BackgroundThere is increased demand for high spatial resolution in domestic and international nano-imaging experimental stations utilizing synchrotron radiation. Nano or nanoradian level positioning accuracy and stability are necessary for focusing mirrors or monochromators on the beamlines.PurposeThis study aims to design and optimize a redundant parallel flexible hinge rotating device to meet this demand.MethodsFirst of all, the kinematics of the flexible mechanism were initially analyzed, and the virtual displacement principle was employed to derive the overall static rotational stiffness of the mechanism. Then, the characteristics of the mechanism, as well as the impact of hinge parameters on stiffness, were investigated. Subsequently, the dynamic model of the mechanism was built by employing the Lagrange equation, and the natural frequency in the motion direction was deduced. Finally, an optimization model was developed for the static and dynamic dual purpose mechanism design, solved through a genetic algorithm accounting for nonlinear constraints. In addition, the first four resonance frequencies and vibration modes of the flexible mechanism were examined using a finite element method for modal analysis followed by the creation and assembly of a high-precision flexible hinge mechanism along with a rotary adjustment device for experimental testing.ResultsTest results indicate that the flexible angular displacement adjustment mechanism achieves a rotation angle of 0.668° and bidirectional repeatability of ±8.91 nrad during fine-tuning. In addition, the angle resolution of 15 nrad and stability of 2.72 nrad (root mean square value) are achieved over a 30 min test in the frequency range of 1~500 Hz. The first natural frequency of the mechanism is approximately 295 Hz, which aligns with the theoretical calculations and finite element analysis results.ConclusionsThe effectiveness and reliability of the optimized flexible mechanism in achieving nanoradian level high-precision angular displacement adjustment is confirmed by this study.

BackgroundDuring the reconstruction of radionuclide activity of waste by conventional tomography gamma scanning (TGS), the measurement time is long due to the division of a large number of invalid voxels, and the reconstruction accuracy is low due to the iterative solution of the pathology equation. The traditional division method of encrypting only circumferential voxels does not significantly improve the reconstruction accuracy.PurposeThis study aims to solve the problem of long measurement time and low accuracy of TGS by elaborating the influence law of voxel division method on reconstruction accuracy.MethodsFor a 400 L cement waste drum with a radius of 35 cm, a coaxial HPGe detector with a detection efficiency of 40% (crystal diameter 6.09 cm, length 5.18 cm) was used at a distance of 74 cm from the center of the drum to measure two types of nuclides, 60Co and 137Cs. The error situation of the optimized method of voxel partitioning and the traditional voxel partitioning method were compared by random point source activity reconstruction experiments. By calculating the condition number and count rate deviation of different voxel partitioning methods, the influence law of voxel partitioning methods on the reconstruction accuracy was investigated.ResultsComparison results show that the radial encryption method leads to an increase in the number of very small value points and a decrease in the value of very large value points in the count rate deviation curve, which improves the reconstruction accuracy. Compared with the conventional division method, the voxel division optimization method reduces the number of voxels and thus the measurement time by about 9/10; at the same time, it reduces the number of conditions and the count rate deviation, which reduces the maximum reconstruction error by about 2/3.ConclusionAn optimized voxel division method proposed in this study makes use of the influence law of different voxel divisions on the reconstruction error through theory and experiment, hence improves the measurement accuracy and reduce the measurement time.

BackgroundThe study of irradiated samples is of considerable importance. Owing to these samples being radioactive, the applicability of conventional characterization methods is limited. Because of the high sensitivity of 3He detectors to neutrons, the small-angle neutron scattering (SANS) technique is nearly unaffected by radiation such as gamma, beta rays, and sample preparation is simple.PurposeThis study aims to develop a device for SANS measurement of nanostructures in radioactive samples.MethodsThe shielding thickness of the device was optimized through Monte Carlo simulations. Experimental efficiency and safety of the device were improved by combining remote-control functionality and automated sample switching among up to 12 samples loaded simultaneously. Finally, the device was used to carry out a SANS experiment for characterizing the radioactive A508-III steel sample.ResultsThe optimized thickness of the lead shielding layer of the device is 7.5 cm, the corresponding maximum dose rate of the measurable radioactive sample is 1.4 mSv·h-1. The results of successful SANS experiment on a A508-III steel irradiation surveillance specimen indicate that low-dose, long-term irradiation has a minimal impact on the nanostructure of pressure vessel steel.ConclusionsThe device and associated technical methods can be applied for nanostructure characterization of radioactive samples.

BackgroundMolten salt electrochemical separation of spent fuel in molten salt medium is one of the most widely studied technologies in pyroprocessing. It is based on the use of electrolysis to achieve the separation.PurposeThis study aims to investigate the separation of actinides and lanthanides in situ using high-temperature molten salt media.MethodsCyclic voltammetry (CV), square wave voltammetry (SWV), and constant potential electrolysis were employed to study the electrochemical behavior and separation feasibility of UF4 and LnF3 (Ln=Eu, Sm, Yb, La, Ce, and Ho) in FLiBe molten salt. Firstly, the eutectic salt was prepared, followed by the precise measurement and mixing of FLiBe molten salt and UF4 or LnF3 powder in accordance with the desired concentration ratio. Then, the electrochemical behavior of uranium and lanthanide elements in FLiBe molten salt was experimentally studied using electrochemical CV and SWV. Subsequently, the constant potential electrolysis method was employed to achieve the electrochemical deposition of uranium or lanthanide elements. Finally, the feasibility of separating uranium and lanthanide elements in FLiBe molten salt was evaluated through X-ray diffraction (XRD) and energy-dispersive X-ray spectroscopy (EDS) analysis of the electrolysis products.ResultsExperimental results show that U4+ undergoes two electron transfer reductions on an inert tungsten electrode, namely U4++e-→U3+ and U3++3e-→U0, while Yb3+, Sm3+, and Eu3+ have only one electron transfer reduction, leading to the +2 valence state of Yb2+, Sm2+, and Eu2+ ions, respectively. In addition, La3+, Ce3+, and Ho3+ ions do not exhibit significant electrochemical signals within the electrochemical window of FLiBe molten salt. The findings from XRD and EDS analyses verify that the electrolysis products presented on the W electrode consist of metal U, UF3, and entrain molten salt, with no lanthanide (Ln=Eu, Sm, Yb, La, Ce, and Ho) electrolytic deposition product detected.ConclusionsThe suitable electrolytic separation method developed in this study provides basic data to separate actinides and lanthanides from molten salts.

BackgroundThe remediation technology for U(VI), Ca2+ and other pollutants in the groundwater of the mining area of in-situ leaching of uranium has become a key technical bottleneck restricting the decommission of uranium mine, and the microbial-induced calcium carbonate precipitation (MICP) technology can well mineralize and remove heavy metals in the groundwater.PurposeThis study aims to remove the pollutants in groundwater of uranium mining areas with MICP.MethodsFirst of all, bacillus pasteurianus was selected, and its acid resistance and tolerance to uranium were analyzed. Then, the effects of initial U(VI) concentration, initial Ca2+ concentration, initial pH value, concentration of bacterial solution, and the time of mineralization on the removal of pollutants by MICP were explored. In addition, scanning electron microscopy (SEM) and X-ray diffraction spectroscopy (XRD) were employed to characterize the composition and microstructure of the mineralization products of MICP, in order to reveal the mechanism of mineralization and decontamination. Finally, water with low uranium concentration was treated by bacillus pasteurianus to verify the application of MICP.ResultsThe results show that Bartonella pasteurii has good urease activity at pH value of 4 and can adapt to the groundwater containing U(VI) concentration of 100 mg?L-1. When the initial pH value is 4, the initial uranium concentration is 50 mg?L-1, the Ca2+ concentration is 8 000 mg?L-1, and the time of mineralization by MICP is 48 h, the removal rates of U(VI) and Ca2+ in the groundwater are 61% and 54%, respectively. The removal rate of U(VI) further increases to 91% when the initial pH value is increased to 7. Meanwhile, the increase of Ca2+ concentration promotes the removal of U(VI) in groundwater by MICP, however, the removal rate of Ca2+ is relatively low if the concentration of Ca2+ is too high.ConclusionsThe composition of MICP mineralization products is mainly uranium-containing calcium carbonate, in which the uranium is immobilized mainly by co-precipitation. Therefore, Bacillus pasteurianus can well remove U(VI) and Ca2+ in the groundwater of the mining area of in-situ leaching of uranium, and has a good prospect of application.

BackgroundWith the development of nuclear energy, spent nuclear fuel reprocessing has generated high-level liquid waste (HLLW) containing large quantities of minor actinides (e.g., Np, Am, and Cm). N,N,N',N'-tetraoctyl-diglycolamide (TODGA), an amide ether extraction agent, has immense potential for the extraction and separation of actinides from HLLW. However, HLLW is characterized by high radioactivity, which can destroy the molecular structure of TODGA and diminish its extraction capacity. In addition, the type of organic medium selected to dilute TODGA has a significant impact on its extraction ability and radiolysis.PurposeThis study aims to evaluate the gamma radiolysis and extraction performance of TODGA in a kerosene medium.MethodsFirstly, several experimental reagent systems were prepared. Organic phase: 50 mmol?L-1 TODGA/kerosene; aqueous phase: 0~4 mol?L-1 HNO3; absorbed doses: 5~1 000 kGy by a 60Co gamma-irradiator (3.7×1015 Bq) at room temperature; absorbed dose rates: 0.6 kGy?h-1 (5~15 kGy) and 5.8 kGy?h-1 (50~1 000 kGy). Then, ultra-performance liquid chromatography (UPLC) and mass spectrometer were used to identify the components in the organic phase and assess the relative concentrations of TODGA and radiolysis products before and after irradiation. Lanthanides (substitutes for actinides) and alkaline earth metals were extracted using TODGA at different absorbed doses (5~100 kGy) and the concentration of metal ions for extraction was 100 μg?mL-1 in 3 mol?L-1 HNO3. Finally, after extraction and dilution, the aqueous phase was detected using an inductively coupled plasma-optical emission spectrometer to estimate the lanthanide and alkaline earth metal concentrations.ResultsThe experimental results indicate that the concentration of TODGA steadily increases with increasing concentrations of HNO3 at a radiation dose of 1 000 kGy. The radiolysis rate of TODGA in the acid-free solution is >80%. Compared to the acid-free solution, the degree of radiolysis can be inhibited at approximately 8% (adding 0.5~3 mol?L-1 HNO3) and 18% (adding 4 mol?L-1 HNO3). Moreover, with an increase in the absorbed doses, the concentration of TODGA decreases in the studied systems. The radiolysis rate of TODGA is estimated to be 10%~40% when the absorbed dose was less than 100 kGy. However, it increases dramatically and reaches over 70% at 1 000 kGy. Breaking the ether bond results in the radiolysis of TODGA, generating two types of radiolysis products (C18H37NO2 and C18H37NO). HNO3 alters the radiolysis path, and breaking of the octyl side chain occurs with the appearance of another radiolysis product (C28H56N2O3). TODGA exhibits extremely poor extraction affinity for Sr, and the extraction rate decreases from 25% to non-extraction at an absorbed dose of 100 kGy. However, TODGA achieves approximately 100% lanthanide extraction (Ce, Eu, and Dy).ConclusionsTODGA has low radiation resistance in a kerosene medium at an absorbed dose of 1 000 kGy. Its radiolysis can be inhibited by the addition of HNO3; a higher concentration of HNO3 leads to a stronger inhibitory effect. The radiolysis of TODGA is insignificant within a 100 kGy absorbed dose, particularly in the range of 50~100 kGy. The molecular structure of TODGA is susceptible to breaking of its ether bond owing to gamma rays or radicals produced by kerosene. These findings demonstrate that TODGA can maintain the ability to extract lanthanides after irradiation at an absorbed dose of up to 100 kGy. Thus, it can be inferred that TODGA has a good extraction capacity for actinides in kerosene under the studied conditions.

BackgroundGaussian shaping is commonly used to filter and extract the amplitude of nuclear pulse signals owing to its high signal-to-noise ratio, resistance to ballistic loss, and ease of amplitude extraction. However, the problems of asymmetry and downwash exist in Gaussian-like signals generated by Sallen-Key and CR-(RC)mfilters.PurposeThis study aims to address the problems of asymmetry and downwash in Gaussian-like signals generated by Sallen-Key and CR-(RC)m filters, this paper presents a trigonometric function-based Gaussian-like pulse shaping algorithm for dual-exponential signals.MethodsFirst, the impact of shaping parameters on the shaping pulse and filtering performance of the algorithm was examined by applying Gaussian-like pulse shaping to simulated nuclear pulse signals. Then, the filtering performances of Gaussian-like and sin pulse shaping algorithms were compared and analyzed using the same shaping width. Finally, the characteristic X-ray signals of manganese (FAST-SDD detector) were collected using a self-made digital nuclear signal acquisition board under various tube flows. The energy resolution and count rate characteristics of the energy spectra obtained from Gaussian-like, sin, and trapezoidal shaping algorithms were comparatively analyzed for different tube flows and shaping times.ResultsComparison results show that Gaussian pulse shaping exhibits superior denoising performance to sin pulse shaping, with an 8.95% improvement in the SNR for the same peak arrival time. Additionally, the energy resolution of the spectrum obtained using Gaussian pulse shaping exceeds that of sin pulse shaping, and its stacking pulse separation ability outperforms that of trapezoidal pulse shaping, making it highly applicable.ConclusionsGaussian-like pulse shaping presented in this study demonstrats better stacking pulse separation capability than trapezoidal pulse shaping, indicating promising application prospects.

BackgroundNeutron localization technology can be used for monitoring spent fuel reprocessing to prevent nuclear criticality accidents.PurposeThis study aims to investigate the effects of different neutron emission modes, 3He pressure, tube diameter, and moderators on the position resolution and detection efficiency of a position-sensitive 3He proportional counter.MethodsFirstly, a 3He proportional counter with a tube diameter, tube length, and 3He pressure of 2.54 cm, 80 cm, and 405 300 Pa, respectively, was taken as object of study, and the Monte Carlo (MC) algorithm was employed to simulate the effects of neutrons with different energies, neutron emission modes, 3He pressures, tube diameters, and moderating bodies on the position resolution and detection efficiency of a position-sensitive 3He proportional counter. Then, the effects of the time difference of arrival (TDOA) on the location of a source in an active zone, source to tube distances, thickness of moderating body, and position resolution of the 3He proportional counter were investigated. Finally, a linear fitting analysis of the active region was performed, and the locations of inactive regions were estimated to determine the positioning ability of the 3He tube.ResultsSimulation results show that the position resolution of this 3He proportional counter is equal in all axial sections, and the position-resolution limit of thermal neutron source is 0.17 cm. When the thickness of the moderating body is 5 cm, and the 244Cm source is close to the external wall of the moderating body, the neutron detection efficiency is 3.0%, and the theoretical position resolution is 9.25 cm whilst experimental result of position resolution is 18.50 cm, half of the theoretical value, and the location resolution limit is 2.64 cm after processed by the linear fitting curve of the active location and a time-difference crest.ConclusionsMC simulation and experimental tests show that the position-sensitive 3He proportional counter exhibits better position resolution and detection efficiency for the 244Cm source when the thickness of the moderating body is 2~3 cm, hence suitable for practical applications.

BackgroundIn the nuclear industry, zirconium alloys, employed as structural materials, undergo irradiation-induced deformation owing to the combined effects of thermo, mechanical stress, and neutron irradiation during the irradiation process, thereby affecting their reliability in use.PurposeThis study aims to explore and predict the irradiation deformation of zirconium alloys to ensure the safety and economical operation of reactors.MethodsFirst of all, a database of experimental data on irradiation deformation of zirconium alloys from publicly available literature was developed by collection, cleaning, re-organization and placement of these data into a well-designed database, accompanied by detailed descriptions of these data. Then, data patterns, model validation and data mining were conducted by combining the database with our own developed mesoscopic model. Specifically, over fifty sets of preliminary data from the data mining library were subjected to data mining to analyze the correlation between model parameters and control variables.ResultsData mining results indicate that the critical resolved shear stress (CRSS) of the slip system increases with Nb content whilst the irradiation creep compliance of zirconium alloys increases with temperature.ConclusionsCompared with conventional data induction methods, this study introduces a novel approach for fitting and data mining using a physics-based mesoscopic model, offering fresh perspectives and methodologies for studying irradiation-induced deformation in zirconium alloys. The impact of this approach will be boosted as the collected dataset expands.

BackgroundMo leaching rate is one of the key product properties for high molybdenum radioactive waste vitrification. Furthermore, soluble molybdenum yellow phase is easy to precipitate, resulting in the enhancement of Mo leaching. It was reported that ZnO can promote the network stability and chemical durability in borosilicate glass to some extent. Whereas, if the composition of solidification glass is complex, then whether ZnO can still improve the chemical stability is worth studying. On the other hand, due to the long period of chemical stability test, it is hoped that the Mo leaching rate can be predicted by mathematical simulation to accelerate the research process.PurposeThis study aims to investigate the effect of ZnO in borosilicate glass on Mo leaching rate in simulated high-level radioactive waste (HLW), and establish the prediction model of Mo leaching rate.MethodsFirstly, a series multi-component borosilicate solidification glass samples with simulated high MoO3 (mass ~3%) HLW and designed mass fractions of ZnO content of 1%, 1.5%, 3.5%, 4%, 4.5%, and 5%, respectively, were prepared for experimental test of Mo leaching rate. Then the glass structure gene modeling (GSgM) were employed to accurately simulate the glass properties using a small amount of data to establish the corresponding structural prediction model. Finally, Infrared spectrometer (IR), X-ray Diffraction (XRD), Differential Scanning Calorimetry (DSC) and thermal dilatometer (DIL) were used to test the transformation temperature, heating rate, coefficient of thermal expansion, etc. of the samples, and the model validation was also performed.ResultsExperimental results show that Mo leaching is mainly affected by Si-O-Si rocking vibration around ~530 cm-1 and νSi-O-Si vibration in Q4 at ~1 180 cm-1, furthermore, B2O3, ZrO2 and alkali variation influence the relative concentration of these two structural units, and then affect the Mo leaching. Mo leaching rate is enhanced with mass of ZnO over 3.5% in this complex glass system.ConclusionsModel validation in this study proves that the structure model of Mo leaching rate has good simulation accuracy and statistical reliability, which are further improved after model iteration, hence can be used for the Mo-leaching prediction in the quick-screening of glass composition development.

BackgroundLaser cladding, recognized for its cost-effectiveness and high efficiency, has become a focal point in the field of laser remanufacturing. GX4CrNi13-4 martensitic stainless steel produced by laser cladding is a widely used structural material in nuclear power plants.PurposeThis study aims to enhance the mechanical properties of GX4CrNi13-4 martensitic stainless steel, fabricated using laser cladding technology, through different heat treatments that cause microstructure modification.MethodsThe GX4CrNi13-4 stainless steel sample was prepared using laser cladding technology, and its heat treatment microstructure was studied in details. Firstly, thermal expansion experiments identified the onset temperature of austenitic phase transformation of sample at 620 °C, serving as a pivotal reference for developing heat treatment schemes. Two distinct heat treatment processes, i.e., solution treatment plus aging (STPA) at 1 050 °C for 1 h followed by a similar treatment at 550 °C for 4 h and single aging (SA) at 620 °C for 2 h, were applied to experiments. The effects of these treatments on the microstructure and mechanical performance of the cladding were comparatively analyzed by using X-ray diffraction (XRD), optical microscopy, scanning electron microscopy (SEM), and transmission electron microscopy (TEM) were employed to characterize the post-treatment microstructure and phase distribution. Tensile tests at room temperature were performed on samples before and after heat treatment.ResultsExperimental results indicate that the as-cladded GX4CrNi13-4 stainless steel exhibits a dual-phase microstructure primarily comprising martensite and ferrite, with continuous network-like ferrite precipitated along martensitic boundaries, accompanied by a minor presence of residual austenite. Post STPA, the matrix still predominantly comprises martensite and ferrite, but the continuous network-like ferrite decomposes, and numerous micrometer-scale transgranular precipitates within the martensite are observed. This led to a slight improvement in plasticity but a significant decrease in strength. The SA treatment of the cladded samples, performed at the critical temperature for austenitic phase transformation, induces the formation of the reversed austenitic phase. This phase, during tensile deformation, triggers the transformation induced plasticity (TRIP) effect. Furthermore, the network-like ferrite precipitated along the martensite decomposes into a dispersed distribution post-SA. The combined effect of TRIP and ferrite decomposition notably enhances the plasticity of the laser-cladded GX4CrNi13-4 stainless steel while effectively maintaining its strength.ConclusionsThe use of austenitic phase transition temperature for aging in this study, coupled with the synergistic effect of reversed austenite TRIP and ferrite decomposition, successfully achieves a balanced strength-plasticity performance in laser-cladded GX4CrNi13-4 stainless steel. Appropriate heat treatment and microstructural control emerge as effective strategies to improve the comprehensive mechanical properties of materials.

BackgroundMolten salt reactors (MSRs) feature high temperature, low pressure, high chemical stability, and nuclear non-proliferation, which make them promising for a wide range of applications. However, owing to the fluid nature of the fuel, changes in the core structure of MSRs are bound to affect the fuel distribution and loading, consequently affecting the physical parameters of the core. Currently, the research on the impact of structural changes of the MSR core, composed of dispersed graphite components, on reactivity is insufficient.PurposeThis study aims to explores the influence of core structure changes on reactivity.MethodsThe core design model of new solid hexagonal graphite module MSR was taken as a reference, the small geometric changes, such as deformation and displacement of the core components, caused by factors including thermal expansion, fluid erosion, core vibration, and graphite irradiation, were analyzed using MCNP code. Additionally, the impact of these structural changes in the core on reactivity was investigated in details.ResultsThe results indicate that the structural changes in the core caused by thermal expansion introduce negative reactivity. The reactivity introduced by fluid erosion and core vibration, causing graphite component displacement, fluctuates. However, the overall trend shows that a shift of the graphite components towards the center of the core introduces positive reactivity, whereas shift towards the outer periphery of the core introduces negative reactivity. Graphite irradiation-induced deformation initially decreases reactivity and then increases it. At the end of the core's lifespan, the reactivity is still lower than at the beginning of the lifespan. This reactivity change can be compensated for by online feed or control rod movement, which has a limited impact on the operation of the MSR. However, the critical issue of control should be considered when this batch of fuel is reinserted into a new core.ConclusionsResults of this study suggest the necessity of constraining the graphite components within a certain range to ensure the safe operation of the reactor, providing an important reference for the design, operation, and maintenance of MSRs.

BackgroundFast neutron pulsed reactors are sensitive to wall scattering neutrons, and their waveforms are changed by reflected neutrons. In addition, their operation may be adversely affected when there are too many reflected neutrons.PurposeThis study aims to solve the problem of fast neutron pulse reactor wall reflection neutron.MethodThe point reactor kinetic method containing the reflection effect, Monte Carlo method, and ANSYS were combined to analyze the Godiva-I transient process in fast neutron pulsed reactor with wall-reflected neutron effect. Firstly, the quench coefficient of the fast neutron pulsed reactor was calculated. Then, the point reactor kinetic method containing the reflection effect was established. Finally, the thermal power obtained from the neutronics calculation was combined with the ANSYS thermal-mechanical module to establish the thermal-mechanical calculation method for fast neutron pulsed reactor, and the effect of wall-reflected neutrons was analyzed and calculated.ResultsThe results show that the reflected neutrons increased the rear edge of the pulse. The reactivity decreases when the flat is washed whilst the core displacement and stress are improved.ConclusionsThe method established in this study can reasonably explain the phenomenon of reduction in attenuation and the increase in power after pulse.

BackgroundSteam generator tube rupture (SGTR) accidents in lead-bismuth cooled fast reactors result in the generation of numerous steam bubbles owing to the interaction of the high-temperature liquid lead-bismuth eutectic (LBE) from the primary circuit and high-pressure subcooled water of the secondary circuit. These steam bubbles carried by the LBE may enter the core, causing local heat transfer deterioration and a power transient, seriously affecting reactor safety.PurposeThis study aims to elucidate the movement and dynamic behaviors of steam bubbles in liquid LBE and develop a drag coefficient model applicable to bubble migration in LBE for assessing the safety of the core during SGTR accidents.MethodsBased on the coupled level-set and volume-of-fluid (CLSVOF) method, a three-dimensional numerical model for calculating the drag coefficient of steam bubbles in LBE was established to study their movement and dynamic behaviors. Firstly, the deformation, velocity, and trajectory characteristics of bubbles in LBE were analyzed, and their drag coefficients were estimated. Then, the simulated drag coefficient values were compared with those from existing models whilst the prediction performance of the model proposed by Tomiyama was superior to those of the others. Finally, Tomiyama's drag model was optimized by introducing the We number, and its applicability was analyzed.ResultsAnalysis results show that the errors of the existing drag coefficient models are large under the condition of bubble–LBE two-phase flow, and the calculation error of the optimized model for the bubble drag coefficient in LBE is within 15%.ConclusionsThe feasibility of drag calculation model optimized in this study is verified, demonstrating its suitability for calculating the drag coefficient of steam bubbles in LBE.

BackgroundCAT-1 (China Astro-Torus 1) is a levitated dipole field magnetic confinement device, which mainly used for dipole plasma physics experiments, requiring a central floating superconducting coil to be stably levitated for at least 5 h without cooling or power supply.PurposeThis study aims to design a levitation control system of coupling superconducting levitation coil and floating coil for CAT-1 dipole device.MethodsAccording to the design parameters of the suspension magnet system of the CAT-1 device, Simulink model of the control system was established and applied to the simulation. Based on Routh-Hurwitz stability criterion, the influence of PID (Proportion-Integral-Derivative) control strategy on stability control was studied. The selection range of stability control parameters was determined to ensure the stable levitation of 1 200 kg, 5 MA floating magnet.ResultsSimulation results show that under ideal conditions, delay time of the PD (Proportion-Derivative) control system is 0.046 3 s, rise time is 0.154 5 s, peak time is 0.628 3 s, adjustment time is 0.084 8 s, and overshot δ=1.6. It means that PD can restore the levitated superconducting ring to the preset balance position in a short time, and the load of the circuit can be greatly reduced by adopting the appropriate starting mode.ConclusionThe results provide key technical support for the design and development of levitated superconducting dipole field devices.

BackgroundPassive safety system reliability is generally evaluated through best estimation plus uncertainty (BEPU) analysis. An important step in the evaluation process is sensitivity analysis of parameters, which is used to identify key system parameters to reduce the complexity of the model. However, local sensitivity analysis methods based on linear or monotonic assumptions may yield incorrect sensitivity results for complex nuclear power systems. Meanwhile, applying the global sensitivity method is difficult in practical engineering because of its high calculation cost.PurposeThis study aims to develop an efficient and low-cost global sensitivity analysis method for passive systems under ocean conditions.MethodsFirstly, the low-rank approximation (LRA) method was employed to improve the Sobol method based on Monte Carlo simulation. The number of unknown coefficients was significantly reduced by using the multivariate-based tensor product. Then, the LRA coefficients were used to calculate the sensitivity index, and the validity of the proposed method was verified by addressing several sensitivity analysis benchmark questions. Finally, taking an integrated pressurized water reactor including the passive residual heat removal system as the object, a simulation program for thermal-hydraulics analysis under ocean conditions was developed. And its sensitivity analysis was conducted using the proposed method.ResultsThe results show that the proposed method can accurately identify system key parameters after only 200 simulation calculations taking about 55 min, and the sensitivity ranking results are consistent with those obtained by Sobol method after 1.0×105 simulation calculations taking about 19 d.ConclusionsThe efficient global sensitivity analysis method established in this study can provide effective guidance for reliability analysis and design optimization of passive systems.

BackgroundThe enrichment of nuclear fuel may significantly influence the fuel performance of a reactor.PurposeThis study aims to explore the effect of U3Si2-Al plate-type fuel enrichment on its performance,MethodsThe fuel performance analysis code BEEs-Plates, neutronics Monte Carlo code OpenMC, and the one-dimensional system analysis code ZEBRA were coupled together within the MOOSE (Multiphysics Object-Oriented Simulation Environment) framework. Then the coupling code was employed to conduct the multiphysics coupling calculation for JRR-3 (Japanese Research Reactor No.3) fuel assembly enriched at 15%, 20%, and 25%. The data exchange among the three codes were realized with the help of interpolation transfer method implemented in MOOSE. Additionally, neutron physical parameters and fuel performance parameters after 231 d of operation were analyzed when the average fuel consumption of the module was 125.71 GWd?tU-1.ResultsThe calculation results indicate that the max power density of fuel assembly enriched at 25% is 18% higher than the assembly enriched at 15%. Due to the high thermal conductivity of aluminum, the temperature difference in the fuel assembly is almost negligible whilst there is a significant difference in the fast neutron fluence. The results of fuel temperature and fast neutron fluence show that the volumetric strain is more obviously affected by the fuel temperature. Specifically, the plastic strain of the assembly with 25% enrichment is approximately 40% higher than that of the assembly with 15% enrichment.ConclusionThe analysis results of this study suggest that the assembly with a higher enrichment is more prone to failure.

BackgroundUranium dioxide (UO2) has been broadly employed as nuclear fuel in nuclear reactors. However, the poor thermal conductivity of UO2 compromises the safety of the reactor owing to possible sharp temperature gradients. Graphene oxide (GO) is a promising additive to improve the thermal conductivity of UO2 owing to its excellent thermal performance.PurposeThis study aims to achieve uniform distribution of GO in UO2 pellets, effectively controlling the doping amounts to enhance the thermal conductivity of UO2 pellets.MethodsGO-doped UO2 powders with different doping amounts were prepared using solid-liquid mixing and ammonium diuranate (ADU) co-precipitation methods. After establishing the optimized powdering process, UO2-GO composite fuel pellets were prepared by spark plasma sintering (SPS). Properties of the UO2-GO composite fuel pellets, such as density, grain size, physical phase, and thermal conductivity, were examined using scanning electron microscope (SEM), energy dispersion spectrometer (EDS), metallographic microscope, laser pulse thermal conductivity meter, etc., and compared with those of conventional pure UO2 pellets.ResultsThe results showed that the density of UO2-GO pellets can reach up to 97.6% T.D.. The thermal conductivity of UO2-GO pellets with 1.5 wt.% doped GO is 85.9% higher than that of conventional UO2 pellets at 1 000 ℃. The grain size of the UO2-GO pellets is uniform, and the GO is homogeneously distributed at the grain boundary to form a bridging thermal conduction network.ConclusionsThe thermal conductivity of UO2 pellets is successfully improved through GO doping.