Monte Carlo simulation of BN-600 LMFR hybrid core

The safe operation of a large fast reactor requires accurate estimation of power produced in different parts of the reactor core and blanket. MCNPX code was used to develop a model to simulate and study the whole core of a prototype LMFR hybrid core; the BN-600. In this model, the core is composed of eight radial zones (typical code model layout is illustrated) the first two inner zones are low enrichment zones (LEZ), followed by a medium enrichment zone (MEZ). In the forth zone is the mixed oxide zone (MOX) composed of (U,Pu)O2 fuel subassemblies, then the outer high enrichment fuel zone (HEZ). The rest of the core are two zones of steel shielding assemblies (SSA) and an outer radial reflector to enclose the whole core. There is also 19 shim and control rods (SHR), and 6 scram rods (SCR). The model also take into account the axial variation in geometry and composition, this is accomplished by dividing the core axially into eight different zones with a definite thickness and composition. Partial insertion of control assembly which distorts the reactor flux and fission rates distribution are simulated using the three dimensional model of the reactor core. The spectrum of neutron flux is divided into 23 energy groups. Through this work several parameters are analyzed including criticality, axial and radial power distributions at different zones of the reactor core and burnup analysis in a typical operating conditions of the reactor core. F4 tally was used to calculate the flux distribution in the core and FM4 card was used to calculate the power distribution which is normalized to a total power of 1470 Mw. The energy release per fission was fixed to 200 Mev, as suggested in the BN-600 benchmark details. The temperature variation inside every cell (assembly) were considered by using the 'TMP' card. All fuel cells are at a uniform temperature 1500 K and all structural and coolant isotopes are at a uniform temperature 600 K, and in our model we assign a cross section library at which neutron cross section are processed at 1200 K ( The nearest one to 1500 K) for fuel isotopes U235, U238, Pu239 which consider the Doppler resonance on the cross section. Most of isotopes cross section were taken from the ENDF/B VI library and the name of library which consider Doppler broadening at 1200 K is 'endf62mt'. The core was divided to 25 zones for depletion (burn-up) calculations, as there is 25 different materials that contains fissionable isotopes. We compromise between the number of burnup zones and the program run time. Fission Products are modeled as MCNPX computer code generates a groups of fission products automatically and its concentrations are varied at each time step. Results from this model were compared with published results of benchmark problems and found acceptable. The control rod worth calculated using the present model, at the beginning of cycle, is 0.065479 compared to 0.0661 calculated using diffusion method and 0.0652 using transport method. The radial power distribution (kW) at BOC (upper values are for the present model) compared to benchmark results (lower values) is illustrated. The difference between the two did not exceed 2.6% in all cases