Sodium-cooled Fast Reactor Research Articles

Thermal conductivity-based hydrogen in argon detector (HAD): Design, development of instrumentation, and performance evaluation.

The Hydrogen in Argon Detectors (HADs) working on the principle of thermal conductivity difference between argon (reference) and argon + H2 (sample) in the cover gas of a sodium-cooled fast breeder reactor using the Wheatstone bridge circuit. The output of HAD is very sensitive to the gas flow rate and to the variation in ambient temperature. The RMSnoise of current was brought within ±1 µA by a constant current source across the bridge. The temperature correction factor fed in the processor brought down the RMSnoise in the HAD signal within ±2.4 mV than the uncorrected one having RMSnoise of ±230 mV. The filtered noise of HAD facilitates enhancing the detection limit of HAD down to 30 ppm (3σ) of hydrogen in argon.

Analysis of the effect of thermal stratification on the natural circulation decay heat removal of sodium-cooled fast reactor

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy.

Challenges in high-fidelity thermal–hydraulic simulation of SFR cores: Insights and PACA-S4FR solutions

Numerical simulations in studying thermal–hydraulic behavior in sodium-cooled fast reactors (SFR) have gained attention for their safety and cost advantages. While low-fidelity numerical simulations commonly used in industry have the advantage of lower computational costs, they face challenges in accurately representing intricate flow phenomena. On the other hand, high-fidelity numerical simulations offer excellent characterization of turbulence but demand significant computational resources. The emergence of exascale supercomputers has provided opportunities and challenges for large-scale high-fidelity numerical simulations. In this paper, firstly, we review the research progress of thermal–hydraulic numerical simulations of SFR cores using both low-fidelity and high-fidelity methods. Secondly, we estimate the computational and storage requirements for high-fidelity numerical simulations of SFR core based on calculations performed on China Experimental Fast Reactor (CEFR). Subsequently, we analyze the challenges of conducting high-fidelity thermal–hydraulic simulations of the entire SFR core, including large-scale mesh generation, appropriate numerical discretization methods, efficient simulation acceleration techniques, coupled simulations, and the complexities of validation and verification processes. Finally, we present the research progress of PACA-S4FR, a high-fidelity thermal–hydraulic simulation software in the China Virtual Reactor (CVR1.0) project, along with the ongoing efforts to address challenges in high-fidelity simulations. This paper contributes to advancing the comprehension of SFR core thermal hydraulics, potentially aiding in safety and performance enhancements in SFR.

CNN-based acoustic identification of gas–liquid jet: Evaluation of noise resistance and visual explanation using Grad-CAM

For the analysis of anomalies in a steam generator (SG) of a sodium-cooled fast reactor (SFR), we evaluate the noise resistance of CNN-based acoustic identification methods of gas–liquid two-phase jets and produce visual explanations for their decisions. First, we introduce the water flow sound and the three types of gas–liquid jet sounds, which simulate the background noise and the anomaly sounds, respectively. Second, we produce time–frequency representations for various signal-to-noise ratios (SNRs) and employ AlexNet, VGG16, and ResNet18 to the identification of the gas–liquid two-phase jets. As a result, the best CNN of ResNet18 achieves more than 92% for SNR=0,−4,−8,and−12 dB and 69% for SNR=−16and−20 dB. This result indicates that our proposed methods can identify the flow states of gas–liquid two-phase jets in low-level noise environments and detect the gas–liquid two-phase jets even in high-level noise environments. Also, Grad-CAM suggests that ResNet18 focuses on one of the spectrum peaks of the water flow sound and all or part of the signal intensity pattern of the gas–liquid jet sounds. Our proposed methods lead to the safe operation and fast, accurate, and accountable analysis of anomalies in SFR.

Angular distribution uncertainty influence in a large sodium-cooled fast reactor with mixed-oxide fuel

Nuclear data uncertainties are considered a main limiting factor in proving safety of advanced reactors. A reasonable amount of available information should be applied to sensitivity and uncertainty analyses for this, and this work considers the influence of usually omitted angular distribution uncertainties. A sensitivity and uncertainty analysis of MOX3600 is conducted using Serpent. The angular distribution influence is assessed by applying covariances of the state-of-the-art libraries. The sensitivity analysis shows that quantifying the influence of higher-order Legendre moment uncertainties on the eigenvalue and void reactivity is of minor importance for MOX3600. The uncertainty analysis shows the P1-coefficients for 238U and 16O introduce uncertainties to the eigenvalue similar to 56Fe and should be accounted for. A library comparison reveals TENDL-2021 and ENDF/B-VIII.0 introduce uncertainties of a few hundred pcm to the eigenvalue, and the data should be reassessed, while the other libraries give an uncertainty of several tens of pcm.

A subchannel analysis code SACOS-Na for sodium-cooled fast reactor

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm.For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes.

Effect of coolant boiling induced by accumulated debris on flow-regime characteristics of debris bed formation behavior for sodium-cooled fast reactor: Experimental and modeling study with gas injection

Understanding the Debris Bed Formation (DBF) behavior is important to the optimized integral evaluation for Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactor (SFR). Aimed at clarifying the characteristics of flow regimes during DBF, extensive experimental works were carried out by releasing solid particles into a water pool. According to the preliminary studies using bottom heating and nitrogen gas injection, it was found that the effect of coolant boiling should be crucial to the flow-regime transition. However, the gas flow rate considered in these studies (0–50 L/min) was not adequate wide to cover the range of vapor flow rate from the possible drastic boiling in actual CDA of SFR. Therefore, in current study, to elucidate the influence of coolant boiling on DBF flow-regime characteristics more comprehensively and provide more extensive data for supporting the predictive model development, a greater number of experiments are conducted with a much-enlarged range of gas flow rates (up to 100 L/min) under varying parametric conditions (including particle size, particle type, water depth and release pipe diameter). Through analyzing the flow-regime transition, an extended functional form for the effect of coolant boiling is successfully proposed. By incorporating this form into our base model, the model predictability can be extended for the DBF at the conditions of coolant boiling induced by accumulated debris in accordance with the verification of the regime map. Present study is believed to be valuable for further developing and verifying the models and codes for SFR safety analysis in China.

Design and dynamic simulation of a molten salt THS coupled to SFR

With the increasing ratio of renewables in the grid, a low-carbon and stable base load source that also is capable of load tracking is in demand. Sodium cooled fast reactors (SFRs) coupled to thermal heat storage system (THS) is a strong candidate for the need. This research focuses on the designing and performance validation of a two-tank THS based on molten salt to integrate with a 280 MWth sodium cooled fast reactor. Designing of the THS includes the vital component, sodium-to-salt heat exchanger which is a technology gap that needs to be filled, and designing and parameter selection of the tanks and related pumps. Modeling of the designed THS is conducted followed by the description of operation strategies and control logics of the THS. Finally, the dynamic simulation of the designed THS is conducted based on Fortran. Results show, the proposed power system meets the need of the design requirements to store heat for 18 h during a day and provide 500 MWth for peak demand for the rest of the day.

A validation study of the SLTHEN code for hexagonal assemblies of wire-wrapped pins using liquid metal heating experiments

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19- and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ∼10 % and maximum errors of ∼20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

Fundamental evaluation of hydrogen behavior in sodium for sodium-water reaction detection of sodium-cooled fast reactor

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR “Monju”. However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H2), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. At the same time, we observed experimentally that the fine H2 bubbles exist stably in the liquid sodium, longer than previously expected. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

High-fidelity numerical investigation on structural integrity of SFR fuel cladding during design basis events

A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.

High-Fidelity Simulations of Shield Assembly Mixed-Convection Flows with Applications Toward Reduced-Resolution Modeling

Flow circulation and heat removal through shield and reflector assemblies can have major impacts on safety in long transients for sodium fast reactors (SFRs). These transients are typically categorized by reduced flow rates and large-scale organized flow patterns, including potential intra-assembly circulation. Such low-flow cases can provide challenges for experiments because of complications in measuring the flow rates and temperatures with high accuracy in different areas. This consequently also raises the uncertainty of many modeling approaches for these phenomena. In an effort to address some of these issues, high-fidelity large eddy simulations are performed using the highly parallel solver NekRS. A 19-pin configuration of a tight-lattice wire-wrapped hexagonal bundle (pitch-to-diameter ratio = 1.07), representing a prototypical internal configuration of a shield assembly, was investigated. The sodium flow was set at a bundle Reynolds number of 2000, with simulations being performed for modified Richardson numbers of 0.0 (i.e., no buoyancy), 0.01, and 0.04, where mixed-convection effects are anticipated. The flow and temperature fields for these cases are discussed in detail. The high-fidelity data should prove useful as reference data for expanding and improving on various reduced-resolution approaches. A basic framework for combining subchannel and computational fluid dynamics methodologies in SFRs is also presented, with preliminary results from simulations of light water reactor bundles and a discussion of changes that need to be made for potential application to SFRs.

Tellurium Corrosion of Type 304/304L Stainless Steel, Iron, Chromium, and Nickel in High-Temperature Liquid Sodium.

Investigating tellurium (Te) corrosion on structural materials is crucial for sodium-cooled fast reactors (SFRs) due to radionuclide presence and knowledge gaps. In this study, Type 304/304L stainless steel (SS304), chromium (Cr), iron (Fe), and nickel (Ni) samples were immersed in low-oxygen environments with Te in liquid sodium at 773 K for 30 days. At 10 ppm oxygen, SS304 showed multiple oxide layers, including a compact NaCrO2 interlayer and porous Na-Fe-Ni-O outer layers. Tellurium penetrated through the porous layers but was hindered by the NaCrO2 interlayer. At 0.01 ppm oxygen, Cr had no oxide layer, while Fe and Ni had unstable ones. Tellurium-induced pitting was deeper in Fe and Ni compared to Cr. Oxygen levels and Cr composition are critical factors affecting stable oxide compound layer formation and mitigating Te-induced pitting.

Cross-scale coupling analysis method for OTSG in SFR and its application

The once-through steam generators (OTSG) in sodium-cooled fast reactor (SFR) are characterized by full-regime flow boiling heat transfer, coupled heat transfer on both sides and complex structure with a large dimensional span. This makes the computational fluid dynamics (CFD) and 3D refinement modeling methods to analyze the 3D temperature field of OTSG containing large-scale heat transfer tube bundles generally face the problems of high computational difficulty, low solution efficiency and poor solution stability. In this paper, based on the full-regime flow boiling heat transfer model and the flow pattern discrimination model, a one-dimensional full-regime flow boiling heat transfer model from subcooled water to superheated steam on the water side (tube side) and its cross-scale coupling solution model with the three-dimensional refined CFD model on the sodium side (shell side) are established. The effectiveness of the coupling analysis model was verified by comparative analysis with experimental data of the Indian 19-tube Steam Generator Test Facility (SGTF). The independent coupling of the water side of each heat transfer tube in the OTSG with the sodium side 3D fluid domain is realized. The results show that the cross-scale coupling method using the one-dimensional full-regime flow boiling heat transfer model for the water side and the three-dimensional CFD model for the sodium side can accurately predict the temperature distribution, the main heat transfer characteristic points on the full height of each heat transfer tube, and can obtain the temperature difference between different heat transfer tubes and tube walls. The maximum error between the calculated results and experimental data for the distribution of sodium temperature along the axial direction of the tubes is less than 3%. The cross-scale coupled solution method avoids the direct adoption of the three-dimensional multiphase flow model, which significantly improves the efficiency and stability of the solution of the three-dimensional temperature field of OTSG. This study provides an efficient 3D refinement numerical analysis method for the design and safety verification of OTSG.

SIMMER-IV application to safety assessment of severe accident in a small SFR

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

Direct Numerical Simulation of Heat Transfer in a 7-Pin Wire-Wrapped Rod Bundle

The Sodium-Cooled Fast Reactor (SFR) is a promising concept chosen in the Generation IV International Forum as a possible design for pursuing the sustainable use of nuclear energy. Its core consists of multiple hydraulically isolated assemblies, with a tightly packed triangular lattice array of fuel pins enclosed in a hexagonal duct present within each assembly. Helical wire spacers are wrapped along the axis of the rods to maintain a gap between them, inducing a secondary flow, increasing the channel mixing, and enhancing convective heat transfer. In this study, a direct numerical simulation campaign is conducted for a simplified 7-pin wire wrapper geometry, with Reynolds numbers ranging from Re = 1000 to 10 000 and a Prandtl number of Pr = 0.005, to investigate heat transfer in low-flow conditions. The wire wrapper case is compared to a bare bundle case with seven pins. The results are discussed, and heat transfer predictions are compared between our numerical results and classic correlations. An anisotropy invariant map is obtained for the above-mentioned cases, and turbulent kinetic energy and turbulent heat flux budgets are computed and analyzed. Our findings provide unique insights into the flow behavior within a wire-wrapped bundle.

Numerical study of debris bed formation behavior for severe accident in sodium-cooled fast reactor by using least square MPS-DEM method

During a Core Disruptive Accident (CDA) in a Sodium-cooled Fast Reactor (SFR), debris beds may form in the lower plenum of the reactor vessel as a result of the accumulation of discharged debris, which is formed from the solidification and fragmentation of core molten materials during their release due to interactions with the sodium coolant. Since the configuration of debris beds plays a vital role in their long-term coolability and neutronic subcriticality for achieving in-vessel retention, it is essential to understand and assess the Debris Bed Formation Behavior (DBFB). In this study, an advanced Moving Particle Semi-implicit (MPS) method, called least square MPS (LSMPS), is utilized and coupled with the Discrete Element Method (DEM) for the numerical investigations of DBFB characteristics in a two-dimensional liquid pool. The applicability and reliability of the proposed LSMPS-DEM method are validated by comparing simulation results with experiments using simulant materials. The LSMPS-DEM method is then employed to study DBFB in sodium. It is observed that debris properties significantly impact the sedimentation and accumulation processes of DBFB, leading to variations in the characteristics of the formed debris beds. The present study is expected to be useful for further investigations of DBFB under more realistic conditions and for the design improvement of safety devices to mitigate the consequences of CDA in SFR.

BISON-FIPD integration enhanced low-burnup SFR metallic fuel swelling model evaluation framework

Experiments indicated that metallic fuel in sodium-cooled fast reactors (SFRs) rapidly swells radially and axially at low burnup. Despite that, prior studies have been focused on describing high burnup axial fuel elongation. With recent conventional and non-conventional metallic fuel concepts being considered for license applications, understanding multidimensional fuel swelling at a wide range of burnup levels is important to fuel analysis and qualification activities. Here, we report the development and demonstration efforts of a low-burnup SFR metallic fuel swelling model evaluation framework using the BISON advanced fuel performance code. The framework leverages the Integral Fast Reactor (IFR) program X423 experiment data through the ongoing integration project to enable standardized and automated use of legacy metallic fuel irradiation data maintained in the SFR fuel irradiation and physics database (FIPD) for BISON metallic fuel model verification and validation. The performance of the framework was demonstrated using the two representative metallic fuel swelling model sets implemented in BISON, with a series of insights about future advanced swelling model development.

Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700°C to 1000°C

ABSTRACT The Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650°C–850°C. However, little data have been obtained above 850°C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700°C–1000°C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix’s phase transformation, and a single equation can express a creep rupture strength from 700°C to 1000°C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.

Deuterium transport behavior of 15-15Ti stainless steel for nuclear structural material in Gen-IV reactors

As a new type of nuclear reactor, the Gen-IV reactor attracts increasing interest from the industry. Tritium is produced in lead-cooled fast reactors (LFR), sodium-cooled fast reactor (SFR), and other Gen-IV reactors. It is necessary to study the tritium permeation behavior of 15-15Ti stainless steel as a candidate cladding and structural material in Gen-IV reactors. In this study, gas-driven permeation experiments were conducted to measure the high-temperature deuterium transport behavior parameters of 15-15Ti stainless steel. The results showed that the deuterium permeation behavior of 15-15Ti and 316L stainless steels is similar. Additionally, this study discussed the effects of microstructures and surface oxidation on the permeability of stainless steels. This study will aid in evaluating tritium permeation in lead-based fast reactors and other Gen-IV reactors and whether tritium permeation barrier or tritium-removal devices are necessary.