2007 International RERTR Meeting

Abstracts and Available Papers Presented at the Meeting

Validation of the MULCH-II Code for Thermal- Hydraulic Safety Analysis of the MIT Reseach Reactor Conversion to LEU

Yu-Chih Ko‡, Lin-Wen Hu*,
‡Nuclear Science and Engineering Department, MIT
*Nuclear Reactor Laboratory, MIT, Cambridge, MA, USA 02139

Arne P. Olson and Floyd E. Dunn
RERTR Program
Argonne National Laboratory, Argonne, IL USA 60439

ABSTRACT

An in-house thermal hydraulics code was developed for the steady-state and loss of primary flow analysis of the MIT Research Reactor (MITR). This code is designated as MULti-CHannel-II or MULCH-II. The MULCH-II code is being used for the MITR LEU conversion design study. Features of the MULCH-II code include a multi-channel analysis, the capability to model the transition from forced to natural convection during a loss of primary flow transient, and the ability to calculate safety limits and limiting safety system settings for licensing applications. This paper describes the validation of the code against PLTEMP/ANL 3.0 for steady-state analysis, and against RELAP5-3D for loss of primary coolant transient analysis. Coolant temperature measurements obtained from loss of primary flow transients as part of the MITR-II startup testing were also used for validating this code. The agreement between MULCH-II and the other computer codes is satisfactory.

 

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