Atomic Energy最新文献

Conventional economic assessments of nuclear power plant projects include a standard set of design technical and economic characteristics using established methods of economic and investment analysis. Evaluating innovative projects of small modular reactors (SMRs) should account for additional factors related to the specifics of projects, their purpose and territorial location, features of operating modes, optimal project capacity for a specific scenario, etc. Additional factors make project economic estimates more realistic. The article examines critical aspects and factors capable of affecting the economic performance of SMR projects operating in different market niches. Moreover, the article touches upon issues related to the risks of implementing these projects. SMR projects are implemented under external conditions complementing the technical and economic assessments of power plant projects. These conditions include scenario factors, site conditions, regulatory aspects, design requirements, as well as organizational, financial, and market mechanisms. The concept of extended analysis in performing economic assessments of innovative SMR projects is urgent due to their various practical applications.

The fire protection system of sodium equipment rooms ensures fire safety of fast neutron reactors by localizing and reducing the consequences of leaks and combustion of sodium during depressurization of equipment or pipelines of sodium systems. The present article describes the composition and operation of the fire protection system used at the BN-350 reactor plant. The efficiency of various subsystems and equipment of the fire protection system is considered as applied to the localization of sodium leaks occurred during the operation of the reactor plant. The performed experiments demonstrate the efficiency of drain fire extinguishing as the element of the fire protection system; requirements for drain fire extinguishing subsystems are provided. The article additionally presents data on the most significant sodium coolant leaks during the operation of the BN-350 reactor and considers trends in the development of fire protection systems for sodium equipment rooms of BN reactors. Technical solutions implemented in fire protection systems for the sodium coolant rooms of BN-600 and BN-800 next-generation reactors are examined based on the experience of using fire protection system elements.

The paper considers the development of an innovative lead coolant sampling device for the BREST-OD-300 pilot demonstration fast neutron reactor developed in the Russian Federation within the framework of the Generation IV international initiative. The sampler is equipped with a lateral inclined channel located above the level of the lead coolant and a curved section for sampling both at an operating reactor and in a shutdown mode. The sampler design is protected by two patents of the Russian Federation. The results of tests on loading/unloading a capillary transport tube into a curved sampling channel and the reverse operation are considered. Moreover, the paper presents the results of tests on filling a sampling capsule with a lead coolant sample and its holding during removal from the reactor. The software of the sampler control system was tested for automatic sampling. The performance of the designed sampler is additionally confirmed by thermal engineering, strength, and radiation calculations.

The paper considers the developed calculation model of the IGR reactor with an upper cover and concrete biological protection. We have simulated photon transport using the MCNP5 code and the ENDF/B‑5, 6 constant libraries. IAEA and JAEA nuclear data libraries were used to describe the characteristics of fission products, as well as the processes of uranium-235 decay and gamma-ray field formation. The proposed approach can be used to spatially visualize the operating conditions of a nuclear reactor after startups. We have assessed the field and rate of the effective dose in the central hall of the IGR reactor during its operation at a constant power of 100 MW for 40 s. The effective dose rate was determined to estimate potential radiation load on small-sized neutron detectors (fission chambers). These detectors are supposed to measure local values of fast neutron flux density during irradiation of devices in the reactor. We have validated the proposed method for calculating the effective dose of gamma radiation using the results of direct measurements in the central reactor hall. Validation results confirm the correctness of the proposed calculation models and methods, as well as the admissibility of their use for assessing the radiation situation in the central hall of the IGR reactor.

Numerical simulation based on contemporary computational codes represents a current trend for the safety justification of existing and planned nuclear power facilities including reactor plants with liquid metal coolants. The present study considers neutron codes developed by IBRAE RAN under the Proryv project direction. The paper discusses CORNER and ODETTA codes based on the discrete ordinate method, the BPSD code for isotope kinetics, activity and decay heat calculation, EVKLID/V1 and EVKLID/V2 integral codes, as well as the COMPLEX software system for substantiating nuclear and radiation safety.

The paper discusses requirements for the development of the Russian energy complex and electric power industry during the transition to accelerated scenarios for the growth of the country’s economy. We have simulated macroeconomic scenarios and determined energy needs of the domestic market and effective export of fuel and energy, as well as the necessary volume and structure of energy resource production. Difficulties in reconciling the objectives of intensive growth and deep decarbonization of the economy are noted. The scenario-optimized structure of electricity production and installed capacity of power plants is presented. The study demonstrates the expediency of increasing the role of nuclear power plants (NPPs) in the electric power industry. Key requirements for the pace and cost of commissioning new NPPs have been identified. The importance of technological learning effects for enhancing the competitiveness of NPPs in the changing national energy system is noted.

The present paper considers thermal-hydraulic tests of an emergency cooldown system for the BN-800 reactor with forced circulation of sodium in the primary and secondary circuits. The operating range of changes in the flow rate of coolants in the system under various operating modes is determined. Heat transfer in the air heat exchangers of BN-600 and BN-800 reactors is examined. The established Nusselt number dependence is in satisfactory agreement with the RD 24.035.05-89 methodological guidelines. A simple relationship has been obtained using the results of tests and calculations for assessing the capacity of air heat exchangers. The possibility of increasing their capacity by 26% compared to the design value is demonstrated. The significance of the thermosiphon effect on the temperature state of fittings in the emergency cooldown system is demonstrated. The experimental models have revealed the effect of pipeline inclination on the intensity of heat transfer.

The paper considers the issues of processing spent alkaline liquid metal coolant of BN reactors. Methods used in the BR-10 research reactor for processing coolant and its residues are described in detail. The main parameters of solidified products including mechanical strength and cesium-137 leaching rate are provided. The obtained data are compared with the indicators of cement compounds immobilizing an alkaline solution obtained by dissolving radioactive sodium in 10% alkali. The efficiency of purifying high-salt concentrates of liquid radioactive waste generated during reactor decommissioning is provided for three options of modified clinoptilolite. An approach to processing liquid radioactive waste concentrates is briefly described. The current state of the BN-350 reactor is considered indicating the need for processing accumulated radioactive waste for further decommissioning.

The paper presents a local system for purifying gas flows from volatile compounds of cesium and iodine released during high-temperature operations of spent nuclear fuel reprocessing. The system consists of two universal cartridge gas purification units. The performed study include data on the efficiency of capturing cesium oxide and molecular iodine by ceramic highly porous block-cell contact elements with an aluminosilicate and silver nitrate active layers, respectively. Recommendations for further conditioning of spent cartridges are provided.

The present paper considers the determination of the carbon monoxide pressure under the coating of fuel particles using the Homan and Proksch formulas. The thermodynamics of the kernel+buffer model system was investigated. For this model, the pressure of carbon monoxide is an order of magnitude higher than the pressure obtained using the Homan or Proksch phenomenological relations. The equilibrium kernel composition was calculated separately from the buffer layer. Unbound oxygen formed during uranium fission in the kernel model system interacts with uranium dioxide to form U₄O₉, U₃O₈, and UO₃ compounds, while its diffusion into the buffer layer is suppressed. Oxygen may release into the buffer layer from the surface layer of the fuel kernel by a mechanism similar to the migration of gaseous fission products under the impact of uranium fission fragments.