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Monte Carlo Modeling of the Texas A&M University Research Reactor

Chan-Hyeong Kim, Siyoung Jang, Warren Dan Reece

Nuclear Technology / Volume 145 / Number 1 / Pages 1-10

January 2004

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The Monte Carlo N-Particle (MCNP) code and a set of high-temperature neutron cross-section data were used to develop an accurate three-dimensional computational model of the Texas A&M University Nuclear Science Center Reactor (NSCR) at full power. The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements and control rods. The most significant approximation was made for entrained fission products because of the lack of knowledge of fission product inventory in the current reactor core. This study used the concept of "average fission product" to model the fission product in the reactor core and determined the concentration of the average fission product by repeating criticality calculations to make the reactor critical for a given critical condition. Finally, the developed model was tested by comparing the calculated results with those of other approaches, i.e., (a) an in-house three-dimensional diffusion code and (b) foil activation measurement. The developed reactor model showed a good agreement with these approaches. The developed model predicted the thermal neutron flux in samples within 11% of difference when compared with the results from the diffusion code and predicted the production of 198Au and 60Co within ~20% of difference when compared with the values measured with foils. The developed model also calculated the neutron energy spectrum very consistently with the other approaches for the entire energy range considered in this study.

 
 
 
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