Sodium-cooled Fast Reactor Research Articles

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

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.

Insight into the Cationic Charge Distribution in U1-y-zPuyAmzO2±x Mixed Oxides.

Uranium-plutonium mixed oxides, containing few mol % of Am, are currently studied as fuel for Sodium Fast Reactors. The study of the O/M ratio of these fuels is of main interest, as its variation can induce issues for reactor safety. For this purpose, four U1-y-zPuyAmzO2±x samples (0.235 ≤ y ≤ 0.39 and 0.005 ≤ z ≤ 0.02) were studied using electron probe microanalysis (EPMA), X-ray powder diffraction (XRD), and X-ray absorption spectroscopy (XAS) experiments. EPMA analyses revealed a homogeneous matrix phase with few U- and Pu-rich agglomerates. A matrix phase of fluorite structure (face-centered cubic), as well as a small contribution of a UO2 phase of the same structure, was evidenced for all of the samples by XRD. The O/M ratios of the (U,Pu)O2±x matrix were calculated based on the obtained lattice parameters. XANES experiments highlighted the simultaneous presence of U5+ and Pu3+/Am3+ for all of the samples, indicating a charge compensation mechanism even for stoichiometric samples. Based on the EPMA and XRD results, it was evidenced that this coexistence of U4+, U5+, Pu3+, Pu4+, and Am3+ was not originating from the U- and Pu-rich agglomerates. It was found to be rather a peculiarity of the U1-y-zPuyAmzO2±x matrix phase, discussed with the help of thermodynamic calculations performed on the U-Pu-Am-O system.

Numerical investigation on the core thermal hydraulic behavior of pool-type sodium-cooled fast reactor (SFR)

A thorough understanding of the reactor core thermal hydraulic behavior is essential for the design and safety analysis of Sodium-cooled Fast Reactors (SFR). Due to the application of hexagonal subassembly, the core thermal hydraulic behavior is significantly affected by the flow field within the subassemblies, the inter-wrapper region and the hot pool. Analysis of the core thermal hydraulic behavior requires a model coupling the three regions mentioned above, which has been identified as one of the thermal hydraulic challenges in SFR. In the present study, a 3D model that covers the three regions was developed for the core of the China Experimental Fast Reactor (CEFR) with the Computational Fluid Dynamic (CFD) code, Fluent. The inter-wrapper region and the hot pool were modeled in detail, while the subassemblies were modeled with a special porous medium model. The core thermal hydraulics behavior under steady state was studied, more specifically, information for the flow field distribution at the core outlet, the inter-wrapper flow and the duct wall temperature distribution was obtained. Under steady state, liquid sodium in the inter-wrapper region is supplied by the inner region of the hot pool. And it enters the inter-wrapper region from the core outer region and returns back to the hot pool inner region from the core central region. The inter-wrapper flow is cooled by non-fuel subassemblies and heated up by fuel subassemblies. For non-fuel subassembly, the ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies could reaches 96%; for fuel subassemblies, the maximum ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies is 2.45%. Significant temperature gradients have been observed on the duct wall, with maximum values of 156.69 K/m in the vertical direction and 2,196.00 K/m in the circumferential direction. The largest temperature gradient appears on the duct of subassemblies adjacent to the transition region of fuel subassemblies and non-fuel subassemblies.

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.

Pareto front of sodium void worth and breeding ratio in metal-fueled sodium fast reactor with axially heterogeneous core design by coupling neutronics and genetic algorithm

For sodium-cooled fast reactors (SFRs), achieving low sodium void worth (SVW) and flexible breeding ratios (BR) is crucial. This study establishes a method coupling neutronics simulations with genetic algorithms (GA) and applies it to a 3000 MWth metal-fueled SFR to optimize the SVW, BR, and the power distribution of axially heterogeneous cores. The Pareto front, i.e. optimal front, between SVW and BR is determined, quantitatively clarifying the trade-off relationship between these two parameters. Axially heterogeneous cores with equal inner and outer fissile heights can achieve SVW adjustments from −71 to 3148 pcm and BR adjustments from 1.26 to 1.62. Designs with unequal fissile heights slightly expand the feasible region, but the impact on the optimal front for equal-height designs is limited. Through the analysis of the spatial distribution of the sodium void effect, it is shown that the bond sodium model significantly influences the SVW in low sodium void cores, while its impact on high breeding cores is minimal. The results indicate that the power distribution can be optimized by adjusting the enrichment of subregions within the constraints of SVW and BR. For low sodium void cores, retaining a certain amount of lower fertile material is necessary to flatten the power distribution. While the criticality search during depletion calculations affects the parameters, its impact on identifying the optimal front is limited.

Enhancing fuel breed-burn performance in a sodium-cooled fast reactor using a novel reactivity control method

This study aims to enhance the fuel cycle and fissile breeding performance of a sodium-cooled fast breeder reactor (FBR) by utilizing minor actinides (MAs) as a means of reactivity control alongside partially-inserted control rods. Choosing the PFBR-500 as the reference design, four core models, designated as Cases A, B, C, and D, utilizing various proportions of minor actinides (MAs) were built and simulated using OpenMC. The MA concentrations were optimized to compensate for the withdrawal of control rods and ensure the same initial reactivity for all cores. Burnup analysis over 365 EFPDs revealed a significant increase in cycle length and burnup for the modified cores along with a modest rise in breeding ratio. Notably, Case C, employing 3.45 wt.% MAs in the 88 inner-core fuel subassemblies achieved an extra 62.25 EFPDs cycle length, a 33.74% rise in single-cycle burnup, and a 3.86% increase in breeding gain compared to the reference. Loading MAs into the inner-core region proved to be more effective in enhancing both fertile-to-fissile conversion (thus extending the cycle length and fuel burnup) and transmutation than utilizing MAs throughout the core due to greater neutron flux at the core center. While Case D utilizing 2.2 wt.% MAs both in the inner and outer core fuel subassemblies had the highest overall MA loading, it demonstrated lower increments in cycle length, burnup, and breeding gain compared to Case C. Case C also exhibited the highest overall destruction rate (approximately 24%/y) for Np and Am isotopes, and successfully transmuted around 24.08 kg 237Np, 9.33 kg 241Am and 3.82 kg 243Am over the course of a single year. The addition of MAs also achieved a slight flattening of the axial and radial flux profile and a decrease in the flux peaking factor. However, it slightly lowered the beta-effective, Doppler constant, and control rod assembly worth, and shifted the coolant void reactivity worth to the positive side.

Long-term thermal aging effects in ferritic-martensitic steel HT9

The ferritic-martensitic steel HT9 is a candidate material for fuel cladding and core components in advanced nuclear reactors, such as sodium-cooled fast reactors, thanks to their high temperature mechanical properties and low susceptibility to irradiation induced swelling phenomena. However, thermal stability and elevated temperature microstructural evolution in these alloys may impact their long-term behavior and reliability. In this work, the effects of thermal aging on the microstructural and mechanical properties of HT9 have been investigated through complementary electron microscopy, synchrotron X-ray diffraction, microhardness, and thermodynamic modeling. Plates of HT9 were aged up to 50 kh at relevant sodium-cooled fast reactor operational temperatures (360 °C - 700 °C). Trends in microstructure as a function of aging time and temperature were apparent from qualitative and quantitative analysis. These observations were further supported by thermodynamic modeling of the bulk and precipitate phases. Specific phases observed include BCC Fe, FCC M23C6, HCP and FCC MX phase and Laves M2X phase. Through the application of our multi-scale and multi-modal approach, clear information on the aging mechanism of HT9 was obtained, allowing for a more informed prediction, and understanding of the long-term behavior, performance and thermal stability of ferritic-martensitic alloys exposed to elevated temperatures.

Knowledge representation to support EMDAP implementation in advanced reactor licensing applications

The Data, Information, Knowledge, and Wisdom (DIKW) pyramid depicts the general principle of a decision-making process. Various levels of abstraction are required to formalize the available data/information/knowledge with respect to the decision being supported. The Evaluation Model Development and Assessment Process (EMDAP) was developed by the United States Nuclear Regulatory Commission (US NRC) to guide the development and assessment of Evaluation Models (EMs) for transient and accident analysis of nuclear power plants under design basis accidents. It concerns the decision of adequacy of an EM with respect to the purpose of analysis (NPP application). However, lack of data, scaling issues, modeling limitations, computation, and experimental overhead are some of the key factors that challenge the EM adequacy decision in EMDAP. Phenomena Identification and Ranking Table (PIRT) which is used for complexity resolution can also introduce noise due to bias and variance in expert opinion and judgment. The uncertainty in the decision due to these issues and challenges can be minimized by enhancing the EMDAP from the DIKW perspective. In this work, we formulate a systematic scheme for knowledge representation, and classification and characterization of evidence to enhance clarity and transparency in the EMDAP implementation. The proposed scheme is governed by the concept of the value of information, where we contextualize, classify, characterize, and evaluate the evidence (relevant data/knowledge/information related to EM development, verification, validation, and uncertainty quantification) with respect to the relevant decision attributes or elements of EMDAP. We make use of the Predictive Capability Maturity Model (PCMM), Argumentation technique, and process quality assurance for knowledge representation to support EMDAP implementation in advanced reactor licensing applications. The main objective of the proposed scheme is to support decision-making by making it more transparent, traceable, and accountable. The illustration of the proposed scheme is provided using a system simulation code (SAS4A/SASSYS-1) as the primary EM and a generic sodium fast reactor under loss of flow accident as the target reactor application.EMDAP involves different elements that require expert input and insights. However, the process of expert elicitation may require significant resources and time. To address these issues, we identify generative artificial intelligence models as potential tool which (after systematic testing and capabilities analysis) can be used for domain specific expert elicitation and information extraction for efficient implementation of EMDAP.

Mitigating fuel cladding chemical interactions by enhancing chromium diffusion barrier performance using pulse reverse electroplating

Fuel cladding chemical interaction (FCCI) is a critical issue affecting the safe usage of metallic nuclear fuel in sodium-cooled fast reactors (SFRs). Although a chromium (Cr) coating diffusion barrier has been developed to effectively mitigate FCCI, Cr coatings formed by direct current (DC) and pulse current (PC) electroplating generate hydrogen-induced micro-cracks that reduce diffusion barrier performance. In this study, electroplating with a pulse reverse (PR) current and optimized anodic time was utilized to minimize the effects of hydrogen. Consequently, this approach enabled the formation of a crack-free and highly stable microstructure in the electroplated Cr coating. A diffusion couple test (FCCI test) was conducted at 923 K for 25 h to evaluate that the diffusion barrier performance of the Cr coating formed by DC, PC and PR electroplating. The test results demonstrate the superior diffusion barrier performance of the Cr coating formed by PR electroplating through a systematic analysis comparing the microstructures of Cr coatings formed by different waveforms.

Vibration test and fatigue test for failure probability evaluation method with integrated energy

The seismic evaluation of key components such as reactor vessel is important for the Seismic Probabilistic Risk Assessment (S-PRA) in a Sodium-Cooled Fast Reactor (SFR). Many components were fractured by integrated damage like fatigue damage during seismic ground motion. In this paper, failure probability evaluation method with integrated energy was developed by comparing the energy with vibration tests and fatigue tests. Vibration tests were performed to evaluate integrated vibration energy at failure by energy balance equation, and fatigue tests were performed to evaluate integrated vibration energy at failure based on experimental results. As results, it is shown that integrated energy at failure time by vibration tests were estimated and its values were in range the energy based on results of fatigue tests.

High-Fidelity Simulation of the Light-to-Dense Stratification Transient in the HiRJET Facility

Density stratification in a large enclosure is a crucial phenomenon to heat transfer and sustainable passive heat removal of a sodium fast reactor during reactivity transients. However, engineering turbulence models were identified to have unsatisfactory performance in predicting propagation of a stratified front. Yet, the scarcity of high-resolution data for stratification hampers the development of models. To explor e applications of leveraging direct numerical simulation (DNS) data to support turbulence model development, this work conducted DNS using NekRS to study a long stratification transient in the High-Resolution Jet (HiRJET) experimental facility. This work considers an experiment run where light fluid is injected into a tank containing a denser fluid with a relative density difference of 1.5%. Formation of the stratified layer is identified as impingement of the buoyant jet promoting mixing of the two fluids. Based on the transient statistics, transport of the concentration can be characterized by regions with dominating effects of turbulent mixing, buoyant dissipation, and molecular diffusion, respectively, as moving away from the elevation of jet impingement. Concentration near the stratified front also exhibits oscillation at Brunt-Väisälä frequency. Preliminary validation of the simulation showed encouraging agreement of the concentration distribution with the reference experiment.

Solidification-induced nonuniformity in U–Zr–RE metallic nuclear fuel rods

Metallic fuel is being developed for the next generation of sodium-cooled fast reactors due to its safety and fuel cycle economics through spent fuel reprocessing technology. After spent fuel reprocessing, rare earth elements that are immiscible with Uranium and Zirconium are included in the raw material for metallic fuel. Achieving a uniform composition and microstructure of RE inclusions is challenging because phase separation and temperature inhomogeneities during solidification lead to non-uniform composition and microstructure. This study explores the solidification and inclusion formation of U–Zr–RE metallic fuel rods. We show how RE inclusions are formed in the U–Zr matrix and how temperature inhomogeneities during solidification control the inclusion size distribution and edge migration of RE inclusions along the U–Zr–RE rods. Understanding the inhomogeneity of inclusions due to solidification can provide hints for making homogeneous nuclear fuel rods.

Investigation of the fast flux test facility transient behavior during a loss of flow without scram test

A Coordinated Research Project (CRP) on the benchmark analysis of Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) has been held under the sponsorship of the International Atomic Energy Agency. The CRP aims to improve understanding of loss of flow events in fast reactors, as well as to assess capabilities of existing computer codes against experimental data. FFTF is a 400 MWth Sodium-cooled Fast Reactor (SFR) owned by the U.S. Department of Energy, operated from 1980 to 1992. The CRP involves the LOFWOS Test #13, belonging to a series of unprotected transients performed as part of the Passive Safety Testing program. In this framework, the Nuclear Engineering Research Group of Sapienza University of Rome has developed an integrated multiphysics modelling based on a coupled approach with RELAP5-3D© (thermal-hydraulics) and PHISICS (neutron-kinetics) codes. The present paper discusses the results obtained by the authors after the participation in the benchmark open phase. As suggested by the organizers, a two-step methodology was followed. Firstly, calculations were run by imposing the reactor power as boundary condition and focusing on the FFTF thermal–hydraulic behaviour. A good accordance was detected between the numerical results and the experimental data, above all for what concerns the reference core outlet temperatures. Then, the simulations were repeated by using a coupled approach. Calculation outcomes demonstrated the capability of the selected code suite (PHISICS/RELAP5-3D®) to reproduce the reactor transient behavior and its suitability to be used as an effective numerical tool to simulate liquid metal fast reactors accidental scenarios where thermal–hydraulic and neutron-kinetic phenomena are strongly coupled.

Simulation Analysis Based on Sodium-Cooled Fast Reactor Fuel Assembly Under Blockage Accident

This paper employs computational fluid dynamics to investigate the fuel assemblies within a sodium-cooled fast neutron reactor. To begin with, we developed computational models for seven wire spacer fuel rods under both normal operating conditions and transient blockage conditions. We conducted separate analyses to assess the impacts of normal operation, blockage thickness, and blockage area. This work allowed us to acquire data on the heat transfer properties and the flow field variations for the coolant, blockages, and fuel rods across different conditions. Subsequently, we leveraged these flow field alterations to examine the resulting temperature distributions. Analyses were conducted to evaluate the effects of normal operation, blockage thickness, and blockage area. The research acquired the heat transfer characteristics and flow field distribution variations of the coolant, blockages, and fuel rods under different operational conditions and utilized these variations to analyze the temperature distribution. Through research analysis, the following conclusions have been reached. Under normal operating conditions, the temperature and flow fields of the fuel components exhibit cyclic variations along the axial length, corresponding to the pitch of the wire spacers. Heat exchange between the internal and external subchannels occurs independently, which further substantiates that the incorporation of wire spacers strengthens lateral flow disturbances. This effect is more pronounced within the internal subchannels, thereby leading to a marked difference in the flow fields on either side of the wire spacers. In the case of blocked conditions, an increase in blockage thickness and area both lead to higher temperatures for the coolant, the blockages themselves, and the fuel rods. The temperature in the recirculation zone behind the blockages also rises with increasing blockage thickness and area, although the magnitude of this increase is not significant. After the onset of blockage conditions, the instantaneous temperature tends to increase. As time progresses, the instantaneous temperature fields following the blockage do not display greater fluctuations in temperature change than those observed after the system has stabilized. For the temperature parameters of the convective field, differences arising from an increase in blockage area are more significant than those caused by an increase in blockage thickness. When blockage area and blockage thickness are increased by the same multiple, an increase in blockage area results in a higher temperature peak. Increasing the area of blockage necessitates a longer duration for both the temperature and the velocity fields to revert to equilibrium.

Comparative study on tensile and high cycle fatigue behaviour of 316L(N) SS hardfaced with Ni-Cr-B-Si alloy by GTA and laser cladding processes

In sodium-cooled fast reactors (SFR), 316L(N) SS grid plate is hardfaced with Ni-Cr-B-Si alloy to achieve higher wear resistance. Tensile and fatigue forces are acting at the interface between substrate and deposit due to different thermal expansion coefficients of those two materials, which can cause cracking of deposit and fracture during operation. Thus, it is very important to consider appropriate hardfacing method which can provide higher tensile and fatigue strength to avoid cracking/debonding at the interface. To find a solution to this problem, two hardfacing techniques, namely Gas Tungsten Arc (GTA) and Laser cladding (LC), are taken into consideration. Hardfaced specimens are prepared using each process on which tensile and high cycle fatigue tests are conducted. From the experimental testing, stress-strain and S-N curves are generated to predict the tensile and fatigue behaviour of specimens. Fractographic studies are conducted at fractured surfaces to understand the fatigue crack nucleation and propagation characteristics. The experimental results for both processes are compared. Tensile and fatigue strength of LC specimens are ∼11 % and ∼17 % less than those of GTA specimens due to its higher brittleness. Thus, GTA process is recommended as the efficient hardfacing process for grid plate of SFR.

System features of fast sodium reactors in a two-component structure of nuclear power generating

System features of fast sodium reactors in a two-component structure of nuclear power generating

Experimental study on practical application of optical fiber sensor (OFS) for high-temperature system

This study explores the application of Raman scattering-based optical fiber sensors (OFSs) in extreme environments, specifically focusing on a loop heater vessel with temperatures ranging from 200 °C to 680 °C. This condition generally covers the advanced reactor designs, such as Sodium-cooled Fast Reactor and High Temperature Reactor. Various optical fiber combinations were employed for temperature measurements, taking into consideration the operating temperature of the target equipment. Two types of OFSs, gold-coated and polyimide-coated, were utilized. Protective tubes made of stainless steel (STS) and carbon were introduced to ensure reliable temperature data collection in high-temperature settings. Results indicate that the STS tube with a gold-coated OFS exhibited the highest consistency and agreement with thermocouple measurements, making it suitable for extreme environments. The study emphasizes the applicability of this system in high-temperature environments, such as liquid metal reactors, high-temperature thermal energy storage system, and hydrogen production system, for environmental monitoring.

Defect monitoring system of the internal structures of a sodium fast reactor using an artificial intelligence model

This study developed a defect-monitoring system with an artificial intelligence model, YOLOv7, which is tailored for processing image data from an ultrasonic visualization system within sodium fast reactor (SFR) internal structures. For the safety of SFR internal structures, although it is a crucial inspection for defect monitoring, it is difficult to identify structural defects because of the invisible environment. Therefore, we applied the YOLOv7 model in this study; however, we encountered challenges including decreased accuracy with complex defect shapes and complications from data augmentation during pre-training. To solve these problems, we additionally applied the enhanced super-resolution generative adversarial network for higher resolution and a Sobel noise-filtering algorithm to enhance the defect detection accuracy. And we evaluated our system by comparing it with a confidence score. This underscores the effectiveness of the approach in enhancing the defect detection capabilities. Therefore, this defect-monitoring system should be designed to preemptively identify internal structure deformations and enhance SFR safety and maintenance practices.

Review and Modeling of Integrated Energy Systems with Nuclear Reactor Coupled Desalination and District Heating

Detailed reviews of a past advanced nuclear reactor–based integrated energy system, as well as other nuclear reactor and fossil fuel–based integrated energy systems, have been performed for this work. A review of the utilization of heat from nuclear reactors for various applications and cogeneration has been done. The heat can be utilized by extracting the steam from the turbine while the steam is still at a desired temperature. While the use of nuclear process heat for district heating in countries like Finland, France, China, Poland, and elsewhere is discussed, more focus of the review has been given to nuclear desalination processes. Integrated energy systems (IESs), where distinct types of reactors like pressurized water reactors, boiling water reactors, sodium-cooled fast reactors, heavy water reactors, and other advanced reactors are coupled with various nuclear desalination processes, like multi-effect distillation (MED), multistage flashing, and reverse osmosis methods, are discussed. The nuclear desalination plant at Aktau is discussed in more detail due to its decades of successful operation. The IES of the Aktau plant coupled with a five-effect MED desalination plant was taken as a reference for modeling the Open Modelica (OM)–based IES in this work. The OM IES model shows good agreement with the MED plant output of Aktau and can be extended for future applications of IESs.