Neutronic and thermal-hydraulic modeling of HTR-10 reactor via V.S.O.P code

The VSOP'94 (Very Superior Old Programs) code is a system of codes linked together for the numerical simulation of neutronic and thermal-hydraulic behaviors for High Temperature Gas Reactors (HTGR) under normal operating and accident conditions. The code can be also utilized for the calculations of in-core and out-of pile fuel management and fuel cycle cost analysis. The VSOP code obtained from OECD/NEA Data Bank, has been used for the neutronic and thermal-hydraulic prediction calculations for the initial core loading of the HTR-10 reactor. In the neutronic aspect of this study, double heterogeneity for the fuel element, buckling feedback in the spectrum calculation, and mixture of graphite and fuel balls are taken into account. The results, compared to a relevant study based on the first criticality experiment of HTR-10, indicate that relative errors for the multiplication factor are less than 1.34%, being in a reasonable range. In the thermal-hydraulic analysis, the power distribution obtained from the neutronic calculations is used where temperature and flow distributions are calculated for the reactor core. Throughout the analyses, conservative hydraulics and physical parameters, such as minimum mass flow rates in the core, no radiation heat transfer in the axial direction, are used. The maximum fuel centerline temperature is obtained as 927.4 deg C for normal operating conditions and observed not to exceed the safety limit. (author)