Nuclear Science and Engineering / Volume 164 / Number 1 / January 2010 / Pages 53-68
Technical Paper / dx.doi.org/10.13182/NSE09-03
The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU.
The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction of flux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases.
A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. “Foils” (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil.
This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil. Power distributions were obtained for clean core (no poison in moderator and symmetrical rod position at 44.536 cm withdrawn with respect to the core axial midplane) and fully poisoned moderator (1.35 grams of boron per liter in moderator and rods fully withdrawn) conditions. The observed deviations between the experimental and calculated values for both conditions were within the reported experimental uncertainties except for some foils located on the top and bottom edges of the fuel plates.