Search Results (58)

Nuclear energy can help stop climate change by generating large amounts of emission-free electricity. Nuclear reactor designs are continually being developed to be more fuel efficient, safer, easier to construct, and to produce less nuclear waste. The term advanced nuclear reactors refers either to Generation III+ and Generation IV or small modular reactors. Every reactor is associated with the nuclear fuel cycle that must be economically viable and competitive. An important matter is optimization of fissile materials used in reactor and/or reprocessing of spent fuel and reuse. Currently operating reactors use the open cycle or partially closed cycle. Generation IV reactors are intended to play a significant role in reaching a fully closed cycle. At the same time, we can observe the growing interest in development of small modular reactors worldwide. SMRs can adopt either fuel cycle; they can be flexible depending on their design and fuel type. Spent nuclear fuel management should be an integral part of the development of new reactors. The proper management methods of the radioactive waste and spent fuel should be considered at an early stage of construction. The aim of this paper is to highlight the challenges related to reprocessing of new forms of nuclear fuel. Full article

As a promising nuclear technology, molten salt reactors (MSRs) have a bright future in the energy sector due to their unique advantages such as high efficiency, safety, and fuel flexibility. However, the accurate analysis of fission products in MSRs requires reliable fission yield data. Current reactor burnup analysis often uses the ORIGEN2 code, whose fission yield libraries mainly originate from the outdated 1970s ENDF/B-VI nuclear database, thus risking data obsolescence. This study evaluates ORIGEN2’s fission yield libraries (THERMAL, PWRU, PWRU50) against the modern ENDF/B-VIII.0 library. Through a comprehensive comparative analysis of Oak Ridge National Laboratory’s Molten Salt Reactor Experiment (MSRE) model, numerical simulations reveal library-dependent differences in MSR burnup characteristics. The PWRU library best matches ENDF/B-VIII.0 for U-235-fueled cases in keff results, while the PWRU50 library has minimal keff deviation in U-233-fueled setups. Moreover, in both fuel cases, the fission yield library was found to significantly affect the activity of key radionuclides, including Kr-85, Kr-85m, I-133m, Cs-136, Sn-123, Sn-125, Sn-127, Sb-124, Sb-125, Cd-115m, Te-125m, Te-129m, etc. Additionally, the fission gas decay heat power calculated via the ORIGEN2 library is over 20% lower than that from the ENDF/B-VIII.0 library tens of days after shutdown, mainly due to differences in long-lived Kr-85 production. These findings highlight the need to update traditional fission yield libraries in burnup codes. For next-generation MSR designs, this is crucial to ensure accurate safety assessments and the effective development of this promising energy technology. Full article

Conventional sources of electricity are limited and they pollute the Earth, so it is necessary to think about an additional source of electricity in the future. Nuclear power is one of the options. Two scenarios using different shares of nuclear power in the future are described in this paper. Scenario 1 describes a moderate increase in nuclear energy use in the future, but with a tendency for a larger increase over 2050. Scenario 2 describes a significant increase in nuclear energy until 2100. Both scenarios are divided into three sub-scenarios (total six) in which the use of different nuclear technologies is analyzed (conventional liquid water reactors, fast breeder reactors and molten salt reactors using thorium as nuclear fuel). In all scenarios, the phase-out of fossil fuel power plants is assumed. One part of the power system is covered by nuclear power plants, and the remaining part is covered by renewable energy power plants. After 2050, an increasing share of the electricity system will be taken over by RES power plants. Nuclear fuel stocks are also analyzed. It is calculated that currently known nuclear fuel stocks are sufficient to meet the needs in all six scenarios. The carbon dioxide emissions saved due to nuclear energy use instead of conventional energy power plants are calculated. The CO_2eq emission savings for Scenario 1 is 87.4% of the recommended emission savings under the IPCC. The CO_2eq emission savings for Scenario 2 is more than sufficient. A calculation of the economic profitability of nuclear energy use is made in relation to fossil power plants and renewable energy power plants. According to calculations, nuclear energy is profitable compared to other energy sources. Nuclear energy use is positive from all the mentioned aspects. Full article

Nuclear energy cogeneration, which integrates electricity generation with thermal energy utilization, presents a transformative pathway for enhancing energy efficiency and decarbonizing industrial and urban sectors. This comprehensive review synthesizes advancements in technological stratification, economic modeling, and sectoral practices to evaluate the viability of nuclear cogeneration as a cornerstone of low-carbon energy transitions. By categorizing applications based on temperature requirements (low: <250 °C, medium: 250–550 °C, high: >550 °C), the study highlights the adaptability of reactor technologies, including light water reactors (LWRs), high-temperature gas-cooled reactors (HTGRs), and molten salt reactors (MSRs), to sector-specific demands. Key findings reveal that nuclear cogeneration systems achieve thermal efficiencies exceeding 80% in low-temperature applications and reduce CO₂ emissions by 1.5–2.5 million tons annually per reactor by displacing fossil fuel-based heat sources. Economic analyses emphasize the critical role of cost allocation methodologies, with exergy-based approaches reducing levelized costs by 18% in high-temperature applications. Policy instruments, such as carbon pricing, value-added tax (VAT) exemptions, and subsidized loans, enhance project viability, elevating net present values by 25–40% for district heating systems. Case studies from Finland, China, and Canada demonstrate operational successes, including 30% emission reductions in oil sands processing and hydrogen production costs as low as USD 3–5/kg via thermochemical cycles. Hybrid nuclear–renewable systems further stabilize energy supply, reducing the levelized cost of heat by 18%. The review underscores the necessity of integrating Generation IV reactors, thermal storage, and policy alignment to unlock nuclear cogeneration’s full potential in achieving global decarbonization and energy security goals. Full article

The high-dimensional integral–differential nature of the neutron transport equation and the complexity of nuclear reactors result in high computational costs. A set of reduced-order modeling frameworks based on Proper Orthogonal Decomposition (POD) is developed to improve the computational efficiency for neutron diffusion calculations while maintaining accuracy, especially for small samples. For modal coefficient calculations, three methods—Galerkin, radial basis function (RBF), and Deep Neural Network (DNN)—are introduced and analyzed for molten salt reactors. The results show that all three reduced-order models achieve sufficient accuracy, with neutron flux L₂ errors below 1% and delayed neutron precursor (DNP) L₂ errors below 2.4%, while the acceleration ratios exceed 800. Among these, the POD–Galerkin model demonstrates superior performance, achieving average L₂ errors of less than 0.00658% for neutron flux and 1.01% for DNP concentration, with an acceleration ratio of approximately 1800 and excellent extrapolation ability. The POD–Galerkin reduced-order model significantly enhances the computational efficiency for solving neutron multi-group diffusion equations and DNP conservation equations in molten salt reactors while preserving the solution accuracy, making it ideal for a liquid fuel molten salt reactor in the case of small samples. Full article

The liquid-fueled molten salt reactor (MSR) is one of the candidate reactors for the Generation IV advanced nuclear power systems, which utilizes flowing liquid molten salt as both fuel and coolant. In transients of liquid-fueled MSRs, the distribution change in the delayed neutron precursors (DNPs) in the primary loop has an important impact on system safety analysis. In order to analyze and evaluate this effect, the RELAP5-TMSR code with a 1-D DNP transport model was used to model the Molten Salt Breeder Reactor (MSBR), and several representative transient scenarios, including the loss of primary flow, increase in primary flow, loss of secondary flow, reactivity perturbation, and load change, were simulated and analyzed. The results show that the DNP distribution changes obviously during primary flow transients, especially during the loss of primary flow. Besides, the power response trends at different power levels during the loss of primary flow are different. The analysis results reveal the steady-state and dynamic characteristics of the DNP distribution, indicating that the DNP distribution, temperature feedback, and reactor power are strongly coupled, which has significant implications for the design and safety analysis of liquid-fueled MSRs. Full article

Nuclear power plants have the lowest life-cycle greenhouse gas emissions intensity and produce more electricity with less land use compared to any other low-carbon-emission-based energy source. There is growing global interest in Generation IV reactors and, at the same time, there is great interest in using small modular reactors. However, the development of new reactors introduces new engineering and chemical challenges critical to advancing nuclear energy safety, efficiency, and sustainability. For Generation III+ reactors, water chemistry control is essential to mitigate corrosion processes and manage radiolysis in the reactor’s primary circuit. Generation IV reactors, such as molten salt reactors (MSRs), face the challenge of handling and processing chemically aggressive coolants. Small modular reactor (SMR) technologies will have to address several drawbacks before the technology can reach technology readiness level 9 (TRL9). Issues related to the management of irradiated graphite from high-temperature reactors (HTR) must be addressed. Additionally, spent fuel processing, along with the disposal and storage of radioactive waste, should be integral to the development of new reactors. This paper presents the key chemical and engineering aspects related to the development of next-generation nuclear reactors and SMRs along with the challenges associated with them. Full article

This work presents the synthesis, purification, and characterization of a molten salt fuel for the irradiation experiment SALIENT-03 (SALt Irradiation ExperimeNT), a collaborative effort between the Nuclear Research and Consultancy Group and the Joint Research Centre, European Commission. The primary objective of the project is to investigate the corrosion behavior of selected Ni-alloy based structural materials which are being considered for the construction of fluoride molten salt reactors. During the test, these materials will be exposed to selected liquid molten fuel salts under irradiation in the High Flux Reactor in Petten, the Netherlands. In addition, the properties and distribution of the fission products formed in the fuel salt during burn-up will be studied by various post irradiation examinations. In the SALIENT-03 fuel, U and Pu fluorides, as fissile material, are dissolved in a carrier melt based on a 78⁷LiF-22ThF₄ eutectic mixture to form fuel salts with four different compositions, containing PuF₃, UF₄, UF₃, and CrF₃. This article comprehensively describes all the steps of this fuel synthesis process: the synthesis of the required pure fluoride powders (⁷LiF, ThF₄, UF₄, UF₃, and PuF₃); the mixing, melting, and purification of the different fuel salt compositions; and the fabrication of solid ingots to be loaded into the irradiation capsules. The characterization of the intermediate and final products is also carried out, following a rigorous quality assurance protocol. The quality assurance is achieved using an analytical scheme consisting of mass balance-based conversion efficiency evaluation, X-ray diffraction, and differential scanning calorimetry analyses. All experimental goals were successfully achieved, highlighting promising prospects for advancing future research and development regarding fuel production methods for fluoride-based molten salt reactors. Full article

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR). Full article

Nuclear waste is one of the most important environmental problems of nuclear power plants. A novel renewable distillation method has been proposed for the direct on-site recycling of spent nuclear fuel and the separation of its valuable components from fissile isotopes, which is especially applicable for reactors using liquid fuels. This dry separation technique can be applied in two single, parallel total-reflux columns with integrated separation stages for chlorinated nuclear waste. According to theoretical calculations, high separation accuracy of the UCl4-NpCl4, PuCl3-UCl3, CmCl3-SmCl3, and EuCl3-CsCl fractions could be achieved using twenty-six separation stages and five total-reflux repetitions, demonstrating the high efficiency of the method proposed. A scheme of the future pyroprocessing separation plant is also presented. Full article

The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10⁻⁴); the splitting method could reach a precision of 10⁻² level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors. Full article

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. Full article

Structural materials in fast reactors with harsh radiation environments due to high energy neutrons—compared to thermal reactors—potentially suffer from a higher degree of radiation damage. This radiation damage can change the thermophysical and mechanical properties of materials and, as a result, alter their performance and effective lifetime, in some cases leading to their disintegration. These phenomena can jeopardize the safety of fast reactors and thus need to be investigated. In this study, the effect of radiation damage on the vessels of molten salt fast reactors (MSFR) was evaluated based on two fundamental radiation damage parameters: displacement per atom (dpa) and primary knock-on atom (pka). Following the previous part of this article (Parts 1 and 2), an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) with stainless steel and nickel-based alloy material vessels, respectively, were considered as case studies. The SPECTER and SPECTRA-PKA codes and a PTRAC card of MCNPX, integrated with a module which has been developed in MATLAB, named PTRIM and SRIM-2013 (using binary collision approximation), were employed individually to calculate and compare dpa and PKA (this master module containing all three tools has been appended to the iMAGINE-3BIC package for future use during reactor operations). Additionally, SRIM-2013 was applied in a 3D simulation of a radiation damage map on a small sample of vessels based on the calculated PKA. Our results showed a higher degree of radiation damage in the iMAGINE vessel compared to the EVOL one, which could be expected due to the harder neutron flux spectrum of the iMAGINE core compared to EVOL. In addition, the nickel alloy vessel showed better radiation damage resistance against high energy neutrons compared to the stainless steel one, although more investigations are required on thermal neutrons and alloy corrosion mechanisms to determine the best material for use in MSFR vessels. Full article

With growing interest in molten salts as possible nuclear fuel systems, knowledge of thermophysical properties of complex salt mixtures, e.g., NaCl-CeCl₃, NaCl-UCl₃ and NaCl-UCl₄, informs understanding and performance modelling of the zero power salt reactor. Fuel density is a key parameter that is examined in a multiscale approach in this paper. In the zero power reactor ‘core’ (cm level), the relative fuel density is estimated for the fuel pin disposition, as well as a function of their pitch (strong effect). Fuel density of the ‘pellet’ (mm–µm level) is first estimated on a geometrical basis, then through tracking pores and cracks using 2D (SEM) and 3D (laser microscopy, LM) techniques. For the nanoscale level, ‘grains’ analysis is done using X-ray diffraction (XRD), revealing the defects, vacancies and swelled grains. Initially, emphasis is on the near-eutectic composition of salt mixtures of CeCl₃ with NaCl as the carrier salt. Cerium trichloride (CeCl₃) is an inactive surrogate of UCl₃ and PuCl₃. The results were measured for the specific salt mixture (70 mol% NaCl and 30 mol% CeCl₃) in this work, establishing that microscopy and XRD are important techniques for measurement of the physical properties of salts component pellets. This work is of significance, as densities of fuel components affect the power evolution through reactivity and the average neutronic behaviour in zero power salt reactors. Full article

The analysis of liquid-fueled molten-salt reactors (LFMSRs) during steady state, operational transients and accident scenarios requires addressing unique reactor multi-physics challenges with coupling between thermal hydraulics, neutronics, inventory control and species distribution phenomena. This work utilizes the General Nuclear Field Operation and Manipulation (GeN-Foam) code to perform coupled thermal-hydraulics and neutronics calculations of an LFMSR design. A framework is proposed as part of this study to perform modeling, design optimization, and uncertainty quantification. The framework aims to establish a protocol for the studies and analyses of LFMSR which can later be expanded to other advanced reactor concepts too. The Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) statistical analysis tool was successfully coupled with GeN-Foam to perform uncertainty quantification studies. The uncertainties were propagated through the input design parameters, and the output uncertainties were characterized using statistical analysis and Spearman rank correlation coefficients. Three analyses are performed (namely, scalar, functional, and three-dimensional analyses) to understand the impact of input uncertainty propagation on temperature and velocity predictions. Preliminary three-dimensional reactor analysis showed that the thermal expansion coefficient, heat transfer coefficient, and specific heat of the fuel salt are the crucial input parameters that influence the temperature and velocity predictions inside the LFMSR system. Full article