Compared with turn-by-turn beam diagnostic techniques widely used in electron storage ring, bunch-by-bunch diagnostic technology allows the measurement and analysis of each bunch, offering a more comprehensive understanding of the internal state of the electron beam with results that are closer to the true physical model. Recent advancements in data acquisition equipment and signal processing algorithms have laid the foundation for the continuous development of bunch-by-bunch diagnostic techniques. This article provides an overview of the basic principles and architecture of bunch-by-bunch diagnostics, summarizes exploratory work and research achievements in this field by major domestic and international research groups, highlights the research approach, latest findings, and technological applications explored by the Shanghai Synchrotron Radiation Facility (SSRF) team. The future research directions worth attention and development trends are discussed, offering valuable insights for researchers dedicated to the field of beam measurements.

BackgroundIn the process of nuclear fuel generation, neutron poisons are added to enhance performance of nuclear fuel. Erbium is a common neutron poison, and its content needs to be measured and analyzed during the production of such nuclear fuels. The traditional methods have limited penetration and can only analyze the surface of the samples, unable to penetrate the bulk nuclear fuel materials for internal component analysis. Prompt Gamma-ray Neutron Activation Analysis (PGNAA) is a non-destructive testing technique, which is suitable for detecting large samples.PurposeThis study aims to explore the feasibility of determining the erbium in large samples based on PGNAA technology.MethodsFirstly, a deuterium-tritium (D-T) neutron generator and a high-purity germanium (HPGe) detector were employed to establish a measurement platform. Erbium oxide was selected as the sample, and measurements were conducted utilizing the 815.9 keV peak emitted from the reaction of fast neutrons with erbium. Then, the neutron yield of D-T neutron generator was calculated using copper foil activation and Monte Carlo simulations, and the neutron spectrum at sample position was calculated using Monte Carlo simulation for observing the thermal and fast neutron fluxes. Finally, the calibration curve and mass detection limit were analyzed.ResultsMeasurement results show that the neutron yield of D-T neutron generator is (2.34±0.01)×106 s-1 and the fast neutron flux at sample position is 106 cm-2?s-1. Analysis results demonstrate that there is a good linear relationship between the 815.9 keV peak counts and the mass of erbium. The mass detection limit for erbium is 28 g. In addition, there is no interference between the intrinsic gamma ray of 238U and the 815.9 keV Er peak.ConclusionsThis study veri?es the feasibility of PGNAA technology for the erbium determination, which can be used for further analysis of erbium in nuclear fuel.

BackgroundMulti-Wires detector (MW) is widely used in beam profile measurements. However, wire deformation and even wire broken have also happened frequently during the MW operation due to beam power deposition on the wire under high beam power environment.PurposeThis study aims to investigate the influences of the beam parameters and detector design, especially wire tension structure, on the wire temperature and wire deformation arising therefrom.MethodsFirstly, based on the backward Euler method with adaptive steps, a numerical algorithm was developed to conduct temperature simulation of MW. Then, verification experiments with various beam parameters and detector design of MW were performed in an ion source platform at Institute of Modern Physics (IMP), Chinese Academy of Sciences, and the wire deformation caused by temperature was reproduced and observed at HIMMWW (Heavy Ion Medical Machine at WuWei city, China) complex. Finally, comparative analysis was conducted on the relevant results to find the appropriate beam parameters and detector design.Results & ConclusionsExperimental results show that temperature plays an essential role on wire deformation if none tension mechanism is implemented on wire structure. Based on numerical simulations, experiments verifications and operation experiences, a fixed wire tension maintained by welding is appropriate while the wire temperature is below 1 300 K, which also provides a simple construction and a low cost. After exceeding 1 300 K, pre-tensioning by a spring is essential to support the wire with a constant tension to avoid deformation.

BackgroundThe Super Tau Charm Facility (STCF) is a new generation electron-positron colliders at the forefront of high precision, and its high brightness requirements pose a major challenge to accelerator technology. Resonant cavity-based monitors utilize characteristic mode signals for non-intercepting, high signal-to-noise ratio measurements, hence may meet the online high-resolution measurement requirements of various high-quality linear accelerators.PurposeThis study aims to address the challenges posed by the short lifetime and the small dynamic aperture of the storage ring beams in the STCF by developing high-resolution monitoring techniques for bunch length and charge to ensure efficient injection and precise measurement of these parameters.MethodsAccording to the beam parameters and measurement requirements of the STCF injector, the physical design and simulation of the resonator bunch length and charge monitor were carried out. Two Pill-Box cavities were designed by using Computer Simulation Technology (CST) modeling, and their structures were optimized. Subsequently, the beam load in the CST particle studio for simulation was conducted to analyze influences of beam tilt and lateral offset on the measurement accuracy, and the measuring resolutions of bunch length and charge were evaluated using cavity beam position monitor (CBPM).ResultsSimulation results show that the measurement errors of bunch length and charge are 3.3% and 0.02%, respectively. According to the online test results of the same type monitor, it is estimated that the resolution of bunch length of the monitor is expected to reach 100 fs@1.5 nC, and the relative resolution of charge measurement is better than 0.07%.ConclusionsThe currently designed monitor meets the diagnostic requirements of bunch length and charge of STCF, it will be manufactured in future for online testing.

After decades of research, the problems and behavior of the corrosion of reactor alloy materials under service conditions are clearly understood. However, some problems in corrosion of reactor materials have not been clarified, including the critical corrosion process of the reactor materials under operational conditions, the role of a single factor in the corrosion process, and the prediction of corrosion behaviors of new materials in extreme environments. The density functional theory (DFT), which is based on quantum mechanics, can be employed to accurately predict the motion process of atoms and the change in the relevant energy within a very short period. The DFT has become an important auxiliary method for investigating the corrosion process of reactor alloy materials in recent years and can help solve the above problems. In this paper, the DFT is firstly introduced, which mainly includes the theoretical basis, development process, and mainstream computational software. Then, a comprehensive discussion and analysis are conducted on the current research status of DFT applied to the corrosion of reactor alloy materials, including the adsorption, separation, combination, and internal diffusion of the reactor alloy material surfaces in the environments of water-cooled reactor, liquid-metal-cooled reactor, and molten salt reactor. Finally, future development trends of DFT application in the corrosion of reactor alloy materials are prospected.

BackgroundAccumulation of α-synuclein is a major hallmark of Parkinson's disease (PD). The development of PET tracers to visualize aggregated α-synuclein is useful for early diagnosis and treatment of PD.PurposeThis study aims to design and synthetize a novel PET tracer, i.e., 18F-YM, for for alpha-synuclein PET imaging, and conduct preliminary biological evaluation.MethodsFirstly, based on benzothiazole scaffolds, 2-((3-fluorobenzyl)thio)-6-(3-[fluorine-18] propoxy)benzo[d]thiazole, a small molecule compound, denoted as 18F-YM, was prepared as PET tracer and labeled using Cu(II) mediated radiofluorination methods. The imaging properties of the tracer were primarily evaluated through PET imaging at A53T mice and normal mice. Additionally, the imaging properties of the tracer were also investigated through biodistribution experiments as well as ex vivo autoradiography and pathological analysis. Then, through chemical synthesis, compounds Sn-YM and 19F-YM were obtained, and the Sn-YM was labeled with 18F using organic tin fluoride method whilst the resulting product 18F-YM was verified by high performance liquid chromatography. The in vitro stability and octanol-water partition coefficient of 18F-YM were determined. Finally, small animal Micro PET imaging was employed to assess the affinity of 18F-YM for α-synuclein, and autoradiography, pathological analysis, and biodistribution were used to validate the results of small animal Micro PET imaging.ResultsObserved results show that 18F labeled small molecule compound is prepared in nearly 1 h with an undecayed yield greater than 10% and a radiochemical purity greater than 95%. In vivo PET imaging of 18F-YM reveals that more radioactivity is significantly detected in the brains of A53T mice than those of normal mice after administration of 18F-YM. Quantitative analysis shows that the uptake values in the brain of A53T mice and normal mice are (2.35±0.06) %ID/g and (1.38±0.15) %ID/g, respectively while ex vivo autoradiography and histological examination confirm the possibility of detection of aggregated α-synuclein in thalamus, substantia nigra and striatum using 18F-YM. Furthermore, a biodistribution study in normal mice reveals that 18F-YM can be quickly cleaned from the brain of normal mice, indicating that the non-specific binding of 18F-YM in the brain is low, which allowed for obtaining good contrast images.ConclusionThis preclinical study demonstrates that the benzothiazole analog, 18F-YM, owns preferable imaging prosperities, hence be a new candidate for α-synuclein PET imaging.

BackgroundCement serves as an essential cementitious material required for the construction of repositories for radioactive waste. Over the operational lifespan of such repositories, environmental CO2 infiltrates the cement, leading to its carbonization and alteration of its physicochemical properties. This, in turn, affects its efficacy in blocking radionuclides.PurposeThis study aims to assess the long-term safety implications of cement carbonization on radioactive waste repositories.MethodsThe carbonization patterns of cement and its adsorption capabilities regarding the fission nuclide 137Cs was taken as investigating object. Characterization techniques, such as X-ray fluorescence spectrometer (XRF), X-ray diffractometer (XRD), scanning electron microscope (SEM), Fourier transform infrared spectroscopy (FT-IR), etc., were employed to analyze the changes in physicochemical properties of cement before and after carbonization. Furthermore, batch adsorption experiments were conducted to examine the adsorption behavior of 137Cs in carbonized cement, elucidating the impact of carbonization on cement's adsorption performance.ResultsAnalysis results reveal that the carbonization process in cement primarily involves the conversion of hydration products such as Ca(OH)2 and hydrated calcium silicate into CaCO3, resulting in an increase in the specific surface area of cement with higher degrees of carbonization, hence significantly enhance the adsorption capacity for 137Cs due to carbonization. Interestingly, the adsorption capacity exhibits an initial increase followed by a subsequent decline with increasing degrees of carbonization, surpassing that of non-carbonized cement.ConclusionResults of this study implicate that cement's adsorption of 137Cs operates via chemical single-layer adsorption, and the mechanism remains unchanged by carbonization.

BackgroundThe C50 concrete is used as a concrete structural materials for engineering disposal units in cavern-type low and intermediate radioactive waste repositories in order to meet the needs of nuclear power plant operation and decommissioning.PurposeThis study aims to prepare C50 concrete material and investigate its adsorption performance for radioactive nuclides Ni2+ and Cs+.MethodsFirstly, according to the design requirements of C50 concrete, C50 concrete specimens were prepared, and then crushed and ground into powder with a particle size less than 75 μm as experimental samples. Then, series of characterization analysis were performed to observe the crystal structure and content of the C50 concrete sample to reveal the main components of the C50 concrete that comprised silicate minerals with high aluminum content and carbonate minerals with high calcium content. Finally, adsorption experiments were carried out to examine the adsorption performance of C50 concrete for radioactive nuclides Ni2+ and Cs+. Considering the widespread presence of ions in groundwater, the effect of the coexistence of different ions on the adsorption of Cs+ and Ni2+ in the C50 concrete was tested, and the effect of temperature on adsorption was investigated with concern of the radiation exotherms of radioactive waste.ResultsThe analysis results show that the main components of the C50 concrete are SiO2, CaO, and Al2O3 with and specific surface area of 2.786 m2·g-1, and irregular polyhedral and lamellar structures with strong amorphous characteristics are noted at the microscale. The adsorption experiments reveal that the C50 concrete have a good adsorption capacity for Cs+ and Ni2+. The Kd of Cs+ adsorption on the C50 concrete reaches 19.400 L·mg-1 with adsorption capacity of 0.316 mg·g-1 whilst the Kd of Ni2+ adsorption on the C50 concrete reaches 465.142 L·mg-1 with adsorption capacity of 96.375 mg·g-1. With the increase of initial metal ion concentration, the Kd value of Cs+ adsorption on the C50 concrete gradually decreases whilst the adsorption capacity increases. The Kd of Ni2+ increases steadily after a gradual decrease, whereas the adsorption capacity gradually increases with the increase of initial metal ion concentration. Experiments results on the influence of environmental factors demonstrate that the effect of pH on the adsorption of Cs+ in the C50 concrete is relatively small, and the order of inhibition of ions on adsorption is K+ > Ca2+ > Mg2+ > SO42-> Cl-> NO3-. Increasing the temperature leads to a slow decrease in the absorptivity. Meanwhile, as for the adsorptivity of Ni2+ in the C50 concrete increases with pH, and the order of inhibition of ions on adsorption is Mg2+ > Ca2+ > SO42- > CO32-. The adsorptivity of the C50 concrete for Ni2+ increases slowly with an increase in the temperature.ConclusionsThe results of this study provides the basic data for the engineering geological disposal of nuclear waste.

BackgroundNuclear sites require monitoring of artificial radioactive aerosols, but existing instruments often encounter the problem of "measurement uncertainty (radon progeny interference)" in the measurement of artificial radioactive aerosols, requiring effective calibration of radon daughter interference levels.PurposeThis study aims to accurately calibrate the radon progeny measurement instruments and effectively calibratt the radon progeny aerosol interference levels for artificial radioactive aerosol monitoring instruments.MethodsBased on the behavior patterns of aerosol particles and radon along with its progeny, a small volume radon progeny aerosol control device was developed, which was composed of aerosol dilution loop, radon progeny regulation loop, and controller with programmable logic controller (PLC) control system. In the aerosol dilution loop, varying concentrations of carrier aerosols was achieved by utilizing a circulating pump in conjunction with a aerosol generator. In the radon progeny regulation loop, different levels of radon sources were employed to achieve regulation of radon activity concentration. Radon progeny aerosol with stable radon progeny state parameters was achieved by adjusting the radon concentration, aerosol concentration and air exchange rate, and the performance of this device was verified by experiments.ResultsExperimental results show that the stable regulation range of equilibrium equivalent concentration (EEC) is 3.3×102~9.4×103 Bq?m-3, stable regulation range of equilibrium factor: 0.12~0.58, stable regulation range of unbound fraction: 1.4%~62.7%. The variation range of EEC within 4 h is within 10%, and the relative standard deviation of the uniformity experiment is less than 7%.ConclusionsThe stable regulation range of the radon progeny state parameters of the small volume radon progeny aerosol control device proposed in this study is wide, and the uniformity and stability of the radon progeny are good. It can be used to effectively simulate the field measurement environment, achieving the development purpose of this device.

Background222Rn/220Rn and their daughters are widely present in the atmosphere and indoor environment. The presence of high-concentration 220Rn can affect the accuracy of 222Rn concentration measurements.PurposeThis study aims to measure radon concentration more accurately by investigating the variation in the response coefficients of the AlphaGUARD PQ2000 radon monitor to 220Rn under thoron gas concentrations with different wind speeds and directions within a standard radon chamber.MethodsThe single-scintillation proportional counter flow gas static method was employed to determine the thoron concentration in this study. Firstly, a self-made and stable high emissivity solid 220Rn source at the University of South China was applied to the generation of 220Rn gas in the chamber, a temperature and humidity control system was adopted to regulate the inside temperature and humidity of 220Rn whilst the wind speed was controlled by changing the operating frequency of the variable frequency fan. Then, the readings of the AlphaGUARD PQ2000 were observed at various wind speeds and directions under controlled temperature and humidity conditions. Subsequently, the response coefficients of the AlphaGUARD PQ2000 to thoron in diffusion mode were calculated according the observed results.ResultsWhen the wind speed increases from 0.05 m·s-1 to 3.50 m·s-1, the response coefficient of AlphaGUARD PQ2000 to 220Rn increases from 0.044 to 0.126. When there is no wind, the response coefficient is 0.049. When the wind speed is fixed at 0.71 m·s-1 and the angle between the wind direction and AlphaGUARD PQ2000 diffusion window changes from 0° to 180°, the response coefficient decreases from 0.083 to 0.051. When the wind speed is fixed at 1.43 m·s-1 and the angle between the wind direction and AlphaGUARD PQ2000 diffusion window changed from 0° to 180°, the response coefficient is reduced from 0.115 to 0.081.ConclusionsThe response coefficients obtained in this study can provide a reference for correcting the interference of 220Rn in the measurement results of the diffusion mode of the AlphaGUARD PQ2000 radon detector.

BackgroundThe high frame rate area detector is the core detector for the major imaging-based experimental stations at the Shanghai HIgh repetitioN rate xfel and Extreme light facility (SHINE), and its data throughput is expected to reach more than 20 GB·s-1. For the real-time receiving and processing of tens of GB·s-1 raw data, traditional single-machine systems are difficult to cope with.PurposeThis study aims to propose a multi-node distributed data acquisition and processing software architecture for high frame rate area detector at imaging-based experimental stations of SHINE.MethodsFirstly, the performance of different network libraries was investigated, and the synchronous transmission method combined with CPU thread binding was found to have the best single-thread data receiving performance. Then, a parallel event building method was introduced by simultaneously dispatching and combing different module data across multiple nodes based on Bunch ID. Furthermore, the data calibration and the bitshuffle/LZ4 compression algorithm were implemented and tested.ResultsTest results show that the highest single-thread data receiving rate is achieved at nearly 3 GB·s-1, a parallel event building data rate of approximately 23.5 GB·s-1 is achieved by using 4 server nodes, and the realized compression ratio is about 5.7.ConclusionsThe feasibility of the multi-node distributed parallel data acquisition method for high frame rate area detector is verified in this study, providing a foundation for the subsequent development of high-throughput data acquisition software for area detectors.

BackgroundEtched track dosimeters (ETDs) based on CR-39 foils are the most frequently used passive detectors for neutron personal dosimetry at various nuclear facilities. That applying a pre-treatment in carbon dioxide (hereafter, CO2 pre-treatment) can improve the CR-39 detection sensitivity and enlarge the tracks is extremely valuable and warrants further investigation.PurposeThis study aims to investigate the effect of CO2 pre-treatment on the sensitivity and the track size distributions of CR-39 detectors, and to obtain an optimal CO2 pre-treatment condition.MethodsA parametric study was conducted to evaluate the effect of CO2 pre-treatments at different pressures and durations on CR-39 detectors. The detectors were firstly irradiated by a standard 252Cf neutron source, delivering a personal dose equivalent Hp(10) of 2 mSv. Then, the detectors were subjected to carbon dioxide treatment prior to undergoing chemical etching for durations ranging from 6 h to 168 h at partial pressures varying between 0.1 MPa to 1.6 MPa. Finally, the correlation between detector sensitivity and pre-treatment condition was obtained by analyzing the number of registered tracks and track size distribution under each pre-treatment condition.ResultsThe obtained results indicate that the CO2 pre-treatment significantly improves sensitivity and maximum track size, with the sensitivity increasing by up to 870% and the maximum track size expanding from 20 μm to 40 μm. The sensitivity increases linearly with pre-treatment pressures, and the saturation point of the maximum track size appears around 40 μm. As the time and pressure continue to increase, the neutron spectrometry information contained in the track size distributions will be erased, the optimal sensitivity for CR-39 to retain the spectral information in the track size distributions is 6.25 tracks·mm-2·mSv-1.ConclusionsThis study is of significance for improving the sensitivity of CR-39 detectors in neutron dosimetry and selecting CO2 pre-treatment conditions, providing a processing method to enhance the track readability whilst preserving the spectral information in the track distributions for practical applications of CR-39 detectors in neutron dosimetry and spectrometry.

BackgroundThe Method of Characteristics (MOC) is widely applied to high-fidelity numerical simulations due to its robust geometric processing capabilities, as well as its ability to balance computational costs and accuracy during calculations. In addition to MOC, common neutron transport calculation methods also include the Collision Probability method (CP) and the Interface Current method (IC). In MOC calculation, different parameter selections will lead to different values of calculation cost and accuracy.PurposeThis paper aims to evaluate the ability of MOC, CP and IC methods for pin-by-pin calculation, and conduct sensitivity analysis to find the best parameter setting for MOC method.MethodsThe three aforementioned calculation methods were compared from the perspective of theory and numerical calculation. Subsequently, numerical calculation and preliminary analysis of the sensitivity of MOC parameters were conducted based on the 2D C5G7-MOX reference problem.ResultsNumerical calculation results show that the computation time and memory cost incur by the MOC are 23.9 min and 37.5 MB, respectively, and the relative error between the MOC results and reference solutions is only 6.04×10-4. The computing times of the CP and IC methods are 56.7 times and 15.6 times that of the MOC, and the memory costs are 407.7 times and 32.8 times that of the MOC, respectively. As a result of the sensitivity analysis of MOC parameters, the following set of parameters is suggested: a grid division of 6×6, a pole angle of GAUS, a pole number of 2, and an azimuth angle of 30°. The calculation time and the memory cost of this set of parameters are 45.4 min and 264.7 MB, respectively, with the relative error of 5.9×10-5 and the normalized RMS error of 0.002 55.ConclusionsThe results of this study indicate that the MOC is superior to the CP and IC methods in accuracy, efficiency, and memory cost, and the grid division of MOC has the greatest influence on the calculation memory cost and calculation time whereas the choice of polar angle has the greatest influence on the calculation accuracy. With its powerful geometric processing ability and consideration of the calculation cost and accuracy, the MOC is more widely used in high-fidelity numerical simulation for neutron transport calculation.

BackgroundThe development of high-throughput reactors is of great significance for supporting the development of nuclear science and technology, improving the efficiency of nuclear energy utilization, meeting the needs of radioactive isotope production, and carrying out irradiation tests and performance tests of new nuclear fuels and structural materials in reactors. Due to the high power density of the core fuel and the large demand for thermal cooling, the nuclear-thermal coupling phenomenon in the high-throughput lead-bismuth reactor (HT-LBR) is more significant than that in conventional lead-bismuth reactor (LBR). When the design optimization of high flux LBR is carried out, it is necessary to carry out collaborative optimization of multiple core parameters, improve the neutron flux density, and meet the physical / thermal constraints such as core refueling period, fuel cladding temperature and coolant flow rate. Therefore, the design optimization of high flux lead-bismuth cooled reactor is a complex problem of multi-physics, multi-variable and multi-constraint coupling.PurposeThis study aims to improve the neutron flux level of LBR and solve the optimization design problem of HT-LBR.MethodsFirstly, a HT-LBR training database was constructed to contain different core design parameter combinations and corresponding objective function response values and constraint condition response values. Based on the reactor Monte Carlo code RMC and sub-channel Code Subchanflow, a Back-Propagation (BP) neural network prediction model was established as a proxy model for reactor physical calculation and analysis to achieve rapid prediction of core neutron flux density and effective multiplication factor using aforementioned training database. Secondly, an updated iterative optimization method based on BP neural network Dynamic Surrogate Model (DSM) was proposed to improve the optimization efficiency and global optimization ability, and search for the optimal HT-LBR core design parameter combination within the design range. Thirdly, based on the open-source machine learning platform TensorFlow, coupled with the reactor physical and thermal calculation and analysis program, an iterative optimization method based on BP neural network prediction model was proposed. Combined with the sensitivity analysis method of core design parameters based on Sobol index method, a HT-LBR optimization design platform was developed to cover five functional modules: training database generation, physical and thermal parameters calculation and analysis, BP neural network model construction, core parameters sensitivity analysis, and core parameters optimization analysis. Finally, a multi-functional ultra-high-throughput reactor was used as a prototype to establish a model to be optimized, collaborative optimization verification of multiple core parameters including core grid diameter ratio, fuel pellet diameter, active zone height, and radial reflector thickness, was conducted.ResultsVerification results show that the prediction accuracy errors for core neutron flux density and effective multiplication factor are maintained within 0.1%. The optimized neutron flux density is 15.41% higher than the original design. The influence degree of the four groups of core design variables on the maximum neutron flux is arranged in the order of reflector thickness < gate diameter ratio < active zone height < fuel pellet diameter. At the same time, the maximum temperature of the fuel pellet and the maximum temperature of the cladding are reduced by 23.57 ℃ and 8.20 ℃, respectively. The optimized core design scheme has a larger steady-state thermal safety margin.ConclusionsThe HT-LBR optimization design platform developed in this paper is effective and reliable.

BackgroundThe unique core and reflector structure of the inverse flux trap research reactors has raised a challenge to the traditional deterministic neutronics calculation methods applicable to power reactors. The deterministic codes DRAGON/DONJON with powerful geometric modeling capabilities have been maturely applied to power reactor types such as CANDU (Canadian Deuterium Uranium) and pressurized water reactors (PWR), but not been performed neutronics calculations and feasibility analysis on the China Advanced Research Reactor (CARR) with inverse flux trap design.PurposeThis study aims to verify the feasibility of the DRAGON/DONJON codes in CARR neutronics calculations and analysis.MethodsFirstly, when performing homogenization calculation using DRAGON/DONJON codes on various assemblies of CARR, the multi-assembly method was adopted to improve the surrounding impact. The OPTEX reflector optimization method was used to modify the homogenization constants of the reflector. Then, the commonly used multigroup cross-section libraries in DARGON were compared and screened, and SHEM-295 group-structure library was selected. Finally, the calculation results of DRAGON/DONJON codes were compared with the Monte Carlo references and the conventional three-step method, and analysis was conducted on the reasons for the significant deviation in the calculation results.ResultsThe results indicate that the deviation of parameters such as keff eigenvalue near critical operating condition, thermal neutron flux distribution in the active zone of the core and the middle position of the heavy water tank, and power distribution of standard fuel assemblies are relatively small whilst significant deviations in the calculation results appear at the junction of the core and heavy water tank, the vacuum boundary outside the pool, and the follower assemblies.ConclusionsThis study verifies that it is feasible to use the DRAGON/DONJON codes for CARR neutronics calculations and achieve a certain degree of accuracy, meeting the needs of experimental schemes design and rapid calculation and analysis of operating parameters.

BackgroundBy simulating and augmenting human intelligence, artificial intelligence can address challenges such as predicting keff and neutron flux of a reactor.PurposeThis study aims to apply the optimized extreme learning machine model to the prediction of reactor neutron flux and keff.MethodsFirst of all, a three-dimensional IAEA reactor was selected as the research object, with the steady-state neutron flux and keff as the predictive variables. and the core physics analysis program ADPRES was employed to generate data samples. Then, the basic neural network models for neutron flux and keff were constructed using Extreme Learning Machine (ELM), and the importance of feature values was evaluated using Random Forest (RF) to establish the optimal input feature subset for each model. Subsequently, the optimal number of neurons in the hidden layer was determined using a traversal method. Finally, the Whale Optimization Algorithm (WOA) was used to optimize the initial weights and thresholds for further improvement of the model accuracy.ResultsThe evaluation results show that after dimensionality reduction and optimization processing, the predictive accuracy of keff has improved by two orders of magnitude, and the prediction error of neutron flux has decreased by 87.24%, and the model training time is also reduced.ConclusionsThe model method constructed of this study has certain reference significance for solving reactor keff and neutron flux.

BackgroundAs a new decontamination approach, the self-embrittle decontamination technology is that the membrane will crack and peel off naturally to achieve the purpose of decontamination after the decontamination agent membrane adsorbs pollutants.PurposeThis study aims to prepare several surface radioactive decontaminants and analyze their performance of decontamination.MethodsThe composite detergent with the self-embrittle function was prepared by compounding nano-SiO2 modified with γ-mercaptopropyltriethoxysilane with random copolymer P(MMA-co-MAA-co-HFBMA) emulsion. The structure and properties of the composite emulsion were analyzed using infrared spectroscopy (IR), thermogravimetric analysis (TGA), X-ray diffraction (XRD), and transmission electron microscopy (TEM).ResultsObserved results show that the control of the self-embrittle morphology of the detergent can be realized by adjusting the ratio of MMA/MAA in the random copolymer. When MMA/MAA/HFBMA=7/5/0.1, the detergent has the good self-brittleness and forms uniformly sized and moderately sized embrittle flakes on the surface of the material, which facilitates subsequent collection and processing. And the decontamination rate of simulated radioactive fallout ash (mixed with elemental K) is greater than 95%.ConclusionsThe composite detergent with the self-embrittle function prepared in this study demonstrates good decontamination effects.

BackgroundNuclear graphite coatings on the surfaces of spherical fuel elements in high-temperature gas-cooled reactors (HTGRs) exhibit a high friction coefficient and low wear resistance. The reciprocating movement of the fuel balls leads to significant friction among the spherical fuel elements and between these elements, the graphite bed, and other components. This friction generates a considerable amount of graphite dust, which poses a risk to the proper functioning of nuclear reactors.PurposeThis study aims to address the issues of friction and wear experienced by nuclear graphite on the surface of spherical fuel elements in HTGR by utilizing surface modification technology to enhance the mechanical and tribological properties of NBG-18 nuclear graphite.MethodsFirstly, NBG-18 graphite, sourced from SGL Group-The Carbon Company, Germany, was cut into blocks with dimensions of 20 mm×20 mm×5 mm, and a polymer-like carbon (PLC) coating was applied to NBG-18 nuclear graphite using a high-energy ion beam deposition (IBD) process with preprocessing of cleaning, sample loading, vacuuming, transition layer deposition, functional layer deposition, and sampling, resulting in a total coating thickness of approximately 400 nm. Subsequently, nanoindentation tests were conducted to determine the hardness and elastic modulus of the sample with a maximum load of 5 mN, while a high-load scratch tester was used to assess the film substrate adhesion of the coating. Then, the coefficient of friction (COF) of NBG-18 with the PLC coatings was examined in a nitrogen environment using a TRB3 friction tester at room temperature with specific testing parameters set for normal loads and sliding frequencies to identify the optimal conditions. Various analyses, including ultra-depth field microscopy, white light interferometry, and Raman spectrometry, were employed to study the microstructure, wear rate, and friction interface characteristics of the coated samples. Finally, comparisons were made between the surface morphology, mechanical properties, and tribological properties of the NBG-18 nuclear graphite before and after coating deposition, highlighting the enhancements brought about by the PLC coating. Simultaneously, the lubrication and failure mechanisms of the PLC coatings were investigated.ResultsThe experimental results demonstrate a significant increase in the hardness of NBG-18 nuclear graphite, from 0.44 GPa to 4.16 GPa, marking an 845% improvement post-PLC coating deposition. The elastic modulus rose from 9.00 GPa to 27.21 GPa, reflecting a 202% enhancement. The optimal conditions of a normal load of 2 N and a sliding frequency of 5 Hz led to a decrease in the friction coefficient from 0.335 7 to 0.006 5, a reduction of 98%. Moreover, the wear rate dropped from 3.71×10-3 mm3·(N·m)-1 to 1.81×10-6 mm3·(N·m)-1, representing a three-order-of-magnitude decrease. The mechanisms behind these improvements involve friction-induced graphitization of the PLC coatings and high hydrogen surface passivation, which play crucial roles in achieving ultra-smooth nuclear graphite. These findings provide valuable theoretical support for the advancement of surface-modified lubrication technologies for nuclear graphite.ConclusionsThe deposition of PLC coatings on the surface of NBG-18 nuclear graphite significantly enhances its friction and mechanical properties. These findings of this study provide valuable theoretical support for the advancement of surface-modified lubrication technologies for nuclear graphite.