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After the Fukushima DAIICHI accident, new safety requirements were imposed in order to reduce risk of severe accident. One of the principles that have been adopted is the introduction of emergency action levels resulting from the expected consequences. They cover a wide range of component and system malfunctions resulting in emergency, incident, and/or accident conditions. To evaluate those emergency action levels, thermal hydraulic (TH) analyses simulating these malfunction/incident/accident conditions are required. This paper describes the simulation of a real operational incident scenario using a standard thermal hydraulic model of the power plant in the TRACE code that was originally intended for simulation of design basis accidents such as large break coolant accident or loss of flow accident. Special attention was given to the methodology, addressing a long duration of an incident with corrective actions of the operators, and to computational issues leading to model modifications caused by a long duration of the incident along with the necessary conservatisms in the estimated results of the simulated radioactivity release.

Research reactors, together with the computational codes, represent a valuable tool combination to investigate over phenomena which can take place in nuclear power plants. In this framework, the Research Centre Řež (CVŘ), which owns two research reactors, Lehkovodni Reaktor 0 (LR-0) and Lehkovodni Reaktor 15 (LVR-15), is moving towards the creation of a code portfolio to be used to relicense them in front of the State Office of Nuclear Safety (SUJB). LVR-15 is a light water tank type research reactor, adopting a sandwich tube type fuel (IRT4M). The first step of this research activity is the preparation of a working model of LVR-15 using SCALE and PARCS codes: SCALE is used to estimate the Cross Sections (XS) based on the real geometry of the reactor and PARCS is used to perform calculation of transients. The results obtained show that the model is adequately accurate for steady state, as soon as the calculation reaches the criticality. The power calculated during the transient calculations, instead, stabilize after about 1000 s and become comparable, within a 2% bias, with the experimental data, extrapolated from the reactor operation. In conclusion, this demonstrates that PARCS code is capable of modelling our research reactor fuel assemblies geometry in a satisfactory way, although further developments would improve the accuracy of the adopted methodology.

Energy Well (EW) is a concept of a small modular reactor, designed as a source of energy for places with low infrastructure. Its design combines TRISO fuel particles, graphite moderator and molten salt coolant. One of key features is that the active core is small and therefore easy to transport. Residual heat estimation of EW fuel assembly is conducted using two different computation codes, SCALE and SERPENT. The obtained residual heat power is then used as an input in TRACE model of the EW to obtain the development of residual temperature in the coolant in the core inlet and outlet and also maximum temperature on the outer side of a fuel slab. The resulting temperatures are below the operational temperature.

This paper focuses on the TRACE code assessment for helium-cooled systems thermal-hydraulic analysis. In the frame of the GoFastR (gas cooled fast reactor) European Collaborative Project, ENEA has offered some selected experimental data for the organization of a benchmark exercise aimed at the validation of the system and CFD codes for the gas reactor transient analyses. One of the Research Center Řež teams participated in it with a CFD code application. Now, the experimental data are used in order to assess the TRACE code for the ongoing high-temperature helium loop (HTHL-2) licensing process. The results of the TRACE calculations agreed very well with the experimental measurements (often within the experimental uncertainties) data provided by the He-FUS3 facility, indicating that the code, despite developed for water coolant applications, if an appropriately tuned input is adopted, it can also be suitable for reasonably accurate gas technology thermo-hydraulic simulations.

Centrum Výzkumu Řež has a vast R&D program which was financed through the SUStainable Energy project. Several state-of-the-art facilities were designed in order to cover the demand of the energy community and to face new challenges for such advanced technologies. Two new experimental loops are to be inserted inside of the LVR-15 Research Reactor. Their purpose is to analyze cladding and structural material behaviour under different pressures, temperatures and coolant media conditions in comparison to the Light Water Reactors. The loops in question are the High Temperature Helium Loop and the Super Critical Water Loop. This paper focuses solely on the supercritical water loop. The first task of the Safety Research Section within the CVŘ was to select and to assess a code in order to simulate the media used in the loops. The second task was to use the selected codes to prepare the facilities models in order to simulate the possible consequences of the accident scenarios on the LVR-15 reactor. Based on the criteria set by the authors of this paper, through a comprehensive selecting procedure the ATHLET 3.1A code was chosen for the super critical water simulation. The paper is presenting the methodology followed in assessing the code’s capability for simulation of SCW.