Browse > Article
http://dx.doi.org/10.5516/NET.2011.43.1.045

DEVELOPMENT OF THE SPACE CODE FOR NUCLEAR POWER PLANTS  

Ha, Sang-Jun (Korea Electric Power Corporation Research Institute)
Park, Chan-Eok (KEPCO Engineering & Construction Company Inc)
Kim, Kyung-Doo (Korea Atomic Energy Research Institute)
Ban, Chang-Hwan (Korea Nuclear Fuel)
Publication Information
Nuclear Engineering and Technology / v.43, no.1, 2011 , pp. 45-62 More about this Journal
Abstract
The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.
Keywords
System Code; SPACE; Nuclear Safety Analysis; Thermal-Hydraulics; Multi-Fields Model; Code Validation;
Citations & Related Records

Times Cited By Web Of Science : 1  (Related Records In Web of Science)
Times Cited By SCOPUS : 0
연도 인용수 순위
  • Reference
1 ˝SPACE code manual: Models and Correlations˝, KAERI, March 8, 2010.
2 Y. H. Kim, et al., V&V Matrix for Safety Analysis Code SPACE, S06NX08-E-1-RD-08, Korea Electric Power Corporation Research Institute, 2009.
3 F. J. Moody, Maximum Flow Rate of a Single Component, Two-Phase Mixture, Trans. of the ASME, J. of Heat Transfer, pp. 134-142, February 1965.
4 R. E. Henry and H. K. Fauske, The Two-Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes, Trans. of the ASME, J. of Heat Transfer, pp. 179-187, May 1971.
5 A. R. Edwards and T. P. O'Brien, Studies of Phenomena Connected with the Depressurization of Water Reactors, Journal of the British Nuclear Energy Society, 9, pp. 125-135, 1970.
6 G. B. Wallis, Flooding Velocities for Air and Water in Vertical Tubes, UKAEA, AEEW-R-123, 1961.
7 S. G. Bankoff and S. C. Lee, A Brief Review of Countercurrent Flooding Models Applicable to PWR Geometries, Nuclear Safety, Vol. 26, No. 2, pp. 139-152, March-April 1985.
8 U.S. Nuclear Regulatory Commission, Reactor Safety Issues Resolved by the 2D/3D Program, NUREG/IA-0127, GRS-101. MPR-1346, July 1993.
9 D. A. Powers and R. O. Meyer, Cladding Swelling and Rupture Models for LOCA Analysis, NUREG-0630, April 1980.
10 J. V. Cathcart et al., Zirconium Metal-Water Oxidation Kinetics IV. Reaction Rate Studies, ORNL/NUREG-17, August 1977.
11 STAR-CD VERSION 4.06 Manual, CD-adapco, 2008.
12 BWR (boiling Water Reactor) Refill-Reflood program Task 4.8 - Model Qualification Task plan, GE Co., San Jose, CA, NUREG/CR-1895, EPRI NP-1527, GEAP-24898, Aug., 1981.
13 A. W. Bennett et al., Heat Transfer to Steam-Water Mixtures Flowing in Uniformly Heated Tubes in Which the Critical Heat Flux has been Exceeded, AERE-R5373, October 1976.
14 E.-L. Pelletier, L.K.H. Leung, A. Teyssedou, and R. Girard, “Comparison and Improvements of Correlations for Film Boiling in Tubes,” Proceedings of the 17th International Conference on Nuclear Engineering (ICONE17), July 12-16, 2008, Brussels, Belgium.
15 D.R. Liles and W.H. Reed, ˝A semi-implicit method for two-phase fluid dynamics˝, J. Com. Phys., vol. 26, pp. 390-407, 1978.   DOI   ScienceOn
16 RELAP5/MOD3.3 Code Manual, Volume I: Code Structure, System Models and Solution Methods, NUREG/CR-5535/Rev 1, December , 2001.
17 TRAC-M/FORTRAN90 (VERSION 3.0) Theory Manual, LA-UR-00-910, July, 2000.
18 Analysis of FLECHT SEASET 163-Rod Blocked Bundle Data Using COBRA-TF, NUREG/CR-4166, January. 1986.
19 M. Robert, M. Farvacque, M. Parent and B. Faydide, CATHARE2 V2.5: a fully validated CATHARE2 version for various applications, Proceedings of the 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10) Seoul, South Korea, October 5–9, 2003.
20 MARS code manual, volume 1: code structure, system models, and solution methods, KAERI/TR-2812/2004, July 2006.
21 Churchill, S. W., ˝Friction-Factor Equation Spans All Fluid-Flow Regimes,˝ Chemical Engineering, pp.91-92., 1977.
22 Claxton, K. T., Collier, J. G. and Ward, J. A., ˝H.T.F.S. Correlation for Two-Phase Pressure Drop and Void Fraction in Tubes,˝ HTFS Proprietary Report HTFS-DR-28, AERE-R7162, 1972.
23 Chisholm, D., ˝A Theoretical Basis for the Lockhart-Martinelli Correlation for Two-Phase Flow,˝ Int. J. Heat and Mass Transfer, vol.10, pp.1767-1778, 1967.   DOI
24 Pan, L. and Hanratty, T. J., ˝Correlation of entrainment for annular flow in horizontal pipes,˝ Int. J. Multiphase Flow, vol.28, pp.385-408., 2002.   DOI
25 J. A. Trapp and V.H. Ransom, A Choked-Flow Calculation Criterion for Nonhomogeneous, Nonequilibrium, Two-Phase Flows, Int. J. Multiphase Flow, Vol. 8, No. 6, pp. 669-681, 1982.   DOI
26 S. Y. Lee et al., ˝Formulation of time and volume averaged two-fluid model considering structural materials in a control volume,˝ Nuclear Engineering and Design, Vol. 239, Issue 1, pp 127-139, January 2009.   DOI
27 S.V. Patankar, Numerical heat transfer and fluid flow, Hemisphere Publishing Corporation. 1981.
28 C.Hirsch, Numerical computation of internal and external flows, John Wiley & Sons, 1991.
29 C. E. Park et al., ˝A Two-Fluid, Three-Field Hydraulic Solver for the Safety Analysis Code, SPACE,˝ ANS winter meeting, Washington D. C., November 2009.