<![CDATA[Criticality benchmark experiment at STACY critical facility is important forvalidation of computation technique and nuclear data library used in design of nuclearfuel cycle criticality safety. This paper discusses criticality safety benchmarkcalculation at STACY facility, which uses uranyl nitrate solution with MCNP-4C MonteCarlo transport code. The continuous energy nuclear data library was utilized inbenchmark calculation to complete criticality safety analysis. The MCNP-4C criticality(keff) prediction indicated overestimated results for all configurations except forconfiguration 131. The biases of calculation with criticality experiment (keff = 1) wereunder 0.26%. Configuration 140 calculation showed the most precisely agreement withC/E value of 1.0001. From these results, it can be concluded that the capability andreliability of MCNP-4C is constantly high in prediction of criticality accuracy for uranylnitrate solution at STACY 280T slab core.]]>
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