Atomic Energy最新文献

Background

A molten salt nuclear reactor used for burning minor actinides from spent fuel assemblies of power reactors represents a promising way to close the nuclear fuel cycle. For testing a molten salt technology, a molten salt research reactor (MSRR) should be designed and constructed.

Aim

To define the engineering and physical design of the MSRR, including key technologies of molten salt reactor burners; to develop a safety concept of the molten salt reactor.

Materials and methods

The present study solves engineering, design, and scientific issues including MSRR design and its neutron-physical and thermal-hydraulic characteristics, selection and justification of construction materials, integrated technology of the operation process, safety concept, etc.

Results

The preliminary engineering and physical design of the reactor, equipment, and heat removal circuits, including emergency ones, is developed and presented. Main technical characteristics of the reactor involve 10 MW thermal power, 73LiF-27BeF₂ + (PuF₃ + AnF_m) fuel salt, 141.6 kg/s salt flow rate in the fuel circuit, 664 and 700 °C salt temperature at the reactor inlet and outlet, respectively, and chromium-nickel alloys as candidate structural materials for the reactor equipment. Calculation justifications are performed to adopt MSRR fundamental design and schematic solutions; the parameters obtained earlier are optimized.

Conclusion

The operating MSRR can be effective to test technological solutions for the development of an industrial high-power 2200–2400 MW(t) molten salt nuclear reactor for burning long-lived minor actinides. The obtained results confirm the feasibility of the design requirements for the MSRR and the possibility of its construction.

Background

An important problem of mass spectroscopy is the accurate determination of the concentration of various ions in the studied substance, which has undergone changes after passing through a high-current ion diode.

Aim

To establish the patterns of changes in the relative concentration of ion beam components passed through an ion diode.

Materials and methods

A self-consistent computer simulation of an ion diode is carried out using the KARAT electromagnetic code based on the particle-in-cell variation of the large particle method. We use the mode of space charge current limitation to measure the concentration of a multicomponent ion beam using mass spectroscopy.

Results and discussion

A significant change in the relative concentration of various ion types on the collector of a high-current ion diode is established compared to the concentrations on its emitter. A correction factor compensating for the detected change is proposed: if the ion diode of the mass spectrometer operates in the mode of space charge current limitation, then the amplitudes of the mass peaks should be corrected by (sqrt{upmu _{i}}) to determine the relative concentration of ion cloud components.

Conclusion

The obtained results are of great importance for interpreting measurements in quantitative mass spectroscopy, as well as for determining the efficiency of separation systems.

The paper considers a method for increasing the accuracy of measuring helium concentration in fuel rods using thermal non-destructive inspection. The results of studying multiple periodic alternating pulsed heating and cooling for the inspected section of the gas cavity cladding in the fuel rod are presented. Thermal non-destructive inspection method increases the reliability and accuracy of measuring helium concentration in fuel rods, thus improving the quality of manufacturing fuel rods and fabricating fuel assemblies under complete inspection. In addition, the method increases the safety of fuel rod operation in the core of nuclear reactors.

As part of a project for a solid blanket of a fusion-fission hybrid plant for neutron irradiation of ²³⁸U and ²³²Th isotopes with their transmutation into ²³⁹Pu and ²³³U consumed by nuclear power industry, studies were conducted to select the optimal source material composition, coolant, and structural materials. Several design options for the blanket module were analyzed and compared; the most promising directions for further development of the blanket design were formulated. The obtained ²³⁹Pu and ²³³U productivity for the developed designs of uranium and thorium blanket modules was used to assess the fusion plant power, which is necessary to achieve an integral productivity comparable to the consumption of the VVER-1000 fuel isotope.

The paper considers the issues of establishing protective action planning zones in case of radiation accidents at nuclear power plants. A deterministic approach is demonstrated as mainly used for solving this problem at the initial stage of radiation accidents. We propose to establish protective action planning zones using the Level 3 Probabilistic Safety Assessment (NF PSA-3) in the initial and recovery periods of accidents.

The paper notes the relevance of developing digital models at the design stage for a detailed analysis of the stress-strain state of the structure. The LS-DYNA dynamic calculation module of the ANSYS software package (USA) was used to study various behavior models of structural materials in the case of an inclined lift trolley emergency descending along guides due to a postulated cable break and impacting on an obstacle. The results of the study have been verified. The research work revealed a less than 10% excess of plastic strain in the bilinear elastoplastic model compared to the Johnson-Cook model. The obtained difference was used to determine the optimal option of using these material models for various structural elements in impact calculations.

The present article assesses the effect of the time nessesary for a hydrogen recombiner to reach its nominal capacity on the formation of explosive concentration for a hydrogen-containing environment in the containment rooms under the conditions of a severe accident at a VVER nuclear power plant. We calculate thermodynamic parameters of the environment in the containment rooms for various times of achieving the nominal capacity of the recombiner. The calculations are performed in the KUPOL‑M integral code. The LIMITS‑V code is used to analyse possible combustion modes of the hydrogen-containing mixture formed during the development of the considered severe accident. An increase in the time for reaching the nominal capacity of the hydrogen recombiner leads to the growth of hydrogen concentration in the containment rooms and emergence of conditions for the detonation of the hydrogen-containing mixture.

The present paper considers two options for the core of the BR-1200 reactor to reduce the fuel loading. The effect of changing the isotopic composition of nitrogen in the fuel on the required launch loading is assessed for each concept. A fuel cycle is assumed closed if irradiated nuclear fuel is reprocessed to recover valuable fissile material for return to the cycle. The introduction of ¹⁵N into the composition of nitride fuel demonstrates reducing the consumption of uranium and plutonium. Prospective isotope separation methods for tonnage production of ¹⁵N are considered. The conventional nitric acid method appears disadvantageous due to the lack of a back flushing necessary for large-scale production. Therefore, we propose the ammonia complex method as a possible option for organizing ¹⁵N production.

The present paper considers the basic diagram of the main operations developed for reprocessing spent fuel particles of a high-temperature gas-cooled reactor. The developed diagram is appropriate for using PUREX hydrometallurgical reprocessing technology in subsequent operations. Main operations of the diagram maintain the integrity of fuel particles for the longest time to ensure the minimal contamination of structural graphite with nuclear materials and fission products. The results of experimental verification of the proposed reprocessing operations on model spent fuel of a high temperature gas-cooled reactor are presented.

Using the example of a power unit with a VVER-1000 reactor, the present paper considers the options for installing a water-to-water heat exchanger on four loops of the main circulation circuit before and after the main circulation pump to increase the power of the nuclear power plant (NPP). The water-to-water heat exchanger performs economizing heating of the secondary circuit feedwater with the coolant at the outlet of the steam generator. The study determines the geometric dimensions of the heat exchanger and its cost. The power generated by a steam turbine due to the increased steam capacity of steam generators is calculated. Specific capital investments in the installation of a water-to-water heat exchanger at an NPP with a VVER-1000 reactor are determined.