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An overview of the TRIGLAV package, a tool for the computational analysis of TRIGA Mark II reactors based on lattice transport code WIMSD for unit cell cross section homogenization and four-group diffusion theory in 2D cylindrical coordinates, is presented. Possible unit cells include four types of fuel elements and different non-fuel elements, such as water, graphite, control rod, etc. Apart from the calculation of the main parameters, i.e. the multiplication factor and neutron flux distribution, secondary parameters, such as fuel power and temperature distribution, xenon defect, etc., are also calculated. The program has been compared with the Monte Carlo code MCNP5. Some possible applications of the TRIGLAV program are also outlined in the paper.

Activation measurements can be used for neutron spectrum determination in irradiation facilities. It is very important that self-shielding is taken correctly into account. Activation measurements of monitor materials were performed at the CEA EOLE facility. Self-shielding factors for the monitor materials were calculated at the Commissariat a l’Energie Atomique with the TRIPOLI 4.6 code using a detailed whole-core model, and at the Jožef Stefan Institute with the MCNP code using a simplified model. The results were compared to assess the relative merits and the achievable accuracy of the calculations. In the end, the results were used to determine the neutron spectrum of the studied location in the EOLE irradiation facility.

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.