MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code

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

OSTI ID:21113492

Dunn, F E; Thomas, J; Liaw, J; Matos, J E ^[1]

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1. RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

For safety analyses to support conversion of MNSR reactors from HEU fuel to LEU fuel, a RELAP5-3D model was set up to simulate the entire MNSR system. This model includes the core, the beryllium reflectors, the water in the tank and the water in the surrounding pool. The MCNP code was used to obtain the power distributions in the core and to obtain reactivity feedback coefficients for the transient analyses. The RELAP5-3D model was validated by comparing measured and calculated data for the NIRR-1 reactor in Nigeria. Comparisons include normal operation at constant power and a 3.77 mk rod withdrawal transient. Excellent agreement was obtained for core coolant inlet and outlet temperatures for operation at constant power, and for power level, coolant inlet temperature, and coolant outlet temperature for the rod withdrawal transient. In addition to the negative reactivity feedbacks from increasing core moderator and fuel temperatures, it was necessary to calculate and include positive reactivity feedback from temperature changes in the radial beryllium reflector and changes in the temperature and density of the water in the tank above the core and at the side of the core. The validated RELAP5-3D model was then used to analyze 3.77 mk rod withdrawal transients for LEU cores with two UO{sub 2} fuel pin designs. The impact of cracking of oxide LEU fuel is discussed. In addition, steady-state power operation at elevated power levels was evaluated to determine steady-state safety margins for onset of nucleate boiling and for onset of significant voiding. (author)

• Research Organization: Argonne National Laboratory, Nuclear Engineering Division, RERTR Department, Argonne, IL (United States); Czech Technical University, Prague (Czech Republic)
• OSTI ID: 21113492
• Report Number(s): INIS-US-08N0001; Contract W-31-109-ENG-38; TRN: US08N0014121030
• Resource Relation: Conference: RERTR-2007: 29. international meeting on reduced enrichment for research and test reactors, Prague (Czech Republic), 23-27 Sep 2007; Other Information: Country of input: International Atomic Energy Agency (IAEA); 6 refs, 14 figs, 9 tabs; Related Information: In: 2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting, vp.
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BERYLLIUM
COOLANTS
FUEL PINS
HIGHLY ENRICHED URANIUM
MODERATELY ENRICHED URANIUM
MODERATORS
NIRR-1 REACTOR
NUCLEAR FUELS
NUCLEATE BOILING
POWER DISTRIBUTION
REACTIVITY COEFFICIENTS
SAFETY ANALYSIS
SAFETY MARGINS
STEADY-STATE CONDITIONS
TANKS
THERMAL HYDRAULICS
TRANSIENTS
URANIUM DIOXIDE
WATER