Giacomo Grasso

To process cross sections libraries with NJoy, in order to obtain a complete thermal treatment of the elements which a material is made of, the single isotopes of interest are identified - unlike the common zaid standard - via a numeric, up to four digit code, the s.c. MAT number.

The simulation of nuclear systems by means of numerical codes, such as ERANOS or MCNP, requires the compiling of proper input files, describing the geometric configuration and material composition of the reactors.

The comparison of the results obtained with different codes (needed in order to design new systems, due to the different transport equation solution models implemented in such codes) may hold if and only if it is found to fulfill equivalence criteria which guarantee the mutual consistency of the simulations.

One of the most important parameters to be considered for the equivalence of the two different calculations is the mass of fissile in the fuel region. In particular, in case of Mixed OXide (MOX) fuel, besides the isotopic characterization of the vectors, the components fractions of the mixture is needed. Within the ERANOS input file, it is possible to define several "Materiaux" which make up a "Milieu" by specifying their volumetric fractions (VFs) in the fuel. On the other side, MCNP needs every cell to be filled with a single material, which composition can be specified by means of a "mat card" defined through either atomic or weight fractions (respectively #Fs and WFs) of the composing isotopes mixture.

The present work describes the creation of a script for automatic processing of cross section libraries in order to create useful datasets for nuclear systems analysis tools, like MCNP, by means of the well-known NJoy nuclear data processing code.

The resulting "njoyinputmaker.sh" script allows the user to automatically process a given list of isotopes at a given temperature by creating multiple appropriated input script for the NJoy code and running the latter to process them.

The development of an European Gen-IV reactor is one of the main object of interest in the European research framework. An European Lead cooled SYstem (ELSY) is therefore under investigation on a joint research committee where the National Agency for the New Technologies, Energy and Environment (ENEA) plays the role of neutronic designer for the so-called square open hypothesis of the core.

Some preliminary Monte Carlo criticality calculations of the square open ELSY core model are therefore here presented: they have been produced in order to verify the main deterministic calculations performed with ERANOS and to give an easier, detailed analysis of the neutronic performances of the system. The latest versions of the core are here presented for the modeling with MCNP5 (whose input files are presented in appendices) and discussed.

Whatever MCNP simulation is completely defined by means of the geometrical, material, operative and numerical positions implemented in the proper input file. As soon as the simulating systems become more and more complex, the input file structure and length grow heavier, making the file itself little readable; such simulations also require great computational powers, which are usually not accessible via graphical interfaces: this implies that it is often required to modify the MCNP input file by means of shell editors like Vim.

To simplify the readability of such files it has therefore been developed a proper syntax highlighter for the Vim editor, supporting all the main functions useful for easy access and browsing of MCNP input files.

The neutronic design of new concept (Gen-IV) nuclear reactors, such as ELSY (the European Lead SYstem), requires deep testing phases and original solutions conceptualization to guarantee the feasibility of such innovative systems. One of the most important questions to be checked is the void effect: the loss of coolant following such an accident is known to introduce into the system a reactivity whose effects must be taken in consideration during the design phase.

The present work represents an introductory investigation of the void effect in fast reactors. The analysis, conducted by means of the general purpose Monte Carlo transport code MCNP5 on the first model of ELSY, aims at an overall understanding of the specific effects caused by the lack of coolant: the hardening of the spectrum, causing a change in the Reaction Rates and the neutronic mean free path; the change of the leakage probabilities and the absence of the absorptions in the coolant.

The present work describes the evolution of the original script njoyinputmaker.sh for automatic processing of cross section libraries. A new graphical, menu-driven interface has been introduced for a more user-friendly approach to the creation of cross sections datasets for nuclear systems analysis tools, like MCNP, by means of the well-known NJoy nuclear data processing code.