BackgroundThe Shanghai High-repetition-rate XFEL and Extreme Light Facility (SHINE) is characterized by high brightness, full coherence, and high repetition rates. For focusing high-repetition-rate hard X-rays, compound refractive lenses (CRL) must withstand significant heat loads, which can potentially lead to optical performance failure due to high thermal stress.PurposeThis study aims to optimize the maximum allowable repetition rate of X-ray beams while maintaining thermal stress within permissible limits to ensure reliable optical performance of CRLs under high-repetition-rate operation.MethodsFirstly, a finite element method (FEM) was employed to perform thermal-structural coupled analysis of beryllium CRLs with various internal spherical radii (0.3 mm, 1.0 mm, 2.0 mm, and 5.8 mm) under different photon energies (5 keV, 7 keV, and 15 keV). Then, the maximum thermal stress and temperature distributions were systematically evaluated under the initial 1 MHz repetition rate condition. Finally, the repetition rates were optimized to keep the maximum thermal stress below the permissible threshold of 120 MPa (50% of beryllium's yield strength), with special focus on the central region of the lens where stress concentration occurs.ResultsAnalysis results show that CRLs with internal radius R=0.3 mm experience a maximum thermal stress of 1 008 MPa at 7 keV photon energy at 1 MHz repetition rate, far exceeding the permissible limit. After optimization, the maximum thermal stress is reduced by 93% to 74 MPa by lowering the repetition rate to 100 kHz. The maximum temperature decreases from 201 °C to 36.6 °C, an 82% reduction. For larger radius CRLs (R=5.8 mm), the optimized repetition rate could be increased to 500 kHz for 7 keV X-rays while maintaining thermal stress below the threshold.ConclusionsBy optimizing the repetition rate according to CRL geometry and beam energy, thermal stress can be effectively controlled within safe limits. Increasing the internal spherical radius of CRLs from 0.3 mm to 5.8 mm allows for higher repetition rates (from 100 kHz to 500 kHz for 7 keV photons), thereby enhancing operational capabilities while ensuring reliable optical performance and structural integrity.

BackgroundMost of the irradiators that undertake the fundamental dosimetry studies are installed with radioactive sources of high-specific activity, which are only approved to be transported by qualified containers before the onsite loading being performed.PurposeThis study aims to design a novel irradiator featuring triple screen functionality to comply with the international and domestic regulation on the transportation of the radioactive source.MethodsFirstly, based on the precise construction of the irradiator, a comprehensive simulation was conducted to analyze the control characteristics of the irradiation device on the beam according to the accurate modelling for the irradiator. Then, detailed simulation and analysis was performed in order to obtain the properties of the field in terms of radiation dosimetry research interests. Finally, a novel irradiator featuring triple screen functionality was implemented to facilitate the onsite installation and the following work.ResultsThis novel irradiator is appropriate to perform the onsite loading for the radioactive source. The scattered components in the spectrum can be reduced by 25% with the presence of a scattering chamber in the irradiator. The leakage radiation in the normal working area can be decreased by 7 orders of magnitude due to the triple screening functionality of the irradiator, the optimization is proposed for installation of the key components in the irradiator. The deformation of the spectra in the radiation field caused by the collimator complex is about 0.2%. The wall effect in the course of the absolute measurement of air kerma for three types of typical ionization chamber is calculated in the 60Co γ radiation field, and the correction factors diverge less than 0.2% from the ones in the earlier publications.ConclusionsThe irradiator designed and implemented in this study can be a proper candidate to support the radiation dosimetry related work under the current regulation. The compact structure enables the irradiator to provide an extended dynamic range of dose rate, leading to a relatively long period of service.

BackgroundActive neutron interrogation (ANI) measurement systems can quantify the fissile masses of special nuclear materials (SNMs) using neutrons and are widely used in nuclear safeguarding fields. However, different hydrogen densities effect the delayed neutron count rate differently, and the hydrogen-containing matrix in the waste drum weakens the signal of delayed neutrons and limits the fissile mass measurement precision of the ANI system.PurposeThis study aims to improve the assay performance of the ANI system in nuclear safeguard measurements by developing a matrix correction method.MethodsBased on the response of the flux monitor, a new matrix correction method was developed based on the conventional correction methods. A WM3210 PAN Shuffler system model was constructed using the Geant4 toolkit, and correction method was explored for common matrix materials. Finally, U3O8 materials with different enrichments and distributions were applied to the effectiveness comparison between new correction method and the traditional method.ResultsThe results show that for U3O8 materials with different enrichments and distribution states, both the traditional correction method and the new correction method can effectively reduce the influence of the matrix material on the measurement of SNM mass. For the case where U3O8 material is located at the center of the matrix, the average relative deviation of the 235U mass obtained by the new correction method is 13.6%, while that of the traditional method is 23.8%. For the case where U3O8 material is uniformly dispersed in the matrix, the average relative deviation of 235U mass obtained by the new correction method is 7.78% whilst that of the traditional method is 20.0%.ConclusionResults of this indicate that the new method demonstrates better correction capability than the traditional one.

BackgroundBoron Neutron Capture Therapy (BNCT) is an emerging radiotherapy technique. However, respiratory motion has a critical impact on the dose accuracy in BNCT treatment of lung cancer.PurposeThis study aims to quantify the dosimetric impact caused by respiratory motion during BNCT treatment of lung cancer.MethodsThis study adopted the Monte Carlo simulation method was adopted to develop a dynamic model that captured the spatiotemporal variations of tumors and organs caused by respiratory motion during lung cancer treatment, and performed dose calculations for BNCT. Firstly, the Multi-function and Generalized Intelligent Code-bench based on Monte Carlo method (MagicMC) was employed to model the adult male phantom provided by Oak Ridge National Laboratory (ORNL). Then, a dynamic dose calculation model was established by incorporating high-order cosine functions that described respiratory motion. Finally, MagicMC was applied to the calculation of the dose errors in tumors and organs resulting from respiratory motion in different directions within three-dimensional space.ResultsThe results indicate that during a respiratory cycle, the tumor in all three motion directions exhibits the largest percentage dose difference at the 50% phase. In the left-right direction (LR), it is 0.310%; in the anterior-posterior direction (AP), it is 5.830%; and in the superior-inferior direction (SI), it is -2.852%. The closer healthy tissues are to the irradiation field, the higher the dose rate they receive. The maximum percentage dose difference for the heart in the LR direction is 2.070%, and the maximum percentage dose differences for the right lung in the AP and SI directions are 4.128% and -11.962%, respectively. During BNCT treatment irradiation, organ motion in the AP direction has the greatest impact on tumor dose, resulting in a dose error of 1.644%. For healthy tissues, the dose errors induced by motion in all three directions remain within ±4%.ConclusionsThe study demonstrates that organ respiratory motion during BNCT treatment for lung cancer affects the doses received by tumors and healthy tissues, the calculation results can provide a reference for precise dose calculation and clinical irradiation dose correction in BNCT treatment of lung cancer.

BackgroundLiCl-KCl eutectic is often used as the electrolyte in pyroprocessing of spent nuclear fuel. The corrosion of structural materials by LiCl-KCl eutectic is primarily affected by impurities. Purifying molten salt to reduce its corrosivity is one of the main ways to solve the material corrosion issue in pyroprocessing.PurposeThis study aims to investigate the effect of electrolytic purification on the corrosion of Inconel 600 by LiCl-KCl Eutectic molten salt.MethodsThe purification of LiCl-KCl salt was conducted by electrolysis method in this study and the cyclic voltammetry (CV) test was performed for monitoring the concentrations of impurities during the purification process. Inductively coupled plasma mass spectrometry (ICP-MS) measurement was performed for the salt to examine the impurity elements before and after purification. Inconel 600 alloy samples were subjected to a 500-h immersion corrosion test with purified and untreated LiCl-KCl salts under Ar atmosphere at 773 K. Scanning electron microscopy (SEM) and X-ray diffractometer (XRD) were employed to characterize the morphology and elemental distribution of the post-corrosion samples.ResultsExperimental results show that the metal ion impurities are removed by electrolytic purification with the removal rates of Fe and Cr reach 46.1% and 51.4%. In argon protected environment, samples are corroded by untreated LiCl-KCl salts and LiCrO2 generates on the surface of samples. Samples exposed to purified salts are less corroded than which exposed to untreated salts.ConclusionsElectrolytic purification has a significant removal effect on metal ion impurities in LiCl-KCl salt, and effectively mitigates the corrosion of Inconel 600 by high-temperature molten salt.

BackgroundBeam loss in a particle accelerator generates a pulsed neutron radiation field. However, the pulse-counting neutron rem counter faces significant issues with missing counts.PurposeThis study aims to address the issue of missing counts and achieve accurate measurements of pulsed neutron radiation doses.MethodsFirstly, based on the physical mechanisms behind low current generation in nuclear reactions, a detection physical model was established that utilized the charge generated by neutron nuclear reactions as a reference quantity for measuring neutron interactions. Then, an integrating electronic system capable of accurately measuring low current and charge was independently developed. Finally, a method for measuring intense pulsed radiation fields was implemented, and experimental tests were conducted at the China Spallation Neutron Source (CSNS).ResultsExperimental results at CSNS demonstrate that the number of nuclear reactions in a single pulsed neutron storm reaches up to103. The dose rate varies from μSv?h-1 to several hundred mSv?h-1, with pulsed neutrons observable on a millisecond timescale. The measured values closely align with simulated values, exhibiting a discrepancy of less than 50%.ConclusionsThis study not only addresses the gap in pulsed neutron measurement capabilities in China, but also fulfills the urgent needs of the CSNS.

BackgroundCurrently, coded aperture γ camera has been extensively utilized in radioactive source location and imaging task. Owing to the advancement of deep learning technology, numerous image reconstruction algorithms have been developed to address the issue of direct convolution algorithm and iterative algorithm for suppressing random noise whilst the convolutional neural network (CNN) algorithm is one of popular candidates, but it has not been fully discussed in existing studies.PurposeThis study aims to explore the influence of CNN algorithm, and the size of dataset on the performance of the imaging model for γ camera.MethodBased on point source imaging process of coded aperture γ camera, Monte Carlo simulation combined with linear random was applied to generate the dataset. The Geant4 software was employed to simulate and encode the aperture imaging process, and the CNN algorithm was utilized to accomplish image reconstruction.ResultsThe image reconstruction results show that, for orphan source imaging, the average Contrast-to-Noise Ratio (CNR) of the model using the 57Co training set for 57Co source location is 75.8, and the average CNR is 24.7 for 137Cs source location, while that of the model using the 137Cs training set for both locations is 43.8 and 44.3, respectively. With the increase of datasets capacity, the generalization ability of 57Co and 60Co training model decreases, while the learning effect of 137Cs training model increases gradually. When reconstructing seven random 60Co sources in the field of view, the CNR of the optimal model is 8.9, and the location of the radioactive source can be clearly identified.ConclusionsThe finding of this study indicates that the capacity and characteristics of the datasets directly impact the learning and generalization capabilities of the CNN model. In high-energy and multi-point imaging scenarios, augmenting the noise in the training set and selecting an appropriate dataset capacity can enhance the effectiveness and accuracy of localizing and reconstructing radioactive sources.

BackgroundThe inner working conditions of a reactor are complicated and affected by many factors. Accurate prediction of the key thermal parameters of the reactor core under various working conditions can greatly improve reactor safety. Most of the existing research focuses on the prediction method that uses a single neural network. In the case of excessive noise, a single neural network cannot sufficiently eliminate noise and accurately detect data change.PurposeThis study aims to propose a novel transient thermal hydraulic parameter prediction method for fast reactor core, making use of a model that is based on the empirical mode decomposition (EMD) and singular spectrum analysis (SSA) combined with an adaptive radial basis function (RBF) neural network.MethodsFirstly, the 1/2 China Experimental Fast Reactor (CEFR) was used as the research object, and the fast reactor subchannel program SUBCHANFLOW was employed to generate a time series of transient core thermal hydraulic parameters. Then, two combined models, i.e., EMD-RBF and EMD-SSA-RBF, were used to predict the core mass flow rate and time series of the maximum temperature on the surface of the cladding. Both the single step prediction and continuous prediction were performed.ResultsThe results show that compared with a single RBF neural network, the single-step prediction errors of mass flow rate with the EMD-RBF combined neural network and EMD-SSA-RBF combined neural network are reduced by 41.2% and 86.7% respectively, whilst the single-step prediction errors of temperature are reduced by 44.7% and 60.5% respectively. Not only the prediction errors are significantly reduced, but also the calculation time for parameter prediction is shortened.ConclusionsThe combined neural network models proposed in this study can make fast and high-precision predictions, providing advantages over the deep neural network. Hence have certain reference value for improving the safety of the reactor in engineering applications.

BackgroundCorrosion products such as iron and nickel ions generated in the steam generator (SG) of a pressurized water reactor (PWR) deposit on the fuel rods in the reactor core, forming Chalk River Unidentified Deposits (CRUD). Activated by neutron irradiation in the reactor core, part of the CRUD layer transforms into radioactive substances, which are mainly 58Co and 60Co. Then the radioactive 58Co and 60Co are carried by the coolant into the entire primary loop. The existing research lacks a comprehensive modeling and discussion on the distribution of radioactive materials 58Co and 60Co in the primary loop. Predicting the content and distribution of radioactive materials 58Co and 60Co in the primary loop and assessing the impact of water chemistry and thermal parameters are of significant importance for radiation protection and core parameter design.PurposeThis study aims to explore the production and distribution of radioactive materials in the primary loop due to the deposition of CRUD in reactor core, with a typical PWR primary loop as the research subject.MethodsFirstly, a predictive model for CRUD deposition and radioactive materials production distribution was established that encompassed CRUD deposition and radioactive material prediction. Then, the primary loop of PWR was simplified into five key nodes, i.e., the core, soluble corrosion products, SG, corrosion particulates, and erosion particles, to address the generation, migration, deposition, and growth of CRUD based on the principles of mass transfer and water chemistry. The proportion of particles returning coolant was controlled by the purification efficiency in the erosion particle node. Finally, the activation of CRUD and the migration, deposition, and erosion of radioactive materials at each node were correspondingly considered on the basis of the activation theory, and the distribution of radioactive materials in the primary loop was obtained by establishing and solving the mass transport balance equations for each node. Based on this established model, a comprehensive analysis was conducted on the influence of coolant flow rate, hydrogen content, and coolant inlet temperature.ResultsThe calculation results indicate that the radioactive materials inventory increases with an increase in coolant flow rate and hydrogen content. The impact of coolant flow rate and hydrogen content on the radioactive materials inventory of steam generators (SG) is 93.9% and 10% greater than the core. As the coolant inlet temperature increases by 8%, the radioactive materials inventory decreases by 9%, and its impact on the core is 19% greater than the SG. The model predictions for CRUD deposition and radioactive materials distribution closely align with the results obtained from the code CRUDSIM (Chalk River Unidentified Deposits SIMulation) with a difference less than 5%.ConclusionsThe results of this study demonstrate a significant influence of coolant flow rate, hydrogen content, and inlet temperature on the radioactive material content and distribution in the primary loop. Moderating coolant flow rates and reducing hydrogen concentrations are beneficial for lowering the content of 58Co and 60Co in SG. Conversely, increasing coolant inlet temperature effectively reduces the content of 58Co and 60Co in the core.

BackgroundMolten salt reactor is a promising type of reactor in the fourth generation advanced nuclear reactor system due to its excellent safety and economy. However, as the coolant for the molten salt reactor system, lithium fluoride beryllium (FLiBe) has a melting point of 460 ℃, which is much higher than the ambient temperature, so there is a risk of coolant solidification in the system.PurposeThis study aims to establish a one-dimensional solidification model with mushy zone effect based on energy conservation and enthalpy porous medium model.MethodsFirstly, based on the energy conservation equation, a solidification layer thickness model was established and a source term model with mushy zone was established based on the enthalpy porous medium model. The velocity and temperature distribution models were obtained on the basis of the boundary layer theory. Secondly, the molten salt solidification experiment was designed to verify these models. Finally, the system safety analysis program ASYST-SF was employed to simulate the filling behavior of FLiBe coolant in the pipe.ResultsThe experimental verification results show that the overall model error is less than ±10%, meeting the requirements of reactor system safety analysis. The evolution behavior of fluid temperature, solidification layer thickness, and pressure drop of the pipe filling behavior under typical working conditions are observed.ConclusionsThe model and calculation results are of great significance for improving the operational safety of molten salt reactors.

BackgroundThe control rod channel tubes of the Thorium Molten Salt Reactor (TMSR) are typical high-temperature, thin-walled, long cylindrical shells designed to withstand external pressure, with creep buckling as its primary failure mode.PurposeThis study aims to use numerical simulation methods to study the creep buckling instability behavior of control rod channel tubes at the elevated temperatures.MethodsFirstly, the Norton creep model and material parameters for the UNS N10003 alloy was obtained on the basis of the high-temperature creep test data. Furthermore, finite element analysis software ABAQUS was employed to assess eigenvalue buckling and creep buckling for TMSR control rod channel tubes. Sensitivity analysis was conducted on the key factors causing buckling instability, and an empirical formula for creep buckling life was obtained.ResultsThe analysis results reveal that temperature, pressure, and structural dimensions significantly influence the tube's creep buckling life, and the derived empirical formulas can be used to verify the durability of the tubes. To ensure a design life of 30 a for the casing at 700 ℃, the tube height needs to be controlled below 3 m. If the design life is 10 a, the tube height can be increased to 6 m.ConclusionsThis study offers engineering guidance for the stability design of TMSR control rod channel tubes and high-temperature structures under complex conditions, and it also serves as a basis for predicting the creep buckling lifespan of other high-temperature thin-walled structures.

BackgroundThe thermal stratification in the upper plenum of lead-bismuth fast reactor after emergency shutdown has a significant impact on the structural integrity of the reactor and the residual heat removal capacity of the natural circulation. The research on thermal stratification based on Computational Fluid Dynamics (CFD) method has the problems of large computational overhead and time-consuming whilst the existing standard dynamic mode decomposition (DMD) method has poor forecasting results on thermal stratification.PurposeThis study aims to solve this problem by proposing a thermal stratification model reduction method for the upper plenum of lead-bismuth fast reactor.MethodsFirstly, the high-precision full-order snapshot was obtained on the basis of the CFD program FLUENT. Then, based on the truncated DMD, the time step samples were compressed according to the characteristic frequency, and the thermal stratification reduction model was constructed by combining the Long Short-Term Memory (LSTM) neural network with DMD. Finally, three methods, i.e., standard DMD, improved DMD and improved DMD-LSTM, were comparatively analyzed in terms of temperature oscillation error and computation time.ResultsComputational results show that the thermal stratification model reduction method based on improved DMD and LSTM in the upper plenum of the lead-bismuth fast reactor achieves best performance, with root mean square error reduced by 46.60% and 30.45% respectively, compared to standard DMD and improved DMD. The computational time of the improved DMD and LSTM is only 4.4% of FLUENT's, significantly improving efficiency and enabling faster emergency response in lead-bismuth reactors.ConclusionsResults of this study verify that the thermal stratification model reduction method proposed in this paper can better simulate the temperature distribution in the upper plenum and realize the rapid prediction of the thermal stratification phenomenon.

BackgroundThe Electron Cyclotron Resonance (ECR) preionization is important for the reliable start-up of spherical tokamaks.PurposeThis study aims to investigate the effects of power deposition, electron density, and electron temperature of ECR pre-ionization process under different power conditions by simulation.MethodThe spherical tokamak device NCST (NanChang Spherical Tokamak) at Nanchang University was selected as research object, and COMSOL Multh-physics, a multi-physics simulation software, was utilized on the basis of the finite element method to simulate the process of ECR pre-ionization forming plasma in the device. Firstly, a three-dimensional model of the NCST device was established by using the parametric modeling method. Then, through the AC/DC, radio frequency (RF) and plasma modules in COMSOL software, and by correctly defining the electromagnetic wave source term, plasma parameters and reasonably setting boundary conditions, the laws of magnetic field, electron density and electron energy changing with time and space were solved through multi-physics field coupling.Results & ConclusionThe results show that increasing the input power can greatly shorten the ionization time of electrons, and greatly increase the peak electron density and electron temperature of plasma. However, too high input power will also cause too large plasma density generated by ionization, making the incident electromagnetic wave difficult to reach the resonance region, thus reducing the heating efficiency of ECR. Higher power can make the heating effect of ECR better, hence greatly shorten the ionization time of electrons, but the duration of this process will also decrease with the increase of power.

BackgroundMegawatt-level nuclear reactor combined with helium-xenon Brayton cycle system can effectively meet the energy needs of large-scale deep space explorers, satellite base, deep-sea unmanned underwater vehicle and other special energy power equipment for high power, small size, highly reliable power supply, which has wide application foreground and research necessity. Currently, the study of the physical properties of helium-xenon gas mixtures in non-ideal state is not sufficient.PurposeThis study aims to establish the thermophysical property model and the thermodynamic model of helium-xenon Brayton cycle, and analyze the effect of the non-ideal gas characteristics to the thermal performance of the cycle.MethodsThe second or third order virial expansion was adopted to construct the helium-xenon mixture physical property model to reflect the deviation caused by the non-ideal gas characteristics. The thermodynamic models of turbine, compressor, mixing chamber, and heat exchanger were conducted on the basis of thermophysical property model. Then, the function models of efficiency and specific work were derived from the thermodynamic models of the above main components, and verified by the submerged subcritical safe space reactor (S4) design. Finally, the influence of the thermophysical properties of helium-xenon mixture on thermal performance of helium-xenon Brayton cycle system such as adiabatic coefficient, pressure loss and relative convective heat transfer coefficient at different temperature, pressure and molar fraction of helium was analyzed, and the influence of He-Xe mixing ratio on the He-Xe thermophysical property under different temperature and pressure was explored.ResultsThe second or third order virial expansion was adopted to construct the helium-xenon mixture physical property model to reflect the deviation caused by the non-ideal gas characteristics. The thermodynamic models of turbine, compressor, mixing chamber, and heat exchanger were conducted on the basis of thermophysical property model. Then, the function models of efficiency and specific work were derived from the thermodynamic models of the above main components, and verified by the submerged subcritical safe space reactor (S4) design. Finally, the influence of the thermophysical properties of helium-xenon mixture on thermal performance of helium-xenon Brayton cycle system such as adiabatic coefficient, pressure loss and relative convective heat transfer coefficient at different temperature, pressure and molar fraction of helium was analyzed, and the influence of He-Xe mixing ratio on the He-Xe thermophysical property under different temperature and pressure was explored.ConclusionsThe proposed model can accurately calculate the thermophysical properties of the He-Xe mixture, including density, specific heat capacity, viscosity, thermal conductivity and Prandtl number can be accurately calculated by the proposed model under different helium molar fraction. The model proposed in this work can be applied to the design and optimization of the He-Xe Brayton cycle systems and direct the device selection of the He-Xe Brayton cycle system.

BackgroundThe fluids on both sides of the helical coil steam generator of the high-temperature gas-cooled reactor (HTGR) are helium and water respectively, whose physical properties are quite different and the transient response time is different. Traditional semi-implicit finite difference scheme applied to thermal hydraulic analysis code can only apply small time steps due to Courant-Friedrichs-Lewy (CFL) condition, which will decrease the computation efficiency.PurposeThis study aims to develop and verify a new transient analysis code using all-implicit algorithm for the helical coil steam generator of the high-temperature gas-cooled reactor.MethodsFirstly, based on homogeneous flow model, a fully implicit finite difference scheme for the convection and diffusion term combined with the full coupling solution algorithm of flow and heat for thermal conductivity process to develop a new transient analysis program, named NUSOL-HTGRSG, for the helical coil steam generators of HTGR. Then, verification of the code was conducted in the four aspects: the design condition of the HTR-PM steam generator used for steady-state calculation validation, the number of grids changed for spatial sensitivity analysis, the time step changed for time sensitivity analysis, and the transient calculation carried out under the same disturbance (the flow rate of the primary side of the steam generator reduced by 10%). and results were compared with that of NUSOL-SG calculation.ResultsSteady-state validation results show that the relative errors of outlet temperatures and pressure drops in the primary and secondary sides are generally within 1%. Transient validation results indicate that, under identical transient conditions, the maximum relative deviation between the transient responses of the two codes is 1.4%.ConclusionsThe validation results demonstrate that the NUSOL-HTGRSG code can effectively predict the operating parameters of the helical coil steam generator in HTGRs under steady-state conditions and accurately capture its transient characteristics with a relatively large time step (5 s).

BackgroundIn pressurized water reactors (PWR), grid to rod fretting (GTRF) is a primary cause of fuel failure due to fuel rod vibrations. Understanding and characterizing the dynamics of fuel rods is essential for analyzing GTRF and ensuring reactor safety.PurposeThis study aims to develop a vibration analysis method that can reasonably represent the dynamics of fuel rods within a reactor, focusing on the vibration characteristics of fuel rods supported by multiple positioning grids.MethodsFirstly, a mechanical model was established for a multi-span elastically supported fuel rod restrained by multiple sets of positioning grids. The restraint effect of the grids was simplified into tension and compression springs and torsion springs within the elastic range. Then, a displacement function based on an improved Fourier series was constructed for the whole beam section, and the modal state was solved using the energy principal method. Subsequently, the Improved Fourier Series Method (IFSM) was used to address boundary discontinuity issues, eliminating the need to reconstruct the model for structural and boundary changes. Finally, the accuracy of this method was verified by comparing with the finite element calculation results, and the vibration characteristics of fuel rods were analyzed.ResultsThe results show that the tension spring stiffness is the dominant factor influencing the overall variation pattern of the intrinsic vibration frequency of the fuel rod. The influence of the torsion spring on the vibration characteristics is dependent on the tension spring stiffness, with minimal impact when the tension spring stiffness is small. Changes in boundary conditions affect the system's stiffness, which in turn influences modal frequency. Increased overall stiffness leads to increased deformation resistance and higher modal frequency.ConclusionsThe study concludes that the strength of the restraint effect of the spacer grid on the fuel rod significantly influences the vibration characteristics of the fuel rod under multi-span elastic support. The developed method provides a reliable tool for analyzing the vibration characteristics of fuel rods with multiple grid constraints, which can be used as a reference in practical engineering applications.

BackgroundIn pursuit of promoting the diversified development of energy cooperation demands among countries participating in the Belt and Road Initiative and address the demand for secure and efficient energy supply along the Belt and Road Economic Belt, Xi'an Jiaotong University has actively innovated and proposed a small modular fluoride-salt-cooled high-temperature advanced reactor FuSTAR.PurposeThis study aims to evaluate the load following capability and safety of the FuSTAR reactor.MethodsThe thermal-hydraulic modeling of the reactor body and the residual heat removal system of a small modular fluoride-salt-cooled high-temperature advanced reactor FuSTAR was conducted using conservation equations in macro form and point kinetics equations. Then, the one dimensional thermal fluid simulation program was used for modeling calculation and a constant coolant outlet temperature scheme was employed in the design of the control system for FuSTAR reactor by coupling simulation program with Simulink. Finally, the disturbance rejection characteristics and load following capability of the FuSTAR reactor were analyzed by inserting reactive disturbances and varying thermal load conditions.ResultsCalculation results show that FuSTAR demonstrates load following capability without relying on an external control system, mainly due to its inherent safety features, which allow the reactor to self-stabilize and regulate under load variations. With the adoption of a constant coolant outlet temperature control scheme, the load following capability of FuSTAR has been further enhanced. In the tests of 10% FP (Full Power) load step change and 5% FP·min-1 rate linear load rise and fall, the overshoot of nuclear reactor power is strictly controlled within 5%.ConclusionsResults of this study indicate that FuSTAR has a good load following capability because of the negative temperature reactivity feedback and the existence of control system, which fully meets the requirements of safety operation of the reactor.

BackgroundThe accurate monitoring of radioactive substances in gaseous effluents from nuclear facilities is critical for ensuring environmental safety. The gas sampling monitoring system facilitates continuous sampling and measurement of these substances. However, the deposition of gaseous effluents within the system can compromise the representativeness of the sampling results if not accurately accounted for.PurposeThis study aims to investigate the penetration efficiency of micron aerosols in horizontal sampling pipes used in nuclear power plant chimneys and to develop a more accurate prediction model for particle deposition.MethodFirstly, the TSI3321 aerodynamic particle size spectrometer was utilized to precisely measure the penetration of the horizontal sampling pipeline for aerosols, encompassing the influence of various pipe roughnesses, aerosol particle sizes, wind speeds, and pipe diameters. Further, a modified prediction model of aerosol penetration was constructed by considering factors such as effective roughness, turbulent diffusion, gravitational settlement, and particle inertia. By comparing the model with this experimental data and the historical experimental data, the accuracy of the predicted settlement rate was verified, and the error was basically controlled within 10%. Finally, in combination with the model and experimental data, the impacts of wind velocity, pipe diameter, and aerosol particle size on deposition velocity and amount were analyzed with comparison against empirical formulas and historical data.ResultsThe experimental results reveal that the revised formula predicts deposition velocity with an error margin of less than 10%, and the surface roughness variations significantly affect the flow field and wall resistance, influencing deposition rates. Even minor changes at the micron level can result in substantial differences in turbulent deposition rates.ConclusionThe results of this study emphasize the substantial influence of surface roughness on particle deposition rates and highlight the existence of an optimal sampling wind speed that maximizes particle penetration in the diffusion-collision zone. For smaller particles predominantly in the diffusion region, sedimentation is primarily governed by gravity and Brownian diffusion, resulting in lower sedimentation velocities. The findings contribute to the enhancement of sampling system design and operation in nuclear facilities for more precise monitoring of gaseous effluents.

BackgroundThe accurate reconstruction of γ radiation fields is fundamental to the digitalization of radiation protection and is a prerequisite for radiation dose assessment and visualization simulation. Traditional interpolation methods and uniform source activity inversion methods struggle to accurately reconstruct high-dose-rate gradient 3D radiation fields in scenarios with high-gradient, non-uniform activity distributions within large volume source terms inside nuclear facilities.PurposeThis study aims to develop an inversion method for non-uniform source activity distribution and apply it to accurately reconstruct 3D gamma radiation fields of the aforementioned types.MethodsBased on multi-objective source activity inversion and the Bayesian Information Criterion, an innovative inversion method for non-uniform source activity distribution was proposed. Then, the accuracy of radiation field reconstruction results in a pipeline simulation case obtained by using this method and ordinary Kriging interpolation method were compared under different source activity distribution conditions in various regions. Finally, the effectiveness of this method was further validated using measured data from inside a nuclear facility.ResultsUnder four different activity distribution conditions in the pipeline simulation case, the proposed method achieves an ARD (average relative deviation) of less than 5% for radiation field reconstruction results in all regions, significantly outperforming the ordinary Kriging interpolation method, especially in high-dose-rate gradient areas. In the real nuclear facility scenario, the ARD between 77 reconstructed dose rate values calculated from 30 measured values and the actual measurements is only 12.69%, much lower than the result of 85.40% by ordinary Kriging interpolation.ConclusionsThe inversion of non-uniform source activity distribution achieved by introducing the Bayesian Information Criterion in this study is very suitable for gamma radiation field reconstruction under complex source term conditions. It provides advanced technical support for the digitalization and simulation of radiation protection based on dynamic data in nuclear facilities, enhancing its effectiveness in practical applications.

BackgroundSupersonic molecular beam injection (SMBI) is a widely used auxiliary fueling method in magnetic confinement devices, significantly influencing plasma flows and turbulence.PurposeThis study aims to investigate the effects of SMBI on flows and turbulence in the edge plasma of a tokamak.MethodsThe experiment was conducted on the J-TEXT tokamak with a major radius of 1.05 m and a minor radius of 0.255 m. The SMBI was injected into the vacuum through one of the bottom ports. A three-step Langmuir probe array was mounted on the top of the tokamak and was employed to measure and analyze edge plasma flows and turbulence with SMBI injection. Particular attention was given to geodesic acoustic mode (GAM) zonal flows, ion-ion collision frequency, turbulent Reynolds works, gradients, and turbulence-driven particle flux.ResultsResults indicate that the GAM, turbulence, and turbulent particle flux are suppressed after SMBI for 40%~60%, 50%~70%, and 60%~70%, respectively, accompanying the increase in ion-ion collision frequency, and the decrease in turbulent Reynolds stress and the turbulent Reynolds work.ConclusionsThe suppression of GAM is related to the decrease in the nonlinear driving from turbulence and the increase in collision frequency; the decrease in turbulence may be the result of the flattening of density and temperature gradients; and the reduction of turbulent heat flux may come from the drop in the fluctuations of radial velocities.