Based on the research of molten salt reactor (MSR), a conceptual design of small MSR core with thermal power of 100 MWt is proposed to meet the power supply demand of small area. By adjusting the initial fuel load of the reactor core, the reactor can operate at full power for 1 250 days without refueling, and then batch process fuel at the end of its life.

This study aims to analyze the yield and source of radionuclides in the main loop during such a small MSR operation by providing the constitutions, main components, and parameters according to the burnup characteristics and fuel salt characteristics of the long refueling cycle.

The calculation software KENOVI for three-dimensional Monte Carlo transportation program and burnup analysis module Origen-S were employed to analyze the fuel consumption analysis module, the storage of radioactive products in the main loop and the neutron energy spectrum and other neutron parameters.

The computation results show that the radioactivity at the end-of-life this small MSR is about 7.36×10¹⁸ Bq, and the radioactivity of fission products in the end-of-life primary loop is about 5.89×10¹⁸ Bq, of which the inert gases, iodine isotopes and the volatile fission metal account for 7.35×10¹⁷ Bq, 9.56×10¹⁷ Bq, 8.17×10¹⁷ Bq respectively. The total radioactivity of actinide nuclides is about 1.47×10¹⁸ Bq, of which the ²³⁹Np accounts for 98%.

This study provides reference for radiation protection design and fuel reprocessing scheme of molten salt reactor.

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 energy deposition behavior of β particles in transducer material is an important issue in the simulation of a betavoltaic battery, which is usually performed by Monte Carlo software Casino and MCNP (Monte Carlo N Particle Transport Code). However, it is difficult to simulate the continuous energy spectrum of β particles by Casino software. The commercial software MCNP is subject to many restrictions on its copyright, it is very necessary to study the energy deposition of β particles in transducer materials based on the universal open-source Monte Carlo program Geant4 (GEometry ANd Tracking, version 4).

This study aims to develop a model based on Geant4 to study the energy deposition of β particles in betavoltaic battery transducer materials.

Based on the framework of Geant4 software, a general simulation model of energy deposition behavior was developed. The continuous β spectrum, the self-absorption of radioactive source, and the backscattering of β particles on the source-transducer interface were taken into account completely in this model for the Monte-Carlo simulation study, and the computation results were compared with that of Casino software.

Simulation comparison results show that this simulation model can accurately simulate the energy deposition behavior of β particles in the transducer materials of betavoltaic battery.

Simulation model proposed in this study can be used to study the energy deposition of β particles in the transducer materials with advantages of convenient application, open software and easy extension to other source-transducer combinations.

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.

As one of the most important issues in the field of gamma spectrum analysis, the main function of spectrum stabilization is to suppress the "spectral line drift" caused by environmental changes. Portable LaBr₃(Ce) gamma spectrometer is mainly used to analyze and process the gamma spectra of radionuclides and obtain qualitative and quantitative information about radionuclides. It often works in the field complex environment where the spectrum drift is inevitable, hence must have the function of spectrum stabilization.

This study aims to improve the performance of spectrum stabilization technique for a portable LaBr₃(Ce) gamma spectrometer.

The intrinsic characteristic peak of LaBr₃(Ce) detector and the peak of ⁴⁰K in the natural background were made of as the reference peaks. Then, the concept of "spectrum similarity" was defined to calculate the similarity between the measured spectrum and the standard background spectrum. Finally, a new spectrum stabilization method using standard reference peak and "spectrum similarity" was proposed to measure and correct the deviation. Experiments were carried out verify the performance of this new method.

Experimental results show that the technology can effectively stabilize spectrum in the temperature range of -30 ℃ to 50 ℃ without calibration of the relationship curve between the reference peak and the temperature, and the peak drift of the reference peak is stable within ±1 channel even if the temperature changes more than 50 ℃?h^-1.

This technology effectively improves the application range of portable LaBr₃(Ce) spectrometers.

Beryllium bearing molten salt has strong chemical toxicity, hence needs to be accurately monitored and effectively removed at high temperature.

This study aims at the characteristics of particle size distribution of high temperature FLiBe molten salt.

Based on the experimental device of low pressure distillation of molten salt and Dekati low pressure impact sampler (DLPI), an experimental device for measuring the particle size of high temperature molten salt particles was set up, and the particle size distribution of FLiBe molten salt under different experimental conditions was studied. The effect of gas purging and different sampling time on the particle size distribution of molten salt particles under high temperature were investigated, and scanning electron microscope (SEM) was employed to examine the experimental measurement results.

The experimental results show that the particle size of the molten salt particles is concentrated above 2.5 μm at room temperature, while at high temperature, it is concentrated at 0.26~2.5 μm. The sampling time has no obvious influence on the particle size distribution of molten salt particles; although the final concentration of molten salt particles is changed by gas purging, it does not significantly affect the particle size distribution. The SEM results of the collected samples verify the accuracy of the particle size scale of the experimental device, and it is observed by SEM that the molten salt particles are in the shape of regular spherical rather than crystalline, which may be caused by the collision and deposition of molten salt particles.

Transuranic isotopes (TRUs) in the spent fuel of pressurized water reactor (PWR) contain more than 50% of fissile fuels and can be used as a starting fissile fuel for thorium fuel cycle.

This study aims to analyze the neutronic performances of a small modular thorium-based molten salt reactor (SM-TMSR) core with TRUs as fissile fuel and Th as fertile fuel.

Considering the batch reprocessing period and the limit value of the inherent safety, proper scenarios for thorium utilization in the SM-TMSR initiated by TRUs were proposed for the evaluation of the neutronic characteristics. Reactor analysis software SCALE6.1 was employed to analyze the influences of two important parameters (fuel volume fraction and initial heavy metal mole fraction) on the characteristics of neutron spectrum, ²³³U production, graphite life, temperature coefficient, TRUs consumption, etc.

The calculation results show that TRUs are an ideal fuel for starting Th-U fuel cycle of the SM-TMSR. Considering the constraints of the graphite's five-year life and the intrinsic safety (-5~-8 pcm?K^-1), the core with 6% heavy metal mole fraction and 25% fuel volume fraction can obtain a preferable fuel utilization efficiency. With the thermal power of 150 MW for 5-year full-power operation, the uranium production is about 16.0 kg?a^-1 and the transmutation rate of TRUs can achieve about 36.1 kg?a^-1, which can effectively reduce the spent fuel containing TRUs.

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.

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.

As a hot material, diamond has been widely used in various civil and military equipment with high hardness (Mohs hardness is 10), compressive strength (greater than 1.2 GPa), excellent thermal (thermal conductivity at room temperature is 20~22 W·cm^-1·K^-1, thermal expansion coefficient at room temperature is only (1.1~1.3)×10^-6 K^-1), optical (higher transparency to X-rays) and electrical properties. Traditional optical window material beryllium is harmful to the human body and industrial hazards, its hardness, thermal conductivity and the thermal expansion coefficient are much lower than that of diamond.

This study aims to explore the performance of micro X-ray tube with diamond optical window by simulation, obtain the best thickness of diamond for X-ray tube optical window and its shielding effect against low, medium and high energy X-ray.

The Monte Carlo method was used to calculate the effective transmission ratio, peak back ratio and transmission ratio of characteristic X rays with diamond optical window in high energy region. The X-ray tube with 50 kV tube voltage and 1.0 mA tube current was taken as an example to determine the optimal thickness of diamond.

Simulation results show that the K_α characteristic X-ray effective transmittance and peak to total X-ray ratio of the silver target increase continuously with the increase of the diamond window thickness, and the optimum diamond thickness is 2.0 mm. Under the optimal thickness of 2.0 mm, the effective transmittance is 154.5%, high-energy X-ray transmittance is 74.5%, and the characteristic X-ray peak back ratio is 27.9%.

Diamond can be used as a substitute for beryllium optical windows and has good application prospect.

Molten salt reactor (MSR), a candidate of the Generation IV reactors, has the advantages of inherent safety and good neutron economy. The circulation flow of fuel, on-line refueling and reprocessing are the three features of liquid fuel in MSR. Due to the features of liquid fuel, the fuel management regulations for conventional reactors that use solid fuels are not applicable to liquid-fueled MSR.

This study aims to develope a new fuel management code named ThorNEMFM to analyze liquid fuel management of MSR.

First of all, the self-developed deterministic code ThorCORE3D was coupled with the cross section processing module and a burnup calculation module. Then a fuel management analysis program ThorNEMFM was developed for liquid fuel molten salt reactor to realize on-line feeding and fission product processing. Finally, the ThorNEMFM was verified on molten salt reactor experiment (MSRE), molten salt breeder reactor (MSBR) and molten salt fast reactor (MSFR).

The results of benchmarks for MSRE, MSBR and MSFR indicate that ThorNEMFM has a high accuracy with benchmarks in comparison with the reference.

ThorNEMFM is suitable for calculation and analysis of fuel management of MSR.

Micro target X-ray source has attracted much attention in X-ray microscopy imaging. The focus spot size of X-ray directly affects the spatial resolution of sample.

This study aims to investigate the influence of target materials and target microstructures on the emission spot of X-ray tube by simulation. The influence of different target materials on the focus size of X-ray tube is also discussed.

Monte Carlo simulation code MCNP5 was employed to calculate the focus size with five different target materials, i.e., Mn, Mo, Silver, Rt, Wu and three different target microstructures, i.e., hemisphere, cone and cylinder. The full width at half maximum (FWHM) of the light intensity distribution line was used as the standard to measure the spot size.

The results show that the minimum focus size can be achieved by micro cone structure target at the incident electron energy of 100 keV. Moreover, different target materials have obvious influence on the focus size behind the target. The focus size decreases with the increase of the atomic number of the target material at the beginning, but expands to a certain extent with the further increase of the atomic number, showing a suppression trend. The influence of different target microstructures on the focus size is obviously related to the target type. The difference of focus size between cylinder and cone is up to 30 μm. Compared with the cylindrical tungsten target, the spot size decreases by 17.93% by using micro cone target structure and the silver target.

Compared with the traditional transmission X-ray tube, the micro target X-ray tube is easy to implement because it does not need strong electron focusing equipment. The methodology and the Monte Carlo simulation results are valuable for guiding the design and optimization of micro target X-ray tube source.

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.

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.

The phase-II beam line project is currently under construction at Shanghai synchrotron radiation facility (SSRF), and one of its important goals is to have the ability of pumping experiments with high time resolution. High time resolution experiments require the purity of the probe bunch to reach the order of 10^-5. However, the traditional bunch-by-bunch charge measurement based on the beam position monitor has a resolution of only 10^-4.

This study aims to propose and implement a bunch purity measurement system for high time resolution at SSRF.

Based on the time-correlated single-photon counting technology, a bunch purity monitor was developed and installed on the diagnostic beam line of SSRF. An evaluation experiment was designed to verify its performance and optimize this bunch purity measurement system.

The experimental results show that the resolution of the new bunch purity monitor can reach to 10^-7 in a few minutes.

The new bunch purity measurement system is applicable to measuring the longitudinal distribution of electron bunches,and studying the effect of filling patterns on purity.

Critical extrapolation is necessary during the restart of China advanced research reactor (CARR) after a long shutdown period. The photoneutron in the heavy water tank is not only the neutron source during the reactor restart, but also the background of the detector. During the extrapolation process, the background neutron makes the extrapolation curve "convex", and the extrapolated critical rod position is overtakes the actual critical rod position, so the extrapolation process is dangerous.

This study aims to optimize critical extrapolation by making use of neutron background value of detector.

Based on the control rod grid worth curve in CARR and extrapolation formula, the background value was obtained by fitting method to improve the extrapolation process. The critical extrapolation formula of control rod was derived, and the background value was obtained by fitting with the measured neutron numbers corresponding to different rod positions in the experiment. Then, the background value was applied to the correction of extrapolated critical rod position value.

The extrapolation results show obvious improvement, and the modified extrapolated rod position basically satisfies the linear relationship between the experimental rod position and the reciprocal of detection neutron count.

The data fitting method can get the background value of the detector, which can significantly improve the extrapolation process.

During the radioactive source inspection or the orphan source disposal, it is essential to estimate the activity of the radioactive source quickly and effectively. However, this may be difficult, especially when the source is under shield with unknown shield thickness.

This study aims to establish an activity estimation method of shielded multi-energy gamma-ray sources.

First of all, based on the relationship between the material attenuation coefficient and energy of the gamma-ray source, a method for the activity estimation of shielded multi-energy gamma-ray source was established. Then, influences of the position bias between the radioactive source and the detector, the characteristic energy selection, and the effect of statistical fluctuations were investigated by experiments. Finally, the optimal characteristic energy of common multi-energy gamma-ray sources and the maximum measurable shield thickness of the radioactive sources with different activities were analyzed.

The experimental results show that the relative errors between the estimated thickness and the real thickness is less than 4%, and the relative errors between the estimated activity and the real activity is less than 5%.

The method proposed in the paper can be applied to the radioactive source inspection or the orphan source disposal.

With the increasing complexity, integration level and data volume of nuclear instrumentation and control system, conventional data buses, such as peripheral component interconnect (PCI) or controller area network (CAN), are difficult to meet the needs of rapid data transmission.

This study aims to design and implement a Gigabit Ethernet transmission system based on the Xilinx's Zynq-7000 system on chip (SoC) and user datagram protocol (UDP).

First of all, the system architecture of UDP Gigabit Ethernet transmission system for nuclear instrumentation and control system was introduced in detail. Then the ModelSim toolkit was employed to verify designed system functions by simulation. Finally, the Gigabit Ethernet transmission system was implemented using ZC706 development board, and tested by experiment.

Experimental results show that the data transmission speed can reach 870 Mbps.

The Gigabit Ethernet transmission system can meet the actual transmission requirements of most nuclear instruments and control system.

In the synchrotron radiation experiment, the position of the beam and the stability of the flux seriously affect the result of experiment. Quantitative evaluation of beam stability is one of key problems in the experiment. It is essential to develop the accurate beam position measurement solution that meets the experimental requirements of the X-ray beams.

This study aims to detect beam position for synchrotron radiation X-ray experiments by using position sensitive semiconductor detector (PSD) miniature detector.

Pin-cushion PSD produced by Hamamatsu company of Japan was adopted for experimental test, and a rectangular beam spot of 5.9 mm×2.0 mm was taken for measurement on BL09B1 beamline of SSRF. When the X-ray beam come into the detector's sensitive region, a current was generated due to photoelectric effect collected by the four electrodes. The location of the incident X-ray beam was determined according to the value of the current given by the galvanometer. In addition, the change of the beam flux was monitored through the total current simultaneously.

Experiment data show that PSD-based miniature detector can meet the needs of synchrotron radiation experiment. For a rectangular spot with a size of 5.9 mm×2.0 mm, the minimum resolution in X and Y direction are 1.5 μm and 1.4 μm, respectively. The signal/noise ratio of PSD detector for beam flux monitoring is more than 4.3×10⁶.

The PSD-based miniature detector can effectively meet the monitoring requirements of spot position, spot size and luminous flux.

After the Fukushima nuclear accident, China's nuclear power authorities have put forward new requirements for the safety management of nuclear facilities in operation. A variety of radioactive inert gases are produced in nuclear power plant (NPP) accidents.

The aims of this study is to develope a special inert gas monitoring device for NPP.

With the combination of flow gas differential detection component and local processing/display unit, a fixed inert gas monitoring device was developed by integrating the relevant detection device, filter device, gas sensing instrument (flowmeter, pressure gauge), sampling pipeline, valve (stop valve and regulating valve), sampling joint and various components together on this fixed equipment platform. The sensitivity of the detector was predicted by using simulation software MCNP to ensure the satisfaction of requirements.

The fixed inert gas monitoring device developed in this study not only preserves all functionalities, optimizes the measurement index, but also reduces the volume and weight. Compared with the similar equipment, the integrated layout is convenient for maintenance and management with better performance index, smaller volume and lighter weight.

This device canbe employed to monitor the activity concentration of radioactive inert gas in the environment more quickly and accurately, find the leakage of radioactive gas in the nuclear power plant in time, and ensure the safe operation of the nuclear power plant and the radiation safety of the field personnel.

Super absorbent polymer (SAP) is a new type of polymer material. SAP is rich in hydrophilic groups, hence can absorb at least one hundred times of its own volume and swell while keeping the water from losing for a certain period of time.

This study aims to synthesize a new SAP for soil and forestry water conservation.

In this paper, sodium alginate (SA) was used as the raw material, acrylic acid (AA) was used as the monomer, and N,N-methylene bisacrylamide (MBA) was used as the crosslinking agent. SA-g-AA SAP with three-dimensional network structure was prepared by co-radiation grafting. The chemical structures and surface morphologies of the SA-g-AA SAP were characterized by fourier transform infrared (FTIR) spectroscopy, thermogravimetric analysis and scanning electron microscope (SEM).

The results showed that when the temperature was 65 ℃, the absorbed dose was 5 kGy, the mass ratio of AA to SA was 5:1, the amount of the crosslinking agent MBA was 1.25 wt%, and the AA neutralization degree was 75%, the maximum water absorption ratio of the SA-g-AA SAP reached 579.8 g?g^-1. After the water absorption reused for 7 times, its water absorption ratio maintained 92.5% of the initial value. It showed that the SA-g-AA SAP had good reusability of water absorption performance; after reaching the saturated water absorption ratio, water-retaining property could be retained for up to 15 days, which showed it had good water retention performance.

This SAP material shows potential application in water retention and desertification control.

Space neutrons can cause damage to the electronic systems of orbiting spacecraft and may pose a serious threat to the lives of astronauts on mission.

This study aims to accurately analyze the characteristics of the space neutron environment to ensure the reasonable design of spacecraft's anti-radiation reinforcement and the safety of astronauts.

Monte Carlo method based Fluka software was employed to simulate, analyze and design a composite structure neutron detector based on plastic scintillator. Simulation analysis from the aspects of probe size, structure position, detection efficiency, anti-coincidence structure, etc., was performed to determine the 0.5 mm, 3 mm gradient composite structure of the detector by using 14~25 MeV single energy neutrons for experimental verification.

The simulation and experimental results show that the designed detector can be used for space station cabin neutron environment and near-Earth space neutron detection.

This study provides theoretical and technical support for subsequent engineering load design.

Traditional γ energy spectrum analysis methods have the disadvantages of long time, many steps and low accuracy.

This study aims to propose an improved gray wolf algorithm based on qualitative analysis to obtain nuclide information faster and more accurately.

The convergence factor based on the inverted S-curve was adopted in the improved gray wolf algorithm to maintain the balance between the global search and the local search of the algorithm, and solve the problem of local optimum that the original gray wolf algorithm fallen into easily. A dimension-by-dimension update strategy was introduced to improve the accuracy and efficiency of optimization. Finally, this improved gray wolf algorithm was applied to experimental test on the gamma spectra of "Soil standard source" and "Marin cup standard source".

Experimental results show that the improved gray wolf algorithm has a better analysis effect on multi-nuclides mixed gamma spectra. The analysis speed is 59% faster than the original gray wolf algorithm.The analytical error of all nuclides in the soil standard source can be controlled within ±10%. The average absolute error of 8 nuclides is within 4%. Except for ²²⁶Ra and ²³⁵U in the Marin Cup standard source, the errors of other nuclides are less than 10%. The average absolute error of the 8 nuclides is within 7%.

The improved gray wolf algorithm provides a new method for rapid analysis of nuclides.

The β-γ coincidence technique is a key method to measure the radioactivity of radioxenon isotopes for comprehensive nuclear-test-ban treaty (CTBT) radionuclide verification. The β-γ coincidence spectrum is usually analyzed by the net count calculation method (NCC) which systematically subtracts gas background from sample spectra regardless of memory effect of the previous sample. Negative net gas background counts are very frequently observed in β-γ coincidence data from International Monitoring System (IMS) radionuclide stations. Gas background subtraction in such cases can potentially lead to overestimated of the activity concentration and even false positives.

This study aims to solve the problem of gas background subtraction in NCC.

First of all, the judgment condition of gas background subtraction was proposed to avoid negative net gas background counts. Then the memory effect of the previous sample was eliminated by modifying the processing flow in NCC algorithm for analyzing β-γ coincidence spectrum. Finally, historical β-γ coincidence data from IMS radionuclide stations were taken for verification of modified NCC and comparison with original NCC.

Comparing the analysis results of original NCC and modified NCC, the false identification number of ¹³¹Xe^m, ¹³³Xe^m, ¹³³Xe and ¹³⁵Xe is reduced by 63%, 68%, 72% and 26%, respectively.

The modified NCC algorithm can decrease the positive fault of identification of xenon isotopes and improve the accuracy of calculation of their radioactivity concentration.

In the shielding design of γ-ray, it is very important to select the appropriate buildup factor for the shielding effect. As far as the current data are concerned, when designing the shielding material, the effect of bremsstrahlung is considered for the single-layer materials, but not for composite materials.

This study aims to improve the accuracy of the shielding thickness calculation with the consideration of γ-ray buildup factors by simulation.

The Monte Carlo method was used to simulate the buildup factors in different γ-ray energies and different shielding material thickness. Buildup factors of single-layer materials and multilayer composite materials were calculated with the consideration of bremsstrahlung and compared with those without bremsstrahlung.

Simulation results show that the contribution of bremsstrahlung to the buildup factor is lower at low energies and thin shielding thicknesses for single-layer materials and composite materials, but it is more obvious at high energies and high shielding thicknesses. The buildup factors show a decreasing trend with the increase of γ-ray energy, and show an increasing trend with the increase of shielding thickness.

This study provides a reference and calculation method for the buildup factor for shielding design of high-energy γ-ray in actual situations.

Tokamak is a possible implementation of controlled nuclear fusion whilst the particle transportation plays an important role in maintaining the progress of nuclear fusion reactions.

This study aims to investigate the effects of ripple field and neoclassical tearing mode (NTM) on the migration of deuterium ions in Tokamak by numerical simulation. and provide a reference for the confinement of high energy particles.

The guiding center orbit code ORBIT was employed for numerical simulation. Effects of NTM and ripple field on particles with different positions, pitch angles and energies, were studied in details.

Simulation results show that the particle loss rate is high under the condition of high magnetic surface and low pitch angle. NTM mainly causes the loss of passing particles, while the loss of trapped particles is mainly caused by ripple field. When both perturbations are considered, the loss of particles is approximately the sum of losses by two perturbations acting alone. Ripple field will significantly increase toroidal angular momentum diffusion of trapped particles without affecting the passing particles, whilst NTM mainly influences toroidal angular momentum diffusion of passing particles. The particle loss situation basically keeps the original characteristics regardless of NTM frequency varition, while the momentum diffusion condition of particle loss changes with the increase of NTM frequency.

Thorium-uranium fuel cycle has the advantages of making use of abundant thorium resources, producing less nuclear waste and low toxicity, as well as nuclear non-proliferation. Molten salt reactor is an ideal type of nuclear energy utilization of thorium resources. The amount of thorium added in the core fuel can affect the thorium-uranium conversion performance in the molten salt reactor.

This study aims to understand the effect of thorium addition on the conversion performance of thorium and uranium in molten salt reactor.

Based on MCNP (Monte Carlo N Particle Transport Code) program, the core physical model was established. The neutron physical characteristics and thorium-uranium conversion characteristics of 2 MW (thermal power) liquid fuel experimental reactor under different thorium additions running at full power for 300 days were analyzed by using MOBAT (burnup code coupled with MCNP and ORIGEN2 using BATch language) burnup program. Analysis objectives included the energy spectrum changes at the beginning and the end of burnup, Xe equilibrium, as well as the consumption ratio and transfer characteristics, and the evolution of important nuclides in thorium uranium chain.

The results show that the consumption rate of thorium decreases with the increase of added amount. CR raises with the increase of Thorium addition, but the growth rate decrease gradually. The rate at which ²³³Th concentration reaches the maximum value is linearly related to the thorium mass, and the ²³³Pa has the same evolution trend as the ²³³Th. The higher the thorium content in the fuel, the higher the concentration of ²³³U, and the accumulation of ²³³U increases with the running time. Based on the radiochemical detection limit, 80 kg thorium can meet the minimum detection value after 4 days.

Based on the data of thorium-uranium conversion, the amount of uranium fuel needed to maintain the back-up reactivity, the limit of radiochemical detection and other factors, the experimental reactor can carry out the utilization of thorium in the order of tens to hundreds of kilograms and the verification experiment of Thorium-Uranium cycle, with the priority of about 80 kg. In addition, the more thorium was added, the more Uranium needed to be added. The increase of ¹³⁵I concentration will increase the equilibrium concentration of Xe. The more thorium is added, the harder the neutron spectrum and the higher the ²³³U output will be, which means that it is easier to achieve transformation or proliferation under the fast neutron spectrum.

The effect of high-energy electrons bombardment on the integrated circuit (IC) for aerospace can be studied by the simulated space electron environment device on the ground. In the device, the high-energy electrons are provided by the electron linear accelerator, and the output beam spot is only on the order of millimeters while the experiment requires a large-area uniform electron irradiation area.

This study aims to process the beam transmission line to meet irradiation area and uniformity requirements of short beam transmission line and large irradiation area.

Two-dimensional homogenized beam expansion scheme was designed with the combination of beam expansion and scanning. According to the beam parameters output by the linear accelerator, the solenoid, α-magnet, scanning magnet, quadrupole magnets, focus sextuple magnet and other beam transmission line elements were taken into consideration simultaneously. By adjusting the element parameters through the electron beam tracking code, the electron beam with different energies of 0.5 MeV, 2.5 MeV and 5 MeV were matched with the required irradiation area and uniformity on the irradiation plane.

Simulation results show that the beam uniformities in different irradiation areas for 0.5 MeV, 2.5 MeV and 5 MeV beamline are better than 90%.

The beam homogenization design proposed in this study for ground simulation of space electron environment meets the requirement of effect of high-energy electrons bombardment in extraterrestrial space.

The protection of secondary neutrons in the complex secondary radiation of the 100~250 MeV proton therapy room has always been the focus of shielding design. Currently, iron plates embedded in concrete are often used to reduce neutron transmission, but this will cause significant neutron reflection (toward the patient). And neutrons have a high radiation weighting factor which may cause harmful effects on patients.

The study aims to investigate the reflection distribution of secondary neutrons in the proton therapy room and its impact on patients.

First of all, a simple model of the proton therapy room was established, and Monte Carlo program FLUKA was used to calculate the effects of the secondary neutron reflection from different directions and at different embedding positions of the iron shield in the 245 MeV proton therapy room. Then, under the three irradiation modes of AP (Antero-posterior), PA (Posterior-anterior) and LAT (Lateral), the effective dose of the patient was calculated according to the energy of the reflected neutrons reaching the position of the patient whilst the neutron external radiation dose conversion coefficient given in report of ICPR (International Commission on Radiological Protection) Publication 116 was used.

The simulation results show that the main shielding wall contributes more than 95% to the reflection distribution of the secondary neutrons in the treatment room. The radiation dose level outside the treatment room is the lowest when the embedded depth of the iron plate in the wall is 0 (the iron plate is exposed in the room), but highest effective dose rates caused by reflected neutrons are 1.99 μSv?s^-1, 1.37 μSv?s^-1 and 1.00 μSv?s^-1, respectively, under the three irradiation modes of AP, PA and LAT. With the increase of embedding depth of iron plate, the effective dose rate of reflected neutron decreased gradually, but the decreasing trend becomes slower and slower.

The structure of the main shielding wall of the treatment room and the embedded position of the iron plate have a significant impact on the reflected secondary neutron energy spectrum and the patient's radiation health. This feature needs to be considered in the shielding design of the proton therapy room.

Tritium is one of the radionuclides released during the operation of nuclear facilities. The radiation hazard of tritium to human body is internal exposure, which is not only related to the activity of tritium, but also closely related to the chemical form and intake route of tritium. In view of the current standards involving tritium, there is an inconsistency in the control limit of tritiated water (HTO), as well as elemental tritium (HT).

This study aims to derive and standardize the concentration limits of HTO and HT, as well as the discharge activity concentration limits of HTO.

The development of tritium concentration control limits and the basis for the formulation of relevant standards were analyzed. According to the current radiation protection standards of China, combined with the latest biodosimetry parameters of tritium in publications of International Atomic Energy Agency (IAEA) and International Commission on Radiological Protection (ICRP), the control limits of derived air concentration (DAC) and water activity concentration for tritium were recalculated. When calculating the recommended limit of air concentration derived from HTO, the skin absorption was converted into the inhalation, and the complete skin absorption was set to half of the inhalation.

Since the dose conversion coefficient used in the current standard is close to that in the new international publications, the recommended limits of HTO are slightly smaller than that specified in the standard, and the corresponding values are basically in the same order of magnitude.

The derived results can provide reference for the standard formulation of radiation protection management. The applicable objects of control limits are also clearly exhibited and distinguished. These objects are only available for people, not for facilities or equipment.

Molten salt reactor (MSR) is an ideal advanced reactor with good neutron economy, inherent safety and online refuelling.

This study aims to evaluate the transient behavior of reactive insertion transients of MSR.

Based on the RELAP5/FLUENT coupling program and the establishment of the primary circuit model, the graphite-moderated channel type molten salt reactor was studied, and transient analysis was conducted when the reactor core was inserted with 0.000 1, 0.000 2, and 0.000 5 reactivities.

The results show that under the concerned reactivity insertion, the maximum temperature of the reactor core and the outlet temperature of the salt are both lower than the safety allowable limit due to the negative temperature reactivity of the reactor core.

This study verifies the good response ability of MSR to reactivity insertion events.