In the Shanghai High Repetition rate XFEL (X-ray free electron laser) and Extreme Light (SHINE) facility, the vertical linear polarization laser is generated by using 40 planar superconducting undulators (SCUs) with a period length of 16 mm, length of 4 m, and a gap of 4 mm. At present, the Hall probes are the most reliable method for measuring the undulator magnetic field whilst the positioning accuracy of the sensitive center of the Hall probe is one of the main factors affecting the accuracy of magnetic field measurement.

This study aims to calibrate the position of the Hall probes' sensitive region for magnetic field measurements of SCU with high-precision.

The experimental platform for magnetic field point measurement of SCUs was introduced in details, a sledge with three mounted Hall probes and a retro-reflector were applied for magnetic field measurement. By flipping the sledge, the lateral distance between the sensitive centers of the Hall probe and each other were obtained, so did the lateral distance between the sensitive centers of the Hall probe and the apex of the pyramid prism. Therefore, the position of the Hall probes' sensitive region and center of the retro-reflector were calibrated.

The positional calibration of the Hall probes has an accuracy higher than ±10 μm, which meets the requirements for magnetic field measurement.

Lost radioactive sources needs to be quickly retrieved, positioning of radioactive source in complex environment is the key to find the lost radioactive source. [Propose] This study aims to develope a novel approach for the rapid positioning of orphan sources using a NaI(Tl) array detection device.

First of all, by leveraging the shadow effect between array detectors, a response curve between gamma-ray incidence angles and counts was obtained through the use of Monte Carlo simulation software. Then, the support vector machine (SVM) method was employed to establish a predictive mathematical model for the counting rate of array detectors as a function of gamma-ray incidence angle, utilizing. Finally, a radioactive source localization physical experiment platform was constructed, and a series of incidence angle response experiments were conducted for the validation of this approach applied to radioactive source localization under varying conditions.

Eexperimental results demonstrate that, through the use of the SVM regression prediction model, the maximum average deviation of the angle is 9.21° whilst the minimum is 1.77° for the angle prediction of an orphan ¹³⁷Cs point source.

This method can achieve rapid and accurate localization of an orphan radioactive source.

The solar X-ray detector (SXD) is the main scientific instrument onboard the Macau Science Satellite-1B (MSS-1B). It consists of two parts—a soft X-ray detection unit and a hard X-ray detection unit—with a dual-channel design comprising a silicon drift detector (SDD) and a cadmium zinc telluride detector (CZT). Both the precise energy spectrum and intensity of the Sun can be simultaneously obtained by the SXD, hence to quantify the level of solar flares and study their evolutionary process.

This study aims to calibrate the detection efficiencies of the SDD and CZT, so as to invert the observed data for obtaining real solar X-ray data.

The Monte Carlo code MCNP5 (Monte Carlo N-Particle 5) was employed to calculate the SDD and CZT efficiencies by simulation. Soft and hard X-ray detection efficiency calibration experiments were performed using a monochromatic X-ray ground calibration facility via relative measurement methods.

The experimental results for the SDD-1 and CZT-1 efficiency calibration agree well with the predicted results of the simulation. In particular, the maximum relative error between the experimental and simulated efficiencies of SDD-1 dose not exceed 3.59%@16 keV, and the maximum relative error between the experimental and simulated efficiencies of CZT-1 dose not exceed 9.54%@120 keV. The relative expanded uncertainty of the monochromatic X-ray flow intensity measurement is 3.8% (k=2), and the uncertainty of the simulation results for the SXD is 0.12%.

This study provides not only data support for SXD onboard MSS-1B satellite, but also valuable guidance for the calibration of other astronomical satellites' detectors in the future.

Nuclear science and technology are closely related to the lives of people. However, nuclear radiation may harm the health of the general public; hence, nuclear radiation monitoring must be strengthened. A wired nuclear radiation monitoring system has the characteristics of complex wiring, a long construction period, high cost, poor mobility, and more difficult troubleshooting.

This study aims to address the demand for convenient measurement and monitoring of gamma radiation fields.

Based on the LoRa wireless communication technology, a γ radiation monitoring system using silicon photomultiplier (SiPM) tube-scintillator detector was designed. The main functions of the system included data collection from the detector, data processing and transmission using a STM32 single-core processor. The collected data packaging and transmission were processed in STM32 microcontroller using the LoRa wireless communication module. Considering the possible channel congestion in the monitoring system and the frequent data transmission, a dynamic optimal path communication algorithm was designed to find the optimal reconnection path and realize the priority allocation of data transmission.

The test results show that the data transmission stability of γ radiation monitoring system based on LoRa is higher than 99.57%, and has the advantages of flexible networking, a further distance of transmission, low cost, and substantial expansion, hence has a broad reference prospect.

Accelerator-driven subcritical systems (ADS) are among the most promising options for next-generation nuclear power systems. Various radionuclides are produced during the process of protons bombarding the target in the ADS, and the cross-sections of various long-lived radionuclides have not been accurately measured. These long-lived nuclides are related to ADS radioactive waste treatment, therefore, accurate evaluation of long-lived radionuclides generated in ADS spallation targets is a key topic in applied research.

This study aims to determinate production cross-sections of ^natPb(p,x)²⁰⁷Bi and ^natPb(p,x)¹⁹⁴Hg reactions according to measurement data, and compare them with existing experimental and theoretical results.

The proton activation method was employed to effectively estimate the cross section of long-lived nuclides produced by the interaction between protons and spallation target materials. Four proton-irradiated natural lead samples were irradiated with protons at energies of 40 MeV, 70 MeV, 100 MeV, and 400 MeV for 90 min, 75 min, 40 min, and 25 min, respectively. After cooling for approximately 20 a, the samples were measured using an ultralow background gamma spectrometer GeTHU in the China Jinping Underground Laboratory (CJPL), and the GeTHU detection efficiency was calculated using the Simulation and Analysis for Germanium Experiments (SAGE) simulation framework. Combined with the irradiation parameters of the samples, the total production cross-sections of the two nuclides were calculated using a cross-section calculation formula. Experimental results are evaluated and compared with those of existing studies.

The production cross-sections of ^natPb(p,x)²⁰⁷Bi reaction in the natural lead samples irradiated by protons with four different energies (40 MeV, 70 MeV, 100 MeV and 400 MeV) are calculated as (40.70±3.59) mb, (19.31±1.43) mb, (13.15±0.96) mb, and (2.90±0.22) mb, respectively. The calculated production cross-section of ^natPb(p,x)¹⁹⁴Hg reaction in the natural lead sample irradiated by protons with an energy of 400 MeV is (57.07±7.83) mb. Based on the same samples, the measurement results of cross-section remain consistent within the error range. The cross-sections of ^natPb(p, x)²⁰⁷Bi are closer to TENDL's evaluated cross-sections. The cross-section of ^natPb(p, x)¹⁹⁴Hg is consistent with the theoretical expectation of INCL++/ABLA. In addition, different sources that contribute to the total uncertainty of both reactions are explained in detail.

The production cross-sections of ^natPb(p,x)²⁰⁷Bi and ^natPb(p,x)¹⁹⁴Hg reactions measured herein were calculated independently and showed good agreement with existing results. These results demonstrate that GeTHU is capable of measuring low-activity and long-lived radionuclides in the CJPL. Finally, the results of this study also provide the latest experimental evidence for the evaluation of radioactive waste in ADS.

Low-level radioactive wastewater (LLW) is generated during the operation of nuclear facilities. Usually, LLW is discharged directly into the ocean in the form of liquid effluent after purification under the discharge management limits. However, discharging LLW into inland water bodies is difficult for inland nuclear facilities because of the poor dispersion and the lack of public acceptance. Thus, LLW disposal has become one of the challenges limiting the development of inland nuclear facilities. Liquid-to-gas discharge, which is based on high-pressure spray evaporation technology, is an alternative solution for LLW disposal for inland nuclear facilities.

This study aims to develop and validate a model for simulating spray flow and evaporation to assess the design feasibility.

A numerical method coupling a two-phase flow model, mass transfer model, and heat transfer model were established to describe droplet evaporation during the flow process. To validate this numerical method, a sample high-pressure spray evaporation system was developed which included three key subsystems: carrier gas generation, source term generation, and measurement systems. Finally, considering evaporation and deposition factors, three experimental cases were designed for experimental comparison of the droplet diameter, number, and deposition rate among these cases.

The comparison results show that the numerical method is highly consistent with the experimental results, with a maximum uncertainty of 15%.

The numerical model developed in this study can be used for the technological design of liquid-to-gas LLW discharge based on high-pressure spray evaporation technology.

Nuclear high-temperature resistant ceramic materials have been widely used in nuclear energy, military, and aerospace fields in recent years owing to their excellent thermal insulation and high-temperature oxidation resistance.

This study aims to investigate the mass and heat transfer process between plasma fluid and flying particles in supersonic plasma spraying during the preparation of yttrium-stabilized zirconia thermal barrier coatings, so as to reveal the process parameters of flying particles.

Firstly, the computational fluid dynamics (CFD) approach was employed to simulate the interaction between flying particles in the plasma spraying process. Then, a three-dimensional mathematical model of the plasma spraying flow field was established, and the jet characteristics of different spraying parameters in the de Laval nozzle and the melting and stress state of flying particles were analyzed by using this model. Furthermore, the online monitoring device Spray Watch 2i (Osier, Finland) was used to compare the online measurement of the velocity and temperature of flying particles obtained with the simulation results.

The comparison results show that relative errors are within 15%, verifying the simulation results effectively by experimental results. When the spraying power is reduced from 71 kW to 36 kW (i.e., reduced by 49.2%), the maximum velocity of the plasma jet is reduced by 8.5%, and the maximum temperature is reduced by 22.2%.

A correlation between plasma spraying parameters, jet characteristics, and melting of flying particles is revealed in this study, providing theoretical guidance for the precise control of high-performance thermal insulation coating structures required for accident resistant fuel cladding in nuclear reactions.

Due to the complex structure of the ventilation ducts in nuclear facilities, the concentration distribution of radionuclides such as aerosols in the ducts is uneven. The inhomogeneity of aerosol distribution brings great challenges to the sampling representativeness of radiation monitoring. In chemical processes, static stirring devices are commonly used to enhance the homogeneity of the product mix. However, this device has not been applied in the nuclear power field.

Accordingly, this study aims to improve the mixing uniformity of aerosols in air supply pipelines by using static stirring devices, so as to provide a reference for the representativeness of radiation monitoring sampling.

The stirring effects of three different stirring devices were investigated through numerical simulations. The RNG k-ε model was used to simulate the gas phase flow field, and the discrete phase model (DPM) was employed in ANSYS CFX software to simulate the behavior of aerosol particles. Selection of the particle size of aerosols followed the recommendations in the sampling representative standards, with the specific size of 10 μm. The other boundary conditions in the simulation were based on the actual operating conditions of a nuclear power plant. As a result, the effects of different stirring devices on the flow field and aerosol concentration distribution were obtained.

The static stirring device can form strong swirls, thereby improving the uniformity of aerosol distribution. Increasing the twist angle of the blades and the proportion area of the inner blades strengthened the generated vortex field, further affecting the diffusion of aerosols. The static stirring device with an increased inner blade area exhibited a moderate swirl intensity, and better stirring effect than those of the other two structures. The coefficient of variation of the aerosol concentration decreased by 30.60%.

Installing a static stirring device in an air supply duct is a feasible method to improve the mixing uniformity of aerosols. Owing to the complex structure of ventilation ducts in nuclear facilities, the concentration distribution of radionuclides, such as aerosols, in the ducts is uneven. This inhomogeneous aerosol distribution poses significant challenges to the sampling representativeness of radiation monitoring. In chemical processes, static stirring devices are commonly used to enhance the homogeneity of the product mix. However, these devices have not yet been applied in the field of nuclear power.

The calculation error of the stacked pulse amplitude generated by traditional pulse shaping methods leads to distortion in the X-ray fluorescence spectrum; thus, it is difficult to accurately analyze the spectrum measured in a high-stacking rate background.

This study aims to propose a transformer model based on deep learning for the pulse amplitude estimation of radiation measurements using high-performance silicon drift detectors.

Firstly, multi-head attention was applied to the transformer model, and an encoder-decoder structure with embedded positional encoding was employed to estimate the amplitude of stacked pulses. Then, a predefined mathematical model was used to simulate the pulse signal output by the detector for model training, and Gaussian noise corresponding to thermal noise and shot noise was added to the signal to simulate real nuclear pulses. Finally, experimental verifications were carried out on powdered iron ore samples and powdered rock samples, and relative error, corresponding to the accuracy of pulse amplitude estimation, was used as a model performance evaluation indicator.

Experimental verification results show that the average relative error obtained for eight offline pulse sequences of powdered iron ore samples and powdered rock samples is 0.89%, which means that the model can accurately estimate the amplitude of stacked pulses.

The annular fuel has a closely arranged structure, and the coolant flow at both the gap between the stringers and the near wall surface is small, which is unfavorable to the coolant mixing between the subchannels and the uniform circumferential temperature distribution.

This study aims to explore the effect of the ratio of gate spacing to gate diameter on the distribution of temperature along the circumference direction.

Based on the software code ANSYS FLUENT, a computational fluid dynamics (CFD) analysis model for annular fuel assemblies was established. Then, the calculations in hydromechanics and the numerical simulation using operating parameters of typical pressurized water reactor (PWR) were performed to analyze the coolant flow and heat transfer characteristic when the annular fuels in square or hexagonal arrangement under different grid ratios. The circumferential non-uniformity of annular fuel outer temperature distribution was investigated under circumstances of various pitch-to-diameter ratio.

Computational results show that an appropriate increase of grid ratio is beneficial to the uniform circumferential temperature distribution of stringers. The appropriate grid ratio of square component is between 1.07 and 1.09, and the non-uniformity of circumferential temperature distribution of triangle component is slightly lower than that of square component. Therefore, the appropriate grid ratio is between 1.06 and 1.09.

The temperature distribution at the bar gap is improved most obviously by increasing the grid ratio and the improvement in the near surface takes the second. The results of this study provide a reference for the subsequent optimization design of the grid ratio of annular fuel.

During reactor operation, zirconium (Zr) alloy cladding is continuously oxidized as it gets in contact with fuel, and combines with the fuel to form a firm chemical interaction layer. This affects the thermal conductivity of the fuel gap, the mechanical properties of the cladding, and the mechanical interaction of the fuel cladding.

This work aims to obtain relevant analytical data on the chemical interaction layer between the irradiation Zr-alloy cladding and uranium oxide (UO₂) pellets in a pressurized water reactor (PWR).

First of all, the D13 intact fuel rod with a burnup of 45 GWd·tU^-1 for PWRs in a nuclear power plant was chosen as UO₂ pellets with a pellet enrichment of 4.45 wt%, and M5 Zr-alloy was used as the cladding materials. Then, a series of operations (cutting, pellet separation, inlaying, secondary cutting, inlaying, and polishing of cladding tubes) was conducted in the hot cell. The polished sample was transferred to the lead chamber and the UO₂ fuel pellet was removed using 4 mol?L^-1 of nitric acid solution. The cladding tube was separated from the chemical interaction layer, and a low-speed cutter was used to cut the cladding tube to a width of 2~3 mm in the glove box. Finally, the morphology and structure of the chemical interaction layer were analyzed using metallographic microscopy, scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), and hot cell Raman spectroscopy.

The analysis results show that the gap between the fuel pellet and cladding is approximately 14~19 μm after the fuel running to a burnup of 45 GWd·tU^-1. In the chemical interaction layer, the time sequence of mechanical contact at different locations is different, resulting in discontinuity of the interaction layer. The SEM-EDS results show that the chemical interaction layer is in the shape of "worms" composed of U, Zr, and O to form a mixed phase (U,Zr)O_x compound.

The results of this study indicate that the chemical interaction layer is mainly composed of tetragonal zirconia (t-ZrO₂) and monoclinic zirconia (m-ZrO₂).

Silicon carbide (SiC) composite claddings are candidate solutions for accident resistant fuel claddings in light water reactors.

This study aims to estimate the failure probability of a double-layer structured SiC cladding under a loss-of-coolant accident (LOCA).

Based on a failure probability calculation method for SiC composite cladding, a quasi-steady state method was used to simulate and calculate the SiC composite cladding failure probability under transient conditions. Sensitivity analysis of the two characteristic parameters of Weibull distribution was performed by analyzing the proportion of various stresses under accident conditions. The effects of different burn-up conditions on the failure probability were investigated, and the failure probability of the cladding under different layer thickness ratios was simulated.

Simulation results indicate that the transient failure probability of SiC composite claddings is significantly affected by changes in the proportion of the composite layer and Weibull parameter, as well as the occurrence of LOCAs under different burn-up conditions. This study makes contribution to the development and design of accident resistant fuel claddings, providing reference for further investigations on the failure probability of SiC composite claddings.

Because of the excellent properties of lead-based materials as reactor coolants, lead-based fast reactors have become a key type of fourth-generation advanced nuclear energy systems. A small passive long-life Lead–bismuth -cooled fast Reactor (SPALLER) is designed by the University of South China for profound research.

This study aims to improve the inherent safety and cost-effectiveness of lead–bismuth-cooled fast reactors, and determine the maximum core power of this kind of reactor.

Firstly, the SPALLER was taken as research object, and five steady-state limitations and two accident limitations were proposed to meet the transportation size, material durability, and long-term operational stability of the reactor core and ensure safety under accident conditions. Then, a neutronics maximum power calculation platform was built through Latin hypercube sampling and a Kriging proxy model whilst the steady-state limitations were considered as multi-objective optimization problems with complex multidimensional nonlinear constraints. Meanwhile, the neutronics maximum power and natural circulation power of SPALLER-100 at different core heights were calculated by taking the natural circulation ability of SPALLER-100 into account. Finally, a design scheme was obtained to meet the requirements of neutronic and thermal-hydraulic assessments of this reactor while producing maximum power. Consequently, during the full life-cycle of SPALLER-100, a safety analysis of three typical accident scenarios (loss of heat sink, transient over power, and coolant inlet temperature undercooling) was performed using a quasi-static reactivity balance approach.

The results show that the maximum core power can be increased from 100 MW to 120 MW, and the neutronics maximum power calculation platform has high accuracy with safe and economical maximum power scheme.

This study can provide reference for other types of natural circulation reactors to maximize power output.

The medical isotope production aqueous reactor (MIPR) has advantages of small size, low power, and high inherent safety, hence is one of better candidate reactor types for the production of ⁹⁹Mo and other medical isotopes.

This study aims at the effects of extraction methods and reprocessing capacity on the production efficiency of ⁹⁹Mo based on low-enriched uranium MIPR designed with neutronic optimization.

First, the calculation method was verified according to existing experimental data, and the neutronic optimization of the MIPR was performed for core design by using SCALE6.1 code and ENDF/B-VII database with 238 groups. Then, the ⁹⁹Mo production efficiency under different extraction methods as well as processing capacities was investigated based on the optimized core structure. The range of achievable critical uranium concentration and enrichment was determined.

There is a minimum critical mass under different enrichment, and with the increase of ²³⁵U enrichment, the uranium concentration at the minimum critical uranium mass decreases. The effective multiplication factor decreases linearly with an increase in nitric acid concentration, and the corresponding nitric acid reactivity coefficient is approximately -1.400×10^-2 L·mol^-1. With an increase in uranium concentration, the void and temperature reactivity coefficients decrease, and the corresponding reactivity coefficients are approximately (-100~-250)×10^-3 ℃^-1 and (-18~-30)×10^-5 ℃^-1.

The production efficiency of the MIPR production of the ⁹⁹Mo online extraction method is slightly greater than that of the offline batch processing method with an increment of about 16% under a five-day production cycle. The reprocessing capability has a greater impact on the production efficiency of the online extraction method. If the reprocessing rate is increased five times, the increment of production efficiency is about 113% under a five-day production cycle.

The neutronics and thermal-hydraulic characteristics of lead-bismuth cooled reactors are significantly affected by the geometric configuration of fuel assembly and lattice parameters. Reactor cores loaded with different geometry type fuel assemblies have different critical dimensions and fuel loadings under the same refueling cycle and thermal-hydraulic constraints.

This study aims to analyze these key factors and select a geometric structure of fuel assembly that is conducive to miniaturization and lightweight of lead-bismuth reactor.

First of all, the core model of a 4 MWt small lead-bismuth reactor was established, and simulation analysis of reactor physical characteristics was conducted using the RMC Monte Carlo program developed independently by the Reactor Engineering Calculation and Analysis Laboratory of Tsinghua University and the nuclear database ADS-2.0 released by the International Atomic Energy Agency (IAEA) in 2008. Then, three fuel assembly schemes of rod bundle type, annular type and honeycomb coal type were selected for comparison and analysis in term of fuel consumption characteristics, reactivity coefficient and steady-state thermal parameters under the same core size, fuel loading, coolant flow area, cladding and air gap volume, 10-year refueling cycle and basically consistent steady-state thermal safety margin.

The results show that compared with the rod bundle fuel assembly and the annular fuel assembly, the honeycomb coal fuel assembly has good steady-state thermal characteristics and hard neutron spectrum. The core of the honeycomb coal fuel assembly can realize smaller core size and fuel loading, and has obvious expansion negative feedback, and can effectively flatten the power distribution and reduce the core pressure drop. It is a fuel assembly solution that is conducive to the miniaturization and light weight of lead-bismuth reactors.