Development of WCOBRA/TRAC-TF2 computer Code: Coupling of the 3D module (COBRA-TF) with the 1D module of TRAC-PF1/MOD2

International Conference on Nuclear Engineering, Proceedings, ICONE

Volumn 3, Issue , 2009, Pages 829-841

  Frepoli, Cesare ^a   Ohkawa, Katsuhiro ^a   Nissley, Mitch ^a  

^a WESTINGHOUSE ELECTRIC COMPANY   (United States)

Author keywords

[No Author keywords available]

Indexed keywords

COBRA-TF; COMPUTER CODES; D-MODULES; DRIFT-FLUX MODEL; FULL SPECTRUM; PACIFIC NORTHWEST; REALISTIC SIMULATION; THERMAL-HYDRAULIC PHENOMENA; TWO-FLUID; WESTINGHOUSE;

MULTIPHASE FLOW; NUCLEAR ENGINEERING; NUCLEAR INDUSTRY; REACTOR CORES;

LOSS OF COOLANT ACCIDENTS;

EID: 77952791820     PISSN: None     EISSN: None     Source Type: Conference Proceeding    
DOI: 10.1115/ICONE17-76044     Document Type: Conference Paper

References (37)

1. Anklam, T. M., et al., 1982, "Experimental Investigations of Uncovered Bundle Heat Transfer and Two Phase Mixture Level Swell Under High Pressure Low Heat Flux Conditions," NUREG/CR 2456.
2. H. ASAKA, at. al, "Improvement of TRAC-PF1 Interfacial Drag Model for Analysis of High-Pressure Horizontally-Stratified Two-Phase Flow," Journal of Nuclear Science and Technology. Vol. 28, No. 1, pp. 33-44 (1991).
3. Bajorek, S. M., et al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis," Technical Report WCAP-12945-NP-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, Westinghouse Electric Company, Pittsburgh, PA.
4. Downar-Zapolski, P., Bilicki, Z., Bolle, L. and Franco, J. "The Non-equilibrium Relaxation Model for One-dimensional Flashing Liquid Flow," International Journal of Multiphase Flow, Vol. 22, No. 3, pp. 473-486, 1996. (Pubitemid 126380264)
5. EPRI NP-1692, January 1981, "Heat Transfer Above the two-Phase Mixture Level Under Core Uncovery Conditions in a 336-Rod Bundle," Volumes 1 and 2.
6. Findlay, J. A. and Sozzi, G. L., 1981, "BWR Refill Reflood Program B Model Qualification Task Plan, NUREG/CR 1899.
7. C. Frepoli, "Need of a Coherent and Consistent Phenomena Identification and Ranking Table (PIRT) To Address Small, Intermediate and Large Break LOCA in PWRs", Transaction American Nuclear Society, Winter meeting and Nuclear technology Expo, November 12-16, 2006, Albuquerque, NM.
8. C. FREPOLI, "Review Article - An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model," Science and Technology of Nuclear Installations, Vol. 2008, Hindawi Publishing Corporation, Article ID 498737.
9. (Frepoli et. al., 2008b) - C. Frepoli, J. Liao and K. Ohkawa, "An Improved Horizontal Flow Regime Map For The 1-D Module of WCOBRA/TRAC-TF2", Proceedings of ICAPP '08 Anaheim, CA USA, June 8-12, 2008 - Paper 8369
10. Guo, Z., et al., December 1993, "Critical Heat Flux for Rod Bundle Under High-Pressure Boil-off Conditions," JAERI-M 93-238.
11. Hewitt, G.F., Delhaye, J.M., Zuber, N., "Multiphase Science and Technology," Vol.6, 1992.
12. Kukita, Y. et al., 1990, Loop Seal Clearing and Refilling During a PWR Small Break LOCA, Nuclear Eng. and Design, 121, 431-440.
13. KWU, 2D/3D Program Upper Plenum Test Facility Experimental Data Report, 1988, "Test No. 8, Cold/Hot Leg Flow Pattern Test," U9 316/88/12, KWU.
14. J. LIAO and C. Frepoli., "Development of a condensation heat transfer model for the analysis of ECC injection in Cold Leg during Small and Intermediate Break LOCA," Proc. Of the 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12), Log Number: 170 Sheraton Station Square, Pittsburgh, Pennsylvania, U.S.A. September 30-October 4, 2007.
15. Liles, D. R., et al., 1981, "TRAC-PD2, An Advanced Best Estimate Computer Program for Pressurized Water Reactor Loss-of-Coolant Accident Analysis," NUREG/CR-2054, Los Alamos National Laboratories, NM.
16. J. M. MANDHANE, G. A. Gregory, and K. Aziz, "A Flow Pattern Map for Gas Liquid Flow in Horizontal Pipes," Int. J. Multiphase Flow, 1, pp. 537-553 (1974).
17. Y. M. MOON and H. C. No, "Off-Take and Slug Transition at T-Junction of Vertical-Up Branch in the Horizonthal Pipe," Journal and Nuclear Science and Technology, Vol. 40, No 5, pp. 317-324, 2003.
18. H. NAKAMURA, Y. Anoda and Y. Kukita, "Flow Regime Transition in High-Pressure Steam-Water Horizontal Pipe Two-Phase Flow," ANS Proceeding of 1991 National Heat Transfer Conference, Minneapolis, 269-276, 1991.
19. H. NAKAMURA, Y. Kukita and K. Tasaka, "An Evaluation of Bernoulli Effect on Slugging in Horizontal Two-Phase Flow," Journal of Nuclear Science and Technology, Vol. 31, No. 1, pp. 113-121, 1994.
20. H. NAKAMURA, Y. Kukita and K. Tasaka, "Flow Regime Transition to Wavy Dispersed Flow for High-Pressure Steam/Water Two-Phase Flow in Horizontal Pipe," Journal of Nuclear Science and Technology, Vol. 32, No. 7, pp. 641-652, (1995)
21. H. NAKAMURA, Y. Kukita and K. Tasaka, "Interfacial Friction Factor for High-Pressure Steam/Water Stratified-Wavy Flow in Horizontal Pipe," Journal of Nuclear Science and Technology, Vol. 32, No. 9, pp. 868-879, (1995)
22. NUREG-0800, "Standard Review Plan - 15.0.2 Review of Transient and Accident Analysis Methods", December 2005.
23. (Ohkawa, 2007a) - Ohkawa, K., "ASSESSMENT OF HOMOGENEOUS NON-EQUILIBRIUM RELAXATION CRITICAL FLOW MODEL," 15th International Conference on Nuclear Engineering, ICONE15-10708, Nagoya, Japan, April 22-26, 2007.
24. (Ohkawa, 2007b) - Ohkawa, K., "DEVELOPMENT OF HOMOGENEOUS NON-EQUILIBRIUM RELAXATION CRITICAL FLOW MODEL WITH NON-CONDENSABLE GAS," The 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-12) Pittsburgh, Pennsylvania, USA, September 30-October 4, 2007.
25. Paik, C.Y. and Hochreiter, L.E., 1986, "Analysis of FLECHT-SEASET 163 Rod Bundle Data Using COBRA-TF," NUREG/CR-4166.
26. Sackett, Kenneth E. and Clegg, L. Bruce, "Experiment Data Report For Semiscale MOD-3 Small Break Test S-07-10D (Baseline Test Series), December 1980, prepared for the U.S. Nuclear Regulatory Commission under Department of Energy Contract No. DE-AC07-76IDO1570.
27. C. SMOGLIE and J. Reimann, "Two-Phase Flow Through Small Branches in Horizontal Pipe With Stratified Flow," Int. J. Multiphase Flow, Vol. 12, N. 4, pp. 609-625 (1986).
28. J. W. SPORE, J. S. Elson, S. J. Jolly Woodruff, T. D. Knight, J. C. Lin, R. A. Nelson, K. O. Pasamehmetoglu, R. G. Steinke, C. Unal, J. H. Mahaffy and C. Murray, "TRAC M/FORTRAN 90 (VERSION 3.0) Theory Manual," LA UR 00 910 (2000).
29. Y. TAITEL and A. Dukler, "A model for predicting flow regime transitions in horizontal and near horizontal gas-liquid flow," AIChE Journal, Vol. 22, No. 1, pp. 47-54, 1976.
30. M. J. THURGOOD et al., "COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems," NUREG/CR-3046 (1983)
31. M. J. THURGOOD et al., "COBRA-TF Development," Proc. Of 8th Water Reactor Safety Information Meeting. (1980).
32. TRAC-PF1/MOD2, Theory Manual // NM 87545. Los Alamos. 1993.
33. Tuomisto, H. and Kajanto, P., 1988, ATwo Phase Flow in a Full Scale Loop Seal Facility,@ Nuclear Eng. and Design, 107, 295-305
34. USNRC, "Regulatory Guide 1.157 (Task RS 701-4), Best-Estimate Calculations of Emergency Core Cooling System Performance", May 1989.
35. USNRC, "Transient and Accident Analysis Methods," Regulatory Guide 1.203, December 2005.
36. M. Y. YOUNG, S. M. Bajorek, M. E Nissley and L. E Hochreiter, "Application of code scaling applicability and uncertainty methodology to the large break loss of coolant," Nuclear Engineering and Design, 186, pp. 39-52, (1998).
37. WCAP 9764, 1980, "Documentation of the Westinghouse Core Uncovery Tests and the Small Break Evaluation Model Core Mixture Level Model," Proprietary.