Abstract.One of the key areas of the development of Accelerator Driven Systems (ADS) are reactivity monitoring techniques. Since the measurement in the future industrial reactor have to be made in the real time applied methods should be accurate, simple and robust. Therefore methods based on point kinetic model are considered. Necessary experimental validation of selected methods was carried out within research project FP7 FREYA using VENUS-F reactor. This paper presents results obtained using the Sjöstrand method and the source multiplication method. Since ADS core behaviour differs from the point kinetics obtained reactivity value depends on the detector position in the system. From the results it is clear that measurement results strongly depend on the position of the detector in the system. For the Sjöstrand method these spatial effects can be successfully corrected using MCNP-calculated correction factors. Those correction factors do not change within the range of reactivity changes covered in the experiments. Spatial effects affecting source multiplication method are more complex and they depend also on neutron flux distribution in the core.

Abstract.Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom -FIMA). Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/t HM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.