With the in-depth exploration of space, the power supply requirements for energy supply are gradually increasing. Space nuclear reactor power supply stands out in the deep space exploration mission due to its advantages of passivity, long life and high reliability, hence heat pipe nuclear reactor has become the focus of research in the field of space nuclear reactor.

This study aims to evaluate the progress and technology of heat pipe space nuclear reactor used as power supply.

First of all, the conceptual design of heatpipe reactor in the early stage and its application in Kilopower was summarized. Then the details of heatpipe power system (HPS), the heatpipe-operated mars exploration reactor (HOMER), safe affordable fission engine (SAFE) and Kilopower were investigated with emphasis on the structure design, fuel selection, heatpipe arrangement and power design of each reactor.

This survey provides ideas and references for future research and design of heatpipe space nuclear reactor power system.

The design of heatpipe reactor is highly flexible and innovative. The use of different alkali metal heat pipes and thermoelectric conversion methods has a direct impact on the system power and the quality of the entire system. The combination of heat pipe and various thermoelectric conversion methods can be attempted.

As the coolant of the reactor, lead and lead-based alloy has excellent thermal properties. Developing lead cooled reactor attracts has been practicing with major investment in many countries, such as European union countries, Russia, Korea, Japan, China and so on.

This study aims at the preliminary design of modularized lead cooled small reactor.

First of all, the neutron transport, depletion, reaction coefficient and kinetic parameters were calculated by using SARAX, a fast reactor neutronics calculation software developed by Xi’an Jiaotong University. High enrichment fuel was adopt for compact light-weight, and two different enrichment fuel was chosen to flat the power distribution. Two group control rods and one group of emergency shutdown rods were placed in the core to satisfy the requirement of reactivity and emergency shutdown. The B₄C with good absorbing capacity in fast and thermal energy region was used as the neutron absorb material of control rods while tungsten with high density and excellent neutron absorb capacity was mixed with B₄C as the neutron absorber material of the emergency shutdown rods. Then neutronics parameters of reactor core were calculated from the beginning of full power operation to the whole life cycle.

The calculation results show that the 6EFPY life length of this design is achieved, mass of actinide elements decreases and mass of ²³⁹Pu was increases apparently. Important reactivity feedback coefficients, such as coolant density coefficient, coolant void value and expansion coefficient are negative, and the value of control rods and safety rods are enough to shutdown reactor in the whole life cycle. Highest clad temperature and fuel center temperature, line power density are within allowed limits.

The main neutronics parameters in the whole life cycle of this preliminary design meet the safety requirements.

In synchrotron radiation facilities, it is important to keep beamlines operating in optimal conditions. The debugging process is normally very time consuming due to the irregular light source beam point, and it is not easy to get global optimum.

This study aims to develop an intelligent debugging system based on Experimental Physics and Industrial Control System (EPICS) and differential evolution algorithm for synchrotron radiation beamline commissioning.

First of all, based on the differential evolution algorithm, intelligent optimization model of beamline was established. Then the automatic optimization of beam adjusting process was implemented by using LabVIEW program and communication with the EPICS-based control system was achieved by CaLab interface module. Functions of the user interface, motion control, algorithm implementation and evolution processing were integrated in the LabVIEW program. Finally, this intelligent commissioning system was tested at the X-ray diffractive (XRD) beamline of SSRF (Shanghai Synchrotron Radiation Facility) for optimization of the beam flux at sample position by adjusting the beamline optical components.

Online tests results show that this intelligent commissioning system converges to the optimal solution quickly, and the convergence time is about 30 min, more than one order of magnitude higher than manual optimization efficiency.

Field programmable gate array (FPGA) is widely used in many places, including readout system of large-scale high-energy physics experiments. This kind of experiments usually has large electronic systems with thousands of channels implemented together with FPGAs. Furthermore, these detectors and electronics are often located in special environments such as radiation field, under water or underground, hence the traditional single FPGA update method with joint test action group (JTAG) cable becomes unsuitable or inapplicable.

This study aims to realize FPGA remote firmware update without additional Ethernet chips or network protocol, and ensure the security mechanism of the process and multi-FPGA updating extendibility in the future.

A new approach using silicon transmission control protocol (SiTCP) technique was proposed. The update firmware in the UDP packages was sent under the SiTCP protocol by MATLAB in host computer to FPGA, and the process of Flash memory programming to update firmware was dominated by FPGA, placing the role of PHY and MAC chips. The secure rollback function was implemented in the modified structure of FPGA firmware with separate update data area and original data area.

UDP broadcasting has the ability to multi-FPGA updating, remote firmware update is achieved and tested the secure rollback mechanism well.

This method is suitable for remote operation with advantages such as no need for additional Ethernet chips, built-in safety function to fall back to a known situation, and extendibility of multi-FPGA updating.

Mars exploration has recently become a mainstream trend in space research. The establishment of a Mars base is an inevitable choice for human to study and develop Mars. Compared with the solar energy storage system, the nuclear reactor as a Mars base energy system has significant advantages in terms of system mass, operational flexibility, and environmental robustness.

This study aims to propose core design scheme of Mars molten salt reactor (M²SR-1) and calculation model in aspects of reactor physics, critical safety and thermal.

The Monte Carlo N Particle Transport Code (MCNP) and Oak Ridge Isotope Generation and Depletion Code (ORIGEN) were employed to establish core physical calculation model. Calculation and analysis were performed in aspects of physical, safety, thermal and so on.

Computational results reveal that the scheme of M²SR-1 meets the requirement of 8-year-life under full power operation. Under different hypothesis falling environments, the effective multiplication factor of nuclear reactor is less than 0.98, satisfying the criticality safety requirements.

This study provides basic theoretical reference for the design of planet surface molten salt reactor.

As a new solid-state photosensor, silicon photomultiplier (SiPM) has high gain (10⁶), low power consumption (supply current is several μA), small size (several square milimeters), low operating voltage (less than 100 V) and other advantages. Compared with traditional photomultiplier tube, SiPM does not affected by magnetic fields. However, there is a disadvantage that the gain of SiPM is greatly affected by temperature, the gain will decrease by about 5×10⁵ for every 10 °C rise in temperature, which will have a greater impact on energy spectrum measurement and radiation imaging applications.

This study aims to design and develop a small programmable SiPM power supply for temperature independent gain.

First of all, a high-voltage power supply module was designed for SiPM, making use of the gain regulation by voltage. Temperature sensor, DC-DC power chip, etc., were integrated into an ARM-based micro control unit (MCU) module for automatic SiPM power supply. Then the effect of voltage compensation to gain in the temperature range of 5 ℃ to 40 ℃ was verified by using digital multi-channel analyzer with comparison of the channels of full energy peak on the energy spectrum of ²⁴¹Am source.

The channel of full energy peak which indicates the gain of the SiPM is stable by using the voltage compensation (gain drift correction), and the maximum drift of the SiPM gain from 5 ℃ to 40 ℃ decreases from 87.3% before compensation to 2.76%.

Gain drift correction achieved by a small intelligent programmable SiPM power supply broadens the scope of SiPM application.

The number of radioactive practitioners has increased rapidly due to the wide application of X-ray, resulting in widespread attention to radiation protection. Lead is commonly used as a shielding material, but is extremely toxic and harmful to human health. The lead-based protective clothing on the market is too heavy and uncomfortable to wear.

This study aims to design flexible composites of lead-free for shielding X-ray of different energy ranges and evaluating its shielding effect.

The X-ray mass attenuation coefficients of high atomic number materials were calculated according to National Institute of Standards and Technology (NIST) method. Hence, the ideal functional fillers were selected as shielding materials instead of lead. The MCNP5 code was employed to evaluate protective performance of lead-free flexible composites consisting of various doped metal elements with high atomic number.

Simulation results show that Bi doping is an ideal substitute for Pb doping, and Bi₂O₃ can be used as the first-rate functional filler in lead-free flexible composites. Shielding performance of the composite doped with Bi₂O₃ and Gd₂O₃ is obviously improved during the energy of 54~66 keV whilst the composites doped with Bi₂O₃ and W functional fillers achieve the best X-ray attenuation at around 70 keV of X-ray.

A variety of bismuth-based flexible composites have been designed to replace lead-based materials. Bi₂O₃ is the optimal filling material for X-ray scattering energy less than 54 keV, incorporation of two functional materials, Bi₂O₃ and Gd₂O₃, can effectively improve the absorption effect of the composites for X-ray scattering energy in range of 54~66 keV. In the energy range of 66~100 keV, incorporation of two functional materials, Bi₂O₃ and W, is the optimal filling method whilst it is best to fill Bi₂O₃ alone to shield X-ray scattering energy greater than 100 keV.

In ⁶⁰Co dual projection radiation imaging system, the radiation sources are arranged at the bottom and side while their corresponding ionization chamber detectors are set respectively in the gantry. With the synchronous movement of the sources and the gantry, the system can obtain both side-view and upward-view images of the objects simultaneously. However, due to the presence of scatter, the imaging of the two projection planes will interfere with each other.

This study aims to reduce the effect of scattering between two projection planes and improve the imaging quality of the system by using Monte Carlo method.

Firstly, a simulation model based on Monte Carlo method was established, and its reliability was verified by comparing the simulation data with the experimental data of the scattering distribution at no-load. Secondly, the scattering effects of different projection plane spacings, different shielding materials (tungsten, lead, steel, and aluminum) around the detector with different installation positions, and thickness of the shielding layer on the scatter were calculated based on the model built. Finally, shielding correction effect of 2.5 cm thick lead layer was evaluated.

Simulation results show that when the projection plane spacing is 40 cm and 80 cm, the scattering can be reduced to less than 55% and 40% of that of the 10 cm plane spacing respectively. When projection plane spacing is 40 cm, a lead shield placed between the gantry support and the side-view ionization chambers can achieve best shielding effect at the same thickness compared with steel and aluminum. When the thickness of the lead is 2.5 cm, scattering will be further reduced by 80%.

This study provides a reference and guidance for the hardware improvement and upgrade of the dual-projection system for scattering correction.

The new generation of pin-by-pin calculation method has gradually become the state-of-the-art approaches along with the improvement of computer technology and the development of core design.The difference between the pin-by-pin calculation and the traditional two-step scheme leads to the homogenization technique difficult to implement.

This study aims to evaluate super homogenization techniques for pin-by-pin core calculationwith emphasis on equivalent homogenization constants.

Super homogenization method was used as one of the main homogenization techniques of pin-by-pin calculation. For fuel assembly, the traditional super homogenization method was employed to generate the group constants. For reflector assembly with neutron leakage, the super homogenization method related to space leakage was applied to the generation of equivalent homogenization constant including super homogenization factor.

Based on the three-dimensional C5G7 benchmark, the application performance of the super homogenization method in pin-by-pin calculation is evaluated. The eigen-value relative error of pin-by-pin calculation is -0.001 6, assembly-power and pin-power relative errors are less than 5.5%.

Numerical results showed that the pin-by-pin calculations have the higher accuracy than the assembly-homogenized results.

The traditional gamma camera cannot locate the contaminated area of nuclear accident, and cannot obtain the information of the contaminated area and degree in real time.

This study aims to develop a coded aperture gamma camera detector system with large area and high sensitivity.

A detector module of large area CsI (Tl) crystal array coupled with photomultiplier tube array was proposed. It consisted of 121 CsI (Tl) crystals at the size of 15 mm×15 mm×15 mm, and 2×2 array composed of 4 detector modules was taken as γ camera detector. Multi-channel readout electronics composed of amplification, filter shaping and single-to-differential conversion circuit was independently designed for readout processing of detector signals. Finally, the performance of this detector system was evaluated by experimental tests.

The results indicate that the detector module can achieve the position resolution, and its average energy resolution is 9.4% under the excitation of 662 keV γ-ray.

The detector system can satisfied the design requirements of coded aperture γ camera of high sensitivity in terms of position resolution, energy resolution, uniformity of peak position and detection efficiency.

With the improvement of computing power and the rise of high-throughput computing, researchers usually need to perform a large number of repeated instructions to submit scientific tasks and further check their computational results during material theoretical calculations. Generally, these instructions are simple, cumbersome and time consuming.

This study aims to improve the efficiency of scientific calculations by using VASP code.

The material theory calculation assistant program, VaspCZ, developed by using Python languages, is summarized to form a general github open source project, which provides detailed user documents, examples and application programming interfaces (API).

VaspCZ, an efficient VASP computation assistant program includes two parts: the software part and API part. The software part provides the command line user interface, which enabled the researchers to complete the VASP calculations with the basic Linux command. The API part provides the bottom layer library to realize the custom calculation (such as high-throughput calculation, writing advanced application, etc.) for the researchers with python experiences.

It is proved by practice that the efficiency of theoretical calculation of materials using VASP can be significantly improved by aid of this assistant program.

The scram rod drop time is an important parameter for reactor safety analysis, which is of great significance to the safety of molten salt reactor (MSR). The process and the consequences of the design basis events of the MSR will be significantly influenced by the dropping time of scram rod.

This study aims to analyse the effect of the scram rod drop time on the consequences of MSR reactivity insertion transient.

First of all, the RELAP-MS based RELAP5-TMSR (reactor excursion and leak analysis program-thorium molten salt reactor) code was employed to establish the safety analysis model for the transient simulation of TMSR-LF. Then, the sensitivity of control rod withdrawal speed to transient consequences was analyzed, next, reactivity introduction event caused by the withdrawal of a control rod at full power was taken into account to analyse the vartions of power, temperatures of fuel salt and structure materials. Influence of scram rod dropping time in emergency shutdown on transient consequences of reactivity introduction in MSR was analyzed.

Even if the scram rod drop time reaches as long as 10 min, the maximum temperature of Harrington alloy is 708.2 °C, and the maximum temperature of fuel salt is 709.2 °C, which is below the safety limit.

The molten salt reactor has good safety characteristics, low requirements for scram rod drop time, and the ability to deal with reactivity insertion event.

Mobile digital radiography (DR) is a hot research direction in the field of modern medical imaging.

This study aims to develop a new-type X-ray tube for mobile DR devices.

Firstly, in order to solve the problems of slow start-up and high dose of ineffective radiation of hot cathode ray tube, carbon nanotube (CNT) was employed to replace traditional filament as the cathode of the X-ray tube. Then flexible adjustment of focal spot size was realized through active focusing mechanism, so as to meet various clinical application needs. Finally, experimental tests were carried out to compare with the simulation results.

Both the simulation and experimental results show that the maximum anode voltage of the finally built X-ray tube is 110 kV, the maximum anode current is 70 mA, and the focal spot size varies between 0.9 and 1.9 (IEC 60336 nominal focal spot value). The different working parameters of the corresponding X-ray tube in four different clinical application scenarios are given.

The X-ray tube has many advantages, such as fast startup, low power consumption, and wide application scenarios, which makes it very suitable for mobile DR devices.

Synchrotron radiation technology has a wide application prospect. X-ray absorption contrast imaging can not only obtain the three-dimensional (3D) morphological structure of neurons, but also improve the effective spatial resolution of detection, which meets the requirements of micro-neural network imaging research. It is feasible and effective to construct high-resolution 3D visualization network of brain neural network when combined with Golgi-Cox staining.

This study aims to explore the application of synchrotron radiation absorption contrast imaging technology in the observation of 3D imaging of Golgi-Cox-stained mouse brain micro-neural network.

Twelve normal C57 mice were randomly divided into two groups, A and B, and perfused with brain tissue. Group A was treated without staining. The brains of group B were stained with Golgi-Cox and dehydrated in gradient ethanol. Specimens were scanned for absorption contrast imaging at X-ray imaging and biomedical application beam line station (BL13W1) in Shanghai Synchrotron Radiation Facility (SSRF), and data processing was performed by using software such as Amira.

The imaging results show that both the cerebellar neural network structure and single neuron morphology in mouse are accurately presented. X-ray absorption contrast imaging technology, as an advanced three-dimensional rendering method of microscopic neural network, provides a new three-dimensional visualization method for morphological research of mouse brain neurons.

Synchrotron radiation can achieve high-precision 3D neuroimaging, which overcomes the disadvantages of traditional staining 2D histology, such as the damage of sections to specimens. This method combined with Golgi-Cox staining, can be used for the construction of regional neural networks and non-destructive research on the integrity of the brain.

Aluminizing is often used as a protective coating for aero-engine turbine blades, but the current nondestructive testing equipment can not achieve accurate measurement of the coating thickness. The uniformity of aluminizing thickness has a great impact on the performance of turbine blades and the safety and stability of engine.

This study aims to develope a nondestructive measurement method for the thickness of diffusion aluminized layer based on X-ray fluorescence absorption.

First of all, according to the theoretical calculation formula of X-ray fluorescence absorption, the linear relationship between the logarithm of the fluorescence intensity ratio (lnR) and the thickness of the fluorescence penetrating material (x) was obtained. Then a unitary linear regression model related to the main elements of K403 alloy and a multiple regression model based on the forward selection variable method were established. Finally, the comparative study of the prediction results of the two models was carried out.

The experimental results show that the multi regression calculation model is relatively stable in comparison with the unitary regression calculation model, and the average relative error of measurement results under different thicknesses is only 3.2%.

The proposed method provides a convenient and feasible guiding idea for solving the problem of measuring the thickness of diffusion permeable layer.

The radiation field of the radioactive inert gas generated in a nuclear power station is mainly β-γ mixed field. The phoswich detector has good ability to discriminate β and γ rays.

This study aims to develop a set of 4π phoswich detector to accurately measure the energy spectrum of the mixed field and distinguish β-γ rays.

The phoswich detector mainly consisted of a hollow cylindrical plastic scintillator EJ-200 with layer thickness of 1 mm and a cesium iodide scintillator CsI (Tl) with side wall thickness of 20 mm overlay on the outer layer of EJ200. The two photomultiplier tubes (PMT) were placed at both end surfaces of the cylinder respectively, and coupled by using silicone oil. The DT5790 dual digital pulse shape analyzer of CAEN company and the digital pulse processing-pulse shape discrimination (DPP-PSD) were used to obtain the detector output signals, and results were processed by ROOT and MATLAB software. Finally the detector was calibrated by use ¹³³Ba and ¹³⁷Cs radiation sources.

The β-γ signal discrimination is realized by mathematical methods in back end computer software, indicating that the detector is capable of measuring radioactive inert gas.

The cold cathode X-ray tube is a key component of static CT system. Since the discovery of super-aligned arrays of carbon nanotubes, their applications in X-ray tubes have been gradually developed.

This study aims to design and implement a cold cathode X-ray tube based on super-aligned arrays of carbon nanotubes.

The cold cathode X-ray tube was developed by designing a new X-ray tube structure and adopting a new X-ray tube preparation process. The acetylene was taken as carbon source gas, the super-aligned arrays of carbon nanotubes were prepared by using the chemical vapor deposition method. Then the cold cathode X-ray tube based on the technology of super aligned carbon nanotubes was developed by taking the nanotubes as electron source, and the performance of the developed X-ray tube was tested. Both the I-V characteristic curve and the X-ray focus size were obtained.

Results shown on the I-V characteristic curve indicates the maximum emission current of this developed cold cathode X-ray tube is 5.0 mA when the gate voltage is 2 373 V. The focus size determined by the pinhole imaging is 0.6 mm×1.6 mm. When the anode high voltage is below 130 kV, the X-ray tube works stably.

This study lays a foundation for the development of pulsed X-ray source and static CT security system.

During the uranium enrichment process, the generated wastewater containing fluoride and uranyl ions, has potential threat to aquatic ecosystem and human health. Thus, the uranium concentration in the wastewater and its secondary sources must be reduced to an acceptable value before being discharged into the environment.

This study aims to graft acrylonitrile (AN) and methacrylic acid (MAA) onto ultra-high molecular weight polyethylene (UHMWPE) fiber to synthesize amidoxime (AO)-based sorbent for removing uranium (U(VI)) from simulated wastewater containing fluoride ions.

The AO-based sorbent was prepared by radiation grafting of AN and MAA onto UHMWPE fiber, followed by amidoximation. The chemical structures and surface morphologies of the pristine and modified UHMWPE fiber were characterized by Fourier transformed infrared spectroscopy (FTIR), scanning electron microscopy (SEM) and X-ray photoelectron spectroscopy (XPS), respectively. The uranium adsorption performance of the sorbent was investigated by batch adsorption in simulated wastewater.

1) The uranium sorption kinetics and isotherm of AO-based UHMWPE fiber were in congruence with the pseudo-second-order model and the Langmuir model, respectively. 2) The maximum U(VI) adsorption capacity in simulated wastewater (U(VI): 40 mg?L^-1; sodium fluoride (NaF): 10 g?L^-1; sorbent dosage: 0.2 g?L^-1; time: 21 d) was 151.98 mg?g^-1, which is basically consistent with the theoretical saturated adsorption capacity (153.85 mg?g^-1). 3) The removal ratio of U(VI) in simulated wastewater reached 99.93% by increasing sorbent dosage (1.5 g?L^-1), and the residual uranium (28 μg?L^-1) can meet the national discharge standard (< 50 μg?L^-1).

At present, most of the physical protection systems of nuclear facilities at home and abroad use the path analysis method to evaluate the effectiveness. The performance of the system is evaluated by calculating the cut off probability, hence the evaluation index is single and cannot reflect factors such as the system reliability, the command strategy, the information security design and other factors.

This study aims to analyze the comprehensive effectiveness of the physical protection system by using neural network method.

The comprehensive effectiveness evaluation index system of the physical protection system was established by using the analytic hierarchy process (AHP), and the comprehensive effectiveness evaluation model of the physical protection system for nuclear facilities was established by using the composite ability of the nonlinearity of the BP (back propagation) neural network. The comprehensive effectiveness of the physical protection system for the marine nuclear power platform was evaluated.

The evaluation results show that the BP neural network evaluation model converges fast, avoids the shortcomings of the majority of sample concentration effects in the AHP evaluation process, and has better adaptability and stability. It provides a kind of field for physical protection effectiveness evaluation.

Boron neutron capture therapy (BNCT) is a promising targeted cancer radiotherapy. The neutron source is pivotal to BNCT. The epithermal neutron flux is one of crucial characteristics for the BNCT neutron source. Therefore, in order to evaluate the quality of the BNCT neutron source, neutron flux detectors have been developed to measure the epithermal neutron flux accurately in our previous work.

This study aims to design a moderator for the D-D neutron source for evaluating the performances of the developed BNCT neutron flux detectors.

The MCNP5 (Monte Carlo N Particle Transport Code, version 5) was employed for the BNCT moderator design based on the D-D neutron source. First of all, the neutron spectrum of a D-D neutron source passing through a single-layer moderating material with different thickness was simulated. Then combined calculation was carried out for three moderator layers configured by different materials with various thickness.

A suitable configuration of the BNCT moderator based on the D-D neutron source is obtained, that is, the combination of 5 cm thick polyethylene (PE) layer + 24 cm thick titanium trifluoride (TiF₃) layer + 22 cm thick magnesium fluoride (MgF₂) layer is adapted as the moderator material, 20 cm thick nickel (Ni) layer is employed as the reflector and 0.03 cm thick cadmium (Cd) sheet is used as the thermal neutron filter.

The moderated neutron beam can be used to evaluate the performances of the developed BNCT neutron flux detectors.

The enhanced X-ray Timing and Polarimetry mission (eXTP) is a space science mission designed to study fundamental physics under extreme conditions of density, gravity and magnetism. The spectroscopic focusing array (SFA) is one of eXTP's four payloads, and its main function is to achieve spectrum and timing measurements with high dynamic range and high signal-to-noise ratio (SNR). The SFA uses multi-pixel silicon drift detector (SDD) as the focal plane detector.

This study aims to design an SDD signal emulator for the performance test of SFA readout electronics.

The field programmable gate array (FPGA) was used as the core device of this emulator, making use of its abundant programmable logic resources to generate multiple groups of oscillatory loops. First of all, the configurable logic blocks inside the FPGA were designed as multiple oscillating rings, and its outputs were XOR-ed as true random numbers generator (TRNG). By Bernoulli trials, the uniform random numbers were converted into exponential pulse sequence. Then, the NIST (National Institute of Standards and Technology) suite was employed to test the quality of the TRNG output. Finally, the time interval distribution and counting rate characteristic of output pulses for the emulator were tested.

The sequence of true random numbers of 1 000 Mbits have passed the NIST randomness test. The time interval of the output pulses for the emulator follows the exponential distribution. The dynamic range of counting rate is 1.6 ks^-1 to 813.8 ks^-1. The voltage range of the output pulse is 2.5 mV to 50 mV, and the minimum time interval is 9.6 ns. The signal emulator can work steadily for a long time.

The SDD signal emulator proposed in this paper can meet the requirements of performance tests for the SFA readout electronics.

BaF₂ scintillation detector is commonly used in low intensity γ-ray detection experiments, and it is easy to pileup during high intensity γ-ray detection.

This study aims to investigate the performance of Yttrium doped BaF₂ (Yttrium doped with 1 at%， atomic percentage) scintillation detector with main concern of pileup and time resolution for high intensity γ-ray detection.

The energy resolution and time properties of Yttrium doped BaF₂ (Yttrium doped with 1 at%) scintillation detector were experimentally tested by using 0.511 MeV γ-ray generated after positron annihilation from ²²Na source. As a comparison, the normal BaF₂ detector was also tested.

The experimental results show that the ratio of fast and slow components of Yttrium doped BaF₂ crystal (1:1.1) is much higher than that of BaF₂ crystal (1:5), but the energy resolution of Yttrium doped BaF₂ detector (38%) is worse than that of BaF₂ detector (19%). For high intensity γ-ray detection, the Yttrium doped BaF₂ detector have better time resolution than the BaF₂ detector.

The decrease of the slow component in Yttrium doped BaF₂ crystal can shorten the time width of the detector output signal, hence reduce the signals' pileup and improve the time resolution of the detector in high count rate environment.

The molten salt fast reactor is one of the advanced nuclear energy systems which has attracted much attention because of its outstanding advantages in fuel proliferation, nuclear waste transmutation and inherent safety. Liquid fluorine salt or chlorine salt is usually used as fuel carrier salt and coolant in molten salt fast reactor, which has high breeding properties.

Based on the dual-fluid molten salt reactor core structure, a comprehensive simulation program SCALE (Standardized Computer Analyses for Licensing Evaluation) were adopted to simulate and calculate the U-Pu fuel breeding ratio of two kinds of fluorine-salt fast reactors and one chlorine-salt fast reactor under the same metal solubility.

The breeding ratios of different breeding and reflecting layers were simulated, and the changes of neutron energy spectra in fission and breeding zones were analyzed. The breeding ratio of chlorine salt fast reactor increases with the increase of the thickness of the breeding layer and the reflector layer, but the growth rate decreases. When the thickness of the breeding layer of the fast reactor is small, the variation of the breeding layer has little effect on the breeding ratio. When the thickness increases to 60 cm, the variation of the breeding layer thickness has almost no effect on the breeding ratio. The variation of reflector size has no effect on the breeding ratio. The critical state of the reactor core and the neutron energy spectrum in the fission region are not affected by the thickness of the breeder layer and the reflector layer.

At the same temperature and molar ratio, chlorine fast reactor has higher U-Pu fuel breeding ratio than fluorine fast reactor. This provides a theoretical basis for basic salt selection and size design for molten salt fast reactor diffusion.

The pipe of thorium molten salt reactor-liquid fuel (TMSR-LF1) has been designed to operate at 650 ℃ for a long time, therefore, it is extremely important to analyze its creep-fatigue damage. At present, the provisional creep-fatigue damage evaluation is only stipulated in ASME-BPVC-III-5-HBB, and the evaluation is too cumbersome to realize for complex piping systems.

This study aims to apply PepS to analysis and evaluate class 1 nuclear piping at elevated temperature.

The analytical predictions of stress components and associated linearization of stress results through wall thickness of the pipe paths were successfully benchmarked with those given by finite element analysis (FEA), and good matching results was obtained. Finally, the structural analysis and structural integrity assessment procedures for the design structure of TMSR-LF1 on PepS are performed.

The study establishes effective connection between the ASME evaluate method and the pipe analysis software and clarifies the evaluation method. Hence, the creep-fatigue evaluation of class 1 nuclear complex piping system at elevated temperature is realized.

The rapid development of small-angle scattering of X-ray (SAXS) has massively increased the speed of experiments to be performed and produced a large amount of raw data collected during each experiment. Although some software can available for batch processing, there are some difficulties in batch processing for deluged raw data, and there is no domestic batch processing available before.

This study aims to develop a package for single and batch processing to pre-analyze SAXS/WAXS patterns and the deluge data.

Based on MATLAB platform, a semiautomatic data processing program PreSAXS 1.0 was implemented. It composed of four main capabilities: background subtraction, beam-center determination, sample-to-detector distance (SDD) calibration and the one-dimensional integrated intensity curves obtaining from deluged two-dimensional patterns. A comprehensive standard database was developed for fast indexing, and the circle fitting method was used to simulate the scattering/diffraction rings to find beam-center and calibrated SDD. The interpolation method from MATLAB was embedded in the package to improve the calculation speed.

The beamcentre find, SDD calibration, and I-q curves are perfectly appropriate to Irena, Raw software. After pre-data analysis, the data are saved as ASCII code and can be used for director read or further analysis by other software.

PreSAXS 1.0 is easy to operate with user friendly interface, satisfies majority requirements of batch processing for SAXS data pre-analysis. However, PreSAXS 1.0 still needs to be improved, such as function extension, faster calculation speed and data analysis for GISAXS, etc.

In clinical application and scientific research of positron emission tomography (PET), the standard uptake value (SUV) is commonly used for disease assessment. In order to reduce the deviations of SUV among clinical PET images, standard uptake value ratio (SUVR) based on a specific reference region is substituted for SUV. Accurate SUVR is hardly achieved directly from PET images due to the difficulty of accurately segmenting reference regions from PET images with low resolution and high noise. The integrated positron emission tomography-magnetic resonance imaging (PET-MRI) can simultaneously acquire MRI images and PET images, hence the combination of MRI image and PET image is expected to improve the accuracy of SUVR.

This paper study aims to investigate the effect of MRI segmentation and PET segmentation on SURV calculation to improve the PET-MRI brain quantitative accuracy.

First of all, integrated PET-MRI was used to acquire MRI and PET images of 12-age-controlled healthy volunteers simultaneously. Then the MRI images and PET images were segmented into reference regions by the atlas-based method. Based on reference regions derived from MRI and PET respectively, the SUVRs of left / right cerebrum, cerebral cortex, cerebral white matter, frontal lobe, parietal lobe, temporal lobe and occipital lobe were calculated. For each group of SUVRs, the coefficient of variation (CV) of SUVRs was calculated to measure consistency level. Finally, quantitative analysis was performed to evaluate which segmentation method offered better SUVR consistency among volunteers.

CV of SUVRs based on cerebellar gray matter from MRI images and PET images were 0.020/0.021, 0.023/0.026, 0.031/0.028, 0.028/0.031, 0.033/0.036, 0.022/0.024, 0.044/0.045 and 0.052/0.055, 0.054/0.055, 0.058/0.063, 0.053/0.058, 0.063/0.065, 0.048/0.050, 0.059/0.065 respectively. When the cerebellum was chosen as reference region, MRI results and PET results were 0.017/0.019, 0.021/0.025, 0.029/0.025, 0.026/0.030, 0.032/0.036, 0.020/0.022, 0.044/0.045 and 0.046/0.050,0.049/0.051, 0.052/0.057, 0.048/0.054, 0.059/0.061, 0.044/0.045, 0.056/0.062 respectively.

Compared with SUVRs based on reference region segmented from PET images, SUVRs based on reference region segmented from MRI images have better consistency with smaller coefficient of variation.

Thorium-uranium sustainability of molten salt reactor usually relies heavily on online reprocessing technology, which is still immature at the present.

The study aims to use reactor-grade plutonium online to achieve²³³U self-sustaining and burn reactor-grade plutonium based on the FLiBe carrier salt and Th/²³³U starter reactor.

Based on Scale6.1, the program of molten salt reactor feeding and reprocessing sequence (MSR-RRS) was developed. The database ENDF /B-VII Library of group 238 was adopted for simulation.Single cell models were used to analyze the performance of²³³U self-sustaining and plutonium utilization at different molten salt volume fraction (VF) and neutron loss ratio.

It is found that²³³U self-sustaining could be achieved in a large range from 10%VF to 85%VF, and²³³U have the best breeding ability at about 43%VF, of which the utilization of plutonium is worst. When the VF becomes smaller or larger, the utilization of plutonium always becomes smaller. It is more beneficial to burn plutonium in the low VF region. When VF is 10%~15%, the burning ratio of plutonium is the highest, corresponding to 75%. In addition, the neutron loss ratio is positively proportional to the burnup of plutonium, and has little influence on the self-sustaining performance of²³³U.

Neutrons play an important role in nuclear detection applications, and make up for the shortcomings of X-ray detection insensitivity to the atomic number. In a specific energy range, the cross section of neutron reaction varies drastically with the neutron energy, and resonance occurs. The resonance spectrum of nuclides is unique, and qualitative and quantitative analysis of nuclides can be carried out based on this. Among them, the resonance energy range of medium-heavy nuclei is in the eV~keV energy range, which is an important part of neutron resonance analysis.

Due to the lack of conclusive articles on neutron resonance in medium-heavy nuclei, this paper summarizes the current research on neutron resonance.

By querying a large number of journal papers and books related to neutron resonance, the theory and application of neutron resonance of medium and heavy nuclear are summarized.

At present, there have been many research results, and they have been widely used in temperature measurement, imaging, non-destructive analysis, nuclear data measurement, and other aspects. In this regard, by reviewing the literature and books on neutron resonance, the research status of neutron resonance at domestic and overseas are summarized and analyzed, and its development tendency is forecast.

Radio frequency quadrupole (RFQ) field accelerator causes temperature rise and thermal deformation during high power operation. Among many multi-physical analysis methods for RFQ cavity, these models are limited to 2D RFQ section or 3D RFQ local structures whilst the fluid flow motion was ignored.

This study aims to propose a noval three-dimensional multi physical field numerical method for coupling analysis of RFQ cavity.

Combined with computational fluid dynamics (CFD) and heat flow coupling, electromagnetic, fluid, thermal, structure were integrated into electromagnetic multi physical field coupling analysis. A 81.25 MHz RFQ cavity was taken as sample for verification and the heat transfer effects of three (Standard, RNG, Realizable)k-ε turbulence models under the same experimental conditions were analyzed.

The results show that the CFD numerical simulation method can better simulate the steady-state heat transfer of RFQ accelerator. The feasibility and reliability of the method were verified by comparison with the experimental results. The variation of cavity temperature, structure and frequency under different parameters were analyzed using a suitable turbulence model. Hence this study provides an effective numerical simulation platform for the structural optimization and performance improvement of the accelerators in future.

120 MeV electron linear accelerator has high requirement for beam emittance, and the traditional optical collimation cannot meet the requirement.

This study aims to use beam-based alignment (BBA) method to achieve higher precision collimation and obtain better beam quality.

According to the configuration of 120 MeV electron linear accelerator, all of beam position monitor (BPM) measurement data of the corresponding position under different energy were collected, then, the dispersion-free steering (DFS) algorithm was used to simulate the correction of accelerator matching transmission section in different beam jitter conditions.

The simulation results indicate that the algorithm can minimize the transverse offset of quadrupole to 50 μm.

The DFS algorithm can reduce the trajectory distortion in the low beam jitter accelerator.

When an accident occurs in a pressurized water reactor (PWR), the radioactive nuclides are released into the environment and endanger the safety of the environment and the staff. Therefore, the analysis of the radioactive source term released into the environment after the accident is critical to the radiation protection of PWR.

This study aims to calculate and analyze the source term of the steam generation tube rupture (SGTR) accident in PWR power plant.

According to the frequency of the accident and the severity of the consquences, the accident was divided into two conditions: pre-accident the iodine peek release and accident concurrent iodine peek release. Calculation models were established for radio-nuclide migration and diffusion after accident, and the parameters of advanced pressurized water reactor AP1000 were used for calculation and verification with emphasis on the radioactivity of inert gases and volatile nuclides iodine in the environment. [Results and

The calculation results show that the radioactive source term calculated by proposed model in this paper is quite consistent with the design source term. The release activity of noble gas is in good agreement with the design source term under two accident conditions, but the release activity of iodine is quite different.

Computational time and calculation accuracy are key features of the application of neutron transfer method.

This paper aims to reduce the computational time on the premise of ensuring the accuracy of calculation.

Based on a neutron transport calculation program for hexagonal nodal, a fast solution method was proposed to solve the 1/6 core is solved with 60° periodic symmetric boundary conditions, and calculated in parallel. The natural decoupling between the response matrices corresponding to different energy groups and nodes, and the parallelism between non-overlapping spatial regions was taken into account to solve the response matrix equation. Based on message passing interface (MPI), parallel program of the response matrix construction was developed to solve the matrix equation. Calibration calculation for the TAKEDA4 benchmark was carried out by parallel computing.

The computational results show that the 60° periodic symmetric boundary condition and the parallel computing function can greatly improve the computational efficiency while ensuring the same calculation accuracy, and verify the correctness of the fast solution strategy. Compared with the full core serial calculation, the total calculation time overhead of the 1/6 core under 9 processors can be reduced by 26, 25, and 22 times under P1, P3, and P5, respectively.

A quick solution for hexagonal nodes is initially achieved.

An optical vortex is an electromagnetic wave with a phase that varies azimuthally along the direction of propagation. The generation and application research of X-ray vortices are becoming more and more popular in recent years.

This study aims to propose a generation scheme of a compact and inexpensive X-ray vortices source based on nonlinear laser Compton scattering (LCS).

Laser Compton scattering (LCS) was caused by high power circularly polarized light and relativistic electron beam. This LCS provided a practical method for generating tunable, near-monochromatic, well-collimated X-rays in a compact, relatively inexpensive source.

Photons with energy of 1 keV is produced by head on nonlinear LCS interaction of 6.5 MeV electron beam and 800 nm Ti: Sapphire laser beam.

Compared with synchrotron, X-ray vortices source based on nonlinear LCS is a compact and inexpensive source. It provides a practical method for the laboratory to research the generation and application of X-ray vortices.

In synchrotron radiation facilities, the insertion devices are installed on the storage ring to produce high brightness, high coherence, and adjustable polarization synchrotron radiation, which are composed of periodic permanent magnet arrays. The permanent magnets are continuously exposed to beam loss electrons due to the Touschek effect. As a result, the permanent magnets can be more or less damaged and deteriorates the magnetic field of insertion devices. Radiation-induced demagnetization of permanent magnets in the insertion device is one of the critical issues for synchrotron radiation facilities.

This study aims to simulate the irradiation field and radiation damage of the Nd₂Fe₁₄B permanent magnet insertion devices during the routine operation of the high energy photon source (HEPS).

First of all, Monte Carlo code FLUKA was utilized to calculate the relationship between the absorbed dose and demagnetization of Nd₂Fe₁₄B magnet array based on the Spring-8 experiment. Then the different electronic sampling modes were set to calculate the irradiation field in the insertion device according to the actual beam loss conditions at HEPS. Finally, the demagnetization and countermeasures were analyzed according to the spatial distribution of absorbed dose.

When HEPS works in top-up mode, the irradiation of the magnetic pole in the upstream end of the insertion device is more severe. Most of the absorbed dose (about 99.7%) is contributed by photon, electron and positron. According to the beam loss parameters by physical calculation, the expected life of the insertion device is 2.1 to 4.6 years.

Placing local shielding layer and increasing the gap are the effective measures to protect the insertion devices, hence slow down radiation demagnetization of permanent magnet.

Because of its superior performance, molten salt reactor has been selected as one of the candidates for the fourth-generation reactors, and has become the focus of international attention.

In the molten salt fast neutron reactor, we try to find the most suitable molten salt fuel scheme to achieve the purpose of high proliferation of fissile nuclides.

Based on the two-fluid cooling cycle scheme, independent cooling cycles of fission molten salt fuel and breeding molten salt, a comprehensive simulation program SCALE (Standardized Computer Analyses for Licensing Evaluation) was employed to achieve high breeding ratio (BR) in the molten salt fast reactor by utilizing the high solubility characteristics of thorium and uranium heavy metal salts in fluorinated or chlorinated molten salts at high temperatures. Three feasible molten salt fuel schemes (LiF+ThF₄+UF₄, NaF+ThF₄+UF₄, and NaCl+ThCl₃+UCl₃) were compared by calculating the neutron energy spectrum and the reactivity temperature coefficient. Influence factors of BR, such as the sizes of fission zone, breeding zone and ZrC reflective layer, the isotopic abundance of ⁶Li and ³⁵Cl in the molten salt, molten salt density error on the calculation accuracy, and the evolution of fissile nuclides with reactor operating time, were computationally analyzed.

The breeding ratio (BR) of reactor reaches about 1.2 in all three molten salt fuel schemes when the diameter and height of the fission zone are both 260 cm.

With optimization of the reactor geometry, and the composition and isotopic abundance of the molten salt fuel, a high breeding ratio of the reactor is achievable for the thorium-uranium fuel cycle of molten salt fast reactor.

Stirling heater is used for heat capture in the molten salt space reactor power generating system, and its heat transfer performance has a great influence on the output power and efficiency of the space reactor energy conversion system.

This study aims to improve the heat transfer performance of the heater by numerical simulation of designed fins in molten salt channel and porous Stirling internal thermal receiver.

The computational fluid dynamics program Fluent was employed to design and optimize the structure of the heater. Thermal-hydraulic characteristics of different fins and molten salt channel types, such as flow field, pressure field and heat flux distribution were simulated and compared.

Adding fins in the molten salt channel can improve the heat transfer performance of the Stirling heater, and with the increase of fin height and the decrease of fin spacing, the heat transfer performance further improves. Compared with radial inlet and tangential inlet, the double-radial inlet and outlet channels have more uniform heat flux distribution and better heat transfer performance.

The results of this study provide useful references for the design of the Stirling heater in molten salt space reactor.

The background count in airborne gamma-ray spectrum measurement is mainly caused by Compton scattering effects, small angle scattering of gamma rays in the sensitive volume of the detector and interference caused by electronic noise in the energy range of the energy spectrum. Background subtraction is one of the important tasks of instrument spectral resolution in airborne gamma-ray measurement.

This study aims to propose an improved background subtraction method to achieve accurate and efficient spectral analysis of airborne gamma-ray instrument spectrum.

As one of the best background subtraction algorithms, the statistics-sensitive nonlinear iterative peak-clipping (SNIP) algorithm with simple mathematical structure and reliable background subtraction effect, was easily affected by the parameters of peak width, and lead to longer running time due to slow convergence speed of iterative process. Motivated by the idea of the Simpson formula with higher algebraic precision, the second step of the SNIP iteration in the original algorithm was improved by using three points of interval bisection for integral interpolation. Comparison of SNIP algorithm before and after the improvement was performed for background subtraction of airborne gamma-ray spectrum with different iteration times and appropriate width of transformation.

The comparison results show that the accuracy and the overall operating efficiency of the improved algorithm are improved by 41.8% and 2.53%, respectively, compared with the original SNIP algorithm.

The improved Simpson-SNIP algorithm achieves fast convergence speed and better accuracy of calculation results while preserves the advantages of simplicity and efficiency of the original, hence better background deduction effect.