Fast Reactor Research Articles

CFD simulation of injection point design for emergency core cooling system of ALLEGRO

The Emergency Core Cooling System (ECCS) is one of the critical safety systems in the Gas-cooled Fast Reactor (GFR) demonstrator ALLEGRO. The ECCS design must be improved to enhance fundamental safety functions, particularly in removing heat from the core under postulated plant conditions. In this paper, we focused on exploring the effect of coolant injection point on the performance of ECCS, for more efficient core cooling under Loss Of Coolant Accident (LOCA) scenarios. We initially performed a 1D primary system loop thermal hydraulic simulation to determine appropriate boundary conditions in case of a LOCA. These conditions were then applied to a CFD model of the ALLEGRO reactor pressure vessel, including injection nozzles at two positions. Transient 3D CFD simulations of LOCA in ALLEGRO were carried out using Fluent to determine location of the most effective emergency coolant injection point. We simulated the flow distribution through the core up to around 30 s after both small- and large-break LOCA events. The results suggest that bottom injection of the emergency coolant is preferred in an updated ECCS design. It provides a more uniform flow distribution across the core and therefore more advantageous than the previously suggested location in the downcomer.

Novel Nuclear Power and Fossil Fuel Hybrid Propulsion Systems for Long-Endurance Unmanned Aerial Vehicles: Configuration Comparison and Performance Optimization

Aircraft nuclear propulsion system can significantly enhance the endurance, but its large size and weight become disadvantages against the conventional fossil fuel propulsion system. In this paper, a novel nuclear power and fossil fuel hybrid propulsion system, which combines the lithium-cooled fast reactor and the kerosene-fueled turbine engine, is proposed to meet the power requirements of different flight segments. Performance assessment and comparison among different hybrid configurations is performed. Results indicate that the configuration combining the nuclear power module with a high bypass ratio turbofan engine is the most promising for aircraft. The primary optimization shows that there is a trade-off between the weight of the nuclear power module and the specific thrust, and the optimal weight of nuclear power module is 3942 kg, with the specific thrust of 195.8 N·kg-1·s-1. The aircraft equipped with the hybrid propulsion system operates at the maximum altitude of approximately 18.2 km and reaches a maximum Mach number of 0.71, allowing for sustained flight for several months in cruise segment. The turbofan engine, nuclear reactor, ancillary systems, shield, and heat exchanger account for 11%, 23%, 11%, 17% and 1% of the fuel mass of the original unmodified aircraft.

Experimental and model development of lead-bismuth eutectic solidification phenomenon flowing inside the tube

Lead-bismuth fast reactor may solidify under accident conditions due to overcooling in components such as steam generators, waste heat exhaust heat exchangers, etc. In order to investigate the impact of the solidified layer adhering to the wall surface on the transient flow heat transfer and pressure drop characteristics of lead-bismuth eutectic, an experimental study of lead-bismuth eutectic solidification phenomenon flowing in a vertical circular tube was conducted. Moreover, a rapid prediction model for lead-bismuth eutectic solidification phenomenon was developed and verified with experimental results. The experimental results demonstrate that the formation of solidification layer results in a reduction in time-averaged Nu by approximately 43%, while concurrently exhibiting a sustained increase in pressure drop of the fluid flow. Additionally, the maximum thickness of the solidification layer on one side is 6.4 mm (inner diameter of the round tube is 20 mm), while the average thickness is 1.97 mm. The findings demonstrate that the formation of the solidification layer not only markedly diminishes the heat transfer efficiency between the lead-bismuth eutectic fluid and the wall but also augments the flow resistance.

Experimental study on laser cutting process of simulated fast Reactor fuel rods

The cladding of fast reactor fuel rods, made of stainless steel, presents significant challenges in cutting due to its ductility, which leads to increased tool wear and poor cut quality with traditional mechanical methods. Laser cutting has emerged as a superior alternative, offering non-contact precision, high efficiency, and suitability for radioactive environments. This study systematically investigates the effects of laser cutting parameters—cutting speed, focal position, power, and gas pressure—on the cutting quality of simulated fast reactor fuel rods. The results show that optimal cutting is achieved with a cutting speed of 1 m/min, a focal position between − 20 and − 25 mm, a laser power between 7200 and 9600 W, and a gas pressure of 10 MPa. These parameters provide the best balance between cutting efficiency, surface roughness, and minimal slag formation. This study contributes valuable insights into optimizing laser cutting technology for nuclear fuel rod processing, with potential applications in fuel reprocessing and decommissioning.

Optimization of the route of the refueling machine to reduce the refueling time: Case study of the BN-800 reactor

Sodium-cooled fast reactors (SFR) are included in the list of fourth-generation nuclear energy systems, the so-called generation IV (GEN-IV). It is the only technology of GEN-IV that possesses the significant practical experience of design, construction, and operation of high-power reactors. Due to the high temperature of the coolant at the BN-600 and BN-800 reactor core outlets, the steam generator produces steam with higher enthalpy, and this significantly increases the efficiency of cogeneration at SFR power plants in comparison with pressured water reactor (PWR) or boiling water reactor (BWR) units. The primary nuclear fuel effectiveness rises due to cogeneration with higher efficiency of a sodium-cooled fast reactor nuclear power plant (NPP) in comparison with thermal reactors. In addition, it is necessary to reduce the reactor refueling outages to improve the nuclear power plant utilization factor.It is possible to reduce the duration of the reactor refueling by the route optimization of the refueling machine movement. The prevailing in the world thermal neutron reactors with water cooling have a refueling machine that points the certain fuel assembly using two coordinates. The fast neutron reactors use sodium as a coolant; thus, there is a problem with its violent reaction with water and air oxygen. Therefore, it is necessary to exclude the contact of sodium and surrounding air during the refueling. To achieve this, the system of refueling machine pointing is used, consisting of two or three eccentrically located rotating plugs with hydraulic locks. The article presents the results of the creation of mathematical models of refueling machine movement and fuel assembly gripping. The time-optimal algorithms for the operation of refueling machines with three rotating plugs are proposed. The use of the new algorithm allows to reduce the time of movement of the grip of the refueling machine by 30–37%.

Self-protection of fast lead-cooled reactors against the partial blockage of the core flow area

Self-protection of fast lead-cooled reactors against the partial blockage of the core flow area

Actual problems of modeling thermal-hydraulic processes in fast neutron reactors

The results of studies on hydrodynamics and heat transfer processes in fast neutron reactors are presented. Data on turbulent momentum transfer in rod bundles are analyzed. It is shown that the intensification of turbulent momentum transfer in the rod bundle channels is due to large-scale turbulent momentum transfer (secondary currents). The intensification of interchannel turbulent exchange in close-packed lattices of rods is explained. A dependence is obtained for the dissimilarity coefficients of interchannel convective exchange forced by wire wrapping in bundles of rods. The methods and results of numerical modeling of thermal hydraulics using the Monte Carlo method, thermomechanical analysis of the temperature field in fuel rod assemblies in the lifetime process are presented. The results of modeling based on a water model of temperature fields and the structure of coolant movement in the primary circuit of the reactor in various regimes are presented. A stable temperature stratification of the coolant was revealed in the peripheral zone of the upper chamber of the reactor above the side screens. It is shown that the process of boiling liquid metals in fuel assemblies has a complex structure, characterized by stable and pulsating regimes and a heat transfer crisis. The agreement between the results of experimental and numerical modeling is shown. A cartogram of the flow regimes of a two-phase flow of liquid metals in fuel rod assemblies has been plotted. The influence of the surface roughness of fuel elements on the boiling process and heat transfer during boiling of liquid metals is analyzed. Long-term cooling of a fuel assembly with a “sodium cavity” above the reactor core in accident regimes with boiling of liquid metals is shown. The objectives of further research are formulated.

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

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).

Calculation analysis of a severe accident at a nuclear power plant with a sodium-cooled fast reactor

Calculation analysis of a severe accident at a nuclear power plant with a sodium-cooled fast reactor

A multidisciplinary framework from reactors to repositories for evaluating spent nuclear fuel from advanced reactors

This study presents a multidisciplinary reactor-to-repository framework to compare different advanced reactors with respect to their spent nuclear fuel (SNF) disposal. The framework consists of (1) OpenMC for simulating neutronics, fuel depletion, and radioactive decays; (2) NWPY for computing the repository footprint given the thermal constraints; and (3) PFLOTRAN for simulating radionuclide transport in the geosphere to quantify the repository performance and environmental impact. We first perform the meta-analysis of past comparative analyses to identify the factors that led previously to their inconsistent conclusions. We then demonstrate the new framework by comparing five reactor types. Our analysis highlights the granularity and the specificities of each reactor and fuel type so that we should avoid making sweeping conclusions about advanced reactor SNF. Significant findings are that (1) the repository footprint is neither linearly related to SNF volume nor to decay heat, due to the repository’s thermal constraint (2), fast reactors have significantly higher I-129 inventory, which is often the primary dose contributor, and (3) the repository performance primarily depends on the waste forms. The TRISO-based reactors, in particular, have significantly higher SNF volumes compared to the others but result in smaller repository footprints and lower peak dose rates. The open-source framework ensures proper multidisciplinary connections between reactor simulations and environmental assessments, as well as the transparency/traceability required for such comparative analyses. It aims to support reactor designers, repository developers, and policymakers in evaluating the impact of different reactor designs, with the ultimate goal of improving the sustainability of nuclear energy systems.

Multi-principal element alloys for fast reactor cladding applications

Given the extensive list of multi-principal element alloys (MPEAs) within literature and the overwhelming number of alloys that can be made from only a handful of elements, this article proposes a methodology for prioritizing alloys for use as cladding within advanced reactors. This paper applies neutronic, mechanical, and chemical assessments to a collection of MPEAs available within literature for use as advanced reactor cladding. The results are compared to a reference design of HT9, a ferritic/martensitic steel, for employed in sodium-cooled fast reactor. Mechanical assessments determined pressure limits for a thin-walled tube pressure boundary, thus relating the material's mechanical properties to a minimum acceptable wall thickness. Neutronic analyses reveal a maximum allowable wall thickness that a given material must meet to provide a level of neutronic economy that is equivalent to than that of HT9. Lastly, each alloying element is compared to typical fission products found in a fast reactor fuels to mitigate deleterious phenomena such as fuel-cladding chemical interactions (FCCI). These analyses indicate that Mo-Nb-Ti-V-based alloys are likely to be advantageous and that the inclusion of elements with a high neutronic penalty (e.g., Hf) could be considered with minimal consequences.

Corrigendum to “Alternative core configurations analysis to improve the neutronics performance of modular gas cooled fast reactor” [Ann. Nucl. Energy 211 (2025) 110951

Corrigendum to “Alternative core configurations analysis to improve the neutronics performance of modular gas cooled fast reactor” [Ann. Nucl. Energy 211 (2025) 110951

Deposition behavior of PbTe doped LBE aerosol and Te valence prediction: platform test and first-principles calculation

In fast reactor investigation with lead-bismuth eutectic(LBE) coolant, understanding the source term within the reactor and its environmental migration is crucial for managing radiation hazards from 210Po aerosols. The numerical simulations using empirical parameters have proffered insights into the theoretical migration and settling rates of 210Po aerosols. However, the scarcity of platform tests has impeded the acquisition of particle size distributions and settling velocity, thus weakening the mutual confirmation between experimental and theoretical validation. In this study, an LBE aerosol testing platform (LATP) was designed and established to obtain the particle concentration data to predict 210Po migration, where Te was employed as an experimental surrogate. The particle concentration and size distribution function of PbTe-doped LBE aerosol were measured by an aerosol spectrometer and a universal scanning mobility particle sizer, revealing the particle size distribution spanning from 0 to 800nm. Under normal operating conditions (873K), the pinnacle particle size of the aerosol concentration is 47nm, which shifted to 41nm under accident conditions (1223K). Notably, the highest mass concentration of particles under both circumstances falls within the 200-300nm range. The settling velocity of PbTe-doped LBE aerosol increase with the particle size, and ranging from 5.0×10-7 to 7.1×10-5m/s. First-principles calculations and X-ray photoelectron spectroscopy results indicate that PbTe-doped LBE aerosols should preferentially generate TeO2 during the interaction with oxygen. This work provide a reasonable prediction method for the migration characteristics of polonium under severe accident.

Numerical study on the interaction of high-pressure water injection with lead-bismuth in the SGTR accident of LFR

The lead-bismuth cooled fast reactor (LFR) is one of the fourth-generation advanced reactors, which has the advantages of enhanced safety and high efficiency. However, the steam generator tube rupture (SGTR) accident is an important design-based accident in safety analysis, which may cause a severe threat to the integrity of the reactor system. In this paper, a numerical method was established to simulate the interaction of high-pressure water injection with lead-bismuth, and validated against the LIFUS5 experiment. A sensitivity analysis of the effect of water and lead-bismuth parameters on the water flow rate, vapor generation, and pressure evolution was conducted. The results indicated that the vapor generation at the early time was mainly caused by the water flash evaporation, followed by an accelerated vapor generation rate due to the enhanced heat transfer resulting from the vapor bubble fragmentation and the direct contact of lead-bismuth with water. It was found that a significant temperature gradient existed at the vapor and lead-bismuth interface. There was a pressure spike at the impact of water injection to the lead-bismuth, followed by a continuous increase of the overall pressure and the final stable pressure in the lead-bismuth vessel exceeded the initial water injection pressure. Sensitivity analysis results indicated that, as the water injection pressure increased, the water flow rate, vapor mass, and pressure in the vessel increased significantly, but the temperatures of water and lead-bismuth had little influence on the interactions.

On plutonium-241 and americium in the two-component nuclear energy system

The paper deals with the influence of the strategy for using separated plutonium from spent fuel of thermal and fast reactors, as well as homogeneous burning of americium in the BN reactor core on the balance of americium in the Russian nuclear fuel cycle throughout the 21st century. The assessment is carried out with the use of mathematical modeling of nuclear materials movement including nuclide transformations throughout the nuclear energy system based on the CYCLE code. Scenario modeling of the americium and Pu-241 accumulation was carried out in Russia’s two-component nuclear energy system model with thermal (VVER) and fast (BN) reactors. In so doing, spent nuclear fuel (SNF) reprocessing was simulated in 2 options: as priority reprocessing of SNF of VVER reactors (1) or SNF of BN reactors (2). In addition to americium accumulation in the system without burning, the accumulation of this actinide was studied taking into account its homogeneous burning in the MOX-fuel of fast reactors at the level of its equilibrium content of ~1%. It has been shown that the priority reprocessing of VVER spent fuel makes it possible to reduce the americium accumulation by the end of the century by ~8 tons, and the effect is achieved by using freshly separated plutonium with short-term cooling, thus, by priority the americium source is eliminated, without directly handling it. Homogeneous addition of americium to the fuel of fast reactors of the BN-1200 type at a level of ~1% makes it possible to stop accumulation of americium in a two-component energy system by 2070, stabilizing it at a level of ~40 tons in the scenario with priority reprocessing of VVER SNF and ~50 tons in the scenario with priority reprocessing of BN SNF.

LES and RANS Modeling of an 84-Pin Hexagonal Rod Bundle with Spacer Grid and Experimental Validation

As part of an ongoing integrated research project (IRP), detailed investigations of the flow characteristics of an 84-pin hexagonal rod bundle representing a potential modular gas-cooled fast reactor design have been performed. The rod bundle features a pitch-to-diameter ratio of 1.5 and a large central guide tube, along with simple spacer grids and an outer enclosure. The primary focus of the current work is modeling and simulation of this geometry using multiple computational techniques. These include high-fidelity large eddy simulation (LES) and advanced Reynolds-averaged Navier-Stokes (RANS) approaches. The flow predictions are validated at a Reynolds number of 12 000 using matched index of refraction particle image velocimetry experimental data that were also generated by Texas A&M University as part of the IRP. The comparison data include velocity magnitude and turbulent kinetic energy (TKE) at different lateral locations and at multiple distances downstream of the spacer grid. Many characteristics of the experimental flow are replicated in the LES and RANS runs, and a general agreement between simulation and experiment is shown. Except for the immediate vicinity of the grid, LES is able to capture the velocity magnitude within a 10% error and the TKE to within the experimental uncertainty. The LES shows notably better agreement with the experimental data than RANS, particularly regarding the TKE decay behavior downstream of the grid. Both methods show significant departures from the experiment in their predictions close to the central guide tube. The experimental and high-fidelity data will be used to inform closures for novel multiscale methodologies. The long-term goal of the work is to use the high-fidelity data to yield improved multiscale thermal analysis techniques for solving fuel performance problems of direct relevance to industry.

Development of fuel depletion code for molten salt reactor with very deep burnup

A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.

Development of a source-term migration model for a large bubble formed in a core disruptive accident

Because sodium-cooled fast reactors are designed with high inherent safety in mind, the probability of a core disruptive accident (CDA) is extremely low. However, from a defense-in-depth perspective, the study of CDA sequences is still worthwhile to assure the safety and reliability of reactors. During a CDA, a large bubble rapidly expands inside the sodium pool, rises from the core, and covers the gas region, providing a potential migration path for source terms (radioactive materials present within the containment barriers). Source terms released initially within the cover-gas region after a few hundred milliseconds are called instantaneous source terms. We propose here an instantaneous source-term migration model that provides a simplified evaluation of the amount of source terms absorbed by coolant sodium during the ascent of the CDA bubble. In the model, the particle motion within the CDA bubble obtained from the basic momentum equation is used to calculate the amount of source terms escaping from the bubble interface. In addition, a model analogous to aerosol scavenging by precipitation is used to assess the amount of source terms absorbed by droplets present in the bubble, especially the entrained sodium droplets that form during rapid expansion of the CDA bubble. The model is further validated by a previous source term migration experiment in which a large high-pressure bubble expands and rises in a sodium pool. Good agreement with the measured retention factor of a source term demonstrates the reliability of the developed model. Given these results, some key parameters are selected for a sensitivity analysis.

Structural, Electronic, and Mechanical Properties of α(U) and α(U-Zr) Alloy Fuels

Uranium-zirconium (U-Zr) alloy fuels have been taken into consideration for fast reactors because of their superior reactor safety, high uranium (U) density, and excellent thermal conductivity. In this paper, the structural and mechanical properties of metallic U and U-Zr alloy fuels are calculated at the atomic level by density functional theory–based calculations with Hubbard U (density functional theory + U) corrections. Several structure-property relations, such as lattice volume, bulk modulus, Young’s modulus, shear modulus, electronic density of states, etc. are calculated for U metal and U-Zr alloy fuels. In addition, the Zr content in metallic U fuel is adjusted from roughly 5 at. % to 15 at. % in order to determine how the Zr content affects the characteristics of U-Zr fuel material. A linear relationship between the volume and Zr concentrations is observed. The concentration of Zr in the fuel affects the mechanical characteristics of U-Zr alloy fuel. It is also observed that the electronic structure of the α-U phase is not changed significantly in the presence of Zr. Our computations provide insight into how U-Zr alloy fuels behave in reactor conditions.

Development of active and passive reactivity control systems for a fast spectrum small modular reactor

This paper presents the development of active and passive reactivity control systems implemented in a fast-spectrum small modular reactor (SMR). The design of the control assembly serving as an active system still relies on a conventional rod-typed geometry. To enhance its worth, a thermalizing ZrH1.6 pin is introduced, which softens the neutrons around the control assembly, leading to increased neutron absorption in the absorber material. Additionally, a control assembly with a partial-length absorber is strategically placed at a certain location to mitigate high peak power levels. Two passive systems have been developed to enhance reactor safety: the gas expansion module (GEM) aimed to mitigate the unprotected loss of flow (ULOF), and the liquid lithium-6 absorber intended to address unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOSH) scenarios. With a total of 24 GEMs and 18 lithium-6 channels installed, they provide negative reactivity ranging from about 932 to 693 pcm (1.17$ to 1.03$) and 521 to 602 pcm (0.65$ to 0.89$), respectively.