Atomic Energy最新文献

Background

Increasing the efficiency of using reprocessed uranium can partially solve the problems of resource availability for thermal neutron power reactors, as well as the continuous accumulation of spent nuclear fuel.

Aim

To develop a method for including the double cascade subproduct in the production of nuclear fuel without additional costs for separative work by replacing part of the natural uranium in producing low-enriched uranium for thermal neutron reactors.

Materials and methods

The possibility of involving the subproduct of a modified double cascade for enrichment of reprocessed uranium in the production of nuclear fuel is considered. A method is proposed for including the subproduct of such a double cascade in the production of nuclear fuel.

Results

The study demonstrates this method saving natural uranium by 4.3% compared to the same cascade without using the light fraction flow.

Conclusion

The proposed method increases the proportion of ²³⁵ U returned to the fuel cycle and provides additional combined savings of natural uranium and costs of separative work compared to an open nuclear fuel cycle.

The article presents estimates of long-term static strength for high-temperature equipment of a commercial reactor plant with a BN-1200M fast reactor at the design service life increased from 60 to 80 years enhancing its economy. The features of loading the equipment of sodium reactor plants and the history of extending the service life of the existing BN-600 reactor are described. Sufficient margins for operating stresses relative to allowable values for most equipment of commercial sodium-cooled fast reactor plants are demonstrated.

Background

The project of the Kazakhstan NPP planned for construction in the Zhambyl district of the Almaty region assumes environmental safety in accordance with the concept for transition of the Republic of Kazakhstan to a “green” economy. Therefore, the impact of this nuclear fuel cycle facility on the environment and the population should be thoroughly analyzed.

Aim

To assess the exposure doses to population from radionuclides potential to release into the environment during the normal operation of the planned Kazakhstan NPP.

Materials and methods

The initial data of the study include characteristics of radioactive emissions and discharges from supposed VVER-1200, APR-1400, and HPR-1000 reactor plants. Calculations were performed using the CROM software (LABI, Kingdom of Spain) and additional modules for assessing the radiation dose from tritium (³ H) and ¹⁴ C. The models of the study are parameterized taking into account regional data.

Results and discussion

The total radiation doses from radionuclides of NPP origin for various age groups of the population were calculated; the contributions of individual radionuclides and exposure routes were assessed. The highest dose was recorded for the HPR-1000 reactor; while for the VVER-1200 and APR-1400, this value is 3–4 times less. Children aged 1–2 years will receive the maximum dose of radiation from NPP radionuclides. In all cases, the predominant contributors to the radiation dose of the population for all types of reactors are ¹⁴ C, inert radioactive gases (mainly ⁸⁸ Kr), and ³ H. The total dose is formed primarily through food consumption: the contribution of this route is 60, 61, and 89 % for VVER-1200, APR-1400, and HPR-1000 NPPs, respectively. Despite differences in reactor types and levels of radiation exposure among age groups, the total dose in all cases falls within established limits.

Conclusion

The performed study proved the normal operation of the Kazakhstan NPP planned for construction in the Zhambyl district of the Almaty region posing no hazard to public health and complying with established radiation safety standards regardless the type of used reactors: VVER-1200, APR-1400, or HPR-1000. Calculations based on emergency scenarios will be presented in the next publication.

Background

The lack of training simulators for the process of dissolving uranium compounds in nitric acid, as well as for many other chemical production processes in the non-reactor part of the nuclear fuel cycle, poses an urgent task of developing simulators to improve the efficiency of training students of chemical and digital specialties, as well as personnel of nuclear industry enterprises, especially taking into account the growing trend of staff turnover.

Aim

To develop a computer training simulator reflecting the specifics of uranium oxide dissolution in nitric acid for controlling the technological equipment of the dissolution unit in normal mode, as well as for preventing and localizing emergency situations.

Materials and methods

The educational version of the simulator is based on an algorithmic and scenario approach using the C++ computer software No. 2024614688 “Computer technology for simulating the nuclear fuel cycle and automation (KT-Nimfa)”. The simulation model calculates dynamic changes in the gas pressure, volume and temperature of substances in the vessel, operation of shut-off valves, interlocks, and automation in normal and emergency operating modes of model elements. The calculation is based on the numerical solution of systems of ordinary differential equations using the Euler method.

Results and discussion

The process flow diagram for the educational version of the computer simulator is based on the analysis of the dissolution process for various compositions of uranium raw materials. A simulation model ensuring the operation of the dissolution unit in various modes has been developed. The simulation results including 3% root mean square error indicate the adequacy of the model and the possibility of its application in the simulator. The developed training method implements automatic assessment of the student’s actions. The hardware and software architecture of the simulator ensures simultaneous training of 12 students at a significantly reduced period of education compared to that without a simulator. In 2024, the developed educational version of the training computer simulator was installed at the National Research Tomsk Polytechnic University (Tomsk, Russian Federation) and included in the educational process for 18.05.2002 and 14.05.2004 specialties.

Conclusion

The developed educational version of the simulator based on a simulation model of uranium oxide dissolution in nitric acid under normal and emergency conditions allows 12 students to be simultaneously involved in the educational process. In addition to a shortened time of education, students can effectively assimilate information about the specifics of radiochemical technology, basics of industrial safety culture, as well as knowledge and skills in managing the technological equipment of the dissolution unit.

The article presents a concept of an option for heterogeneous burning of americium and neptunium in the core of a BN-1200 reactor. The option was developed based on the design layout of the BN-1200 core with mixed uranium-plutonium fuel and an axial blanket with 20% of fuel assemblies replaced by burning ones with actinides and inert filler evenly distributed in the active part. Americium, neptunium or their mixtures can be located in burning assemblies. Burning of americium supposes the annual transmutation volume of ~100 kg. The advantages of the considered disposal method are its high intensity and possibility of localizing actinides in separate burning assemblies. The disadvantages include a significant deterioration in the operational characteristics of the core due to the increased fuel rating and high rate of reactivity loss for burnup.

The paper presents the results of SOCRAT-V1/V2 code simulation of QUENCH-10, -16 experiments for oxidation of zirconium fuel rod claddings in a steam-air environment. The simulation results are consistent with experimental data at the stages of heating, preliminary oxidation, and cooling. However, at the flooding stage after oxygen starvation, the calculated mass of released hydrogen is lower than measured one. The differences are due to the fact that simulation does not take into account the formation of zirconium nitrides at the stage of oxygen starvation and their effect on the oxidation process. Zirconium nitride in the zirconium oxide layer significantly intensifies the diffusion of oxygen, which, in turn, leads to an abrupt increase in the mass of hydrogen at the flooding stage.

The construction of the BN-800 reactor opens the way to the development of a closed nuclear fuel cycle on an industrial scale. After mastering the industrial production of mixed uranium-plutonium fuel, BN-800 was transferred to a full loading with this fuel. The present article demonstrates that plutonium of any isotopic composition can be used for fuel production, provided its content adjusted using a specially developed method. The features of handling fuel assemblies (FAs) due to high radiation of mixed uranium-plutonium fuel are considered. The technology for reprocessing spent FAs with this fuel was successfully tested during the reprocessing of experimental BN-600 FAs. Preparations for the reception and processing of spent FAs at the Mayak Production Association (Ozersk, Russian Federation) are underway. To demonstrate the possibility of burning long-lived actinides in BN-800, reactor tests have begun on three FAs of mixed uranium-plutonium fuel with partially added americium and neptunium.

Justification of a reactor fuelled with circulating molten metal fluoride salts requires the development of special tools for multiphysics simulation. The NRC “Kurchatov Institute” (Moscow, Russian Federation) has developed the MULTIMSR software package for simulating neutron-physical and thermal-hydraulic processes of molten salt nuclear reactors in a 3D setting. The software additionally includes the MSR-NODES tool, which simulates the system in a simplified one-dimensional formulation using the point kinetics approximation to calculate the reactor neutron power. The developed software package was used to simulate transient processes with the loss of forced cooling in the intermediate circuit of a molten salt reactor. The simulation results indicate a similar oscillatory nature of the transient change in the average temperature of the core fuel salt in 1D and 3D simulations. The possibility of using simplified models to study transient processes in molten salt reactors is considered.

In order to improve the resource characteristics of main circulating pumps (MCPs) for nuclear reactors, the JSC “Afrikantov OKBM” conducted calculation studies and experimental work for optimizing the blade profile by improving the hydrodynamics of the flow around the inlet section. The studies aimed to develop impellers of the BN-600 MCP‑1 that are resistant to cavitation erosion. As a result, their cavitation-erosion wear was significantly reduced. The design of BN-1200 MCP‑1 and MCP‑2 ensures criteria based on the operating experience of the BN-600 MCP‑2 without any traces of cavitation-erosion wear after ~167,000 h of operation. In addition, corresponding blade profiling insensitive to feed changes was performed. Three-dimensional calculations were used to optimize the geometry of flow sections without developing expensive test benches. The design of BN-1200 MCP‑1 and MCP‑2, based on the experience of operating sodium pumps and CFD optimization calculations, will ensure the required resource of their flow parts.