Abstract
In this work the latest developments a Monte Carlo simulator with continuous energy is reported. This simulator makes use of a sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum and the Maxwell-Boltzmann distribution. The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution, such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies. It is then shown how this procedure can emulate the up-scattering effect by the increase in the kinetic energy of the neutron population. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process.
Highlights
The present work reports on the development of a Monte Carlo simulator for neutron propagation and associated nuclear reactions either for transient or stationary problems and considering genuine continuous energy dependence by cross section function calls
This procedure executes more efficiently than other models with respect to computational time that are found in the literature (GEANT [1], Serpent [5], MCBEND [6], MCNP [7], OpenMC [8], KENO [9], TRIPOLI [10]), where the cross sections are determined from interpolation of cross section from data bases
In this work we showed the state of developments of a new Monte Carlo transport code for neutron interactions
Summary
Neutron transport is relevant in a variety of applications as for instance industrial applications, such as structural examinations, in nuclear medicine with diagnostic and therapeutic procedures, or radiation protection and last not least nuclear energy production. The present contribution is a part in a larger project, where in the long term we intend to provide an alternative program package for neutron transport and interactions for the GEANT platform [1], which, currently, utilizes the MORSE [2] code for neutron transport In this context, the present work reports on the development of a Monte Carlo simulator for neutron propagation and associated nuclear reactions either for transient or stationary problems and considering genuine continuous energy dependence by cross section function calls. The present work reports on the development of a Monte Carlo simulator for neutron propagation and associated nuclear reactions either for transient or stationary problems and considering genuine continuous energy dependence by cross section function calls This feature was the novelty of the first step of the development [3], which was optimized and extended by new modules, the consequences of up-scattering were included. The Maxwell-Boltzmann distribution is the one where up-scattering effects are considered, and all distributions are continuous over the whole range of energy present in the simulation
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