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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

The high energy photon source (HEPS) currently under construction in Beijing will be the brightest synchrotron radiation facilities in the world. The high energy electrons which generate the desired synchrotron radiation interact with matter (for example, residual gas molecules in the vacuum chamber), to produce a spectrum of bremsstrahlung photons with energies up to the electron energy – in HEPS 6 GeV. The bremsstrahlung is highly penetrating, and it can produce exotic reactions in any material that it strikes, leading to the production of other radiations, especially photoneutrons. The beam stops are made of either lead or tungsten of rectangular shape in order to prevent the gas bremsstrahlung when transmitting through beamline to an experimental enclosure.

This study aims to design a new gas bremsstrahlung absorber to reduce the dose equivalent rate by simulation calculation based on the spectrum of scattered photons and photoneutrons for HEPS beamlines.

Based on the technical specification, layout and related parameters of HEPS and the first optical enclosure (FOE), Monte Carlo simulation software FLUKA was mainly utilized to investigate the gas bremsstrahlung production in HEPS straight sections. A complete treatment of electron, positron and photon interactions at energies above 100 eV were calculated for the gas bremsstrahlung generated in the HEPS storage ring straight section and the gas bremsstrahlung transmitted through the beamline and scatters from components and the radiation emanating through the hutch wall.

The results show that the beam stop is effective in shielding photons, but the photoneutrons in HEPS beamlines is inevitable. The equivalent rate of neutrons outside the shield wall is still high. The maximum energy of neutrons in photonuclear reaction reaches upto 300 MeV. Giant resonance neutrons (0.1~5 MeV) account for 91% of the total flux of photoneutrons. Using polyethylene as local shield is a very effective method.

The combined shielding of lead and boron containing polyethylene can effectively reduce the neutron dose equivalent rate inside and outside FOE of HEPS beamlines.

Flow rate distribution is a significant study object in the reactor core design of China initiative accelerator driven system (CiADS). Reasonable flow rate distribution will ensure reactor core work safely.

This study aims to optimize flow distribution in the coolant inlet area of each fuel assembly of CiADS so that the outlet temperature distribution is flat.

Firstly, the top and bottom nozzle flow field were calculated by computational fluid dynamics (CFD) software. Secondly, based on the above results, the porous media model was employed to establish the flow and heat transfer analysis model of the whole reactor for numerical simulation of the flow distribution in the core, and the power coefficient was used as the reference for reactor core flow distribution.

These calculation results show that the flow distribution and power coefficients are almost consistent, and the outlet temperature distribution for the fuel assemblies has been flattened. Both the pressure drop characteristics and coefficient of resistance properties of the top and bottom nozzle obtained by simulation provide the necessary parameters for analyzing the flow distribution of the core.

Based on the calculations, the coolant inlet area of each fuel assembly has been optimized. These calculation results show that the flow distribution and power coefficients are almost consistent, and the outlet temperature distribution for the fuel assemblies has been flattened. Thus, the fuel assembly could be operated under the safety conditions. Meanwhile, these results can provide some reference data for the subsequent hydraulic analysis.

Pulse height weighting technology is a data processing method for measuring neutron capture cross section with C₆D₆ detector.

This study aims to measure the neutron capture cross section of ¹⁹⁷Au on the back-n target station of China spallation neutron source (CSNS) by using Pulse height weighting technology, and verify the feasibility and data processing of this method.

First of all, detection efficiency of the detector, the pulse height weighting function of C₆D₆ and other basic terms under different target conditions were simulated by Geant4 Monte Carlo code, so that the weighted detector efficiency was proportional to the gamma energy. Then experimental measurement of neutron capture cross section of ¹⁹⁷Au was carried out by the Au target calibration data in the experiment.

The experimental results show that the measured neutron capture cross section data are consistent with the ENDF/B-VIII.0 evaluation data. Inconsistent size of the carbon target and the sample target, and the increase of the proton beam power will cause a large error in the background subtraction.

Compared with the traditional continuous-wave laser heating method, the pulse laser heating (PLH) technology can minimize the heating time to prevent the potential chemical reactions in the sample chamber of diamond anvil cell.

This study aims to construct an in-situ pulsed laser heating method which can be applied in high pressure X-ray diffraction experiments.

Based on the continuous-wave laser heating system of the high-pressure beamline of Beijing synchrotron radiation facility (BSRF), a signal generator was used to synchronize the pulsed laser, charge coupled device (CCD) spectrometer and X-ray detector, made it capable to heat the sample with pulsed laser, and collect the thermal radiation spectrum and high-pressure diffraction data during heating. The temperature stability, repeatability and temperature gradient of Pt samples heated by pulsed laser under high pressure were measured, and the in-situ X-ray diffraction experiments were completed by using the cumulative time method.

Experimental results show that the axial temperature gradient is larger when the temperature is low, but decreases with the increase of temperature. The PLH high-pressure XRD experimental data of Pt with good quality are collected by using the in-situ X-ray diffraction patterns.

The in-situ PLH method for high-pressure X-ray diffraction experiments enables the pulsed laser to have heating ability at the high-pressure beamline of BSRF, and lays a foundation for the future research on the application of related methods at the new High Energy Photon Source (HEPS) in Beijing.

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.

Tune is one of the most important characteristic parameters of storage ring, it seldom changes over time, hence can be obtained by a simple harmonic analysis on the beam position data. In some particular process, tune will change with time.

This study aims to explore and optimize the time-varying tune calculation algorithms for the tune drift caused by the nonlinear effect of Lattice during the decay of the residual transverse oscillation damping, after the injection of the storage ring.

First of all, the applicability of the wavelet algorithm and numerical analysis of fundamental frequencies (NAFF) algorithm for time-varying tune calculation was analyzed, and their performances were on above physical process of nonlinear effect of the storage ring were compared. Then the better Morlet wavelet algorithm was employed to quantitatively analyze the drift of tune during the injection of the Shanghai Synchrotron Radiation Facility (SSRF) storage ring and evaluate the nonlinear intensity changes of the SSRF during different operating period.

Algorithm comparison results show that both algorithms can realize the dynamic analysis of time-varying tune, but the wavelet method shows better performance. Evaluation results of nonlinear strength change of SSRF storage ring show that when the data analysis window is selected as 6 000 cycles, a track offset change 0.3 mm, and tune change as small as 0.000 1 can be detected. During the period of storage ring injection, the tune changes obviously.

By applying the morlet wavelet algorithm, the complete trajectory of the tune drift can be tracked.

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.

During the Fukushima accident, the shortcomings of the existing UO₂-zirconium alloy fuel form in resisting severe accidents were exposed. Fully ceramic microencapsulated fuel (FCM) disperses tri-structural iso-tropic (TRISO) fuel particles in a SiC matrix and has the ability to contain fission products, hence can effectively improve the ability of nuclear fuel to maintain structural integrity under severe accidents. FCM fuel can reduce the risk of large quantities of radioactive materials leakage in nuclear power plants, it is one of the main research objects of accident tolerant fuel (ATF). Compared with the traditional UO₂ ceramic fuel pellets, the U loading of FCM fuel has better moderation ability and less U loading. When the FCM fuel is used in a commercial pressurized water reactor (PWR), it may cause the positive moderator temperature coefficient (MTC) and loss of inherently safety of the core.

This study aims to analyse the neutronics characteristics of commercial PWR using FCM fuel from the perspective of inherent safety and economy.

The UO₂-Zr alloy fuel assemblies in the form of standard AFA3G 17×17 grid were taken as reference objects. The NESTOR software independently developed by China national nuclear corporation (CNNC) was employed to analyse the neutron characteristics of FCM fuel (UN core) assembly in forms of 17×17 grid and 13×13 grid. The overall physical performances of the reactor core composed of FCM fuel (UN core) in the form of a 13×13 grid were evaluated.

Computational results show that FCM fuel (UN core) assembly in the form of 13×13 grid containing gadolinium poisons meets the requirements of under moderation, and the FCM fuel (UN core) assembly in the form of 13×13 grid can be adapted by large commercial PWR with negative MTC at the initial stage while the fuel consumption depth and cycle length of the first cycle core are basically equivalent to those of the reference core.

FCM fuel assemblies used in commercial PWR can initially meet the inherent security and economic requirements. Further research work will be carried out to optimize the analysis of core power flattening.

Facing the internationally serious shortage of ³He gas, a novel neutron detector based on ⁶Li+CdTe was proposed in 2018, which has high neutron detection efficiency and low γ sensitivity, showing a good application prospect.

This study aims to conduct an optimal design of the neutron detector based on ⁶Li+CdTe and explore its application to the special nuclear material monitoring system.

First, the structure and operating principle of a typical ⁶Li+CdTe unit was analyzed. Then the Monte Carlo simulation of MCNP6 was employed to optimize the key parameters of this detector. Finally, integrated analysis was performed using this model for the special nuclear material monitoring system.

The optimal thickness of the CdTe absorption layer is 10 μm. When the number of ⁶Li+CdTe detection units reaches 6 layers with the optimal ⁶Li layer of 60 μm, the detection efficiency is more than 50%. A detector system with an area of 4 m×32 cm, surrounded by polyethylene of 3 cm in the front and 10 cm in the side and back, can meet the requirements of the special nuclear material monitoring system.

The novel neutron detector with optimal design can be used as the alternative of ³He based neutron detector in the special nuclear material monitoring system.

Small modular fluoride-cooled high temperature reactor (SM-FHR), with inherent safety and high temperature output ability, will accelerate the diversified development of world energy and meet the demand of comprehensive utilization of energy.

This study aims to simplify the reactivity control of SM-FHR, one of the key design goals, by using burnable poisons of boron carbide to lower the excess reactivity of SM-FHR.

Based on the SM-FHR design model, burnup code MOBAT was used to analyze the excess reactivity curves under different loads, different particle sizes and different space distribution of burnable poisons in fuel compacts.

Calculation results show that an optimal scheme exists when volume ratio of fuel to burnable poison (F/P) is 52, and particle size is 200 μm with lower loading in some outer assembles. Its excess reactivity is reduced from 38 000 ×10^-5 to 2 500 ×10^-5, and the power peak factor after layout of burnable poisons is only 1.26. The burnup time is decreased, but it still meets refueling cycle expectation for more than 2 full power years.

The layout scheme of burnable poisons proposed in this paper makes the core burnup depth and power distribution flattened, hence improving the core safety.

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.

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.

Space nuclear reactor (SNR) has attracted more and more attention worldwide due to its significant advantages in deep space exploration. Different from the traditional liquid metal, gas and heat pipe cooling methods, molten salt coolant can dissolve fission materials with good heat transfer properties, hence can be used as a coolant in SNR scheme.

This study aims at the preliminary neutronics design for molten salt cooled SNR.

Based on the design scheme of SNR, Monte Carlo code SERPENT and database ENDF/B-Ⅶ.1 were employed for preliminary neutron design of SNR with molten salt cooling. The effects of different fuel, cladding materials, rod pitch on the kinf of fuel rods and effects of different molten salts, reflector materials on the k_eff of core were investigated. Finally, a preliminary SNR core scheme based on fluoride salt ⁷LiF-BeF₂-UF₄ (66.4-32.7-0.9 mol%) to cool UC fuel (²³⁵U, mass fraction at 80%) was proposed.

The results show that k_eff of fuel rods is closely related to the fuel material, it increases with the ²³⁵U enrichment of fuel and decreases with the rod pitch.

Different molten salts and reflector materials have great influences on k_eff of reactor core, but the ²³⁵U enrichment of UF₄ in coolant is not sensitive to k_eff.

Molten salt reactor has been selected as one of the candidates for the fourth generation reactors due to its superior performance, and molten salt fast reactor has become one of the hot research subjects in the world.

This study aims to find the most suitable molten salt fuel scheme in the uranium-plutonium cycle to achieve the purpose of high proliferation of fissile nuclides.

Based on the two-fluid cooling cycle scheme, independent cooling cycles for fission molten salt fuel and breeding molten salt, a comprehensive simulation program SCALE (Standardized Computer Analyses for Licensing Evaluation) was employed to calculate the neutron spectrum and reactivity temperature coefficient of three feasible molten salt fuel schemes (LiF+PuF₄+UF₄, NaF+PuF₄+UF₄ and NaCl+PuCl₃+UCl₃). High breeding ratio (BR) in the molten salt fast reactor was realized by utilizing the high solubility characteristics of plutonium and uranium heavy metal salts in fluorinated or chlorinated molten salts at high temperatures. Influence factors of BR, such as the sizes of fission zone, breeding zone and ZrC reflection layer, the isotopic abundances of ⁶Li and ³⁵Cl in the molten salt, as well as the dynamic change of BR with running time, were computationally analyzed.

The BR of two fluorinated molten salt schemes reaches about 1.06 whilst BR of the chlorinated molten salt scheme reaches 1.46 when the diameter and height of the fission zone are both 260 cm.

Combined with the phase diagram of molten salt, the variation curve of BR with the molar concentration of heavy metals and the variation curve of the maximum value of BR with the average operating temperature of molten salt, the operating temperature of molten salt, the molar concentration of heavy metals and the reactor BR can be be quickly determinated in the conceptual design of the molten salt fast reactor.

The molten chloride salt fast reactor (MCFR) has advantages of high solubility of nuclear fuel and fast neutron spectrum, which makes it possible to achieve high breeding and transmutation efficiency.

This study aims to investigate the neutronics characteristics of Th-U fuel cycle and U-Pu fuel cycle in a 2 500 MWth MCFR, including the critical characteristic, burnup evolution, and breeding and transmutation performance.

The ²³³U and ²³⁹Pu were used as ignition fuels of Th-U fuel cycle (U3+Th) and U-Pu fuel cycle (Pu9+DU) respectively, and their fertile materials were ²³²Th and depleted uranium (DU), correspondingly. The analysis of the transition modes TRU+Th and TRU+DU with TRU as ignition fuel are also introduced.

Considering the neutron absorption rate of actinides, neutrons produced per fission (ν) and the conversion ratio (CR), nuclear fuel cycle mode of U3+Th needs a variable feed of fissile fuel to maintain the critical operation of the core, whereas Pu9+DU, TRU+DU and TRU+Th can be operated for a long time without on-line refueling, that is 46 a, 50 a and 29 a, respectively.

TRU+Th fuel cycle mode is promising to MCFR due to its high self-sustaining breeding and high effective transmutation.

The self-amplified spontaneous emission (SASE) free-electron laser beamline, dedicated for live-cell imaging, is under construction in the Shanghai soft X-ray free-electron facility. The energy range of the beamline is from 100 eV to 1 000 eV and the beamline covers two branch lines, including frontend, attenuator, plane mirror PM1, variable-line-spacing grating PM2, monochromator, bendable mirror, Kirkpatrick-Baez (KB) mirrors and ellipsoidal mirror etc. The Kirkpatrick-Baez mirrors of the north branch line consisting of a plane mirror and two elliptical cylinder mirrors are used for biological-cell imaging. The surface of mirrors should be highly reliable and precise in the system of the free-electron laser beamline. Slope error of optical elements during the production processing is an important criterion for measuring the surface quality of optical elements. The slope error of mirror is associated with the height error, and affects the coherent and wavefront of the beam, and further affects the final light spot.

This study aims to explore the variation of the slope error on X-ray coherent and wavefront after propagating in the beamline, and choose appropriate mirror whose slope error has little influence on the beam.

The software SHADOW and SRW were employed for the simulation to trace the X-ray propagation in the KB mirror and the whole beamline.

When the slope error of plane mirrors is controlled between 0.1 μrad and 0.2 μrad, the slope error of elliptical cylinder mirrors is less than 0.4 μrad, and the height error is approximately 3 nm, the high-quality image with spot size less than 4 μm is obtained, satisfying the imaging needs of biological imaging experimental station.

Results of this study provide comprehensive detailed data for the optimal design of soft X-ray free-electron laser beamline.

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.