pp.230-238. https://doi.org/10.1115/1.3240969
  26. 2008, Coste, P., et al., "Status of a two-phase CFD approach to the PTS issue," XCFD4NRS Workshop, Grenoble, France.
  27. 1979, Drew, D., Cheng, L.Y., Lahey, R.T., "The analysis of virtual mass effect in two-phase flow," Int. J. Multiphase Flow, Vol.5, pp.233-272. https://doi.org/10.1016/0301-9322(79)90023-5
  28. 2002, Tomiyama, A., Tamia, H., Zun, I., and Hosokawa, S., "Transverse migration of single bubbles in simple shear flows," Chemical Engineering Science, Vol.57, pp.1849-1858. https://doi.org/10.1016/S0009-2509(02)00085-4
  29. 1991, Antal, S.P., Lahey, R.T., Flaherty, J.E., "Analysis of phase distribution in fully developed laminar bubbly two-phase flow," International Journal of Multiphase Flow, Vol.7, pp.635-652.
  30. 1993, Lahey, R.T., Lopez de Bertodano, M., and O.C. Jones, O.C., "Phase distribution in complex geometry conduits," Nuclear Engineering and Design, Vol.141, pp.177-201. https://doi.org/10.1016/0029-5493(93)90101-E
  31. 1998, Lopez de Bertodano, M., "Two fluid model for two-phase turbulent jets," Nuclear Engineering and Design, Vol.179, pp.65-74. https://doi.org/10.1016/S0029-5493(97)00244-6
  32. 1986, Michiyoshi, I., and Serizawa, A., "Turbulence in two-phase bubbly flow," Nuclear Engineering and Design, Vol.95, pp.253-267. https://doi.org/10.1016/0029-5493(86)90052-X
  33. 1981, Sato, Y., Sadatomi, M., Sekoguchi, K., "Momentum and heat transfer in two-phase bubble flow - I," Int. J. Multiphase Flow, Vol.7, pp.167-177. https://doi.org/10.1016/0301-9322(81)90003-3
  34. 2010, Yun, B.J., Splawski, A., Lo, S., and Song, C.-H., "Prediction of a subcooled boiling flow with mechanistic wall boiling and bubble size models," Proc. CFD4NRS-3, Washington D.C., USA, 14-16 September.
  35. 1997, AEA, CFX-4 Solver Manual, UK.
  36. 1969, Cole, R., and Rosenhow, W., "Correlation of bubble departure diameters for boiling of saturated liquids," Chem. Eng. Prog. Symp. Ser. Vol.65(92), pp.211-213.
  37. 1978, Liles, D.R., and Reed, W.H., "A semi-implicit method for two-phase fluid dynamics," J. Comput .Phys., 26, 390-407. https://doi.org/10.1016/0021-9991(78)90077-3
  38. 2010, Yoon, H.Y. and eong, J.J., "A continuity based semi-implicit scheme for transient two-phase flows," Journal of Nuclear Science and Technology, Vol.47, No.9, pp.779-789. https://doi.org/10.1080/18811248.2010.9711654
  39. 2009, Park, I.K., Cho, H.K., Yoon, H.Y., and Jeong, J.J., "Numerical effects of the semi-conservative form of momentum equations for multi-dimensional two-phase flows," Nuclear Engineering and Design, 239, 2365-2371 (2009). https://doi.org/10.1016/j.nucengdes.2009.06.011
  40. 2010, Cho, H.K., Lee, H.D., Park, I.K., Jeong, J.J., "Implementation of a second-order upwind method in a semi-implicit two-phase flow code on unstructured meshes," Annals of Nuclear Energy, Vol.37, 606-614. https://doi.org/10.1016/j.anucene.2009.12.015
  41. 2010, Lee, H., Jeong, J.J., Cho, H.K, Yoon, H.Y., "An improved numerical scheme to evaluate the pressure gradient on unstructured meshes for two-phase flow analysis," International Communications in Heat and Mass Transfer, Vol.37, pp.1273-1279. https://doi.org/10.1016/j.icheatmasstransfer.2010.07.008
  42. 1994, Frink, N.T., "Recent progress toward a three-dimensional unstructured Navier-Stokes flow solver," AIAA Paper 94-0061.
  43. 1989, Barth, T.J., and Jesperson, D.C., "The design and application of upwind schemes on unstructured meshes," AIAA Paper 89-0366.
  44. 1995, Venkatakrishnan, V., "Convergence to steady state solutions of the Euler equations on unstructured grids with limiters," Journal of Computational Physics, Vol.118, pp.120 -130. https://doi.org/10.1006/jcph.1995.1084
  45. 1999, Weiss, J.M., "Implicit solution of preconditioned Navier-Stokes equations using algebraic multigrid," AIAAJ., Vol.37, No.1.
  46. 1997, Mathur, S.R., "A pressure-based method for unstructured meshes," Numerical Heat Transfer, Part B 31, pp.195-215. https://doi.org/10.1080/10407799708915105
  47. 1992, Lee, S.Y., Jeong, J.J., Kim, S.H., and Chang, S.H., "COBRA/RELAP5; A merged version of the COBRA-TF and RELAP5/MOD3 codes," Nuclear Technology, Vol.99, pp.177-187.
  48. 2000, Boer, R., Finnemann, H., and Knoll, A., "MSLB exercise 2: 3-D kinetics results with RELAP5/PANBOX," Proc. Int. Topl. Mtg. Advances in Reactor Physics and Mathematics and Computation into the Next Millennium(PHYSOR2000), Pittsburgh, Pennsylvania, May 7 -12, 2000, American Nuclear Society.
  49. 2003, Joo, H.G., Jeong, J.J., Cho, B.O., Lee, W.J., and Zee, S.Q., "Analysis of the OECD MSLB benchmark problem using the refined core thermal-hydraulic nodalization feature of the MARS/MASTER code," Nuclear Technology, Vol.142, pp.166-179. https://doi.org/10.13182/NT142-166
  50. 2010, Jeong, J.J., Lee, S.W., Cho, J.Y., Chung, B.D., and Lee, G.C., "A coupled analysis of system thermal-hydraulics and three-dimensional reactor kinetics for a 12-finger control element assembly drop event in a PWR plant," Annals of Nuclear Energy, Vol.37, pp.1580-1587. https://doi.org/10.1016/j.anucene.2010.06.004
  51. 2010, Lee, J.R., Cho, H.K., Jeong, J.J,. "Coupled simulation of component thermal hydraulics and neutron kinetics for a nuclear reactor core with CUPID and MASTER," Transactions of the Korean Nuclear Society Autumn Meeting, Jeju, Korea, October 21-22.
  52. 2009, Cho, H.K., Park, I.K., Lee, H.D., Jeong, J.J., "Constitutive relations and a second-order interpolation scheme for the semi-implicit two-phase flow solver, CUPID," Proceedings of the 17th International Conference on Nuclear Engineering (ICONE17), July 12-16, Brussels, Belgium.
  53. 2008, Nagatake, T., Kawara, Z., and Kunugi, T., "Establishment of experimental data base on dam-breaking problem for validating interface tracking methods," NTHAS6: Sixth Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety, Okinawa, Japan, Nov. 24-27.
  54. 2005, Huh, B.K., Experimental and Analytical Study of Interfacial Are Transport Phenomena in a Vertical Two-phase Flow, Ph.D. Thesis, Seoul National University, Korea.
  55. 2010, Euh, D.J., et al., "Transport of local two-phase parameters in vertical air/water flow for bubbly and slug flow regime," Transactions of the Korean Nuclear Society Spring Meeting, Pyeongchang, Korea, May 27-28.
  56. 2010, Cho, H.K., Yun, B.J., Park, I.K., and Jeong, J.J. "Computational analysis of downcomer boiling phenomena using a component thermal hydraulic analysis code CUPID," Proc. the 18th International Conference on Nuclear Engineering, Xi'an, China, May 17-21.