The CANDLE burnup strategy applied to small modular pressurized water reactor loading with fully ceramic microencapsulated fuel

For post-Fukushima nuclear power plants, there has been interested in accident-tolerant fuel (ATF) since it has better tolerant in the event of a severe accident. The fully ceramic microencapsulated (FCM) fuel is one kind of the ATF materials. In this study, the small modular pressurized water reactor (PWR) loading with FCM fuels was investigated, and the modified Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor (CANDLE) burnup strategy was successfully applied to such compact reactor core. To obtain ideal CANDLE shape, it’s necessary to set the infinity or enough length of the core height, but that is impossible for small compact core setting infinity or enough length of the core height. Due to the compact and finite core, the equilibrium state can only be maintained short periods and is not obvious, other than infinitely long active core to reach the long equilibrium state for ideal CANDLE. Consequently, the modified CANDLE shape would be presented. The approximate characteristics of CANDLE burnup are observed in the finite and compact core, and the power density and fuel burnup are selected as main characteristic of modified CANDLE burnup. In this study, firstly, lots of optimization schemes were discussed, and one of optimization schemes was chosen at last to demonstrate the modified CANDLE burnup strategy. Secondly, for chosen compact small rector core, the modified CANDLE burnup strategy is applied and presented. Consequently, the new characteristics of this reactor core can be discovered both in ignition region and in fertile region. The results show that application of CANDLE burnup strategy to small modular PWR loading with FCM fuels suppresses the excess reactivity effectively and reduces the risk of small PWR reactivity-induced accidents during the whole core life, which makes the reactor control more safety and simple.

Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

Application of Direct Numerical Simulation to Determine the Correlation Describing Friction Losses during the Transverse Flow of Fluid in Hexagonal Array Pin Bundles

This paper presents the results of a numerical simulation to determine the hydraulic resistance for a transverse flow through the bundle of hexagonal rods. The calculations were carried out using the precision CFD code CONV-3D, intended for direct numerical simulation of the flow of an incompressible fluid (DNS-approximation) in the parts of fast reactors cooled by liquid metal. The obtained dependencies of the pressure drop and the coefficient of anisotropy of friction on the Reynolds number can be used in the thermal-hydraulic codes that require modeling of the flow in similar structures and, in particular, in the inter-wrapper space of the reactor core.

Titanium beryllide as an alternative to beryllium in nuclear and thermonuclear engineering, capabilities of UMP JSC in the technology development and beryllides products manufacture

For many decades, beryllium has been used as a structural element in nuclear installations as a moderator / breeder of fast neutrons. The consequence of neutron irradiation is a significant production of gas products in the form of helium and tritium, which leads to swelling and loss of strength properties of beryllium reflectors. The relatively low melting point of beryllium also imposes restrictions on the high-limit temperature regimes of the reactor core. As an alternative to pure beryllium, it is necessary to consider intermetallic compounds based on it, in particular titanium beryllide. Preliminary studies on the thermal desorption of helium and tritium from titanium beryllide have shown that this material has a much lower retention tendency and a lower release temperature. The higher melting point of titanium beryllide compared to pure beryllium is also an advantageous characteristic.Over the past years, UMP JSC, thanks to its research in this area, has achieved significant success in the development of technology for obtaining intermetallic billets and articles based on titanium and chromium beryllides. As a technology demonstrator, prototypes of structural elements of a helium-cooled blanket breeder module of the projected DEMO reactor were made by order of the Karlsruhe Institute of Technology, Germany.The advantages of titanium beryllide, as well as the success achieved in the production of billets and products from it, open up opportunities for a more extensive study of the nuclear, physical and mechanical properties of this material with the possibility of further use in nuclear technology, including thermonuclear reactors, and in high-temperature instrumentation.

Localization of Plastic Deformation in the Copper and Stainless Steels Samples, Irradiated with Neutrons

The reduction of ductility of austenitic stainless steels as a result of long-term operation in the nuclear reactor core is an important problem of modern radiation materials science. Understanding the mechanisms of the effect of neutron irradiation on the mechanical properties of austenitic steels is impossible without research of localization processes occurring during the deformation. In this paper, it was found that the value of the true local deformation corresponding to the onset of neck formation in face-centered cubic structured metals decreases with an increase in the radiation dose, while the true stress remains almost constant. Additional hardening of AISI 304 steel due to the intensive formation of the martensitic α’-phase increases not only the stress at which a neck is formed in this alloy, but also the true local deformation. As a result, the uniform elongation increases and remains high after neutron irradiation to 0.05 dpa. The forehanded formation of the martensitic α’-phase in sufficient quantity before the necking onset can be considered as an additional deformation mechanism that will increase the ability of the material to deform uniformly.