Browse > Article

HOT CHANNEL ANALYSIS CAPABILITY OF THE BEST-ESTIMATE MULTI-DIMENSIONAL SYSTEM CODE, MARS 3.0  

JEONG J.-J. (Korea Atomic Energy Research Institute)
BAE S. W. (Korea Atomic Energy Research Institute)
HWANG D. H. (Korea Atomic Energy Research Institute)
LEE W. J. (Korea Atomic Energy Research Institute)
CHUNG B. D. (Korea Atomic Energy Research Institute)
Publication Information
Nuclear Engineering and Technology / v.37, no.5, 2005 , pp. 469-478 More about this Journal
Abstract
The subchannel analysis capability of MARS, a multi-dimensional thermal-hydraulic system code, has been enhanced. In particular, the turbulent mixing and void drift models for the flow-mixing phenomena in rod bundles were improved. Then, the subchannel analysis feature was combined with the existing coupled system thermal-hydraulics (T/H) and 3D reactor kinetics calculation capability of MARS. These features allow for more realistic simulations of both the hot channel behavior and the global system T/H behavior. Using the coupled features of MARS, a coupled analysis of a main steam line break (MSLB) is carried out for demonstration purposes. The results of the calculations are very reasonable and promising.
Keywords
MARS; MASTER; Coupled Calculation; System T/H Analysis; Hot Channel Analysis; 3D Reactor Kinetics;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
연도 인용수 순위
1 A. Tapucu, M. Geckinli, N. Troche, and R. Girard, 'Experimental Investigation of Mass Exchanges between Two Laterally Interconnected Two-Phase Flows,' Nuclear Engineering and Design, 105, pp. 295-312 (1988)   DOI   ScienceOn
2 D. H. Hwang et al., 'Assessment of the Interchannel Mixing Model with a Subchannel Analysis Code for BWR and PWR Conditions,' Nuclear Engineering and Design, 199, pp. 257-272 (2000)   DOI   ScienceOn
3 H. Herkenrath et al., Experimental Investigation of the Enthalpy and Mass Flow Distribution in 16-Rod Clusters with BWR-PWR-Geometries and Conditions, ISPRA Report, EUR 7575 EN (1981)
4 R. T. Lahey, Jr. and F. J. Moody, The Thermal Hydraulics of a Boiling Water Reactor, 2nd Ed., American Nuclear Society, La Grange Park, Illinois USA, pp. 168-184 (1993)
5 J.-J. Jeong, S. W. Bae, D. H. Hwang, W. J. Lee, and B. D. Chung, 'Subchannel Analysis Capability of the Best-Estimate Multi-Dimensional System Code, MARS 3.0,' Proc. KNS 2004 Fall Meeting, Korean Nuclear Society (2004)   과학기술학회마을
6 J. E. Kelly, Development of a Two-Fluid, Two-Phase Model for Light Water Reactor Subchannel Analysis, PhD Thesis, Department of Nuclear Engineering, MIT (1980)
7 J.-J. Jeong, H. G. Joo, B. D. Chung, K. S. Ha, W. J. Lee, B. O. Cho, and S.-Q. Zee, 'MARS/MASTER Solution to OECD Main Steam Line Break Benchmark Exercise III,' J. Korean Nuclear Society, 32, no. 3, pp. 214-226 (2000)   과학기술학회마을
8 N. Todorova, Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark. Volume IV: Results of Phase III on Coupled Core-Plant Transient Modeling, NEA/NSC/DOC (2003)21, OECD NEA (2003)
9 H. G. Joo, J.-J. Jeong, B. O. Cho, W. J. Lee, and S. Q. Zee, 'Analysis of the OECD MSLB Benchmark Problem using the Refined Core Thermal-Hydraulic Nodalization Feature of the MARS/MASTER Code,' Nuclear Technology, 142, pp. 166-179 (2003)
10 B. O. Cho et al., MASTER-2.0: Multi-purpose Analyzer for Static and Transient Effects of Reactors, KAERI/TR-1211/99, Korea Atomic Energy Research Institute (1999)
11 W. J. Lee et al., Development of Realistic Thermal-Hydraulic System Analysis Code, KAERI/RR-2235/2001, Korea Atomic Energy Research Institute (2001)
12 R. T. Lahey, Jr. et al., Two-phase Flow and Heat Transfer in Multirod Geometries: Subchannel and Pressure Drop Measurements in a Nine-rod Bundle for Diabatic and Adiabatic Conditions, GE report, GEAP-13049 (1970)
13 REALP5 Code Development Team, RELAP5/MOD3 Code Manual, NUREG/CR-5535, Idaho National Engineering Lab., USNRC (1995)
14 M. J. Thurgood et al., COBRA/TRAC-A Thermal-Hydraulic Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems, NUREG/CR-3046, USNRC (1983)
15 J.-J. Jeong et al., 'Development of Subchannel Analysis Capability of the Best-Estimate Multi-Dimensional System Code, MARS 2.3,' Proc. KNS 2004 Spring Meeting, Korean Nuclear Society (2004)   과학기술학회마을
16 D. C. Groeneveld, S. C. Cheng, and T. Doan, '1986 AECLUO Critical Heat Flux Lookup Table,' Heat Transfer Engineering, 7, pp. 46-62 (1986)   DOI   ScienceOn
17 KEPCO, Final Safety Analysis Report of Yonggwang Nuclear Units 3 and 4
18 W. J. Lee et al., 'Coupled Analysis of YGN 3/4 Single RCP Locked Rotor Accident Using MARS/MASTER,' Proc. KNS 2000 Spring Meeting, Korean Nuclear Society (2000)
19 J.-J. Jeong et al., Main Steam Line Break Analysis of YGN-3/4 Using the Coupled Multi-Dimensional Thermal-Hydraulics - Reactor Kinetics Code, MARS/MASTER, KAERI/TR-1989/2001, Korea Atomic Energy Research Institute (2001)