Browse > Article

A Subchannel Analysis Code for LMR Core Subassembly Thermal Hydraulic Analysis: The MATRA-LMR  

Lim, Hyun-Jin (Ajou University, Korea Atomic Energy Research Institute)
Kim, Young-Gyun (Ajou University, Korea Atomic Energy Research Institute)
Kim, Yeong-Il (Ajou University, Korea Atomic Energy Research Institute)
Oh, Se-Kee (Ajou University, Korea Atomic Energy Research Institute)
Publication Information
Journal of Energy Engineering / v.12, no.4, 2003 , pp. 281-288 More about this Journal
Abstract
The MATRA-LMR code has been developed based on a subchannel analysis method for LMR (Liquid Metal Reactor) core subassembly thermal hydraulic design and analysis. The code was improved to allow a seven assembly calculation and can account for inter-assembly heat transfer based on a lumped parameter model. This paper describes the main modifications and improvements of the code and shows reference calculation results which compared single assembly calculation with seven assembly calculation cased for driver and blanket subassemblies of the KALIMER 150 MWe breakeven conceptual design core. KAL- IMER is a pool-type sodium cooled reactor with a thermal output of 392.0 MWth, which have inherently safe, environmentally friendly, proliferation-resistant and economically viable reactor concepts.
Keywords
Citations & Related Records
연도 인용수 순위
  • Reference
1 MacDougall, J.D. and Lillington, J.N.: 'The Sabre Code for Fuel Rod Cluster Thermohydraulics', Nuclear Engineering and Design, 82, 171 (1984)   DOI   ScienceOn
2 Cheng, S,K, and Todreas, N.E., 'Hydrodynamic Models and Correlations for Bare and Wire-wrapped Hexagonal Rod Bundles - Bundle Friction Factors, Subchannel Friction Factors and Mixing Parameters', Nuclear Engineering and Design, 92, 227 (1986)   DOI   ScienceOn
3 Kim, W.S., et al.: 'A Subchannel Analysis Code MATRA-LMR for Wire-Wrapped Liquid Metal Cooled Reactor Subassembly', Annals of Nuclear Energy, 29(2) (2002)
4 George, TL, et al.: 'COBRA-WC: A Version of COBRA for Single-Phase Multi subassembly Thermal Hydraulic Transient Analysis', PNL-3259, PNL. (1980)
5 Novendstern, E.: 'Turbulent Flow Pressure Drop Model for Fuel Rod Assemblies Utilizing A Helical Wire-wrap Spacer System', Nucl. Eng. and Des., 22, 19 (1972)   DOI   ScienceOn
6 Park, C.K.: 'Development of Korea Advanced Liquid Metal Reactor', IWGFR (1998)
7 Yang, W.S.: 'An LMR Core Thermal-Hydraulics Code Based on the ENERGY Model', Journal of Korean Nuclear Society, 29(5), 406 (1997)
8 Yoo, Y.I. and Hwang, D.H., 'Development of Subchannel Analysis Code MATRA a-Version', Proc. of KNS Autumn Meeting, Taegu, Korea, October 2425 (1997)
9 Lim, H.L., et al.: 'Analysis on Temperature Profiles of KALIMER Breakeven Core Subassemblies', Proc. of KNS Spring Meeting, Gwangjou, Korea, May 24-25 (2002)
10 Tang, Y.S., et aI., 'Thermal Analysis of LiquidMetal Fast Breeder Reactors', ANS (1978)
11 Wheeler, CL, et al.: 'COBRA-IV-I : An Interim Version of COBRA for Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores', BNWL-1662 (1976)
12 Chiu, D., et al.: 'Turbulent Flow Split Model and Supporting Experiments for Wire-wrap Core Assemblies', COO-2245-56TR, MIT. (1978)