Prompt gamma-ray activation image (PGAI) is a non-destructive element imaging method for large volume samples. Most of PGAI platforms are located in research reactors, which limit their applications. From the perspective of in-field applications, attractive alternative neutron sources are isotope neutron source and neutron generator. However, the neutron fluxes of these sources are much lower than that of reactor neutron source, which leads a poor spatial resolution.

This study aims to solve this problem by implementing an approach based on multi coded-aperture collimators.

First of all, the Monte Carlo code MCNP5 was employed to calculate spatial distribution of Cl in a known sample, and the characteristic gamma rays were produced by the thermal neutrons absorbed by the sample. Then, 36 coded-aperture collimators with random holes were used to collimate gamma rays, and 36 gamma signals were collected by high-purity germanium detectors (HPGe). Finally, the imaging of Cl was reconstructed through these data and maximum likelihood expectation maximization (MLEM) algorithm, and the relative deviation (d_f) and structural similarity (SSIM) were chosen to evaluate the image quality.

The spatial resolution of the imaging is 1 cm×1 cm, and the relative deviation and SSIM between the reconstructed image and the original image are 0.065 8 and 0.952 1, respectively. After neutron self-shielding correction, the relative deviation and SSIM between the reconstructed image and the original image are 0.002 3 and 0.998 4, respectively, which shows a good agreement.

The proposed approach is efficient to measure the distribution of Cl element, hence for element imaging of plate samples, and the reconstructed image is consistent with the set sample image.

The Fukushima nuclear accident in 2011 exposed the shortcoming of high temperature oxidation resistance of zirconium alloy cladding. For this reason, the concept of accident tolerant fuel was proposed in the international nuclear fuel field. Cr-coated zirconium alloy cladding, as an accident tolerant fuel cladding near-term commercial technology approach, has received extensive attention.

This study aims to study the high-temperature oxidation behavior of Cr-coated Zr-4 alloys at different temperatures.

Cr-coated Zr-4 alloy was prepared by multi-arc ion plating, and oxidized in air atmosphere at 800?1 200 ℃ for 4 h. Scanning Electron Microscope (SEM) and Energy Dispersive Spectrometer (EDS) were used to analyze the surface and cross-sectional micro-morphologies of Cr coated Zr-4 alloy samples before and after high-temperature oxidation and the distribution of elements on the cross section. The orientation image microscopy (OIM), inverse pole figure (IPF) and pole figure (PF) of Cr coated Zr-4 alloy samples were obtained by electron backscatter diffraction (EBSD). The phase of the samples was obtained by glancing angle X-ray diffractometer (XRD).The effect of oxidation temperature on the microstructure, phase, Cr-Zr diffusion layer thickness and oxidation weight gain of Cr-coated Zr-4 alloy were investigated.

The results show that there are a large number of droplets of different sizes on the surface of the Cr-coated Zr-4 alloy samples prepared by multi-arc ion plating, and the Cr coating has a columnar crystal morphology and preferentially grows along the (110) crystal plane. After high temperature oxidation at 800~1 100 ℃ for 4 h, the surface of the sample is oxidized to different degrees, but the un-oxidized Cr coating still remains inside the coating, and no micro-cracks appear on the surface and cross-section of the sample. The thickness of the Cr-Zr diffusion layer increases linearly with the increase of the oxidation temperature, and the oxidation weight gain increases slowly. However, after high temperature oxidation at 1 200 ℃ for 4 h, the Cr coating on the surface of the sample was completely oxidized, a large number of cracks appeared on the surface and cross-section, and the thickness of the Cr-Zr diffusion layer and oxidation weight gain increased sharply.

Therefore, the Cr-coated Zr-4 alloy prepared by multi-arc ion plating exhibited good high temperature resistance at 800~1 100 ℃, while accelerated oxidation occurred at 1 200 ℃.

High-performance accelerators have higher requirements for operational reliability and stability. By analyzing the historical data that is routinely saved during accelerator operation, most failures can be judged. However, when some rapid failure processes occur, due to the insufficient granularity of the historical data stored conventionally, it is impossible to effectively analyze such rapid failure processes. When a failure occurs in a particle accelerator, fast acquisition techniques are needed to collect large amounts of data from various devices with precise timestamps. The failure occurrent process can be rapidly reconstructed by using these data to locate and judge the root cause of the failure. In order to obtain data accurately when a failure occurs, hardware devices with data cache function can be used at the front-end devices, and data can be locked and obtained in the synchronous trigger mode. That is, after receiving the synchronous trigger signal, data in the cache area of the front-end hardware device can be locked, and then read and stored.

This study aims to design a failure analysis system prototype based on the event-timing technique.

Two core parts of the prototype were implemented: global high-precision timestamp implementation and data assembly and acquisition analysis. As one of the key factors, the global high-precision time stamping of failure data was applied to analyzing failure causes. Based on a high-performance rubidium atomic clock and the event-timing system, high-precision time stamps were implemented in this prototype with synchronization accuracy better than 16 ns to provide global high-precision time stamps for time data. Structured data based on the normative type of EPICS 7 was adopted for assembling and publishing the data. Essential information, including the system name, the subsystem name, the device name, the device card number, the data sampling frequency, the event timestamp, and the latched data, was obtained from the structured data.

The prototype experiment results show that the failure sequence of different equipment can be distinguished by the obtained high-precision time data, confirming the high feasibility of our proposed failure analysis system.

The prototype designed in this study meets the requirements for rapid failure analysis of particle accelerators. And this prototype will be applied to the CSNS accelerator in the near future. In addition, it can also be applied to EPICS-based and event-timing based accelerator control systems.

The solid fuel thorium element molten salt reactors (MSR) have attracted more attention recent years. A3-3 graphite is chosen as the fuel matrix for MSR, thus its irradiation behavior and mechanical property is very important before the application.

The study aims to observe the irradiation defects and hardness of A3-3 matrix graphite after ion irradiation by slow positron beam and nano-indentation, respectively.

The matrix graphite of fuel elements was irradiated with 1 MeVXe ions to fluence of 5.8×10¹⁴ ions·cm^-2 and 2.9×10¹⁵ ions·cm^-2 respectively at room temperature. The slow positron beam and nano-indentation were employed to investigate the effect of Xe ions irradiation on vacancy defects and hardness of matrix graphite. The changes in irradiation induced defects distribution with depth and fluence were analyzed according to the obtained positron annihilation S parameters versus positron incidence energy or depth curves, compared to SRIM (Stopping and Range of Ions in Matter) calculation.

Results from slow positron beam measurement show that 1 MeV Xe ions irradiation in matrix graphite introduces a damage layer with depth of about 600 nm, and the damage peak locates at about 250~350 nm in depth, consisted with SRIM simulation. The S parameters in irradiation samples increase significantly compared to virgin sample, which suggests that a high concentration of vacancy-type defects appeared within irradiation damage layer. In addition, the S parameters increase with the irradiation fluence, which shows that the concentration or size of vacancy-type defects increases. The nano-indentation results show that the hardness of irradiated graphite matrix is enhanced.

The enhanced hardness of A3-3 matrix graphite after ion irradiation is ascribed to the pinning of basal plane dislocation by the high concentration of vacancy type defects introduced by irradiation, consisted with the slow positron beam analysis. Slow positron beam is a very sensitive tool to study the irradiation defects.

Digital measurement system based on ADCs (analog-to-digital converter) has higher requirement on the signal to noise ratio (SNR) of sampled data. Among all the factors, the jitter of sampling clock has the most prominent effect on SNR.

This study aims to design a clock circuit based on dual-loop phase-locked loop to reduce the jitter of digital measurement system input clock.

First of all, the influence of clock jitter on digital measurement system was analyzed. Then, the LMK04610 chip with dual loop PLL architecture of Texas Instruments was employed to design and implement a dual-loop phase-locked loop jitter cleaner circuit. The cores of this design were power supply design and the loop filter design. At last, the performance of the circuit was tested by using Rodschwarz phase noise analyzer.

After testing, the dual-loop phase-locked loop jitter cleaner circuit can reduce the jitter of the 62.475 MHz source clock from more than 7 ps to less than 2 ps with output frequency of 499.8 MHz. The SNR of the sampled data is close to the theoretical value.

Dual-loop phase-locked loop jitter cleaner circuit has a good result and can provide reference for designers of digital measurement system.

In recent years, with increased awareness of environmental protection and safety, the development of nuclear logging tools using non-chemical sources such as X-ray instead of chemical sources like ¹³⁷Cs has become a new trend. However, X-ray source usually has a lower energy level compared to chemical source, therefore the measurement accuracy is hardly satisfying the demand of density logging tool.

This study aims to investigates the detector spacing design of a X-ray source tool based on an existing multi-detector gamma density tool.

Based on a 215.9 mm diameter borehole filled with water where the logging tool was eccentrically placed in, Monte Carlo software Geant 4 was employed for the simulation of the X-ray density logging in the formation density range of 1.7~3.0 g?cm^-3. According to density sensitivity, detection efficiency and depth, a series of models of this logging tool with detector-to-source distance between 135 mm and 430 mm were simulated to analyze the detector responses. Finally, based on above data, the design of source spacing for detectors was determined for the X-ray density tool.

The finalized tool includes three NaI detectors with detector-to-source distances of 160 mm, 270 mm, and 344 mm, respectively. Simulation results show that the maximum wall detection depth reaches 120 mm with the vertical resolution of 344 mm, and the density measurement accuracy is 0.014 g?cm^-3.

The feasibility of developing a potential X-ray density logging tool is validated by this study, providing reference for future design of nonchemical source density logging tool.

At present, most of the contamination detection equipments for small items in domestic nuclear power plants are manually putting in and taking out the testing items, and the operation time is long and the steps are relatively tedious. According to the on-site demand of nuclear power and feedback on the use of similar imported equipment, contamination detection equipment based on the conveyor belt can effectively solve these drawbacks.

This study aims to design a small item γ pollution measuring instrument that can be used to transmit items with a conveyor belt, hence effectively improve the detection efficiency and save the cost of manual operation.

Semi-automatic two-channel control design scheme was adopted, one through the motor driver to control the conveyor belt motor running mode, the other control the digital circuit of the radiation detector and the corresponding electrical parts through the main control board. Linkage of contamination monitoring status and item convey was achieved by the status control board associated with the motor drive. The background counting rate of the equipment in the background environment, the counting rate of the radioactive source at rest in the center of the measuring chamber and the dynamic counting rate of the radioactive source moving through the measuring chamber with the conveyor belt were tested and analyzed.

The results show that the minimum net count value of the detector is 81.6% of the average count in the static state. The net value of the minimum peak value of the detector at motion state is 89.3% of the average peak count, and the minimum detectable limit is 111 Bq.

The test performance of prototype is better than the reference standard and meets the design requirements.

Solid oxide cell (SOC) is the core converter for hydrogen production by high temperature electrolysis of water vapor and hydrogen fuel utilization.

This study aims to develope two kinds of aqueous casting pastes of NiO-YSZ with different components for the batch preparation of SOC without the usage of a large number of organic solvents.

A 10 cm×10 cm large-scale full-scale cell was prepared by screen printing the hydrogen electrode functional layer, electrolyte layer, barrier layer and oxygen electrode layer with NiO-YSZ support film at one time casting of about 450 μm. The effect of dispersant on the microstructure of hydrogen electrode support and the stability of the pastes were analyzed by scanning electron microscope (SEM). The performance of the SOCs were tested by I-V curve and electrochemical impedance.

Based on the optimized NiO-YSZ supports, the prepared planar SOCs delivers a peak power density of 0.36 W·cm^-2 at 750 ℃. The electrolysis current density of SOC can reach -0.68 A·cm^-2 at 1.30 V in solid oxide electrolysis cell (SOEC) model.

The performances of the aqueous-based SOCs can be considered highly remarkable, thus supporting the success in scaling the fabrication of SOCs using more environ-mentally friendly processes than conventional ones.

Decommissioning of nuclear facilities generates large quantities of different types of radioactive material. According to the requirement of the waste packaging, most incompressible waste must be put into steel boxes directly, which makes it difficult to measure and obtain the activity of the radioactive material in the steel boxes.

This study aims to establish an approach for measuring radioactive waste in steel box based on the in situ objects counting system (ISOCS).

In this work, ISOCS was used to characterize the radioactive material in steel boxes. The corrections for the absorption coefficients and the geometry factors of the big bulky sources were calculated using the ISOXSW (ISOCS Calibration SoftWare) in situ efficiency calibration without a radioactive source software. The verification experiments were carried out using standard sources with similar size and geometry.

The results show that the measurement error of six symmetrical positions of steel box with standard ¹³⁷Cs and ⁶⁰Co radioactive source by ISOCS is less than 30%.

The study verifies that ISOCS software is able to accurately estimate the composition and activity of radioactive material in a steel box.

The debris motion is an important phenomenon of a high-altitude nuclear detonation, which is also a foundation for the study of the geophysical phenomena such as the ionosphere effect and artificial radiation belt.

The study aims to clarify the debris motion characteristics and laws from a near-space nuclear detonation.

Firstly, a fluid dynamics model of debris motion from a near-space nuclear detonation was established. Many influence factors were considered, such as the variation of energy dissipation, air density varies with height, gravity, air temperature rise caused by X-ray depositions and radiation cooling. Then the parameters of debris motion within the explosion equivalent of 1 kt~10 Mt and the explosion height of 30~80 km were systematically studied. The evolutions of parameters such as center height, horizontal radius, expanding velocity, ascending velocity, and shape of debris were given. Finally, the variation laws of typical characteristic parameters such as maximum ascending height and expanding radius changing with explosion height and explosion equivalent were summarized.

When the explosion height is 30 km, the maximum rising height and the maximum horizontal radius at 5 min for a kiloton-level nuclear explosion debris are about 13~16 km and 4~5 km, the maximum rising height and the maximum horizontal radius at 5 min for a megaton-level nuclear explosion debris are about 20~40 km and 15~30 km. When the explosion height is 80 km, the maximum rising height and the maximum horizontal radius at 5 min for a kiloton-level nuclear explosion debris are about 30~50 km and 20~40 km, the maximum rising height and the maximum horizontal radius at 5 min for a megaton-level nuclear explosion debris are about 200~400 km and 110~220 km. When the explosion equivalent is small and the explosive height is low, the debris evolves into a flat ellipsoid. When the explosion equivalent is large and the explosion height is high, the debris evolves into an inverted pear shape.

The results show that the maximum height, horizontal radius, and speed of the debris cloud increase with the increase in the explosion height and explosion equivalent. The changes of the height, the horizontal radius, the rising time, and the shape of the debris obtained from the study are in good agreement with the literature estimation method. Those obtained motion parameters of debris can provide delayed radiation source information for the study of the geophysical phenomena such as the ionosphere effect and artificial radiation belt of nuclear explosion in near-space.

At present, in the key stage of the construction of China Initiative Accelerator Driven System, various research institutes have also put forward corresponding core schemes for different purposes, including the accelerators drive advanced nuclear energy system (ADANES) proposed by Institute of Modern Physics, Chinese Academy of Sciences. Detailed calculation and analysis of its scheme can provide strong technical support for the sustainable development of nuclear energy and the national energy security strategy.

This study aims to analyze the steady-state neutronics characteristics of ADANES reactor with emphasis on the preliminary typical transient analysis under the typical accident conditions of fast reactor.

The NECP-SARAX (Nuclear Engineering Computational Physics Laboratory, System for Advanced Reactor Analysis at Xi'an Jiaotong University) code system, which was based on deterministic neutron transport theory, was applied to perform the detailed analysis. The main design parameters, including core length, neutron spectrum and reactivity feedback coefficients, were calculated under various fuel types, coolant types, and reactor geometry parameters. In addition, the primary transient characteristics were simulated, including unprotected transient over power and unprotected loss of flow transient. The changes of reactor power and the maximum fuel/coolant temperature were obtained and analyzed.

The steady-state calculation results show that ADANES could achieve 10 effective full power year for each selected case. The reactivity feedback coefficients reach -5.4×10^-5 K^-1 in total so that the core has inherent safety under typical accident condition conditions during the simulated transient.

In accelerator driven sub-critical system (ADS), the high-energy proton beam produced by accelerator is used to strike the target nucleus, and generate spallation neutrons as external neutrons to drive and maintain its operation. The power level and the safety of ADS are susceptible to the instability of proton beam, such as beam overpower (BOP) which is treated as the typical transient accident for ADS system. In ADS core, the sudden increases of power and temperature will be caused by BOP accident, which may exceed the safety limits of materials, threatening ADS safety.

This study aims to investigate the transient safety characteristics of the eXperimental accelerator driven system (XADS) under BOP accident.

The multi-physics coupling code MPC-LBE, in which the fuel pin heat conduction (HC) model and the point reactor kinetics (PK) model were coupled with self-developed computational fluid dynamics (CFD) code, was employed to simulate BOP accident of XADS. Firstly, the MPC-LBE simulation model of XADS was constructed and then the steady state condition was established. On this basis, the double and triple BOP accident cases were simulated, and the safety boundary of BOP accident conditions was also evaluated.

For the simulation results, in the double and triple BOP cases, the reactor powers increase to only 1.88 and 2.7 times of the original ones, respectively. The maximum temperature of the cladding is about 843 K in the triple BOP case, exceeding its safety limit.

Conclusions can be drawn that the negative temperature feedback effect plays an important role in protecting the reactor from power sharp rise, and the double BOP case can be treated as the BOP safety boundary of BOP accident in XADS.

The energy produced by nuclear fusion on a Tokamak device is mainly exhausted through the divertor, its service life is directly affected by the interaction between huge heat flux from core and the divertor target. The large amount of impurity produced by the heat flux hitting the target leads to the reduction of the plasma confinement performance whilst pumping is an important means to control plasma density and impurity density.

This study aims to investigate the influence of pumping on the heat load of the target plate which is of reference significance for the future experiment.

Based on the experimental parameters of the HL-2A Tokamak, SOLPS-ITER code was used to study the effect of pumping on the heat load of the divertor target under different upstream electron densities. Analysis was performed through density scanning to find the sensitive threshold whilst and atom-molecular collision process was applied to the effect of pumping on the distribution of plasma and neutral particle parameters in divertor region at different upstream electron density.

Density scanning results show that pumping near the detachment threshold (TetOSP~5 eV) has a greater effect on the thermal load of the target plate. When the pumping rate is 12 m³·s^-1, 36 m³·s^-1 and 96 m³·s^-1 respectively, the miss threshold and thermal load peak of outer target plate are 1.11, 1.24, 1.39 and 1.37, 1.96, 2.54 times of those without pumping respectively.

It is found that the decreases of deuterium molecular density results in the energy of the collision reaction power decreases when the upstream electron density exceeds the detachment threshold, leads to the increase of the temperature and energy flow of the plasma in the target plate.

Narrow rectangular channels are widely used in major thermal flow fields because of the compact structure and large heat transfer area.

This study aims to improve the prediction method of critical heat flux in the narrow rectangular channel and establishing the critical heat flux (CHF) mechanism model for the enhancement of reactor safety and economy.

CHF experiments was carried out in the present study to identify the dominant mechanism in a narrow rectangular channel at different gap sizes. The visualization experiments were performed at pressures ranging from 1 MPa to 4 MPa, inlet subcooling from 60 K to 120 K, and mass flux from 350 kg·(m²·s)^-1 to 2 000 kg·(m²·s)^-1.

According to the visual experiment results, two typical bubble behaviors are investigated in the narrow rectangular channel. Based on the bubble dynamics characteristics of narrow rectangular channels, a new CHF mechanism model was proposed, and a set of constitutive relations will be provided to close the developed model.

A comprehensive assessment of new model has been conducted and analyzed by using the experimental data for the upward flow in a vertical narrow rectangular channel and it has good accuracies of less than ±30% as relative to the experimental values.