Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU

Conference · Tue Jul 15 04:00:00 EDT 2008

OSTI ID:21113502

Ko, Y -C ^[1]; Olson, Arne P; Dunn, Floyd E ^[2]

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1. Nuclear Science and Engineering Department, MIT, Cambridge, MA 02139 (United States)
2. RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

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. (author)

• Research Organization: Argonne National Laboratory, Nuclear Engineering Division, RERTR Department, Argonne, IL (United States); Czech Technical University, Prague (Czech Republic)
• OSTI ID: 21113502
• Report Number(s): INIS-US--08N0001
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
COMPUTER CODES
COOLANTS
DESIGN
NATURAL CONVECTION
PRIMARY COOLANT CIRCUITS
REACTOR SAFETY
RESEARCH REACTORS
SAFETY ANALYSIS
STEADY-STATE CONDITIONS
TEMPERATURE MEASUREMENT
THERMAL HYDRAULICS
VALIDATION