Computational codes that simulate steady-states and transients in nuclear reactors are key to supporting the safety of nuclear power plants not only for today’s, but also for future projects, such as SMRs. At the same time, they contribute to increasing efficiency by using the project reserves of power plants. Computational tools perform simulation of various physical phenomena of a nuclear reactor. One of the advanced simulation tools are subchannel codes. The subchannel analysis approach is a proven method for the determination of safety criteria margins, resulting in key thermohydraulic parameters of the nuclear reactor, such as the departure from the nucleate boiling ratio. This research simulated a steady state and transient event (loss of flow accident) in the pressurized water reactor VVER-1000, using the codes SubChanFlow 3.5 and VIPRE-01 for one fuel assembly. The results were subsequently compared as a benchmark. Boundary conditions were calculated using data from the VVER-1000 nuclear power plant model in TRACE code. In general, SubChanFlow has been shown to be more conservative than VIPRE and can be used to evaluate and compare future analysis of various transients. Two independent models were developed to simulate the LOFA scenario, taking into account code differences. In general, the differences in the results can be explained by the different approaches of the crossflow models in the subchannel codes. Nevertheless, the departure from nucleate boiling ratio was calculated using the OKB correlation and the results did not exceed the safety limit criteria.

Activity measurement of the primary circuit water of fission reactors is one method that can provide early detection of a damaged fuel assembly in the reactor core. This is an important aspect in the safe operation of the reactor and for radiation protection of staff. Radionuclides in the primary circuit water are produced by the activation of stable nuclides and the fission of fissile nuclides, mainly the isotope 235U. In the LVR-15 research reactor, measurement of the activity of the primary circuit water has been regularly undertaken since 1996. A water sample is taken from the primary circuit every week and the activities are measured four days later using gamma spectrometry. The results of these long-term measurements from 1996 to 2022 are presented. The activity time dependences of the individual radionuclides are discussed in relation to fuel assembly damage and for events connected to contamination of the water by objects inserted into the primary circuit during experiments carried out near the reactor core.

A coupling of advanced reactors with high-capacity energy storage and other innovative non–nuclear energy technologies is a way to increase the flexibility and utilisation factor of the future nuclear systems and to diversify its energy outputs. In this paper, the basic layout of a hybrid system for a gas-cooled fast reactor is proposed. The system is composed of a high-temperature thermal energy storage unit coupled with a supercritical CO2 conversion cycle and both low-temperature and high-temperature hydrogen production and utilization. As a reference GFR reactor, the ALLEGRO concept is considered. The paper describes state-of-the-art and overviews potentially applicable technologies. Selection and justification of concepts of the individual parts of the hybrid system including operational conditions is also given. The R&D needs necessary for the potential implementation of the system are identified and the activities ongoing in the Czech Republic and especially in Centrum Výzkumu Řež are also described.

Since 2014, the irradiation of material samples manufactured from advanced Russian zirconium alloys has been performed in the VVER-1000 reactor core of Temelin NPP Unit 1, focused on the study of irradiation-induced growth (IIG) and related microstructural changes. The material samples differ from each other by alloying elements and final heat treatment, which provides a variety of initial microstructural characteristics. The test matrix allows the evolution of these different types of microstructures to be studied with increasing neutron irradiation and how it corresponds with macroscopic IIG. The expected outcome of this large research program is to obtain new experimental data on the IIG of Zr-alloys and the underlying microstructural changes that occur under standard irradiation conditions in a commercial VVER-1000 reactor. Six material cluster assemblies (MCAs) containing ampoules with longitudinal segments of cladding tubes with a size of 50 × 6 × 0.6 mm were irradiated in five irradiation cycles to achieve six different values of neutron fluence. One/two MCAs were removed from the reactor core after each irradiation cycle and then cooled in the spent fuel pool. A unique cutting device called POMA installed in the transport container pit is used to separate the ampoules from the remaining part of the MCA. Irradiated materials are being evaluated in the hot-cell facilities at UJV Řež and Research Centre Řež. The material tests and analyses include the determination of neutron fluence based on the activities of neutron activation monitors, geometry measurements of samples and their evaluation, TEM, SEM, and LOM analyses and microhardness measurement. The irradiation of all material samples has been completed. The IIG has been measured for the first four batches and related to the accumulated neutron fluence and to the microstructural changes. The alloys of multicomponent Zr-Nb-Sn-Fe systems exhibit higher values of IIG compared with the Zr-Nb-(Fe,O) alloys at the same neutron fluences.

Having knowledge of the spectrum of fast neutrons in the irradiation position of a reactor plays an important role in many experimental settings. If well described, it can be used as a reference neutron field for validating and testing measuring devices or detectors. This paper compares model calculations with measurements of the neutron spectrum in the research reactor VR-1. The fast neutron spectrum was measured in the radial channel of VR-1 reactor using a stilbene scintillation detector and the NGA-01 measuring device. The NGA-01 is a two-parameter spectrometric system for neutron and gamma radiation fields. The calculations were performed in a critical mode of the MCNP6 code using ENDF/B-VII.1 nuclear data library. Generally, a good agreement was achieved. However, a slight discrepancy in the neutron spectrum between the calculated and measured values was found in the 2–3 MeV region. In other regions, the differences between calculated and experimental values are comparable with uncertainties. Since the detector is distanced from the core, the neutron beam can be assumed parallel and therefore possible issues with stilbene anisotropy are eliminated. Thus, this experiment can be used for validation of a reactor leakage spectrum and nuclear cross sections in the energy region above 6 MeV, which is of interest for the nuclear data section of International Atomic Energy Agency.

This study evaluates the types of waste generated by tritium during nuclear fusion. Some methods of reprocessing and decontaminating solid waste using thermal processes are evaluated, and the advantages and disadvantages of different methods are compared. The high-temperature technology selected for this study is intended for use in the EU DEMO project in the area where waste from nuclear fusion reactions is processed. Safety and environmental concerns around the technology are evaluated. The potential for detritiation of solid wastes of various sizes are investigated. The study’s focus is on wastes comprising mostly tungsten dust grains of various sizes. The possibilities and rationale for the use of high-temperature technologies are investigated. Tests conducted focus primarily on tungsten waste in powder form in various atmospheres. Problems related to the induction heating and melting of metals and nonmetals are addressed.

The article deals with increasing the mechanical properties of stainless steel 316 Ln-IG, which is intended for work in cryogenic temperatures (liquid nitrogen and liquid helium), such as conductor conduits for the ITER magnet system. The strength and plastic properties were increased by a combination of cold and cryo-rolling and heat treatment. The mechanical properties of rolled material were investigated at 293 K, 77 K, and 4.2 K. The work-hardening rate of the steel increased continuously with a lowering of the temperature. The maximum yield strength and ultimate tensile strength were achieved by the cryo-rolling process with a total thickness deformation of 50%. The material properties tested at ambient temperature were 0.2YS = 1050 MPa, UTS = 1200 MPa, and at 4.2 K, the values were 0.2YS = 1804 MPa and UTS = 2081 MPa. Two types of long-term heat treatment were applied after experimental rolling (823 K and 1093 K for 10 h). The highest precipitation hardening of steel was achieved at a temperature of 823 K after 50% deformation. The resulting grain size decreased from the initial 216 μm (before the rolling process) to 70 μm after ambient rolling and 72 μm after cryo-rolling.