Abstract
A set of experiment has been done at STACY facility and many fundamental parameters of uranyl nitrate solution have been found out. Criticality is one of main parameters in predicting neutronic characteristic of STACY experiment beside solution level reactivity, void reactivity, kinetic parameter and temperature reactivity which dominates transient phenomenon in abnormal condition. Criticality experiment performed at STACY core uses 9.97% 235U -enriched uranyl nitrate solution with 80-cm-diameter cylindrical and 150-cm-height tank. Eight critical configurations in unrelected and water-reflected conditions were selected in this paper for criticality safety calculation with Monte Carlo transport code MCNPX. For all configurations, MCNPX calculations show good consistency with the trend of producing underestimated k eff. Calculation biases with experimental data ( k eff = 1) for water-reflected configurations, i.e. 0.01-0.18%, were slightly better than those of unreflected configurations (0.14-0.41%). MCNPX calculation results which are better than the prediction of MCNP-4C concludes that MCNPX is more eligible to be applied to criticality safety analysis of uranyl nitrate solution in commercial nuclear fuel cycle facility.
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