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Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

Kwi-Seok Ha, Hae-Yong Jeong, Chungho Cho, Young-Min Kwon, Yong-Bum Lee, Dohee Hahn

Nuclear Technology

Volume 169 / Number 2 / February 2010 / Pages 134-142


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As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

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