The article describes a methodology justifying the serviceability of a nuclear power plant containment for predicting the forces in the anchors of reinforcement bundles and concluding about their performance without using hydraulic jacks and the lift-off method. The calculation methodology is based on the results of monitoring the forces in each reinforcement bundle using PSI-01 force sensors, as well as on readings from control and measuring equipment, including containment movement sensors.
This paper presents key test results of various steam generator models evaluated at SSC RF–IPPE JSC to validate designs for sodium-cooled power units. The tests provided critical operational recommendations, including: the required steam superheat temperature to prevent moisture carryover in BN-600 evaporators, the optimum working pressure for BN-350 evaporators to avoid flow instability, and thermodynamic non-equilibrium data for post dryout heat transfer in sodium-heated steam tubes. The study also established stability limits for Field channels and identified unique heat transfer characteristics in high-steam superheated tubes.
The article examines the effects of the Novovoronezh NPP‑2 industrial development on the external radiation dose to personnel calculated using the ROUZ software for design basis accidents. The ROUZ assessment of the dose is compared with the results of mesoscale models used for calculating radiation safety of personnel at nuclear power plants. We have established that increasing atmospheric stability makes ROUZ estimates of radioactive surface contamination for the ground and buildings at an industrial site and personnel exposure doses from clouds and ground surface more conservative than those using mesoscale models, regardless of the propagation direction and emission height.
This work is devoted to the analysis of an experimental approach to modeling the processes of non-isothermal helium jet mixing at the outlet of the prismatic fuel assemblies and entering the bottom plenum of the high temperature gas cooled reactor core. A review of relevant literature and an analysis of similarity numbers characterizing the investigated process were performed. Based on an integrated approach, the characteristics of the experimental model and the operating parameters of the aerodynamic test facility assembled at the NSTU were selected. Computational studies of the three-dimensional coolant flow in a full-scale reactor fragment under normal operating conditions and in an aerodynamic model have validated the design of the experimental model and confirmed the operating parameters of the test facility.
The present study aims to improve the accuracy of dosing process solutions by measuring their flow rate. The paper provides insights to the measurement of low flow rates of aggressive and radioactive liquids, as well as a description of a developed measurement system and results of its testing; a proposed option for increasing the measurement accuracy was industrially tested. The study was carried out within the framework of the “Advanced Engineering Schools” Federal Project (scientific project PISh-NIR-2023-005 “Development and implementation of algorithms and automated process control systems”).
The article discusses the areas of application and possible limitations of using artificial intelligence in nuclear industry. Evolving artificial intelligence can be used in almost all stages of the nuclear fuel cycle and related areas. Methods of artificial intelligence can improve the efficiency of activities in nuclear medicine, science, as well as in the search, systematization, and processing of large amounts of information. The use of artificial intelligence methods in nuclear industry raises a number of issues that need to be comprehensively considered.
The article presents the results of studying the generation modes of pulsed discharges and neutrons in a system with inertial electrostatic plasma confinement under conditions of pulsed-periodic power supply with preliminary gas ionization. The obtained data show that amplitude-time characteristics of the pulsed discharge depend on the gas pressure, bias voltage polarity, and strength of the pre-discharge current. The operating mode of the generator with multipolar power supply of the preliminary and pulsed discharges provides conditions for increasing the current and reducing the ignition delay time of the pulsed discharge. For stable generation of a pulsed discharge with maximum amplitude values, the current density of the preliminary discharge should exceed the threshold.
The paper proposes a method for calculating a cascade with the same number of gas centrifuges in stages, variable stage separation coefficients, and a given concentration of the target isotope in output streams. The method ensures determining the optimal parameters of a gas centrifuge cascade with a given layout based on the calculated identical feeding streams of stages. The computational experiment was carried out for germanium tetrafluoride. The advantages over cascades with a profiled feeding stream of stages are demonstrated using the example of concentrating 70Ge and 76Ge.
The share of nuclear power in global electricity production has fallen from 17 to 9% since 2000. Moreover, the issue of limited natural uranium resources and their availability for the industry becomes acute.
To analytically determine the possibilities of accelerated and sustainable development of global nuclear power industry (NPI) based on thermal nuclear reactors.
Scenarios for the accelerated development of global NPI based on thermal reactors were calculated using analytical methods developed by the authors for predicting the dynamics of natural uranium resource depletion.
The paper provides the development dynamics of NPI with only thermal reactors of a VVER type and its share in the global electricity production. Given the uranium reserves for currently operating reactors, 7.9 MtU of existing natural uranium resources appear sufficient to build only 490 (290) reactors with an operating period of 60 (80) years. Doubling natural uranium resources to 16 MtU and increasing the annual rate of thermal reactor construction to 4% could increase the contribution of nuclear power plants to the global electricity generation from 9 to 11–12% by 2035, yet followed with a rapid decline as uranium is depleted.
The performed analysis demonstrates inadequacy of a long-term and sustainable development for global NPI based only on thermal nuclear reactors. An analysis of development scenarios based on fast reactors will be presented in Part 2.
Reactors with heavy liquid metal coolant (HLMC) based on a eutectic lead-bismuth alloy (Pb-Bi) require a safety justification. The latter includes the obligatory simulation of normal operating and emergency modes using thermal-hydraulic calculation codes.
To develop a one-dimensional three-field model for the joint flow of three fluids including HLMC, steam, and water for a KORSAR/LMR-based software tool numerically simulating in detail physical processes in reactor plants during inter-circuit leaks with water coolant ingress into HLMC.
The KORSAR/LMR thermal-hydraulic code developed by the Alexandrov Research Institute of Technology was used for calculations; validation was performed using published experimental data of the LIFUS 5 facility.
In the course of the study, we have modified the system of conservation equations, flow mode and heat exchange maps, as well as changed an approach to closing relations. The calculated pressure dynamics of the three-field model turned out to be closer to the experimental one as compared to the two-field model consisting of HLMC and steam-gas mixture.
A three-field one-dimensional model of the joint flow for three fluids has been developed and implemented in a new software tool based on the KORSAR/LMR computer code. This software can be used for the predictive safety analysis of reactor plants with HLMC. Model correlations can be further refined due to the expansion of the experimental base.
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