The volume of aviation gamma spectrum data is immense. If only a central processing unit (CPU) is used for data post-processing, it would be constrained by computational efficiency.

This study aims to propose a CUDA-based graphics processing unit (GPU) parallel solution that optimally accelerates the denoising of airborne gamma-ray spectral data using wavelet transformation.

First, the impact of different block sizes on computational time was tested to determine the optimal block size for processing airborne gamma-ray spectral data. Subsequently, a GPU, instead of a CPU, was used to calculate the acceleration ratio for handling airborne gamma-ray spectral data of different volumes, and wavelet basis functions were used for those with the same data volume. Finally, by introducing white noise to the experimentally measured airborne gamma-ray spectral data, the signal-to-noise ratio of denoised data was calculated to optimize the threshold denoising method suitable for parallel acceleration of the GPU.

The optimal two-dimensional block sizes for denoising airborne gamma-ray spectral data are 64×64 and 128×128. Among the wavelet basis functions, those that achieved a total time acceleration ratio exceeding 100 compared to CPU processing account for 80%, while those that reached an acceleration ratio exceeding 90 constitute 91%. The coif5 function achieves an acceleration ratio of 353 times whilst the acceleration ratio of the threshold denoising function approaches 570.

All wavelet functions exhibit insufficient denoising effects at low signal-to-noise ratios and excessive denoising effects at high signal-to-noise ratios. Significant denoising can be achieved using hard thresholding of coif5, soft thresholding of coif1, and improved thresholding of bior3.7.

Boron neutron capture therapy (BNCT) is a promising tumor therapy method that irradiates ¹⁰B attached tumors with thermal or epithermal neutrons. Specifically, ⁷Li(p,n)⁷Be is one of the main methods for producing thermal or epithermal neutrons. The Li target is a critical component for BNCT device.

This study aims to design a composite material Li target with a semi-tirelike surface structure that rotates around a central axis.

Based on a 2.5 MeV, 20 mA proton beam, the TOPAS Monte Carlo software was used to simulate the influence of curved lithium target with different radial axial ratios on the angular distribution, energy spectrum, and flux of neutrons. The steady-state temperature distribution of the neutron target was simulated by using ANSYS software to investigate the impact of radial axial ratios on heat dissipation.

The simulated results show that the semi-tirelike curved surface enhances the concentration of the output neutron beam. When the radial axis ratio is 1.5, the number of neutrons with an exit angle of 0°~45° is 2.59 times that of the same plane target whereas the number of neutrons with an exit angle of greater than 90° is only 0.29 times that of a planar target in the same situation. The heat dissipation performance is improved by the fold channel design and curved surface structure. The maximum temperature of the lithium layer is only 100 ℃ for a 50-kW proton beam incident, which satisfies the heat dissipation requirement.

The innovative curved neutron target of this study significantly improves the forward performance of the outgoing neutron and heat dissipation performance when compared with the conventional plane target. These types of advancements will lead to good prospects in the field of BNCT.

Within GEN-IV reactors, nuclear graphite plays a crucial role as both a moderator and reflector in an environment characterized by high temperatures and intense fast neutron irradiation. The exposure to fast neutron irradiation induces the formation of numerous Frankel defects in the nuclear graphite. These defects undergo processes of annihilation and diffusion, ultimately giving rise to larger defect clusters. This transformation in the microstructure of nuclear graphite directly impacts its macroscopic properties, necessitating a thorough investigation.

This study aims to comprehensively explore the evolution of defects in nuclear graphite under conditions of high-temperature irradiation which is essential for advancing reactor safety.

Firstly, the 30 MeV ¹⁰⁷Ag⁵⁺ ion source was employe to irradiate IG-110 nuclear graphite at 420 ℃, simulating the defect evolution behavior during fast neutron irradiation of nuclear graphite. Then, the energy loss, defect distribution, and ion implantation profiles of 30 MeV ⁵⁸Ni⁵⁺ and ¹⁰⁷Ag⁵⁺ ion beams bombarding standard nuclear graphite ICRU-906 (density of 2.26 g?cm^-3, displacement energy of 28 eV) were calculated using the full cascade damage model in the SRIM (Stopping and Range of Ions in Matter) software. The cross-sectional structure of IG-110 nuclear graphite was characterized using micro-Raman spectroscopy. Finally, the relationship between the Raman spectroscopic features at various depths of IG-110 nuclear graphite and the irradiation damage dose was compared to investigate the evolution of IG-110 nuclear graphite microstructure with increasing irradiation damage dose (Displacements Per Atom, DPA).

With the increase in particle fluence, the characteristic parameters of the Raman spectra of nuclear graphite, including the I_D/I_G ratio (the ratio of the D peak height to the G peak height), the Full Width at Half Maximum of the G peak (FWHM(G)), and the shift of the G peak, all show significant increments. Compared to samples irradiated with ⁵⁸Ni⁵⁺ at the same irradiation damage dose, the graphite Raman spectra irradiated with ¹⁰⁷Ag⁵⁺ demonstrate higher I_D/I_G ratios and FWHM(G). At the same FWHM(G) level, the I_D/I_G ratio of the graphite Raman spectra irradiated with ¹⁰⁷Ag⁵⁺ is greater than that of the samples irradiated with ⁵⁸Ni⁵⁺.

The results of this study suggest that irradiation with heavier ions induces a higher rate of defect accumulation in nuclear graphite, leading to a more rapid reduction in graphite grain size and promoting the progression towards nanocrystallization.

In the process of accelerated development of nuclear medicine department in recent years, the construction of decay tanks and the storage time of radioactive wastewater containing ¹³¹I have become issues of great concern for environmental regulatory agencies and hospitals. Regulations and standards, such as "Basic Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources" (GB 18871-2002), "Radiation Protection and Safety Requirements for Nuclear Medicine" (HJ 1188-2021), and "Reply to Consultation on Several Clauses of the Nuclear Medicine Standard", have stipulated the compliant discharge methods for radioactive wastewater containing ¹³¹I from hospitals.

This study aims to demonstrate the methods for calculating the optimal volume of the decay tank and the minimum storage time so as to minimize the unnecessary construction costs and land use while ensuring compliance with regulations.

Firstly, according to theoretical induction, a formula for the upper bound of total activity of ¹³¹I was presented for a full decay tank in a hospital. Then, the RJ (Radioactivity Judgement) equation group was put forward to address the calculation of minimum decay time and volume of the decay tank. Finally, the actual monitoring data from four hospitals were applied to verification the recommendation of this study.

Verification results demonstrate that when the temporary storage period for radioactive wastewater containing ¹³¹I reaches the minimum time calculated by the RJ equations, the total discharge activity of ¹³¹I complies with the national environmental protection standards. Among the three compliant discharge methods for radioactive wastewater containing ¹³¹I in the decay tank, the method specified in GB 18871-2002 is advantageous for the operation of the nuclear medicine department in the hospital.

With the methodologies detailed in this paper, it is possible to significantly reduce the required volume of the decay tank and the minimum storage time. These findings provide clear and specific guidance for the construction of decay tanks in nuclear medicine departments and for the supervision and inspection conducted by regulatory authorities.

When an X-ray tube operates, a large number of backscattered electrons are generated inside the tube. Under the influence of the electric field inside the X-ray tube, these backscattered electrons are pulled back to the anode target by the electric field, generating additional X-rays, which ultimately have a negative impact on the quality of the emitted beam.

This study aims at the distribution rules of backscattered electrons in transmission X-ray tubes, and the changes in backscattered electrons caused by target materials and tube voltages as well as their impact on the outgoing beam quality.

Firstly, Geant4 was used to establish the physical model of the transmission X-ray tube, and the theoretical calculations were compared with experimental data to verify the accuracy of the calculation program. Then, C, Si, Cu, Ag, and W were selected as anode target materials, and the influences of different electric field strengths and these target materials on the distribution of backscattered electrons in the X-ray tube were analyzed in details. Finally, the influence of the backscattered electrons on the outgoing beam quality under the effect of tube voltage was investigated.

The full width at half maximum, photon yield, and characteristic peak yield of the outgoing beam of the X-ray tube increase when considering the electric field, indicating that the backscattered electrons have effect on the outgoing spectrum quality. Simultanously, the backscattered electrons pulled back by the electric field in the first and second generations predominantly affect the outgoing beam quality. A change in the tube voltage alters the extent of influence of the backscattered electrons on the outgoing beam quality, and a larger tube voltage leads to a more obvious influence of the backscattered electrons on the outgoing beam quality of the X-ray tube.

The distribution of inside backscattered electron is closely related to the characteristics of the X-ray tube, including the tube voltage and target material. The results of this study may provide a reference for numerical simulation calculations of miniature transmission X-ray tubes.

Laser simulation technology is widely used in the research of transient ionizing radiation effects in semiconductor devices. Fully dielectrically isolated silicon-on-insulator (SOI) devices exhibit different responses to dose-rate gamma irradiation compared to bulk-Si devices.

This study aims to examine the photocurrents of both Si-based and SOI NMOS transistors, and investigate the performance of a SOI MCU with varying dose rates.

An irradiation experiment was conducted on three types of transistors by using a 1 064 nm/12 ns laser device, and the photocurrent was tested under various laser energies. A pulsed γ-ray source was employed to perform the transient γ dose rate radiation test on an SOI-integrated circuit. The function, electrical parameters, and flipflop chain status of the SOI-integrated circuit under different dose rates were measured. Based on theoretical model for the generation of photocurrent in SOI transistor, the dose-rate threshold for logic flipping and corresponding critical charge were estimated based on theoretical model for the generation of photocurrent in SOI transistor.

The results indicate that the peak photocurrent of the SOI transistor is approximately 20 times lower than that of the bulk silicon transistor with the same feature size under identical irradiation conditions. This reduction is attributed to the decreased charge collection sensitive area of the SOI transistor. Within a dose rate range from 1.0×10⁹ rad(Si)·s^-1 to 4.2×10¹¹ rad(Si)·s^-1, the SOI-integrated circuit exhibites no latch-up effect. However, irradiation-induced upsets are observed in the SOI-integrated circuit.

These upsets caused by transient radiation effects manifest as transient functional interruptions, variations in operating current and voltage, and erroneous flip-flop statuses. These irradiation-induced upsets in the SOI-integrated circuit are likely attributable to, among other factors, transistor upsets and circuit-level voltage fluctuations on printed circuit board.

Neutrons/Gamma (n/γ) discrimination is critical for neutron detection in the presence of γ radiation and traditional pulse shape discrimination methods suffer from unstable discrimination accuracy.

This study aims to implement a machine-learning method that combines the kernel principal component analysis (KPCA), marine predator algorithm (MPA), and extreme learning machine (ELM) is proposed to improve the n/γ discrimination efficiency and accuracy against the traditional pulse shape discrimination methods.

The KPCA was used to reduce the dimensionality of the pulse signal characteristics of neutrons and gamma rays. Owing to the randomness in the ELM input layer weight and hidden layer bias, the MPA was employed to optimize the foregoing factors to improve the n/γ discrimination accuracy of the ELM. Finally, experimental data of Pu-C neutron source using BC-501A liquid scintillator detector were applied to effectiveness comparison of training and test with and without KPCA dimensionality reduction.

Comparison results reveal that the average discrimination accuracy of the KPCA-MPA-ELM is as high as 99.07%, which is 12.19%, 2.52%, and 1.56% higher than those of the ELM, MPA-ELM, and KPCA-ELM models, respectively. Compared with the charge comparison method and pulse gradient analysis method, the accuracy is improved by 1.80% and 5.91%, respectively.

The proposed model has a simple structure, exhibits good stability, hence be applied to handling high-dimensional data with good discrimination and generalization ability.

Solid-polymer electrolyte (SPE) water electrolysis has been widely used to concentrate low-level environmental tritium for measurements. It has the advantages of electrolyte elimination, infinite reduction ratio, high current, safety, and environmental friendliness. However, the use of high-activity tritium samples for the calibration of its tritium enrichment factor E_t causes a tritium memory effect in the system.

This study aims to reduce the tritium contamination caused by high-activity tritium samples during calibration.

First, the deuterium method (DM) was used to calibrate the prepared SPE-based tritium enrichment system. The strong correlation between the tritium-to-protium (β) and deuterium-to-protium separation factors (α) during the electrolysis was determined to obtain the electrolytic cell constant, k = αβ-1/βα-1. Then, based on the deuterium enrichment factor E_d, the tritium enrichment factor (E_d)^k of the SPE-based tritium enrichment system was determined. Finally, the influence factors of the k-value were analyzed, and the tritium enrichment factors derived via low-level tritium measurements using the DM and spike-proxy method (SPM) were compared.

The value of k is approximately 1.088 9, with a slight variation of 0.89% among cells. The k-value depends on the electrode material, but it is almost independent of the initial volume of water.

The k-value can be used as a constant for calibration of the tritium enrichment factor for the same types of electrolyzer with high accuracy and precision.

The accurate determination of oil saturation is an essential part of reservoir petrophysical evaluation. Both C/H and C/O logging methods are primarily used to determine oil saturation in low-salinity reservoirs.

The study aims to compare the logging response effects of the C/H and C/O methods.

Firstly, a spherical model and the model set by the C/O logging instrument RPM (Halliburton Company) were applied to Monte Carlo simulations, and the window counts of the atoms formed under different geological parameters and pore structures were calculated by MCNP5. Then, the relationship between C/O and C/H with saturation and the sensitivity to saturation were analyzed, and the response effects of C/O and C/H were compared. Finally, the mechanisms underlying the different response effects of C/H under different models were clarified.

The analysis results show that C/H and C/O exhibit the same relationship with porosity and saturation in casing wells, and both can be used to calculate saturation. However, compared to C/O, C/H is more affected by the borehole environment and is less sensitive to saturation; therefore, it is not suitable for saturation calculations in complex boreholes.

Exposure to high-intensity X-ray irradiation induces numerous defects in the thin films of optical components applied to X-ray free-electron laser (XFEL) facilities. These defects result in structural damage to the material and degradation of macroscopic properties, thereby affecting its service life and significantly compromising the reliability and stability of XFEL facilities.

This paper aims to design and implement an automated irradiation simulation software based on Python called automatic irradiation simulation based on LAMMPS (AISL) to support the simulation research of XFEL radiation damage to materials using molecular dynamics methods.

AISL was designed to facilitate high-throughput automated studies on XFEL radiation damage by exploring the process of radiation-induced defects and promoting the accumulation of data for radiation-resistant materials. This software package established an automated workflow pipeline for simulation tasks, encompassing aspects such as batch submission management, scheduling of simulation tasks, reliable storage of computational data, and post-processing of thermodynamic information. Additionally, the metadata was automatically recorded in a radiation damage simulation database based on MongoDB, so were the workflow information, and calculation result files of the simulation. Finally, the utilization of AISL in simulating radiation damage in metal thin films of XFEL optical components was demonstrated to verify the validation of AISL.

The demonstration results indicate that AISL is an effective, user-friendly software for conducting high-throughput automated irradiation simulation studies. It significantly enhances the efficiency of material irradiation damage simulation calculations using LAMMPS.

Lead-free double perovskite Cs₂AgBiBr₆ has garnered significant attention in the field of nuclear radiation detection as an environmentally friendly material. Experimental observations have revealed that doping Cs₂AgBiBr₆ with Cu⁺ significantly enhances the material's stability and photoelectric conversion efficiency.

This study aims to investigate the impact of Cu⁺ doping on the crystal structure and electrical properties of Cs₂AgBiBr₆.

Based on density functional theory, first-principles calculations were applied to exploring the effects of Cu⁺ doping on the structure and electrical properties of Cs₂AgBiBr₆. Structural properties, such as the stability, doping formation energy, lattice parameters, elastic constants, of Cu⁺ doping on the lead-free double perovskite Cs₂AgBiBr₆ were investigated by simulation whilst the band analysis and density of states analysis were employed to study the impact of electrical performance.

The results indicate that Cu⁺ doping enhances the stability of Cs₂AgBiBr₆. The Cs₂Ag_1-xCu_xBiBr₆ compounds formed by doping, as well as the original Cs₂AgBiBr₆ material, exhibit indirect bandgap semiconductor behavior. The bandgap significantly narrows with an increase in the Cu⁺ doping ratio. Based on an analysis of the density of states (DOS), the bandgap narrowing can be attributed to the downward shift of the conduction band minimum dominated by Bi6p orbitals due to Cu⁺ doping.

Cs₂Ag_1-xCu_xBiBr₆ exhibits greater stability and superior electrical properties compared to Cs₂AgBiBr₆, making it a promising candidate material for semiconductor radiation detectors.

When the suspended dipole field coils experience significant disturbances during the operation of the suspended dipole field device, the Tilt-Slide-Rotate (TSR) coils are used to control the attitude of the dipole field coils, thereby preventing them from losing control.

This study aims at attitude control of dipole fields by coil forces of the TSR coils with relatively complex geometric structures.

Based on the line segment approximation method, the modeling and analysis were conducted for scenarios where the dipole field coils return to their balanced position under conditions of dipole field tilt and offset by controlling the magnetic forces generated by the TSR coils on both opposite and same sides. The coupling magnetic field between the TSR coils and the dipole field coils was calculated, and the magnetic field line structure was analyzed. The excitation magnetic field of the TSR coils, in the form of sinusoidal currents, was simulated to induce azimuthal magnetic disturbances. Analysis was performed on the magnetic field line structure and phase of the perturbation magnetic field on the Poincaré section at the equatorial plane. [Results and Conclusions] If only one group of TSR coils on the same or opposite sides is used to control the dipole field coils to restore equilibrium, exhibited mechanical instability needs necessitating coordinated control by multiple sets of TSR coils. Applying pseudo-sinusoidal currents to all TSR coils induces compression and stretching effects on the background dipole magnetic field. By controlling the current of the TSR coils to control the phase of the compression position, the feasibility of controlling the attitude of the dipole field coils with multiple TSR coils has been validated. However, this process may also cause drift in the background magnetic field and the formation of open magnetic field lines, which could drive plasma instabilities and lead to particle loss.

The printed circuit heat exchanger (PCHE) has high heat transfer efficiency and compact structure, which can be used as the key equipment for heat transfer of small modular molten salt reactor, It is of great significance to the study of its flow heat transfer characteristics.

This study aims to explore the heat transfer characteristics of printed circuit heat exchanger and compare numerical simulation result with experimental data.

The flow heat transfer characteristics of molten salt (FNaBe) -helium heat exchangers with different flow channel structures, fin types and pitch were obtained by CFD numerical simulation, and the results were compared with those of traditional straight channel structures for comprehensive evaluation. The main thermal performance parameters of the heat exchanger with various airfoil fin structures were verified by experiment to find the best structure.

The results show that the numerical simulation results are in good agreement with the experimental results. Compared with the traditional straight channel structure, the fins can strengthen the heat transfer characteristics of the heat exchanger and reduce the flow resistance. NACA0025-8 mm airfoil fin structure has the best flow heat transfer characteristics.

The numerical simulation method established in this study can be used to predict the flow heat transfer characteristics of printed circuit heat exchanger. The empirical correlation formula of NACA0025-8 mm airfoil fin structure is fitted, which provides a theoretical basis for the design of subsequent heat exchangers.

Traditional text-based system engineering in the design and application of passive residual heat removal systems (PRHRS) for lead-cooled fast reactors has several requirement problems, such as low development efficiency, a long iteration cycle, and model ambiguity.

This study aims to effectively address the aforementioned problems encountered in the PRHRS of lead-cooled fast reactors by adopting a model-based system engineering method.

The model-based system engineering (MBSE) method was preliminarily applied to the design requirement analysis of a PRHRS for lead-cooled fast reactors. During the design requirements study, the design process was combined with the preliminary design of the system architecture, comprising three parts: requirements analysis, functional analysis, and design synthesis.

The generated requirement and use case diagrams describe the system requirements and determine the top-level use cases of the system in the requirement analysis stage. The time sequence, activity, and state machine diagrams form the system function model and provide early confirmation and verification in the functional analysis stage. Finally, the white box model realizes the analysis and design of the system architecture in the design synthesis stage.

The system architecture designed by this method ensures the consistency of the design requirements. In the future, it will further reduce the design risk, improve the design efficiency, and provide an application reference for the design and optimization of digital lead-cooled fast reactors' passive residual heat removal systems.

During severe reactor accidents, melten core causes the radioactive source term material to be no longer retained in the fuel but is released into the environment, causing serious radioactive contamination in the surrounding areas.

This study aims to investigate the release of fission products, both inside and outside the pressure vessel, using different models in order to analyze the effectiveness of the spray system in controlling the source term release and the decay heat generated.

Based on a typical mega kilowatt pressurized water reactor (PWR) nuclear power plant model, the integrated security analysis program MAAP was applied to modeling reactor rupture accident. Then, the accident sequence and consequences of the reactor primary circuit heat pipe breakage superimposed high and low voltage safety injection failure were calculated and analyzed under CORSOR-M, CORSOR-O and ORNL-BOOTH source term release models.

The findings indicate that the fission product source term is mainly released in the pressure vessel, and the release amount is significantly higher than that released outside the pressure vessel. Under the CORSOR-O model, the pressure vessel fails the last whereas the containment fails the first; although the pressure vessel fails first in the ORNL-BOOTH model, the containment vessel fails much later than that in the other two models. The difference in source term release leads to different decay heat phenomena in different models, and the main heat source is the volatile fission product. Turning on the spray can not only keeps the suspended iodide in the containment vessel but also effectively removes the decay heat generated by the source items and reduces the pressure of containment.

The ORNL-BOOTH model results in lesser release of source term in the pressure vessel, the greatest variety of source term released, and the maximum time to containment failure. In addition, the opening of the spraying system effectively ensures the integrity of the containment.