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A thermosiphon is a heat transfer device that utilizes the phase change of a liquid and has a single closed-loop shape in a gravity-dominant field. This can be expressed as a single-step thermosiphon because boiling and condensation occur once per cycle. In contrast, the multistep thermosiphon, introduced for the first time in the field of thermal engineering in this study, is a new heat transfer mechanism in which boiling and condensation occur several times per cycle in a single loop with multiple channels. The new mechanism has a superior heat transfer rate compared to the existing single-step thermosiphon, and the operating pressure of the loop can be lowered. However, as the heat transfer rate increases, the circulation flow in the channel tends to pulsate. This thermohydraulic characteristic was confirmed through theoretical and computational analyses of a two-step thermosiphon. In this study, an improved concept of an asymmetric two-step thermosiphon was developed that can be applied to heat exchanger design by eliminating pulsating flow while maintaining the advantages of a two-step thermosiphon. The newly proposed heat transfer mechanism, termed the multistep thermosiphon, can be effectively used in the design of heat exchangers in industrial fields. In particular, if the asymmetric two-step thermosiphon is applied to the design of small nuclear reactor containments currently being developed in several countries, there are several advantages associated with the reduction of the containment volume and design pressure. [ABSTRACT FROM AUTHOR]

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies. [ABSTRACT FROM AUTHOR]

Marine sources contribute approximately 2% of global energy-related CO₂ emissions, with the shipping industry accounting for 87% of this total, making it the fifth-largest emitter globally. Environmental regulations by the International Maritime Organization (IMO), such as the MARPOL (International Convention for the Prevention of Pollution from Ships) treaty, have driven the exploration of alternative green energy solutions, including nuclear-powered ships. These ships offer advantages like long operational periods without refueling and increased cargo space, with around 200 reactors already in use on naval vessels worldwide. Among advanced reactor concepts, the molten salt reactor (MSR) is particularly suited for marine applications due to its inherent safety features, compact design, high energy density, and potential to mitigate nuclear waste and proliferation concerns. However, MSR systems face significant challenges, including tritium production, corrosion issues, and complex behavior of volatile fission products. Understanding the impact of marine-induced motion on the thermal–hydraulic behavior of MSRs is crucial, as it can lead to transient design basis accident scenarios. Furthermore, the adoption of MSR technology in the shipping industry requires overcoming regulatory hurdles and achieving global consensus on safety and environmental standards. This review assesses the current progress, challenges, and technological readiness of MSRs for marine applications, highlighting future research directions. The overall technology readiness level (TRL) of MSRs is currently at 3. Achieving TRL 6 is essential for progress, with individual components needing TRLs of 4–8 for a demonstration reactor. Community Readiness Levels (CRLs) must also be addressed, focusing on public acceptance, safety, sustainability, and alignment with decarbonization goals. [ABSTRACT FROM AUTHOR]

A miniature nuclear reactor is desirable for deployment as a localized nuclear power station in support of a carbon-free power supply. Coupling aspects of proliferation-resistant fuel with natural burnable absorber loading are evaluated for once-through operation cycle performance to minimize the need for refueling and fuel shuffling operations. The incorporation of 0.075 wt.% 237Np provides favorable plutonium isotopic vectors throughout an operational lifetime of 5.5 years. providing 35 MWe. Core performance was assessed using a verification-by-comparison approach for core designs with or without 237Np and/or gadolinia burnable absorber. Burnup Monte Carlo calculations were performed via MCOS coupling of MCNP and ORIGEN to an achievable burnup of ~62.5 GWd/t. The results demonstrate a minimal penalty to reactor performance due to the addition of these materials as compared against the reference design. Coupling of a proliferation-resistant fuel concept with a uniform loading of natural gadolinia burnable absorber for LEU+ fuel (7.5 wt.% 235U/U UO2) provides favorable excess reactivity considerations with minimized concerns for additional residual waste and more uniform distribution of un-depleted 235U in discharged fuel assemblies. [ABSTRACT FROM AUTHOR]

Currently, little is known about the characteristics of oxide films on Ti alloys exposed to the secondary circuit water environment of small modular reactors (SMRs). In this study, multiple microstructural characterization techniques were employed to meticulously characterize the oxide film characteristics of the TA16 alloy after 6000 h and 8000 h of exposure to a simulated SMR secondary circuit water environment. The results indicate that the oxide films on the 6000 h specimen are composed of a microcrystalline anatase TiO2 interior layer and a coarse-grained anatase TiO2 exterior layer, while the oxide film on the 8000 h specimen typically consists of a microcrystalline anatase TiO2 inner layer, an NiFe2O4 middle layer, and a coarse-grained anatase TiO2 outer layer. As the corrosion time is extended to 8000 h, the presence of NiFe2O4 oxide particles can be attributed to oxide deposition by the 316L stainless steel autoclave used in the corrosion experiments, which results in the migration of Ni and Fe ions onto the surface of the Ti alloy.

The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.

A miniature nuclear reactor is desirable for deployment as a localized nuclear power station in support of a carbon-free power supply. Coupling aspects of proliferation-resistant fuel with natural burnable absorber loading are evaluated for once-through operation cycle performance to minimize the need for refueling and fuel shuffling operations. The incorporation of 0.075 wt.% 237Np provides favorable plutonium isotopic vectors throughout an operational lifetime of 5.5 years. providing 35 MWe. Core performance was assessed using a verification-by-comparison approach for core designs with or without 237Np and/or gadolinia burnable absorber. Burnup Monte Carlo calculations were performed via MCOS coupling of MCNP and ORIGEN to an achievable burnup of ~62.5 GWd/t. The results demonstrate a minimal penalty to reactor performance due to the addition of these materials as compared against the reference design. Coupling of a proliferation-resistant fuel concept with a uniform loading of natural gadolinia burnable absorber for LEU+ fuel (7.5 wt.% 235U/U UO2) provides favorable excess reactivity considerations with minimized concerns for additional residual waste and more uniform distribution of un-depleted 235U in discharged fuel assemblies.

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48–97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56–70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.

The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive and Discrete Gadolinia/Alumina Burnable Absorber (HIGA) rods for controlling excess reactivity sustainably. Through comprehensive analysis and simulations, the reactivity behavior with varying quantities of HIGA rods is examined, leading to the development of optimized fuel assembly designs. Furthermore, the integration of HIGA rods with integral gadolinia BA rods is discussed to enhance reactivity control and operational flexibility further. This approach utilizes the spatial self-shielding effect of gadolinia for extended reactivity management, crucial for stable and efficient reactor performance. The paper thoroughly addresses core design considerations, including fuel assembly configurations and control rod patterns, to ensure safety and performance in initial and reload cycles. This research advances the development of SBF operation in i-SMR by offering practical reactivity management solutions.

This paper proposes a flooding safety system (FSS) and its operation strategy that can provide long-term safety and effective maintenance for modules of small modular reactor (SMR) and metal containment maintained at dried environment during normal operation. During hypothesized accidents, the FSS re-collects the evaporated steam into the common pool by the condenser installed above the common water pool and provides an emergency coolant for the cavities and auxiliary pools. This study suggested that the condensate re-collection strategy using the FSS can effectively delay the depletion of available water in response to the accidents. Without re-collection, the achievable grace periods ranged from 44 to 1507 days for six-module and one-module accidents, respectively. However, with a full re-collection (ratio = 1.0), the time to total depletion of emergency coolant was estimated indefinite. Even with a partial re-collection ratio of 0.3, a grace period of 83.5 days could be ensured for a six-module transient. This study reported the effectiveness of condensate re-collection and the FSS as an innovative safety management strategy and system. Employing a condensate re-collection strategy with a high re-collection ratio can enhance the long-term safety and effective convenience of SMR operations and maintenance.

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

A novel microreactor, called MoveluXTM, was previously proposed that utilizes heat pipes as the primary heat transfer device and calcium hydride as the moderator. In this core design, the moderator temperature is the critical core operation limit because at high temperatures above 800°C, the hydrogen dissociates from the calcium hydride. The core temperature distribution, therefore, was previously evaluated. However, this evaluation did not consider gamma heating in the core and assumed that power was produced only in the fuel region. By contrast, the moderator region has a power density under realistic conditions due to gamma heating. Thus, the present work considers gamma heating in the core power distribution calculation and evaluates the impact on the moderator temperature. The power density of gamma heating was 1/10th that of the fuel region and around 1/100th that of the core thermal power. This increased the temperature of the moderator by 10 K from the case without considering gamma heating. In addition, this temperature distribution difference did not have an impact on the core criticality. In conclusion, considering the gamma heating, concerns regarding the core design are not suggested. [ABSTRACT FROM AUTHOR]

As small modular reactors are gaining attention as the future of nuclear energy, it has become increasingly imperative to minimize the number of required operators in order to improve economic competitiveness. Although the tasks of operators have changed and their workloads have been relatively reduced as digital technologies have been applied to advanced main control rooms (MCRs), no change in the number of operators has been made when compared to conventional MCRs in the Republic of Korea. As a solution, the introduction of natural language processing (NLP) technology to replace certain operator tasks in advanced MCRs can be a valuable means of reducing MCR staff. In this paper, we suggest a novel communication framework utilizing NLP technology to minimize the number of operators in advanced MCRs. To do this, we analyze operator tasks in advanced MCRs and select those that can be replaced by NLP technology. We then develop a prototype NLP-based system and analyze the process and characteristics of the suggested communication framework. [ABSTRACT FROM AUTHOR]

In the present-day conditions under which the nuclear power industry is developed, a need arises to diversify the designs of new nuclear power plant units, which should differ from the previously constructed ones by featuring flexibility to the customer requirements and by using safety systems based on fully passive safety assurance principles. In 2022, specialists of Experimental and Design Organization (OKB) Gidropress commenced activities on elaborating the draft design of a new integral pressurized water-cooled reactor plant VVER-I with natural circulation of coolant for a basic thermal capacity of 250 MW. The design incorporates passive safety systems able to provide reliable heat removal from the core under the conditions of a long-term NPP blackout and without the operator's participation. The article presents the results obtained from thermal and fluid dynamic computations of the new reactor plant carried out using the KORSAR/GP code that has been certified for safety analyses. A reactor plant thermal-hydraulic model, which can be used for computations of stationary normal operation conditions and, subsequently, also for simulating the accident scenarios evolvement dynamics, has been developed and tested. Computations carried out using the system code have confirmed a correct choice of the reactor's main geometric parameters and the steam generator's heat-transfer surface for operation at the nominal power. Based on the computation results for optimizing the design, it is proposed to use a jacketed steam generator, which will make it possible to exclude stray coolant leaks in bypass of the heat-transfer surface. It is shown that the newly developed reactor plant has a significant potential for increasing the thermal power capacity up to 400 MW without introducing fundamental changes in the design. The study results can be used in designing new VVER reactors with natural coolant circulation, and also in the development of passive safety systems. [ABSTRACT FROM AUTHOR]

Decarbonization of energy production is key in today’s societies and nuclear energy holds an essential place in this prospect. Besides heavy-duty electricity production, other industrial and communal needs could be served by integrating novel nuclear energy production systems, among which are low-power nuclear devices, like small modular reactors (SMRs). The ELSMOR (towards European Licensing of Small Modular Reactors) European project addresses this topic as an answer to the Horizon 2020 Euratom NFRP-2018-3 call. The consortium includes 15 partners from eight European countries, involving research institutes, major European nuclear companies and technical support organizations. The 3.5-year project, launched in September 2019, investigates selected safety features of light-water (LW) SMRs with focus on licensing aspects. Providing a comprehensive compliance framework that regulators can adopt and operate, the licensing process of such SMRs could be optimized, helping their deployment. In this prospect, as a result of ELSMOR’s work, this article gives an overview of the specific issues that LW-SMRs may bring about in the different domains of nuclear safety, in terms of: • Methodological standpoints: safety goals, safety requirements, safety principles (defence-in-depth implementation); • Main safety functions of reactivity control, decay heat removal and confinement management; • Severe accident management; • Other safety issues particular to SMRs: use of shared systems; performing of multi-unit probabilistic safety assessment (PSA); spent fuel management, transport and disposal management. In this article, adequate methodologies are developed to deal with these issues and to help assess the safety of LW-SMRs. This work gives a precious synthesis of the safety assessment issues of LW-SMRs and of the associated methodologies developed in the context of the ELSMOR European project.

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

Small nuclear power plants are a promising direction of research for the development of carbon-free energy in isolated power systems and in remote regions with undeveloped infrastructure. Improving the efficiency of power units integrated with small modular reactors will improve the prospects for the commercialization of such projects. Power cycles based on supercritical carbon dioxide are an effective solution for nuclear power plants that use reactor facilities with an initial coolant temperature above 550 °C. However, the presence of low temperature rejected heat sources in closed Bryton cycles indicates a potential for energy saving. This paper presents a comprehensive thermodynamic analysis of the integration of an additional low-temperature organic Rankine cycle for heat recovery to supercritical carbon dioxide cycles. A scheme for sequential heat recovery from several sources in S-CO2 cycles is proposed. It was found that the use of R134a improved the power of the low-temperature circuit. It was revealed that in the S-CO2 Brayton cycle with a recuperator, the ORC add-on increased the net efficiency by an average of 2.98%, and in the recompression cycle by 1.7–2.2%. With sequential heat recovery in the recuperative cycle from the intercooling of the compressor and the main cooler, the increase in efficiency from the ORC superstructure will be 1.8%. [ABSTRACT FROM AUTHOR]

AbstractAPOLLO3® is a French deterministic reactor code for lattice and core calculations. It has been developed at CEA (Commissariat à l’Energie Atomique) with the financial and technical support of EDF (Electricité de France) and Framatome since 2007. The main goal of APOLLO3 is to provide a unified deterministic tool that includes the capabilities of the French lattice and core codes of the previous generation, APOLLO2, CRONOS2, ECCO, and ERANOS, introducing many improvements to reactor physical modeling in a modern and flexible software architecture platform for research and development and industrial activities. This paper presents an overview of the main new capabilities of the APOLLO3 code in the lattice and the core components. [ABSTRACT FROM AUTHOR]

AbstractWe have developed a heat transfer correlation for saturated flow boiling of water in a helically coiled tube. Initially, we collected experimental data encompassing a broad spectrum of thermal-hydraulic conditions and geometric configurations, and examined the influences of key dimensionless parameters, such as the convection number and the boiling number. The data analysis showed that the observed trend aligns with previous studies on boiling heat transfer within a straight tube. Also, we investigated the influence of centrifugal force acting on the fluid in a helically coiled tube and confirmed its significant impact on boiling heat transfer. Based on our findings, we propose a new heat transfer correlation that incorporates a dimensionless number of the centrifugal force divided by gravitational force. The basic structure of this correlation was adapted from the Kandlikar correlation for saturated flow boiling. The new correlation demonstrated enhanced accuracy compared to existing ones. Additionally, we showed its applicability to boiling heat transfer within a straight tube as well. [ABSTRACT FROM AUTHOR]

Small nuclear power plants can provide a stable, carbon-free energy supply to civil infrastructure and industrial enterprises in remote regions isolated from unified energy systems. More than 70 projects of small modular reactors are currently being developed by IAEA member countries; several low-power power units are already supplying thermal and electrical energy to consumers. One of the main limitations standing in the way of widespread dissemination of this technology is the high specific capital cost of a low-power nuclear power plant; therefore, new scientific and technical solutions are needed in this industry. Increasing the thermodynamic efficiency of power cycles of small modular reactors can become a driver for reducing the cost of supplied electrical energy. This paper presents the results of a comprehensive thermodynamic analysis of existing and promising power cycles for small modular reactors. In addition to traditional steam power cycles, cycles using non-traditional working fluids, including carbon dioxide, freons, and helium cycles, are considered. Optimal sets of thermodynamic parameters were determined to ensure maximum net efficiency of electricity production. For water-cooled reactor plants, a maximum efficiency of 33.5% at an initial temperature of 300 °C could be achieved using a steam turbine cycle. It was revealed that for reactor plants with liquid metal and liquid salt coolant in the range of initial temperatures above 550–700 °C, the maximum thermal efficiency was provided by the Brayton recompression cycle with a carbon dioxide coolant: the net electrical efficiency exceeded the level of steam turbine plants, with intermediate superheating of the steam, and could reach a value of 49.4% at 600 °C. This makes the use of these cycles promising for low-power nuclear power plants with a high initial temperature. In small gas-cooled reactor plants with a helium coolant, the use of a binary cycle consisting of a helium Brayton cycle and a steam-powered Rankine cycle provided an efficiency of 44.3% at an initial helium temperature of 700 °C and 52.9% at 1000 °C. This was higher than in the Brayton cycle with a recuperator, with a minimum temperature difference in the heat exchanger of 20 °C: the efficiency was 40.2% and 52%, respectively. Also, the transition to power cycles with non-traditional working fluids will lead to a change in the operating conditions of turbomachines and heat exchangers. [ABSTRACT FROM AUTHOR]

The subject of this article is the development of information technologies at the end of the 20th and the beginning of the 21st century for the Fourth Industrial Revolution in the form of Industry 4.0, i.e., Internet of Things (IoT) technologies. Successes in the implementation of this technology have led to the development of new applications in the energy sector, such as energy management systems, small modular reactor (SMR) management systems, and alternative power supply systems based on renewable energy sources. The digital infrastructure of these management systems is characterized by a high "density of knowledge", which requires clarification of fundamental concepts in information theory, namely the content of the concepts "data", "information", "knowledge" and "meaning of knowledge". Special attention was given to defining the role of knowledge. The aim of this study is to further develop methods and models of semiotic theory by determining the role and place of logical as well as eight- and four-factor logical-semantic models of knowledge bases in semiotic space. The tasks: comparing existing knowledge representation methods and models. Research results: We found that the logical models used in the development of knowledge bases are based on the principles of artificial intelligence theory, which relies on sign system hypotheses and formal logic theory. The main drawback is the complexity of practical implementation in the form of expert systems. For logicalsemantic models in the form of eight-vector graphic models, it was found that there is currently no theoretical justification for defining the vectors that form the coordinate axes, making these models unique to specific subject areas. It was determined that the advantage of using these methods is that an expe rt can independently form such a knowledge model. For logical-semantic models in the form of four-factor graphic models, there is a theoretical justification for defining the factors of the model that form the coordinate axes, making these models universal for specific subject areas. It was established that the advantage of these models is that they can be developed by experts without the involvement of a knowledge engineer. Therefore, it is proposed to use four-factor logical-semantic knowledge representation models for further application. It is also proposed to split the element "logical-semantic models" into two elements in the semiotic spatial model in vector K8 "Knowledge Representation Models", namely: "logical-semantic eight-vector models" and "logical-semantic four-factor models". Additionally, it is proposed to add the element "post-Cartesian representation of metaknowledge" to the element "geometric" in vector K5 "Ideal Models". Conclusions: The theoretical basis for developing eight-factor logical-semantic knowledge representation models is the form of connections between adjacent vectors in the form of Cartesian products on elements of the corresponding inter-coordinate matrices. The theoretical basis for the methodology of developing four-factorlogical-semantic knowledge representation models is the form of connections between adjacent vectors in the form of Cartesian products for elements of the corresponding inter-coordinate matrices, as well as for diametrically opposite pairs of factors in the form of dialectical unity of the concepts "general" and "particular." The application of logical-semantic knowledge representation models for alternative energy-source management systems will ensure increased energy efficiency. Other cases related to the development of databases for SMR digital infrastructure are discussed. [ABSTRACT FROM AUTHOR]

AbstractEvaluating atmospheric dispersion and radiological doses in the vicinity of buildings is required for small modular reactors (SMRs) because of the reduced size of their exclusion area boundary. The current Canadian nuclear industry tool for these calculations implements the methodology defined in CSA Standard N288.2-M91, which was written to support large Canada Deuterium Uranium (CANDU) nuclear reactors as opposed to SMRs. The ORCA (On/offsite Radiological Consequences of Accidents) code has been developed to address this technical concern in addition to evaluating atmospheric dispersion and doses in the far field. The code calculates worker and public doses following an airborne release of radioactive material into the atmosphere under postulated accident conditions at a nuclear facility. The current paper presents the key assumptions and methods utilized in ORCA and discusses qualification of the software to the requirements of CSA Standard N286.7-16. The new model is applicable to SMRs and existing reactor designs and reduces conservatisms in the near field (i.e., <1 km from the source) relative to the methods in CSA N288.2-M91. [ABSTRACT FROM AUTHOR]

The Small Modular Reactor emerges as a recommended technology for nuclear power plant (NPP) development in West Kalimantan, particularly at the Pantai Gosong site. NPP demand highly purified water for cooling system, while seawater contains high TDS, so desalination by utilizing heat and power from NPP, offers a viable solution. Additionally, desalination products can supply freshwater to communities within a 10 km radius, including Sungai Raya and Sungai Raya Island Districts. This study aims to determine the techno-economic aspect of desalination cogeneration with SMR-type NPP for West Kalimantan. The methodologies in this research encompass a comprehensive literature review, a qualitative assessment of desalination technology, and a quantitative economic analysis utilizing the DEEP version 5.1. Techno-economic analysis shows that the simple payback period for MED is 12.3 years, while MSF is 13.9 years. Power lost at MED is 5.3 MWe, while MSF is 9.1 MWe. At a temperature of 30.7 °C, the total water cost of the MED is $1.277 $/m3, while MSF is $1.595 $/m3. MED desalination with temperature 70 °C can be effective using residual heat as an energy source. Cogeneration of MED technology can be option for utilization of heat and power in the context of NPP development in West Kalimantan.

Marine sources contribute approximately 2% of global energy-related CO₂ emissions, with the shipping industry accounting for 87% of this total, making it the fifth-largest emitter globally. Environmental regulations by the International Maritime Organization (IMO), such as the MARPOL (International Convention for the Prevention of Pollution from Ships) treaty, have driven the exploration of alternative green energy solutions, including nuclear-powered ships. These ships offer advantages like long operational periods without refueling and increased cargo space, with around 200 reactors already in use on naval vessels worldwide. Among advanced reactor concepts, the molten salt reactor (MSR) is particularly suited for marine applications due to its inherent safety features, compact design, high energy density, and potential to mitigate nuclear waste and proliferation concerns. However, MSR systems face significant challenges, including tritium production, corrosion issues, and complex behavior of volatile fission products. Understanding the impact of marine-induced motion on the thermal–hydraulic behavior of MSRs is crucial, as it can lead to transient design basis accident scenarios. Furthermore, the adoption of MSR technology in the shipping industry requires overcoming regulatory hurdles and achieving global consensus on safety and environmental standards. This review assesses the current progress, challenges, and technological readiness of MSRs for marine applications, highlighting future research directions. The overall technology readiness level (TRL) of MSRs is currently at 3. Achieving TRL 6 is essential for progress, with individual components needing TRLs of 4–8 for a demonstration reactor. Community Readiness Levels (CRLs) must also be addressed, focusing on public acceptance, safety, sustainability, and alignment with decarbonization goals.

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies.

During the Fukushima Daiichi nuclear accident in 2011, hydrogen explosions occurred in all units from Unit 1 to Unit 3. Consequently, one of the lessons learned from the Fukushima Daiichi accident is the necessity of implementing hydrogen control and mitigation strategies for most Nuclear Power Plants (NPPs). This paper focuses on the incorporation of Passive Autocatalytic Recombiners (PARs) during the design phase of a small modular Pressurized Water Reactor (SMR-PWR) project. The numerical analyses are conducted using the MELCOR v. 2.2 code. Two scenarios are compared: the Total Loss of Feed Water (TLOFW) severe accident with and without PARs. Saphiro’s diagram is utilized to investigate whether the mixture's composition (hydrogen, air, steam) is flammable for both scenarios. It has been observed that the inclusion of PARs leads to a reduction in hydrogen risk (detonative or deflagrative) as the final hydrogen concentration values fall below the flammability limit. This study is preliminary, and further research is required.

The corrosion behavior of a TA16 titanium alloy during exposure in simulated water environment of small modular reactor (SMR) at 320°C for up to 5000 h was investigated by employing dedicated microstructure characterization. The corrosion weight gain curve of the alloy is found to follow a parabolic law with an exponential parameter of n = 0.738. Oxide scales formed on the alloy surface generally show a two-layer structure composed of a relatively compact inner layer (dense oxide layer) and a discontinuous outer layer (microcrystallites). The dense oxide in the inner layer corresponds to anatase TiO2 while the surface microcrystallite layer is mainly comprised of ilmenite FeTiO3. The number and size of the microcrystallites in the outer layer gradually increase with the exposure time. Fe ions from the corroded loop pipe materials should have participated in forming the outer layer. The compact inner layer is thickened gradually with the test time with its thickness growth curve featured by l = 4.245t0.588, which could better reflect the corrosion behavior of the TA16 alloy.

The publication is devoted to the basics of calculating the economic effect of providing electric energy storage systems with the service of compensating the surplus of electric energy generation by nuclear power plants with small modular reactors at night. The main components of the mathematical model for calculating the service tariff for the accumula-tion of a surplus of electric energy generation are considered. A simulation model of the operation of the electric en-ergy storage system is given. A mathematical model was developed for calculating the minimum tariff for the service of compensation for the surplus of electric energy generation by nuclear power plants with small modular reactors at night with the help of electric energy storage systems. Bibl. 13, fig. 3.

The use of nuclear energy can contribute to achieving positive socio-economic and environmental benefits, but nuclear power plants are one of the most water-intensive industries in the world. The use of Small Modular Reactor (SMR) technologies is increasing due to their interesting advantages such as reduction of construction costs and use in remote areas, which favors distributed generation. Hygroscopic Cycle Technology (HCT) can be of great interest for power generation in nuclear power plants, due to the potential improvement in terms of energy efficiency and water savings. This study presents the benefits of implementing HCT in an existing SMR, the HTR-10, based on the classical Regenerative Rankine Cycle (RRC). The HTR-10 is used to produce electricity and thermal energy for District Heating (DH). Analytical models of both cycles have been developed to compare them in terms of energy production and water consumption. Sensitivity analyses of the influence of the main variables have been performed. The results show that by varying the condensing pressures, the thermal power for DH and the net mechanical power production of the HCT increase up to 2.5 % and 1 %, respectively, with respect to the RRC. The maximum tolerable ambient temperature for the plant with the HCT is 43.12 °C, increasing the availability of the plant and avoiding water consumption between 70000 and 88000 m3/year, depending on the operating conditions. Extrapolation of the results suggests that HCT can improve the energy production of nuclear power plants in a more sustainable way, contributing significantly to the energy transition.

Abilene Christian University (ACU) is developing a 1-MW(thermal) molten salt research reactor that will be built on the ACU campus. A conceptual reactor core model was developed to facilitate the safety analysis required for a construction permit. A series of scoping studies were performed seeking to define the reactor core design parameters subject to a variety of design requirements. A Pareto curve identifying the tradeoff between uranium and LiF-BeF2 was determined. Within this curve, at least 250 kg of uranium and 700 kg of LiF-BeF2 are needed, albeit for different reactor configurations and fuel salt compositions. The cylindrical reactor vessel associated with the best-performing fuel salt composition is ~130 cm in diameter, ~170 cm tall, and contains ~2.5 tons of graphite. The conversion ratio of the reactor is low and will require regular refueling. The shift in neutron spectrum observed with the changing fuel salt composition does not significantly impact reactivity loss with respect to burnup. [ABSTRACT FROM AUTHOR]

An experiment was conducted that demonstrates a novel core power distribution reconstruction method based on ex-core detectors using time-dependent measurement at the University Teaching and Research Reactor of Kindai University (UTR-KINKI). Although the proposed method PHOEBE was able to identify the power distribution change caused by control rods under static conditions in a previous experiment, time-dependent experiments were not conducted. Hence, the present study measured time-dependent neutron counts using ex-core detectors to reconstruct the power distribution based on PHOEBE. Extraction of the control rods was expected to cause a shift in the reactor power distribution from the north side to the south, and the results of the power distribution reconstruction also demonstrated this power shift. This result experimentally and qualitatively demonstrated the detection of time-dependent power shifts based on PHOEBE. However, quantitative verification was difficult in this study because there are no verified time-dependent three-dimensional neutronics codes available. This issue will be addressed in a future study when a code becomes available. [ABSTRACT FROM AUTHOR]

One of the well-known technologies that fit well into the goal of reduction of greenhouse gas emissions is nuclear energy. In particular, the change in approach to the design and construction of nuclear power plants led to the development of small modular reactors (SMRs), which are characterized by a broader range of possible applications than large nuclear reactors and the ability to flexibly operate as per load demand. This paper presents an analysis of the thermal loads of a steam turbine rotor operating in a power plant with SMR. Steam-water cycle and turbine train of a 300 MW unit are presented. High-pressure steam turbine rotor and its thermal loading due to varying steam conditions are investigated for a cold startup designed with consideration of the thermal characteristics of nuclear reactors. It was shown by numerical simulations that steam condensation on rotor surfaces plays a crucial role in determining its thermal behaviour. Comparison with conventional rotors has shown that the thermal loading of nuclear turbine rotors is lower and more stable than that of conventional turbines. [ABSTRACT FROM AUTHOR]

Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study’s thermal scattering data calculation.

Hydraulic performance and flow resistance tests were performed to confirm the main parameters of the hydraulic instrumentation that can affect the pump performance of the reactor coolant pump. The flow resistance test offers important experimental data, which are necessary to predict the behavior of the primary coolant when the circulation of the reactor coolant pump is stopped. Moreover, the shape of the hydraulic section of the pump, which was considered in the test, was prepared to compare the mixed-flow- and axial-flow-type models, the difference in the number of blades of the impeller and diffuser, the difference in the shape of the impeller blade and its thickness, and the effect of coating at the suction bell. Additionally, five models of the hydraulic part were manufactured for the experiments. In this study, the differences in performance owing to the design factors were confirmed through the experimental results.

Small nuclear power plants are a promising direction of research for the development of carbon-free energy in isolated power systems and in remote regions with undeveloped infrastructure. Improving the efficiency of power units integrated with small modular reactors will improve the prospects for the commercialization of such projects. Power cycles based on supercritical carbon dioxide are an effective solution for nuclear power plants that use reactor facilities with an initial coolant temperature above 550 °C. However, the presence of low temperature rejected heat sources in closed Bryton cycles indicates a potential for energy saving. This paper presents a comprehensive thermodynamic analysis of the integration of an additional low-temperature organic Rankine cycle for heat recovery to supercritical carbon dioxide cycles. A scheme for sequential heat recovery from several sources in S-CO2 cycles is proposed. It was found that the use of R134a improved the power of the low-temperature circuit. It was revealed that in the S-CO2 Brayton cycle with a recuperator, the ORC add-on increased the net efficiency by an average of 2.98%, and in the recompression cycle by 1.7–2.2%. With sequential heat recovery in the recuperative cycle from the intercooling of the compressor and the main cooler, the increase in efficiency from the ORC superstructure will be 1.8%.

Small nuclear power plants can provide a stable, carbon-free energy supply to civil infrastructure and industrial enterprises in remote regions isolated from unified energy systems. More than 70 projects of small modular reactors are currently being developed by IAEA member countries; several low-power power units are already supplying thermal and electrical energy to consumers. One of the main limitations standing in the way of widespread dissemination of this technology is the high specific capital cost of a low-power nuclear power plant; therefore, new scientific and technical solutions are needed in this industry. Increasing the thermodynamic efficiency of power cycles of small modular reactors can become a driver for reducing the cost of supplied electrical energy. This paper presents the results of a comprehensive thermodynamic analysis of existing and promising power cycles for small modular reactors. In addition to traditional steam power cycles, cycles using non-traditional working fluids, including carbon dioxide, freons, and helium cycles, are considered. Optimal sets of thermodynamic parameters were determined to ensure maximum net efficiency of electricity production. For water-cooled reactor plants, a maximum efficiency of 33.5% at an initial temperature of 300 °C could be achieved using a steam turbine cycle. It was revealed that for reactor plants with liquid metal and liquid salt coolant in the range of initial temperatures above 550–700 °C, the maximum thermal efficiency was provided by the Brayton recompression cycle with a carbon dioxide coolant: the net electrical efficiency exceeded the level of steam turbine plants, with intermediate superheating of the steam, and could reach a value of 49.4% at 600 °C. This makes the use of these cycles promising for low-power nuclear power plants with a high initial temperature. In small gas-cooled reactor plants with a helium coolant, the use of a binary cycle consisting of a helium Brayton cycle and a steam-powered Rankine cycle provided an efficiency of 44.3% at an initial helium temperature of 700 °C and 52.9% at 1000 °C. This was higher than in the Brayton cycle with a recuperator, with a minimum temperature difference in the heat exchanger of 20 °C: the efficiency was 40.2% and 52%, respectively. Also, the transition to power cycles with non-traditional working fluids will lead to a change in the operating conditions of turbomachines and heat exchangers.

The growing demand for electricity, increasing rising fossil fuel prices and global warming in Brazil, together with the lack of access to electricity on the part of the population in certain locations in the Northern Region, such as, for example, Rondônia, Pará and Amazonas, is currently seeking for solutions. The objective of the present study is to carry out a Strengths, Weaknesses, Opportunities and Threats (SWOT) analysis of the inclusion of a project of a floating nuclear power plants (FNPP) in Brazil with a view to solving the lack of electricity in homes and in industries located in remote regions of Brazil. Study results identified several socioeconomic and environmental impacts on this regions. The study concludes that in addition to contributing to the solution of the lack of energy in remote regions, the implementation of this proposal could represent in the future a source of inflow of financial resources in the country and of international collaboration. This SWOT study will encourage investors, researchers, engineers and decision makers to open the debate about the benefit of using a FNPP in Brazil, which in addition to providing economic energy rates for the population, would be a profitable business and an international collaboration.

Small modular reactors with power below 300 MW have the advantages of small specific mass, long lifetime, and flexible power supply, and they are suitable for providing power support for small and medium-sized towns with small populations and remote areas without grid coverage. In this paper, a small modular S-CO2-cooled molten salt reactor is proposed, and the design of a 10 MW small modular chlorine salt fast reactor (sm-MCFR) with 20 years of operation without refueling is presented. The neutron feasibility of the S-CO2-cooled small modular chlorine fast reactor is analyzed in terms of neutron energy spectrum, reactivity control, temperature reactivity coefficient, and power distribution. A distinctive feature of the sm-MCFR is the use of chlorine salts with high heavy metal solubility and a hard energy spectrum, allowing the core size to be minimized while maintaining the maximum lifetime. The designed core is about 2.44 m in diameter and 2.24 m in height. Meanwhile, the sm-MCFR uses control drum control as the control system, which can effectively achieve reactivity control without increasing the reactor size. The final optimized sm-MCFR has a negative temperature reactivity coefficient, which is necessary to ensure the safe operation of the reactor. [ABSTRACT FROM AUTHOR]

The main aim of the natural circulation of the primary coolant in a nuclear reactor is to reject heat from the reactor core to the steam generator without using a circulation pump. In this work, a vertical heater–vertical cooler, high-temperature, high-pressure, nonuniform-diameter, single-phase natural circulation loop is proposed. The rig contains a spacer grid assembly of electrical heaters in a core with a conical section in its upper plenum and a double helical coil steam generator. The proposed loop is analyzed using RELAP5 and various analytical models. First, these models are benchmarked with experimental data from the Facility to Investigate Natural Circulation in SMART or FINCLS. The model results are found to be in good agreement with the experimental data. The same models are then employed to investigate the proposed natural circulation facility, named the Integral PWR-type SMR Test Rig (iPSTR), to investigate the mass flow rate as a function of geometric and process parameters. Core power input was varied from 5 to 82.5 kW at a maximum system pressure of 10 bar and a maximum elevation difference of thermal centers of 3400 mm. Elevation differences of the thermal centers and diameter of core are found to be important parameters that affect thermal-hydraulic performance significantly. However, cone angle, spacer grid, and system pressure are found to have no significant effect on the performance of the iPSTR. Moreover, the proposed iPSTR is found to possess higher Reynolds number compared with the existing facilities. [ABSTRACT FROM AUTHOR]

Small modular reactors represent a promising technology for power generation, offering solutions to the energy crisis and mitigating greenhouse gas emissions. As Ukraine considers the deployment of NuScale, UK SMR, and SMR-160, it is crucial to address the safe management of spent nuclear fuel. This study focuses on evaluating the radiological characteristics of spent nuclear fuel from the selected small modular reactors and for comparison, from the VVER-1000 reactor. Using the Monte Carlo code Serpent, depletion calculations were performed for an assembly in an infinite 2-D geometry, and the activity, decay heat, and inhalation toxicity of the spent nuclear fuel were assessed. We determined the main nuclides contributing to the radiological characteristics and quantified the mass content of these nuclides. The total number of spent nuclear fuel assemblies produced during the entire life of each small modular reactor type was estimated. The radiological characteristics assessed for the three small modular reactors do not exceed those observed for WER-1000 reactors currently operating in Ukraine. So, spent nuclear fuel generated by the selected small modular reactors will introduce no new challenges to Ukraine's radioactive waste management system. The results of this work provide valuable insights for identifying the optimal small modular reactor technologies for Ukraine concerning safe spent nuclear fuel management. [ABSTRACT FROM AUTHOR]

Neutron thermalization leads to the complexity of the scattering cross-section calculation, which influences the accuracy of the neutron transport calculation in the thermal energy range. The higher precision of thermal scattering data is demanded in the small modular reactors (SMRs) design, especially for small-sized PWRs and SCWRs. Additionally, the thermal neutron scattering problems in supercritical water have not yet been solved. In this study, the thermal neutron scattering problems in subcritical water are tested. Based on thermal neutron scattering theory, the GA model and IKE model were analyzed. This work selected the corresponding input parameters, such as the frequency spectrum, the discrete oscillator energy, weight parameters and so on, as well as preliminary studies on how to calculate the thermal scattering data for HinH2O to accomplish the calculation at various temperatures by developing LIPER code. The deviation between the calculated and reference results, which were both obtained by the Monte Carlo code, COSRMC, was below 0.2 pcm. The deviation of the scattering cross-section between the calculation results and reference was below 0.1%, indicating the reasonability of this study's thermal scattering data calculation. [ABSTRACT FROM AUTHOR]

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

Most power plants, particularly those that burn fossil fuels such as coal, oil, and gas, create CO2, a greenhouse gas that contributes to climate change. By 2060, the Indonesian government has committed to reach net zero emissions. With the lowest CO2 emissions, nuclear power plants are dependable sources of energy. Small modular reactors (SMRs) are a particular kind of nuclear power plant that has the potential to be Indonesia's first commercial nuclear power plant because of their small size, low capacity, uncomplicated design, and modular characteristics. The purpose of this study is to examine the economics and technological feasibility of SMRs. In this analysis, the levelized cost of electricity (LCOE) comparative method and the technology readiness level (TRL) approach are both applied. The SMRs with a minimum TRL value of 7 were CAREM-25 (TRL7), KLT-40S (TRL8), and HTR-PM (TRL 8), according to the results of this research. Although CAREM-25 and KLT-40S are still in the demonstration stage and have not yet entered the market, their LCOE estimates are greater than 0.07 USD/kWh with a 5% discount rate. Whereas CAREM 100 MW is an economy scale from CAREM-25 and VBER 300 MW is a commercial size from KLT-40S, HTR-PM is already an economy scale. With discount rates between 5% and 10%, the LCOE values of HTR-PM, CAREM 100 MW, and VBER 300 MW range from 0.06 USD to 0.12 USD per kWh. Other than hydropower and coal-fired power plants, these LCOE figures can compete with the local LCOE in Indonesia and the LCOE of a variety of other types of power plants. [ABSTRACT FROM AUTHOR]

This scientific paper discusses the importance of reducing greenhouse gas emissions to mitigate the effects of climate change. The proposed strategy is to reach net-zero emissions by transitioning to electric systems powered by low-carbon sources such as wind, solar, hydroelectric power, and nuclear energy. However, the paper also highlights the challenges of this transition, including high costs and lack of infrastructure. The paper emphasizes the need for continued research and investment in renewable energy technology and infrastructure to overcome these challenges and achieve a sustainable energy system. Additionally, the use of nuclear energy raises concerns, such as nuclear waste and proliferation, and should be considered with its benefits and drawbacks. The study assesses the feasibility of nuclear energy development in Latvia, a country in Northern Europe, and finds that Latvia is a suitable location for nuclear power facilities due to potential energy independence, low-carbon energy production, reliability, and economic benefits. The study also discusses methods of calculating electricity generation and consumption, such as measuring MWh produced by power plants, and balancing supply and demand within the country. Furthermore, the study assesses the safety of nuclear reactors, generated waste, and options for nuclear waste recycling. The transition to a carbon-free energy system is ongoing and complex, requiring multiple strategies to accelerate the transition. While the paper proposes that nuclear energy could be a practical means of supporting and backing up electricity generated by renewables, it should be noted that there are still challenges to be addressed. Some of the results presented in the paper are still based on studies, and the post-treatment of waste needs to be further clarified. [ABSTRACT FROM AUTHOR]