Serial X-ray crystallography has developed rapidly due to its advantages of data collection at room temperature, low radiation damage and time resolution. To solve protein structures by using the serial X-ray crystallography, a large amount of produced diffraction data needs to be screened for finding the effective diffraction patterns. The use of convolutional neural networks (CNN) can not only automate the data screening process, but also improve the accuracy of data classification comparing with the traditional "point finding method".

This study aims to explore five types of popular convolutional neural networks, i.e., AlexNet, GoogleNet, MobileNets, Vgg16, ResNet, for screening crystallographic diffraction patterns, and compare the accuracy and efficiency of them to build up a fast and accurate convolutional neural network tool for screening the diffraction patterns of different protein crystal samples.

Firstly, the primitive data for model training extracted from the coherent X-ray image database, collected by Linac Coherent Light Source (LCLS) and Spring-8 Angstrom Compact free electron laser (SACLA), were pre-processed by gray level equalization and data enhancement. The deep learning models were trained by iteration of the preprocessed data. Then, the selected convolutional neural network through the comparison of accuracy and efficiency was used to process further the experimental data of protein crystals diffractions.

The results show that MobileNets not only has the accuracy similar to large networks such as ResNet, GoogleNet-Inception, but also runs faster.

MobileNets provides an effective and convenient screening tool for serial X-ray crystallography experimental data.

Neutron radiography (NR) is an important nondestructive testing method. NR is particularly useful for detection of light materials in medium and large heavy samples. Especially, the fast neutrons can penetrate the heavy materials and reveal the structure of the light materials. Compared to accelerator neutron sources, the fission neutrons elicited from a reactor are stable and of high quality. The fission neutron imaging is a useful complementary testing technology, especially for industrial applications that require high throughput and large-scale testing.

This study aims to investigate the super field of view neutron imaging by fission neutrons elicited from research reactor.

Based on theoretical analysis and Monte-Carlo simulation, one filter combination was employed to improve the proportion of fission neutrons in the thermal neutron beamline at China Mianyang Research Reactor (CMRR). A fission neutron imaging system was constructed by employing a large field fast neutron fluorescent screen, short focus distance lens, and scientific charge coupled device (CCD) camera. Finally, some samples were tested using fission neutron tomography.

The fission neutron flux reaches up to 3×10⁵ cm^-2·s^-1 when the L/D ratio is about 260. The field of view of NR is up to 400 mm×400 mm with resolution was better than 0.5 mm. Using super field of view method, samples less than 600 mm can be tested with this new system.

Combination of theoretical calculation and experimental methods, fission neutron imaging can be improved to overcome some of the limitations of traditional neutron radiography techniques, and meet the needs of large sample detection in the future.

Stable isotopes play a crucial role in a variety of fields such as energy, military, semiconductor, agriculture, medicine, pharmacology, biology, food industry, and chemistry. With the rapid growth of nuclear science and technology applications in China, there has been an increasing demand for isotopes that cannot be met by current production capacities. Thus, the development of electromagnetic isotope separators capable of producing high yields and high isotopic purity has become necessary.

This study aims to develop an electromagnetic isotope separator based on a 2.45 GHz microwave ion source and isotopic magnet for studying a number of important heavy isotopes, such as xenon and molybdenum isotopes.

Firstly, adjustable axial magnetic field in the source was designed by a double-solenoids to obtain high density plasma, and a high coupling efficiency matching waveguide was optimized by CST microwave computing module. Then. a crucible built in the discharge chamber was used to melt metal oxide for generating heavy metal ion beams. Finally, the discharge chamber, microwave coupling waveguides and heating oven of the ion source were simulated and designed for the generation of heavy ions.

Simulation result shows that the temperature around the crucible is 917 ℃ when the current of heating wire is set to 70 A, and 100 mA hydrogen beam is generated during commissioning. The designed crucible in the discharge chamber can generate metal vapor efficiently for ionization, and achieve producing 20 emA Xe⁺ and 5 emA Mo⁺ respectively at the energy of 40 keV.

The feasible scheme of the magnetic field and microwave coupling design of this study are verified. The design of the 2.45 GHz electron cyclotron resonance (ECR) ion source provides a feasible and effective solution for the high yields isotope ions.

Hard X-ray Imager (HXI) is one of the three scientific payloads onboard of the advanced space-based solar observatory (ASO-S). The calorimeter of HXI consists of 99 LaBr₃ crystal and photomultiplier tube (PMT) detection units. A highly integrated charge-measurement application-specific integrated circuit (ASIC) with model IDE3381 is adopted in the front-end electronics of the calorimeter to process the signals from the 99 detection units on the space-limited and power-limited satellite platform.

This study aims to evaluate the radiation tolerance of model IDE3381 ASIC in a space radiation environment.

A test bench with a flexible structure was designed by separating the device undergoing testing from the data acquisition (DAQ) system, hence shielding DAQ from the radiation environment. The performance of ASIC was automatically tested and monitored in the test bench during radiation tests. Both single-event effect (SEE), including single-event upset and single-event latch-up, and total ionizing dose (TID) tests were carried out by using a heavy ion beam and ⁶⁰Co gamma-ray, respectively.

The test results show that the SEE threshold of model IDE3381 ASIC is greater than 75 MeV?cm²?mg^-1, and the TID capacity is greater than 30 krad(Si).

The radiation tolerance of the charge measurement ASIC (model IDE 3381) meets the requirements of ASO-S HXI flight model.

¹³¹I-BEV-PTX-SPIONs is a type of nanoparticle used in cancer therapy. It is composed of four components: a radioactive isotope of iodine (¹³¹I), a chemotherapy drug called paclitaxel (PTX), a type of nanoparticle called superparamagnetic iron oxide nanoparticles (SPIONs), and a molecule called bevacizumab (BEV) which is an antibody that targets and blocks the growth of blood vessels that supply tumors.

This study aims to investigate the preparation and biological distribution of ¹³¹I-BEV-PTX-SPIONs.

First of all, ¹³¹I-BEV-PTX-SPIONs were prepared, synthetized and identified. The transmission electron microscope (TEM) were employed to observe the particle characteristics. Then, 30 tumor-burdened nude mice were divided into the single targeting group and the dual targeting group for evaluation of ¹³¹I-BEV-PTX-SPIONs distribution in these nude mice, each group was divided into five sub-groups based on time points of 2 h, 6 h, 12 h, 24 h and 48 h, 3 in each sub-group. Finally, ¹³¹I-BEV-PTX-SPIONs were injected into the caudal vein of these mice, and experiments of biological distribution in vivo and SPECT imaging were carried out, and results were analyzed using GraphPad Prism 8.3 software.

The nanospheres in prepared ¹³¹I-BEV-PTX-SPIONs are obtained in good mono dispersion with a diameter of approximately 220 nm by TEM observation. ¹³¹I-BEV-PTX-SPIONs obtained in a high radiolabeling yield is about 81.4% with the radiochemical purity of over 99% and good stability shown in the 0.2 mol·L^-1 PB buffer. And it could attain sustained PTX release in vitro. Comparing with the cellular uptake of ¹³¹I, a higher uptake and sustained PTX release in vitro are shown for ¹³¹I-BEV-PTX-SPIONs. Biodistribution experimental results show: After the injection of ¹³¹I-BEV-PTX-SPIONs, with the extension of time, the radiation count of the tumor is relatively higher, at 12 h reaching the peak. And the T/NT ratio increased gradually, and it reaches 7.8±0.50 at 48 h. The counts and the ratios at 6 h, 12 h, 24 h and 48 h are notably higher in the dual targeting group than the single targeting group (P<0.05), and over time the differences are more significant. SPECT examination results show: At 2 h after the injection of ¹³¹I-BEV-PTX-SPIONs, the tumor site has a radioactive build-up. With the extension of time, the accumulation of radioactivity increased and remained stable, and the T/NT ratio rises steadily.

These results demonstrated the potential of ¹³¹I-BEV-PTX-SPIONs in the diagnosis and treatment of lung cancer, and it was worthy of further study.

Compton imaging technology is a new radiation hotspot location technology that does not require collimation and has a wide field of view, high efficiency, and broad application prospects. With the development of nuclear technology, Compton cameras with the above-mentioned advantages have a wide range of applications not only in the nuclear industry but also in the field of nuclear medicine, hence recently become a popular research field worldwide.

This study aims to develop a double-layer separated Compton camera for far-field imaging of specific radiation scenes in nuclear facilities.

First of all, two pixel-type cadmium zinc telluride (CZT) detectors and an application-specific integrated circuit (ASIC)-based readout electronics system were adopted for the development of a double-layer separated Compton camera. A list-mode maximum likelihood expectation maximization (LM-MLEM) image reconstruction algorithm was implemented in the host computer software. Then, ¹³⁷Cs point source was used for experimental test of imaging performance of the system, and the parameters affecting the imaging performance, such as the detector layer spacing and area of the absorption layer, were optimized. Finally, far-field three-dimensional imaging of the radiation source was performed by moving the measuring position of the detector.

The test results show that the energy resolution of the CZT detectors is approximately 3% (FWHM@662 keV), which can determine the location of the point source at a distance of 5 m, and the angular resolution for θ and φ directions of the optimized system is approximately 10°.

Double-layer separated Compton camera of this study has advantages of adjustable structure, low detector cost, relatively simple readout electronics, and wide imaging field of view. The angular resolution of this double-layer separated Compton camera can be improved by proper adjustment of the imaging influence parameters (such as layer spacing and the area of the absorption layer).

The miniature fission ionization chamber is a widely used neutron detector for the in-core neutron flux monitoring of a nuclear reactor. Typically, the in-core neutron flux rate measurement system of a domestic CPR1000 nuclear power unit adopts a mobile miniature fission ionization chamber as the neutron probe to measure the neutron flux of the reactor and provide an in-core neutron flux distribution map during the operation. Therefore, it is an important piece of safety and control equipment for nuclear power plants.

This study aims to develope a mobile miniature fission ionization chamber neutron detector according to the service conditions and technical requirements of current foreign products.

The nuclear properties of self-made fission ionization chamber neutron detector was developed strictly following the national standard GB/T 7164-2022 and the industry standard NB/T 20215-2013. The gamma sensitivity was tested and compared with a reference commercial fission detector using a ⁶⁰Co gamma source. The thermal neutron detection characteristic, including the length and slope of plateau, the thermal neutron sensitivity and linearity were tested in one test channel of the China Mianyang Research Reactor (CMRR) with neutron flux from 1×10⁹ n?cm^-2?s^-1 to 4×10¹³ n?cm^-2?s^-1.

The test results indicate that "domestic substitution" of this in-core safety product can be achieved, and the nuclear characteristics of self-developed prototypes are comparable to those of foreign products.

Plastic scintillators have potential for application in neutron detection. Two sizes (?2.54 cm×2.54 cm, ?5.08 cm×5.08 cm) of plastic scintillators are self-developed by scientific research team in the school of physics, Sichuan University.

This study aims to experimental test the neutron/gamma (n-) discrimination performance for two self-developed plastic scintillators.

A photomultiplier tube (PMT) was used to build detection systems, and high speed oscilloscope (LECROY HDO6104A) was employed to sample signal of detector for the energy calibration of the self-developed plastic scintillator. The pulse amplitude spectrum of ¹³⁷Cs γ radiation source was measured and compared with the MCNP5 simulation spectrum to obtain the position information of the Compton edge and accurately calibrate the energies of γ rays. The data obtained from a ²⁴¹Am-Be neutron source were analyzed using the charge integration method, and parameters such as the figure of merit (FOM), peak-to-valley ratio for neutrons, and the proportion of leaked neutrons over all neutron events were used to quantify the n-γ discrimination in different energy zones. The detection efficiencies of two self-developed plastic scintillators relative to the Commercial off-the-Shelf (COTS) EJ-299-33A were determined.

The results show that the FOM of ?2.54 cm×2.54 cm self-developed plastic scintillator is higher that of ?5.08 cm×5.08 cm self-developed plastic scintillator, and the detection efficiency of two self developed plastic scintillators relative to EJ-299-33A is about 0.49 and 1.0, respectively.

The performance of the ?5.08 cm×5.08 cm self-developed plastic scintillator is comparable to that of the COTS plastic scintillator EJ-299-33A with near the same discrimination ability.

The pre-equilibrium cluster manifests the nuclear structure and reaction dynamics of collision system. The systematical investigation of cluster emission in transfer reactions is of significance in deep understanding of the synthesis of superheavy nuclei, shell evolution, new isotope production, etc.

The dynamics of pre-equilibrium cluster in a few of nucleon transfer reaction has been described by theoretical model, such as exciton model, cluster model. However, the cluster emission in massive transfer is very complicated because of the emission mechanism associated with the structure properties and also the dynamical process.

In this work, the pre-equilibrium cluster emission in massive transfer reaction has been systematically investigated within the dinuclear system model. The model has been successfully used for describing the massive fusion reaction and multi-nucleon transfer dynamics. The nucleon exchange and energy dissipation take place once the dinuclear system is formed. The nucleon transfer between the binary fragments is governed by the single-particle Hamiltonian and proceeds around the Fermi surface formed by the dinuclear system. The master equation is used for the nucleon transfer dynamics and the relative motion energy dissipation is taken into account. The dynamics of neutron, proton, deuteron, triton, ³He, ⁴He, ^6,7Li and ^8,9Be in collisions of ¹²C+²⁰⁹Bi, ⁴⁰Ca+²⁰⁸Pb and ⁴⁸Ca+²³⁸U near Coulomb barrier energies is analyzed, i.e., temporal evolution of production rate, kinetic energy spectra and angular distribution.

It is found that the emission probability of ⁴He is the same magnitude of proton emission and several orders larger than the one of ³He. Both the nuclear structure and dynamical effects influence the pre-equilibrium cluster production.

The pre-equilibrium clusters are emitted from the 'projectile-like' and 'target-like' fragments and the angular distributions manifest the similar trends. The kinetic energy spectra of clusters are shown as the Boltzmann distribution. The method is also extended to the cluster emission in weakly bound nuclei induced reactions by considering the preformation factor for the cluster construction.

There is usually a strong coupling of neutronics-thermal hydraulics (N-TH) fields inside nuclear reactors.

This study aims to accurately simulate the multi-physics fields in nuclear reactors by developing a three-dimensional N-TH coupling code MORPHY tailored to advanced complex reactors.

First of all, a three-dimensional triangular-z nodal variational nodal method (VNM) was employed for neutronics calculation. and the stiffness confinement method (SCM) was used to solve the neutron temporal-spatial equation; thermal hydraulic calculations were based on the one-dimensional multi-channel model and the one-dimensional cylindrical thermal conductivity model. Then, the accuracy of neutron dynamics was verified by TWIGL benchmark, Dodds benchmark, and the typical pressurized water reactor (PWR) benchmark NEACRP. Finally, the effects of different coupling methods and angle discrete orders on the results were analyzed and compared against reference solutions by PARCS.

Verification results of TWIGL benchmark show that the deviation of relative power from the reference results is less than 0.5%. Compared with the results of Dodds benchmark, it verifies the MORPHY code's ability to describe unstructured meshes. The transient coupling calculation capability of MORPHY is verified by NEACRP benchmark.

Numerical solutions by MORPHY are in good agreement with reference results of the TWIGL, Dodds and NEACRP benchmark problems. It is concluded that MORPHY can adapt to the transient N-TH coupling analysis of nuclear reactor cores.

In the neutron activation calculation, the inherent uncertainty of the input nuclear data will cause a certain impact on the calculation results. The uncertainty of the calculation results plays an important role in the source term analysis and radiation shielding design of nuclear facilities.

This study aims to analyze the uncertainty of neutron activation calculation based on direct derivative method.

Firstly, the direct derivation method and Gear algorithm for uncertainty analysis were investigated, and the activation coefficient matrix and sensitivity coefficient matrix were constructed. Then, the Gear algorithm was employed to solve the activation equation and sensitivity equation simultaneously, and the sensitivity coefficient of nuclide inventory to nuclear data was obtained. The relative uncertainty of nuclide inventory was obtained by combining the relative uncertainty of nuclear data. Finally, This method was integrated into the neutron activation program ABURN, and typical examples was selected to test and verify its performance.

The calculation results of the nuclide inventory and its sensitivity coefficient and relative uncertainty by the ABURN program have little deviation from the analytical solution and the numerical solution of the European activation program FISPACT, most of the deviations do not exceed 0.2%, and the maximum deviation does not exceed 1%.

Verification results show that the method and procedure developed in this paper have the ability to analyze the sensitivity and uncertainty of nuclide inventory with high precision, and can provide tools and data support for the radiation protection of nuclear facilities and the source term analysis.

Under the condition of sub-cooled nucleate boiling (SNB), corrosion products in primary coolant of nuclear reactor will deposit on the outer surface of fuel cladding, which is commonly called fuel crud. Previous literature shown, zinc injection in primary coolant is an important method to inhibit the fuel crud deposition on the fuel cladding surface.

This study aims to investigate the influence of zinc concentration on the behavior of fuel crud deposition, and eventually provide guidance for zinc injection in primary coolant of nuclear power plant.

The fuel crud deposition tests of domestic zirconium alloy fuel cladding in different zinc concentrations were carried out by using a self-made fuel crud deposition device. Tubular crud deposition test specimen with built-in heating unit was designed and prepared for simulation study. After the tests, stereo microscope (SM) and scanning electron microscope (SEM) were employed to observe the macro and micro morphology of fuel crud whilst the composition of of fuel crud was observed and analyzed by the energy dispersive spectroscopy (EDS) with SEM, and X-ray photoelectron spectroscopy (XPS) was used to analyze the contents of Zn and B elements in the crud phase and inside the crud.

Observation results show that the chimney-like crud formed on the fuel cladding surface becomes less obvious with increasing the zinc concentration in the coolant and the crud surface becomes flatter. Simutabeously, the crud thickness, the ratio of Ni/Fe and the boron precipitation mass within the crud are decreasing with increase of the zinc concentration. When the zinc concentration increases to 100 μg?L^-1, new Zn-containing phases precipitate within the crud.

Within the zinc concentration of 0~100 μg?L^-1, zinc injection in primary coolant of reactor can significantly inhibit the crud deposition on the fuel cladding surface.

Core fuel salt emergency drain system designed for fuel salt drain and afterheat removal, provides a safe shutdown mode for a molten salt reactor (MSR). It has important significance to evaluate reliability of the system for the safety of MSRs.

This study aims to quantitatively analyze the failure probabilty of the system and identify the pivotal factors that affecting the system failures, and provide suggestions for optimization of the system in engineering application.

First of all, the fault tree analysis was employed to model the reliability of the core fuel salt drain system of MSRE through RiskSpectrum software. Then, the minimum cut sets and importance analysis was adopted to identify the most important basic event in fault tree of the system. Finally, two optimization methods, i.e., reduce the use of welds in bayonet cooling thimbles, and use different types of valves to isolate cooling gas flow of freeze valve, were proposed.

The results show that failure probability of the system is 5.62×10^-4, and the identified pivotal factors affecting the system failures are welds leakage failures of thimbles and common cause failure of two groups valves of freeze valve. The optimization methods based on results of fault tree analysis can significantly reduce the system failure probability.

This study provides reference value for design and engineering application of the core fuel salt emergency drain system for MSRs.

Molten salt systems have been extensively applied in the generation of nuclear and solar energy owing to their excellent heat transfer and storage performance. Therefore, it is essential to explore their physical and chemical properties, which are largely determined by their composition and ionic structure. In this regard, high-temperature (HT) nuclear magnetic resonance (NMR) has been demonstrated as an effective solution for qualitative analysis. Due to the existence of vent holes on the top of the commercialized standard NMR sample cell, it is not suitable for the study of some volatile, toxic or radioactive molten salt systems.

This study aims to implement ceramic NMR sample cells that are suitable for different molten salt systems.

Firstly, a novel sealed sample cell that meets the requirements of molten salt systems was designed and produced by using an inner tube comprising AlN, BN, and Al₂O₃ ceramic materials, and an outer tube composed of ZrO₂ ceramic materials. Then, with KBr as the standard sample, temperature calibration for this sample cell was conducted on the basis of ⁷⁹Br NMR method. Finally, LiCl-KCl two-component molten salt (59.2 mol% LiCl-40.8 mol% KCl, eutectic melting point 353 ℃) was selected for HT-NMR experiment to verify the accuracy of temperature calibration, and check the performance of the HT-NMR method.

Experimental results show that the sample cells are applicable in molten salt systems at a maximum temperature of 700 ℃, which meets the detection requirements of most molten salt systems. Additionally, the measurement results of the ⁷⁹Br chemical shift of the KBr samples and captured HT ³⁵Cl NMR spectra of the LiCl-KCl molten salt verify the reliability of the sample cell.

Based on the results, the HT-NMR (High Temperature NMR) sample cells proposed in this study can be widely applied in various fields when the molten salt system requirements are met.

Pack cementation aluminizing technology is a common method for preparing tritium barrier coatings, and its relative parameters during the preparation process have an important influence on the microstructure of the aluminide layer and the tritium barrier properties of the in-situ oxidized Al₂O₃ coating.

This study aims to investigate the effects of pack aluminizing conditions on the microstructure of the Fe-Al layer and analyze the related kinetic analysis of the aluminizing process.

First of all, a pack aluminizing process activated by 1 wt% AlCl₃ was used to fabricate aluminide coatings on the substrate of 316L stainless steel in the 923 K to 1 173 K range. Then, scanning electron microscope (SEM), energy dispersive spectrometer (EDS), and X-ray diffraction (XRD) were employed to characterize the cross-sectional microstructure and composition of the aluminized layer. Finally, the effects of aluminizing temperature and time on the microstructure and composition of the aluminized layer were analyzed, and the kinetic parameters of the formation of the Fe-Al layer and the relationship between aluminizing time and the thickness of the aluminized layer were further obtained.

The experimental results show that the main phases of the aluminized layer are Fe₂Al₅ and FeAl with a certain amount of FeAl(Cr,Ni) precipitates. The high aluminizing temperature would accelerate the growth of aluminized layer and lead to the formation of a thick intermediate layer between the substrate and outer aluminized layer above 1 023 K. Simultaneously, extending the aluminizing time could increase the thickness of the Fe-Al layer, but has no effect on the phase composition. The relation between aluminizing temperature and the growth velocity of the Fe-Al layer is in accord with Arrhenius' equation, and the relative activation energy of the aluminizing process is about 79.23 kJ·mol^-1. During the process of pack aluminizing, the relationship between the aluminizing time and the Fe-Al coating thickness is h=14.585t^1/2+19.514.