Fast Reactor Research Articles

Corrosion of F/M steels and austenitic steels in Pb-Mg at different temperatures

Pb83Mg17 is expected to be a potential coolant for lead cooled fast reactors because of its less toxic radioactive products and possible acceptable corrosion rate without additional oxygen control in comparison with LBE. In order to evaluate the compatibility of structural steels and Pb-Mg, two F/M steels (SIMP, T91) and two austenitic steels (316L, 15-15Ti) were tested in Pb-Mg at 350, 450 and 550 °C with exposure time ranging from 100 h to 2000 h in this work. Compared with the slight corrosion or no obvious corrosion of F/M steels, austenitic steels have suffered very serious dissolution corrosion which is significantly influenced by temperature and exposure time. The corrosion rate of austenitic steels is controlled by dissolution or diffusion mechanism at different temperatures.

Research and development of the neutron-photon coupled transport method in the fast reactor code system MOSASAUR

In the high-fidelity reactor physics design of fast reactors, the influence of photon effect needs to be explicitly simulated with the neutron-photon coupled transport method. The neutron-photon coupled transport-burnup method was researched based on the fast reactor physics design software package MOSASAUR. Photon library, photon cross-section production, photon source and photon power calculation methods were researched and the related calculation modules were developed and verified with the RBEC-M benchmark and the MET-1000 benchmark. Numerical results showed the good accuracy of the newly-developed neutron-photon coupled transport ability with MOSASAUR. The photon effect introduced the 80% difference for the assembly power in the RBEC-M benchmark, which showed the importance of the neutron-photon coupled transport in the fast reactor design.

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

Fluid Dynamics Effects of a Porous Blockage on Laminar Flow in the Interior Subchannels of a 61-Pin Wire-Wrapped Rod Bundle

Flow blockages in Liquid Metal Fast Reactor (LMFR) fuel assemblies that can cause fuel pin cladding failures require further investigation for the development and optimization of these Generation IV reactor designs. The objective of this study was to experimentally evaluate the effects of a porous blockage accident scenario on laminar flow behavior in the interior subchannels of a prototypical 61-pin wire-wrapped rod bundle. The laminar flow condition is considered because of its importance to natural circulation and loss-of-coolant-accident operating scenarios as well as the limited availability of experimental data. The matched-index-of-refraction method was utilized to obtain two-dimension two-component time-resolved particle image velocimetry (TR-PIV) measurements at three planes centered on blocked and neighboring subchannel regions. Time-averaged first-order and second-order statistics, computed by Reynolds decomposition, were visualized including the mean and fluctuating streamwise and spanwise velocity components. Line profiles described the evolution of flow from upstream regions, to separated flow, and recombination downstream, while a modified Galilean decomposition distinguished differences between flow in the blocked measurement plane and flow in its neighboring counterpart region. Spatial-temporal cross-correlations were performed for the streamwise velocity fluctuations to characterize the convection velocity and decay of traveling vortices in the wake of the porous blockage. The isothermal TR-PIV measurements from this study provide high-fidelity experimental data sets for the validation of computational models and numerical studies to characterize complex fluid phenomena in wire-wrapped rod bundles during the potential accident scenario of a porous, interior flow blockage.

Lead Fast Reactor Thermal-Hydraulic Testing Facilities in Support of the UK Advanced Modular Reactor Program

The Westinghouse lead-cooled fast reactor (LFR) is a next-generation nuclear power plant whose primary mission is to reduce front-end capital costs and generate flexible and cost-competitive electricity for global markets, while offering mission versatility and satisfying the highest safety and sustainability standards. The LFR is an economical choice for carbon-free power generation to meet the ambitious goals of reducing carbon emissions. The plant utilizes passive safety systems for high reliability and public safety. The LFR testing program aims to fill the technology gaps in the key materials, components, and systems of the LFR plant. It provides demonstrations of LFR engineering that reduce the uncertainty of the LFR design and experimental data for the verification and validation of the modeling and simulation computer codes. With the support of the LFR phenomena identification and ranking table, the phenomena important to plant safety with insufficient states of knowledge have been identified and a Westinghouse testing plan developed to address them. The United Kingdom Advanced Modular Reactor Program Phase 2 was purposed to advance the level of design maturity toward eventual regulatory approval and deployment. Westinghouse and its partners have developed eight state-of-the-art test facilities within the program. Among them, the Passive Heat Removal Facility, Versatile Lead Facility, Lead Water Interaction Facility, and the Lead Freezing Facility are thermal-hydraulic testing facilities, with the other four facilities dedicated to testing lead corrosion and material mechanical behavior in lead. The design, development, and testing of the four thermal-hydraulic facilities are presented in this paper. A variety of computer codes ranging from system-level analysis, component-level analysis, and high-resolution computational fluid dynamics analysis, are utilized to support the development of the facilities. The analyses provide operational and accidental conditions in the LFR to identify major testing conditions and inform the facility design and testing matrix. With a significant amount of experimental data being generated with these highly instrumented test facilities, the computer codes and models will be benchmarked against the experimental results to improve their validation and verification status.

Effects of different load variation rates on the operational stability in an S–CO2 Brayton cycle coupled natural circulation lead-cooled fast reactor

The thermal inertia results in net heat absorption in natural circulation lead-cooled reactors during power variation processes, increasing the risk of core meltdown. In this study, a dynamic simulation model of a natural circulation lead-cooled fast reactor coupled with a supercritical carbon dioxide Brayton Cycle power generation system has been developed based on Vpower. The influence of thermal inertia on system operating temperature under different power variation rates was discussed. The results indicate that during power reduction, different decline rates only affect the temperature variation range of the system without impacting the final stable state. However, during power increase, different ramp rates not only affect the temperature variation range but also lead to varying degrees of system temperature rise. To address this issue, an optimized turbine throttling scheme is proposed. Compared to the original plan, the new plan successfully stabilizes the temperature at the rated value after increasing the system power. This study provides insights for the operation control strategy of similar reactor types.

Mesh morphing-based pre-processing and numerical simulation of blockage accident in lead–bismuth fast reactor fuel assembly

This paper presents a CFD modeling approach for a 19-rod bundle fuel assembly with spiral semi-cylindrical ribs, which potentially can be used for liquid metal cooled fast reactor, as well as a numerical simulation of flow and thermal behavior under partially blocked conditions. The ribs on fuel rod surface are integrally formed with the cladding to enhance heat transfer while preventing detachment. However, the complex channel geometry presents challenges for grid generation, as using a large number of tetrahedral or polyhedral elements would consume significant computational resources. Therefore, this paper proposes a mesh morphing-based pre-processing method to establish a structured hexahedral mesh for such complex geometry. Using this approach, the 19-rods assembly is analyzed, and the result has been verified by conventional mesh scheme as well as experiment.

Improving tally efficiency and accuracy of multi-group scattering matrix calculations in the Monte Carlo code NECP-MCX

Two issues arise in the calculation of the multi-group scattering matrix when employing a continuous-energy Monte Carlo code for generating homogenized multi-group cross-sections. Firstly, the analog estimator is used to evaluate group-to-group elements, which leads to large statistical uncertainty. Secondly, employing the scalar flux as the weighting function in generating the high-order scattering matrix introduces errors in fast reactor calculations. For the first issue, the repeated collision approach and pre-tabulated cross-section approach are adopted to improve the tally efficiency. For the second issue, the average scattering cosine is calculated based on the conservation of the mean square displacement of neutrons, which is then used to correct the first-order self-scattering cross-section. To evaluate the effectiveness of the above approaches, a PWR pin-cell problem and fast reactor core problems are tested. The results demonstrate that: 1) The figure of merit for multi-group scattering matrix calculations was improved by 8–12 times with the pre-tabulated cross-section approach. 2) Biases of keff were reduced from over 500 pcm to less than 300 pcm when using the corrected self-scattering cross-section. 3) The corrected self-scattering cross-section also yielded higher accuracy for the assembly power calculations, where the maximum biases are reduced from 5 % to 1 %.

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.

Neutron balance approach for inline critical rod position search calculation in Monte Carlo reactor analysis

The critical rod position search has been one of the issues in the Monte Carlo (MC) reactor analysis. This paper proposes a practical and simple approach for the critical rod position search based on the neutron balance equation and the monotonic relation between the control rod absorption rate and the total rod insertion length. The proposed method was implemented in an inline manner within the McCARD, the MC reactor design code, so that the critical rod position is calculated and updated for each fission source iteration. The numerical results in a typical fast reactor problem demonstrate the promising performance of the proposed method in the critical rod position search calculation. Furthermore, the cycle depletion calculation was performed to show the capability of the McCARD to simulate the rodded operation condition. By addressing the critical rod position search problem through the neutron balance approach, the applicability of the MC code in the advanced reactor design will be expanded.

Analysis of flow heat transfer characteristics of helium-xenon gas mixtures during natural circulation

Helium-xenon(He-Xe) gas mixture is a commonly used cooling medium in fourth-generation advanced reactor systems. The thermal conductivity of a passive waste heat system is related to the safety and reliability of the reactor system. Therefore, it is crucial to explore the passive characteristics of the helium-xenon gas mixture. The passive waste heat removal characteristics of a helium-xenon gas mixture are influenced by the flow and heat transfer characteristics of the mixture. However, existing research on the flow and heat transfer characteristics of helium-xenon gas mixture mainly focuses on single tube bundles and plate bundles with fixed flow rates and high Reynolds numbers. However, existing research on the flow and heat transfer characteristics of the He-Xe mixtures mainly focuses on single tube bundles and plate bundles with fixed flow rates and high Reynolds numbers, and most empirical formulas apply to Reynolds numbers greater than 10000. However, the natural circulation flow rate under the influence of gravity varies with the variation of heating power, and in most operating conditions, the Reynolds number is less than 10000. Due to the lack of research on the convective heat transfer characteristics of the He-Xe mixtures at low Re, this paper investigates the flow and heat transfer characteristics of the He-Xe under natural circulation conditions. A three-dimensional model of the natural circulation circuit was established using numerical calculation methods. Comparing the numerical simulation results of natural air circulation with empirical formulas, it was found that the error was less than 20 %. This proves that the SST k-ω turbulence model can be used to study the flow and heat transfer characteristics of He-Xe gas mixtures in natural circulation. On this basis, natural cycle numerical simulations of He-Xe mixtures with different mixing ratios were carried out at a heat flux of 5000 W/m2, as well as natural cycle numerical simulations of 40 g/mol He-Xe gas mixture at different heating powers, to study the effects of heating power and He-Xe mixing ratio on the natural cycle heat transfer characteristics of the He-Xe gas mixture. The existing He-Xe gas mixture flow heat transfer equation was modified, and the error of the modified Nu empirical equation was less than 12 %, which can be used for the development of the passive waste heat export system program for subsequent He-Xe gas-cooled fast reactors.

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.

Numerical study on two-phase flow instability in multi-parallel channels of helical-coiled once-through steam generator of lead-cooled fast reactor

The helical-coiled once-through steam generator (H-OTSG) has the advantages of compact structure and strong heat transfer ability, and is appropriate for lead-cooled fast reactor. The two-phase flow instability may cause mechanical vibration and thermal fatigue of heat transfer tube bundles, posing a serious threat to the safe operation of the H-OTSG. In this work, the secondary side of the H-OTSG characterized as multi-parallel channels is modeled by RELAP5/MOD3.4 code, and the oscillation behavior during start-up and the influence of structural and operating parameters on system stability are studied based on time-domain method. The results indicate that the pressure, flow rate, and temperature of the secondary fluid exhibit density wave oscillations at the heating section in a (n-2,2) pattern. In addition, increasing inlet throttling, reducing outlet throttling, controlling subcooling within an appropriate range and avoiding operation under low loads are all beneficial for improving the system stability of the helical-coiled once-through steam generator.

Facile Synthesis of U2Ti Intermetallic by Direct Electrochemical Reduction of UO2-TiO2 Composite in LiCl-Li2O Melt

Alloys of U with Ti are of importance in nuclear industry as fuel for Gen IV fast reactors, hydrogen isotope storage medium for fusion reactors, super conductors, and also as corrosion resistant material for use in various applications. Here, preparation of U2Ti intermetallic compound was investigated by the direct electrochemical reduction of mixed oxide of UO2-TiO2 in LiCl-0.5% Li2O molten salt at 650 °C. The mixed oxide pellet of UO2-TiO2 sintered at 1500 °C was found to be a mixture of UTi2O6 and UO2 as evidenced by X-ray diffraction (XRD) analysis. Direct electro-lithiothermic reduction of UO2, TiO2, and a mixture of sintered UO2-TiO2 and UTi2O6 coupled with cyclic voltammetry of these oxides in the melt was performed to understand the electro-reduction mechanism. Potentials of reduction of these oxides in the melt, w.r.t Ni/NiO reference electrode, obtained by analysis of CV data of these oxides contained in metallic cavity electrodes and XRD analysis of partially electro-reduced oxides were used to arrive at the electro-reduction mechanism. Results indicate that U2Ti can be prepared conveniently by the electro-lithiothermic reduction of sintered pellet of UO2-TiO2 cathode by constant current electrolysis in a two-electrode cell.

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.

Degradation of remaining mechanical properties of bainitic 2.25Cr1Mo steel during creep

In this work, the microstructure and remaining mechanical properties of bainitic 2.25Cr1Mo steel plate used in steam generator of sodium cooled fast reactor during creep were systematically examined. Creep tests at 530 °C with a rupture time up to 18800 h reveals that the creep stress correlates linearly with the Larson-Miller parameter (PLM), while the creep plasticity is rather independent on PLM. The remaining strength parameters, such as yield strength (RP0.2), ultimate tensile strength (Rm) and hardness (H), are found to linearly decrease with PLM, while the remaining plasticity is seen to increase. It suggests that the application of creep stress to the steel accelerates the softening of the material. During the creep process, the dislocation density in bainitic matrix decreases, while the number density of carbides increases, sacrificing the solution strengthening and leading to a continual reduction in the lattice parameter. Although abundant fine carbides precipitate after creep rupture, the remaining strength is however reduced. Instead, the weakening effect due to reduction in dislocation and solute content overwhelms the strengthening of precipitation, resulting in a degradation of the remaining strength. The obtained remaining mechanical properties are attempted to be applied in the design and operation assessment of steam generator.

Safety analysis code development and validation for lead-cooled fast reactor with helical coiled once-through steam generator

LFRs are one of six advanced Generation IV reactor concepts using lead-based coolant with high thermal conductivity, and HCOTSGs are generally adopted to highlight LFR miniaturization characteristics. An inhouse STH code SACLER with the HCOTSG model and the cover gas model was developed specifically for LFRs. NACIE facility was used to demonstrate SACLER accuracy. The HCOTSG model was verified using the RELAP code, experimental data, and ELSY design parameters. The validation results are consistent with the experimental data of NACIE. The HCOTSG model calculation results are in good agreement with the experimental data and the RELAP code. The error between the ELSY simulation result and the design value is within 1.0%. Liquid level simulation of ELSY shows that liquid levels are mainly determined by the pump head and flow resistance. SACLER can accurately simulate the LFR system with HCOTSG, including transient and steady-state conditions.

Numerical investigation of SUPERFACT-1 experiment by a MOX fuel pin thermo-mechanical model

Preparations are currently underway to develop ISAA-Na code, a sodium-cooled fast reactors (SFRs) specific version of an integrated severe accident analysis code. This version aims to evaluate both protected and unprotected accidents in SFRs. Given the crucial role of fuel pin thermo-mechanical behavior in SFR safety analysis, an advanced model for MOX(Mixed OXide) fuel pin thermo-mechanical performance under operation conditions has been developed and validated. This validation involved simulating the SUPERFACT-1 fast reactor irradiation experiment using ISAA-Na. The resulting calculation data were compared with experimental measurements and the outcomes of other codes dedicated to fuel pin performance analysis (TRANSURANUS, MACROS, andGERMINAL) to assess the fuel pin thermo-mechanical model. The overall code assessment reveals that the model achieves satisfactory performance in simulating fuel pin behavior, with predicted errors falling within an acceptable range.

Enhanced irradiation tolerance of the body-centered cubic structured FeCrW medium-entropy alloy as revealed from primary damage process

Multi-principal element alloys have attracted much attention in the development of nuclear reactor materials due to their potentially excellent irradiation resistance. Here, we employed Molecular Dynamics (MD) simulations to investigate displacement cascades resulting from Primary Knock-on Atoms (PKA) energies ranging from 10 to 50 keV in body-centered cubic structured FeCrW, FeCr and α-Fe. The analysis encompasses irradiation-induced defect generation and evolution, aiming to elucidate mechanism of irradiation tolerance. The simulations results indicate that FeCrW medium-entropy alloys generate more Frenkel defects during the thermal spike stage compared to α-Fe, yet fewer residual defects are observed in FeCrW at the end of the cascade. The suppression of the increase in defects number in FeCrW compared to α-Fe is significantly enhanced with the increasing PKA energies. It is foreseeable that the number of residual defects in FeCrW will be much smaller than that of α-Fe with a dramatic increase of PKA energy, which will be attributed to the improved defect self-self-healing ability in FeCrW. Moreover, the mechanism of enhanced defect self-self-healing is attributed to the size of the defect cluster, the formation and distribution of defects, the enhanced thermal spike, and the chemical disorder caused by the increase in composition leading to low thermal conductivity. The spectral energy density (SED) curves spikes of α-Fe, FeCr and FeCrW reveal an important role in the enhanced phonon scattering in improving defect self-repairing ability due to chemical disorder. Our research provides valuable insights for guiding the development of advanced multi-principal element alloy structural materials in nuclear industry, particularly for the fourth-generation fast reactor.

Effect of cooling rate on the precipitation characteristics and growth mechanism of Bi3Ni in liquid lead‑bismuth eutectic

Austenitic stainless steel is a promising candidate structural material for direct contact components with lead‑bismuth eutectic (LBE) in lead-cooled fast reactor (LFR). However, the preferential dissolution of Ni from austenitic steels at low oxygen concentrations will lead to the precipitation of Ni-containing phases in the low-temperature parts of reactors, which has a risk of pipe clogging. In this paper, three samples, comprising one extracted from a corrosion tank after prolonged operation and two prepared, respectively, by quenching in liquid nitrogen and cooling in the furnace, were obtained to investigate the effect of cooling rate on the precipitation characteristics and growth mechanism of Bi3Ni that forms by the reaction of dissolved Ni and Bi in LBE. The results showed that rod-like Bi3Ni was found in all three Ni-containing LBE samples. As the cooling rate increased, the width of Bi3Ni decreased. The Bi3Ni in all three samples displayed a pseudo-hexagonal prismatic shape and had a typical faceted growth character; the cooling rate had no effect on the morphology of them. The crystallographic analysis showed that the growth axis of Bi3Ni was in the 〈010〉 direction, and the prism planes were composed of {001} and {101}. In addition, the precipitation of Bi3Ni inevitably increased the content of Pb7Bi3, which would in turn enhance the volumetric expansion of solid LBE samples. For the growth mechanism, Bi3Ni phases were controlled by the screw dislocation mechanism, with the step growth rate of the dislocation core being greater than that of the outer edges. Given the low cooling rate and low Ni content in the LBE coolant of reactors, the precipitation of Bi3Ni with a rod-like morphology seems to be inevitable.