The preparation of alumina ceramics is usually sintered under atmospheric pressure, and the sintering temperature is usually above 1 800 ℃ to achieve densification of alumina. High temperature and long-term heat preservation often make the grains of alumina ceramics overgrow, and it is impossible to predict what defects and porosity that occur inside the crystal, resulting in the decline of the comprehensive properties of alumina ceramics. Previous work predominantly focuses on studying the stable nano α-phase alumina (α-Al₂O₃) sintering process, microstructure, and abnormal grain growth, but few studies on the variations in microstructure and defects during the sintering process of metastable γ-phase alumina (γ-Al₂O₃), let alone use CaO as an additive in nanoceramic alumina.

This study aims to apply positron annihilation lifetime spectroscopy combined with X-ray diffraction (XRD) to investigating the impact of calcium oxide (CaO) addition on the sintering process of Al₂O₃ nanoceramics.

Firstly, γ-phase nanoscale alumina powder with an average grain size of 20 nm, 99.99% pure and nano-calcium oxide powder with an average grain size of 160 nm, 98% pure was used as raw ceramic materials. Tablet samples of Al₂O₃ nanoceramics and Al₂O₃ nanoceramics doped with CaO (with a mass fraction of 1% CaO) were pressured under consistent experimental conditions. Then, the samples were sintered from 500 ℃ to 1 100 ℃ for 2 h, and cooling to room temperature. XRD was used to characterize the phases and compounds of the sample, and field emission scanning electron microscope (SEM) was employed to observe the microscopic morphology of the sample cross-section, accompanied by an EDS spectrometer for elemental composition analysis. Finally, a radioactive isotope ²²Na with intensity of about 0.5×10⁶ Bq was used as positron source and fast-fast coincidence positron lifetime spectroscopy was applied to characterizing the sample defects. The total count of each lifetime spectrum was over 10⁶, and all samples were measured for 2~3 times. The lifetime spectrum data were fitted and analyzed by LT(V9) program. The average value was taken as the result after spectral deconvolution.

The results of XRD and SEM show that the sintering process of CaO/Al₂O₃ nanoceramics is divided into two stages, no phase transition occurred from room temperature to 900 ℃, while significant phase transition occurred from 900 ℃ to 1 100 ℃. The addition of a small amount of CaO (such as 1% mass fraction) is uniformly distributed in the Al₂O₃ matrix at first, and with the increase of sintering temperature, it reacts with the Al₂O₃ to form a second phase and transform into a liquid phase at higher temperature. Analysis results of positron annihilation lifetime spectrum show that the size and number of vacancy clusters and micropores in CaO/Al₂O₃ nanoceramics before and after the phase transition are different from those of Al₂O₃ nanoceramics. In CaO/Al₂O₃ nanoceramics, vacancy clusters and micropores are more likely to form as the temperature rises, some micropores gradually merge to form macroscopic pores before phase transition whilst micropores in Al₂O₃ nanoceramics disappear with phase transition and grain growth, leaving only a small amount of micropores on the sample surface.

The addition of a small amount of CaO suppresses the growth of Al₂O₃ grains during the sintering process, resulting in a more uniform and denser structure. Simultaneously, the addition of CaO within Al₂O₃ nanoceramics induces the formation of a liquid phase. This alters the mass transfer mechanism from solid-state diffusion to liquid-phase flow, thereby delaying the phase transformation process of Al₂O₃ nanoceramics during sintering and leading to the formation of macroscopic pores.

Facilitating object imaging through the utilization of cosmic-ray muons mandates the precise delineation of muon trajectories, where the pinpoint localization of muon impact points assumes paramount importance for effective muon track reconstruction. Existing muon track detection systems necessitate the integration of multifaceted electronic channels to attain meticulous positioning of muon impact points. The construction of such detection systems is distinguished by its intricacy and entails substantial associated costs.

This study aims to achieve a design for a muon track detection system that is characterized by simplicity, low cost, and high precision.

The Geant4 software was applied to the simulation of detectors comprising square and circular plastic scintillators coupled with silicon photon multipliers (SiPMs) without segmentation. The SiPMs was used to collect the number of photons and the time triggering SiPM responsed as characteristic parameters in the simulation, and a uncut square and circular plastic scintillator detector with an area of 200 mm × 200 mm was constructed, with a thickness of 10 mm. The surface was coated with a TiO₂ reflective coating with a thickness of 0.11 mm and a reflectivity of 95%. Then, three types of artificial intelligence regression algorithms, i.e., extreme gradient boosting (XGBoost), multilayer perceptron (MLP) and long short-term memory (LSTM), were employed as the method for muon localization.

The simulation results demonstrate that LSTM algorithm achieves the highest accuracy among the three regression algorithms when photon number is considered as the characteristic parameter. Specifically, under the LSTM algorithm, the position resolution of a configuration comprising 12 SiPMs coupled to the upper surface of the detector can attain a resolution at the centimeter level. Furthermore, by employing photon number and trigger time as characteristic parameters, the position resolution of a setup involving only 6 SiPMs coupled to the side of the detector also reaches the centimeter level. Remarkably, these results align with the experimental findings obtained from a detector equipped with a photomultiplier tube (PMT) coupled to a large-area plastic scintillator.

This study employs the LSTM regression algorithm as the muon localization method, proposing a detector system structure for plastic scintillators with 6 SiPMs coupled to the side. The proposed structure is characterized by simplicity, low manufacturing cost, and achieves a positioning accuracy at the centimeter level.

A fuel rod is a fundamental unit of a fuel assembly, and it directly impacts the safe operation of nuclear reactors. To efficiently detect internal defects in fuel rods, a high-resolution visual nondestructive testing method, X-ray imaging, i.e., digital radiography (DR), is employed.

This study aims to address the issue of low contrast in fuel rod X-ray DR images by proposing a brightness fusion and multiscale optimized enhancement algorithm.

First, logarithmic and gamma transformations and further refined by incorporating local information fusion were employed to correct the brightness of fuel rod DR image. Subsequently, a wavelet function was applied for multiscale decomposition, enhancing and sharpening low-frequency components with Retinex, and non-local means (NL Means) was applied to filtering high-frequency components. Then, image enhancement was realized via wavelet reconstruction. Finally, quantitative analysis experiments were conducted using the DR images of fuel rods to evaluate the performance of the algorithm by means of two representative image quality assessment metrics, i.e., average gradient (AG) and information entropy (IE), and compared with that of new low-light image enhancement (NLIE) algorithm, homomorphic filtering (HMF) algorithm, and low light image enhancement (LIME) algorithm.

The experimental results demonstrate that quality of fuel rod DR image is significantly improved by image brightness fusion and multiscale optimized enhancement algorithm proposed in this study with the highest information entropy (IE) of 6.834 5, which is 10.2%, 3.3%, and 12.6% higher than NLIE, HMF and LIME algorithms, respectively, hence the internal defects of fuel rods are better highlighted.

This algorithm proposed in this study not only effectively improves the overall and local contrast of the fuel rod DR image, but also significantly highlights edge details, verifying its effectiveness in improving the quality of X-ray DR images.

Large-volume container waste generated from nuclear facility decommissioning exhibits extremely inhomogeneous nuclide distribution. Its activity measured using the traditional tomographic gamma scanning (TGS) technique shows errors of up to six orders of magnitude; this hinders its safe disposal.

This study aims to develop an improved tomographic gamma scanning (ITGS) technique based on the equivalent surface source method and establish a reconstruction algorithm.

Firstly, based on the count rate ratios of different γ-scan positions, the basic distribution of nuclides in the waste container was calculated, and the accurate activity reconstruction results were obtained by the combination of efficiency calibration. Then, γ-scan experiment was simulated for measuring various radioactive source distributions of Co-60 nuclides in large waste containers to verify the accuracy of this method, and the influence law for the nuclide distribution and measurement parameters on the reconstruction results was discovered.

Simulation results demonstrate that a deeper radioactive source location or larger medium density can increase the activity reconstruction error. However, based on the same measurement time, the ITGS measurement error can be controlled within an order of magnitude compared with the conventional TGS.

The proposed method achieves higher reconstruction accuracy under complex conditions such as multiple nuclide distributions and high medium density. Furthermore, it mitigates waste disposal safety hazards to humans and the environment by facilitating rational arrangements of the measurement system and activity reconstruction of radioactive waste containers.

The energy selection system (ESS) is a key part of the cyclotron-based proton therapy facility. In addition to regulating the energy of the fixed-energy proton beam generated by the cyclotron to match the depth of the tumor, the other main role is to control the beam emittance to meet the therapeutic requirements. Existing schemes for controlling the emittance are based on switching combinations of collimators with different apertures, where the adjustable emittance is limited by the number of combinations of apertures, and the need for a small aperture collimator upstream or downstream of the ESS to suppress the beam intensity limits the possibility of increasing the high-energy beam intensity.

This study aims to propose the beam optics design of proton therapy energy selection system (ESS) on the basis of the superconducting cyclotron CYCIAE-230 under development by the China Institute of Atomic Energy (CIAE) that can flexibly control the beam emittance and transmission.

First of all, the main optical design features of this ESS was described, and the Monte Carlo method based FLUKA code was applied to calculating the beam emittance, energy, and energy spread of a proton passing through a degrader with a set of collimators. Then, beam optical software TRANSPORT and TURTLE were combined to design the linear beam optics with consideration of placing a set of width-adjustable slits in front of the achromatic section and combining the beam optical conditions at the corresponding positions to control the beam divergence. Finally, the feasibility of flexible and adjustable beam emittance and transmission efficiency for esigned ESS was verified by calculations.

The results show that the ESS meets the demand of proton therapy, and the rational use of adjustable slit can make the ESS have the characteristics of flexible and adjustable beam emittance, so as to realize the adjustable beam transmission in the energy selection section. This simple and flexible optical design proposed in this study for adjusting the state of the beam can be used for regulating the beam intensity for proton therapy and other research applications, which has a wide range of application prospects.

¹⁴C has become the largest contributor of annual effective dose to the surrounding public in radioactive effluent during normal operation of nuclear power plants whilst ¹⁴CH₄ is the largest and most chemically stable of the ¹⁴C hydrocarbons. However, an effective treatment method for ¹⁴C in the form of hydrocarbons has not yet been established.

This study aims to investigate the treatment performance of a non-thermal plasma/catalytic coupling system for low concentrations of ¹⁴CH₄.

Considering the optimization of radiation protection, ¹²CH₄, which possesses the same physicochemical properties as ¹⁴CH₄, was selected as the experimental subject. The Pd/γ-Al₂O₃ catalyst were prepared by segmental heat treatment and wet impregnation. The discharge behaviors of the plasma were analyzed by the Lissajous figure. The CH₄ treatment performance of the non-thermal plasma before and after the introduction of the catalyst under different discharge voltages, gas flow rates and temperatures were analyzed by the constructed experimental system. The microstructural changes of the catalyst were analyzed by N₂ adsorption-desorption isotherm, X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS).

The introduction of Pd/Al₂O₃ catalyst can significantly improve CH₄ treatment performance in non-thermal plasma, and the CH₄ treatment efficiency and CO₂ selectivity can reach 100% and 83.7% by increasing the discharge voltage, lowering the reaction temperature and reasonably adjusting the gas flow.

The non-thermal plasma and the Pd/Al?O? catalyst exhibit a synergistic effect. The Pd/Al?O? catalyst can reduce the reaction barriers, optimise the reaction paths and significantly improve the CH₄ treatment performance of the plasma. Furthermore, the plasma-excited reactive species are conducive to the formation of PdO, which is the key reactive phase in the catalyst.

High dose rate (HDR) brachytherapy is a widely utilized treatment modality in modern clinical brachytherapy. In clinical practice, accurate dosimetric parameters for ¹⁹²Ir HDR brachytherapy sources are essential. Due to the variation in source designs, each model requires specific dosimetric parameters. Although there is extensive international research on the dosimetric parameters of ¹⁹²Ir sources, simulation studies focusing on the dosimetric parameters of HDR ¹⁹²Ir brachytherapy sources produced by the Atomic Hi-Tech (HTA) Co., Ltd are relatively scarce in China.

This study aims to calculate the dosimetric parameters of a domestically produced HDR ¹⁹²Ir source by constructing a detailed structural model of the source using Monte Carlo simulation software, based on the dosimetric calculation methods recommended in the TG-43U1 report by the American Association of Physicists in Medicine (AAPM).

Based on the dosimetric calculation methods recommended in the TG-43U1 report by the American Association of Physicists in Medicine (AAPM), Monte Carlo software was used to simulate the ¹⁹²Ir brachytherapy source, and a detailed structural model of the domestically produced high dose rate ¹⁹²Ir brachytherapy source was established within the Monte Carlo software to accurately conduct the simulation. Then, the dosimetric parameters calculated in the simulation such as the dose rate constant, air kerma strength per unit activity, radial dose function, and anisotropy function were calculated in the simulation and compared with that reported in the literature.

Simulation results show that the simulated dose rate constant is found to be 1.105 cGy·h^-1·U^-1, with a difference of less than 1.2% compared to values reported in the literature. Additionally, the air kerma strength per unit activity is calculated to be 9.788×10^-8 U·Bq^-1, with a discrepancy of 0.23% compared to the literature values. Furthermore, the dosimetric parameters for the radial dose function and anisotropy function obtained in this study show a high degree of consistency with corresponding data from existing literature.

The domestically produced ¹⁹²Ir source model established using the Monte Carlo software demonstrates good consistent dosimetric parameters with the literature-reported dosimetric parameters, indicating that this model can be used for clinical practice applications of domestically produced ¹⁹²Ir sources and has certain guiding significance.

The presence of technetium in spent nuclear fuel reprocessing waste solution significantly increases the pressure of glass solidification and geological disposal of radioactive waste, and poses long-term potential radioactive hazards to the ecosystem. Therefore, it is necessary to separate and extract it in advance.

This study aims to propose a process for the extraction and purification of technetium that can be well adapted to the plutonium uranium recovery by extraction (PUREX) process, providing support for engineering applications.

Based on the experimental data pertaining to the extraction and separation of technetium utilizing NTAamide(C8), a refined technetium separation process route containing technetium extraction and stripping with ammonium carbonate was meticulously designed whilst impurity ions were washed by oxalic acid mixed with nitric acid. Subsequently, validation of optimized process route was finalized through a series of cascade experiments with several stages' extraction and washing.

Experimental results show that good washing effect on each impurity ion is achieved when the concentrations of oxalic acid and nitric acid are 0.10 mol·L^-1. An exceptional technetium recovery rate of 99.9% is attained after 8-stage extraction and 8-stage washing process while the purification factors of strontium, cesium, zirconium, and ruthenium reach 6.9×10³, 7.9×10⁴, 4.3×10² and 45, separately.

This optimized process proposed in this study provides technical support for the separation and extraction of technetium in post-treatment plants.

Lithium-doped plastic scintillators possess the dual-mode detection capability for both fast and thermal neutrons, and combine the advantages of conventional plastic scintillators such as cost-effectiveness and the ability to fabricate in large sizes, have drawn much attention and shown great potential for thermal neutron detection in recent years.

This study aims to prepare large, trasparent polystyrene (PS) plastic scintillators doped with lithium methacrylate (LiME) via a thermal polymerization method, and to characterize their optical transmittance, luminescence, and pulse height spectrum response properties.

The composition, and structure of synthesized LiME were analyzed using Fourier transform infrared (FTIR) spectroscopy, Raman spectroscopy, and spectrophotometry. The transmittance, photoluminescence (PL) and X-ray excited luminescence (XEL) properties of LiME-doped plastic scintillators were measured and compared with those of pure plastic scintillator. The pulse height spectra (PHS) under excitation of gamma-ray were measured to determine values of the relative light output for the plastic scintillator by using Compton Edge calibration.

The optimal concentrations of AIBN, PPO, and POPOP for polystyrene plastic scintillators were experimentally determined to be 0.1 wt%, and 0.02 wt%, respectively. Characterization results indicate that the LiME-doped plastic scintillators maintain excellent optical quality, with the peak position of the emission spectrum remaining consistent with that of the pure plastic scintillator. However, as the LiME content increases, both the fluorescence intensity and light output of the scintillator gradually decrease. This trend is presumed to be associated with fluorescence quenching effects induced by the doping process.

Large, transparent plastic scintillators doped with up to 12 wt% LiME (equivalent to 0.9 wt% Li) can be successfully synthesized using a thermal polymerization method. While the performance of lithium-doped plastic scintillators still requires further improvement, these materials hold great potential for applications requiring efficient dual-mode detection of both fast and thermal neutrons.

⁴¹Ar is one of the main radionuclides released into the environment by the operation of reactors and accelerators. The measurement of ⁴¹Ar activity concentration is of great significance for ensuring public health. In the field of radioactive gas measurement, the β-γ coincidence method is widely used because it can significantly reduce the background and improve the sensitivity of the detector. However, there is currently little research on β-γ coincidence detectors for ⁴¹Ar measurements in literature.

This study aims to optimize the β-γ coincidence detector composed of plastic scintillator (PS) and CsI(Tl) scintillator to realize high sensitivity measurement of ⁴¹Ar.

An optimization method of detector structure based on minimum detectable concentration (MDC) was proposed. The optimization process of the detector was realized based on Geant4 simulation. Firstly, the penetration percentage of β-rays depositing energy into CsI(Tl) through different thicknesses of BC404 and 0.5 mm aluminum film was simulated. Secondly, the variation of full-energy peak efficiency of 1 293.6 keV γ-ray in CsI(Tl) scintillator with different thicknesses of CsI(Tl) was simulated. Thirdly, simulating with different gas chamber volume, the comprehensive effects of β-rays detection efficiency, γ-ray full-energy peak efficiency, gas chamber volume and sampling time on MDC were analyzed with the assumption that the sampling time of argon was proportional to the volume of the detector's gas chamber. Finally, the influence of measurement time on MDC under different background counting rates was analyzed.

Simulation results show that the penetration percentage of β-rays emitted from the decay of ⁴¹Ar is about 0.78% when the thickness of BC404 is 3 mm. The γ-ray full-energy peak efficiency increases with the increase of CsI(Tl) scintillator thickness. With the argon sampling rate of 600 mL?h^-1, the optimal detector size parameters that minimize MDC are completely determined. When the background count rate is 1×10^-3~1 s^-1, the recommended measurement time for ⁴¹Ar is about 200 min. The MDC of the optimized β-γ coincidence detector for ⁴¹Ar measurement is about 1.7 Bq?m^-3 under the condition of measurement time of 200 min, background count rate of 5×10^-3 s^-1 and sample cooling time of 30 min.

The detector structure optimization method used for β-γ coincidence detectors in this study can be applied to the structural design of other radioactive gas detectors, and the impact of detector structure on background counting rate needs to be considered in further enhancement study.

Differential die-away analysis (DDAA) is a technique utilized for the active detection and analysis of Special Nuclear Materials (SNM). This method can identify the presence of nuclear materials irrespective of shielding by inducing fission through external pulsed neutrons.

This study aims to design and optimize a detection system based on DDAA technology by Monte Carlo simulation to achieve the highest sensitivity in detecting SNM.

The Geant4 Monte Carlo simulation software was employed to design the SNM detection system with optimization of its key parameters. The size, pressure, thickness, and density of the ³He detector, as well as the thickness and density of the moderator were calculated for obtaining the geometric parameters required to achieve optimal neutron detection efficiency. Then, different geometric arrangements of the system were tested to identify the configuration that delivers optimal sensitivity for SNM detection. Finally, based on the DDAA principal and simulation optimization results, the detection system was evaluated using a pulsed neutron with an energy of 14 MeV and a pulse width of 100 μs.

The simulation results indicate that the detection system achieves high sensitivity for detecting 1 g of ²³⁵U nuclear material when configured with the following geometric parameters: a pressure of 0.405 3 MPa, a 5.08 cm diameter ³He tube, an outer wrapping density of 1.10 g·cm^-3 for the ³He tube, a 4 cm thick moderator, and a 7 cm thick moderator with a density of 0.85 g·cm^-3 surrounding the testing sample.

The optimized geometric configuration significantly enhances the sensitivity of SNM detection systems based on DDAA technology. These findings provide crucial insights for the parameter design of effective SNM detection systems, thereby contributing to improved nuclear material security and non-proliferation efforts.

The helical-coiled once-through steam generator (H-OTSG) has the advantages of compact structure and strong heat transfer ability, which is appropriate for lead-cooled fast reactor (LFR). The two-phase flow instability may cause mechanical vibration and thermal fatigue of heat transfer tube bundles, posing a serious threat to the safe operation of steam generators.

This study aims to explore the oscillation modes and influencing laws of two-phase flow instability of H-OTSG, providing reference for industrial design.

Firstly, RELAP5/MOD3.4 code was applied to modelling the helical-coiled once-through steam generator with 14 parallel heat exchange tubes. The primary working fluid of H-OTSG was liquid lead bismuth eutectic (LBE) and the secondary fluid was water. Then, the oscillation behavior during start-up was studied based on time-domain method and the oscillation characteristics and the parameter sensitivity of the stable boundary were analyzed. The limit cycles of the oscillation in each channel were shown on the pressure-drop vs. flow-rate plane. Finally, the influence of structural parameters on system stability were explored, so did that of operating parameters such as the pressure, flow rate, and temperature of the secondary fluid.

The results indicate that the operating parameters exhibit density wave oscillations at the heating section in a (n-2,2) pattern, with superimposed flow pattern transition instability. The smaller the flow amplitude, the shorter and thinner the corresponding limit cycle, and the closer to the circle. In the same channel, as the driving force increases, the flow amplitude gradually decreases, and the limit cycle also gradually shrinks. In addition, as the inlet throttling has been increased from 1 100 to 1 700, the duration of oscillation shortened from approximately 6 000 s to less than 1 000 s, and the amplitude decreased by nearly 30%. With the increase of the outlet throttling from 0 to 200, the duration of oscillation has been lengthened from less than 5 400 s to approximately 18 000 s. In addition, the steam temperature and the power-flow ratio of the channel increase with the increase of outlet throttling, resulting in reduced system stability. As the system pressure is increased from 3.7 MPa to 4.4 MPa, the oscillation duration of the flow curve shortens from 10 000 s to less than 5 000 s, and the amplitude also decreases. The density difference between the liquid phase and vapor phase decrease with the increase of system pressure. As a result, two-phase frictional resistance is decreased, the self-sustained oscillation of mass flow is suppressed, and the system stability is increased.

Results of this study demonstrate that the system stability of the helical-coiled once-through steam generator can be improved by increasing inlet throttling and system pressure and reducing outlet throttling, and involved structural and operational parameters should be focused on during the design process.

Reactor activation-burnup calculation is a crucial component of reactor analysis, involving an iterative process that combines criticality programs with point burnup programs.

This study aims to design and develop a novel lightweight, general-purpose activation-burnup program, named LightAB (Light Activation and Burnup) for activation-burnup calculation.

Burnup databases on the basis of ORIGEN-2 and ORIGEN-S were utilized and the Chebyshev rational approximation (CRAM) algorithm was implemented in LightAB for accurate burnup systems. Point burnup calculations in decay mode, constant flux mode, and constant power mode were supported by LightAB with well-structured program architecture, consisting of a solver module, an I/O module, and a burnup chain module. In the meanwhile, nuclide was used as the fundamental unit of storage, and physical quantities such as burnup database path and sub-burnup step division were specified as the input module of LightAB. Thereafter, the decay of ²³⁷Np and the irradiation of Zr under fixed-flux conditions were calculated using LightAB for accuracy validation, and various reactor burnup models, including pressurized water reactor (PWR) cell, PWR assembly, and Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) fast reactor models, were calculated by coupling LightAB with RMC programs. Finally, LightAB was applied to the irradiation production of transplutonium isotope with comparison to RMC.

Results of LightAB are consistent with that of ORIGEN 2.1 for the calculation of ²³⁷Np's decay and Zr's irradiation. Calculation results of LightAB coupling with RMS programs are consistent with RMC calculations. The errors between LightAB and RMC for production calculation of transplutonium isotope in three cases are within 5%.

LightAB has shown promising application prospects in the irradiation production of transplutonium isotopes compared with RMC simulation calculations.

Solving benchmark problems is a significant step in the validation of numerical simulation programs. The Experimental Breeder Reactor II (EBR-II) is a famous benchmark for sodium-cooled fast reactors (SFR), with a complicated core configuration and spatial distribution of nuclide density, hence the modeling difficulty and computational cost of its fine numerical models are relatively high. Therefore, simplified model that mixing the spatial distribution of nuclide density by the component type is adopted in many studies on EBR-II benchmark calculation.

This study aims to contrast the difference between the results of the fine model and the simplified model, evaluating the rationality of the simplification.

In this study, both the fine model and the simplified one were built using LoongSARAX, a neutronic numerical program for fast reactors developed by Xi'an Jiao Tong University. Some approximations were applied to these two models, i.e., one-dimensional homogenization adopted for the half-worth driver assembly to handle its complex radial geometry and the super-assembly method adopted in the cross-section generation of poison elements. Finally, deviation between neutron physical properties in two models was evaluated in terms of computation time consumption, effective multiplication factor, neutron flux density distribution.

The evaluation results show that the spatial distribution of fuel nuclide density presents strong asymmetry and strong non-uniformity in the simplified model, and calculation time spent in the simplified model is one-tenth of that in the fine model. Compared with the fine model, the effective multiplication factor (k_eff) is 1.383×10^-2 lower than in the fine model and the spatial distribution of neutron flux is lower in the center and higher in the outer core whereas the maximum relative deviation between neutron flux in two models is 4.25%.

This study demonstrates that the simplified model has a much lower calculation cost but limited numerical accuracy in k_eff and neutron flux, hence it is still necessary to adopt the fine model when necessary.

Thermal conductivity of SiC composite cladding significantly decreases after irradiation, leading to cladding failure due to high tensile stress under pellet-cladding mechanical interaction (PCMI).

This study aims to address the challenges of high fuel temperature resulting from low thermal conductivity and cladding failure under PCMI in SiC cladding fuel rods.

Firstly, a design of pellet-cladding gap filled with liquid LBE (Lead-Bismuth Eutectic) duplex SiC cladding fuel rod was proposed, and the gap filling material model was developed based on FRAPCON code. Then, the model was applied to incorporating the influence of LBE filling on gap heat transfer, accounting for changes in immersion height due to variations in gap and LBE volume during operation, and evaluating the impact of LBE volume on gas space and internal pressure within the fuel rod. Subsequently, the performance of this UO₂-SiC cladding fuel rod with LBE filled gap preliminarily analyzed under normal operating conditions with different initial LBE filling heights, using a typical pressurized water reactor fuel rod power history. The effects of different initial filling heights on reducing fuel temperature during operation and their impact on internal pressure were investigated. For high burnup fuel rods, further optimization of parameters including initial internal pressure, plenum length, and gap size was carried out based on the characteristics of the LBE gap and SiC cladding, resulting in an enhanced performance of UO₂-LBE-SiC fuel rod. Finally, fuel performance of three designs of fuel rods (UO₂-SiC, UO₂-SiC with central void, UO₂-LBE-SiC) were compared.

Under the condition of high power and high burnup, the decrease of the gap size and the void volume results in a significant increase in internal pressure. Increasing plenum length can compensate for the gas volume occupied by LBE and maintain the fuel rod internal pressure lower than the original He gap. SiC cladding can withstand large compressive stress and gap heat transfer no longer depends on He, hence the initial internal pressure of fuel rods is optimized to reduce internal pressure during operation. The final optimized design parameters for UO₂-LBE-SiC include a 70% fuel stack height as the initial LBE filling height, atmospheric pressure for the initial gas, a 50% increase in plenum length, and an initial gap size raised to 99 μm. The peak fuel temperature is 1 972 K, with a peak fission gas release of approximately 20% and a peak internal pressure of about 25 MPa. The hoop stress in the Ceramic Matrix Composite (CMC) layer consistently remains below its ultimate tensile strength, and the chemical vapor deposition (CVD) layer is predominantly under compression, with a maximum tensile stress of 5 MPa. The failure probability is less than 10^-6, meeting safety criteria.

The results of this study show that the design of UO₂-SiC with pellet central void cannot avoid cladding failure. Due to the excellent thermal conductivity of the LBE gap, the pellet-cladding temperature difference is minimal, increasing the gap size can weaken PCMI and reduce the probability of fuel failure without affecting the temperature field distribution.

Radiation shielding is crucial for ensuring the environmental safety of personnel and nuclear facilities in the nuclear industry. As it usually consumes a long time in single simulation calculation, the optimization design of radiation shielding is a classical expensive optimization problem.

This study aims to reduce the number of sampling points required in the radiation shield optimization design and improve the efficiency of intelligent optimization algorithms.

A dynamic radial basis function based on particle flying (PF-DRBF) surrogate model was proposed in this study for radiation shielding optimization. Firstly, a radial basis neural network was used to build the initial surrogate model of the actual objective function, and the surrogate model was globally searched for optimality by a differential evolutionary (DE) algorithm. Thereafter new sample points were selected to join the original sample points based on the result of the surrogate model search and the particle flight sample update strategy, and the surrogate model was updated based on the new set of sample points and iterated until the convergence condition was satisfied. Since the flight speed of the original sample point to the random sample point and the optimal predicted sample point based on the fitting accuracy of the surrogate model were controlled by the model, the adaptive balance between the global exploration and the local exploration of the dynamic surrogate model was achieved. Finally, in order to verify the effectiveness of the method, the proposed method was applied to 12 numerical test functions and the optimization design for radiation shielding of marine reactors, and the calculation results of other optimization methods, i.e., mode pursuing sampling (MPS) method and dynamic radial basis function based on trust region (TR-DRBF) method, were compared.

The comparative results show that for numerical test functions, the proposed PF-DRBF method has significant advantages in search accuracy, search efficiency, and algorithm robustness. For the radiation shielding optimization, the neutron transmittance obtained by the proposed method is 48% and 8% of MPS and TR-DRBF methods, and the number of required sample points is 25% of the static surrogate model.

The results of this study indicate that by using the dynamic surrogate model based on the particle flying algorithm, the number of sample points needed to solve the expensive optimization problem is greatly reduced. The dynamic radial basis surrogate model with particle flight is an effective method for radiation shielding optimization.

The excellent performance of graphene has attracted the attention of space scientists, which can be a new space materials that offer high performance and functionality in space application.

This study aims to explore the irradiation effect of 500 keV electron with different dose on reduced graphene oxide film (rGOF).

The rGOF materials with the thickness of 50 μm for experiment were prepared by chemical process, and the electron irradiation experiments were carried out in the electron proton double beam accelerator at Lanzhou Institute of Physics. The electrical properties, surface morphology, structural defects, thermal stability and inter-lamellar spacing of rGOF were measured by using a four-point probe (RTS-9 KEITHLEY 2400), Raman spectrometer (Raman), X-ray photoelectron spectroscopy (XPS), field emission scanning electron microscopy (SEM), thermogravimetric analyzer (TGA) and X-ray diffractometer (XRD), respectively, before and after irradiation. Finally, the FLUKA software was employed to calculate the electron deposition of 500 keV electron irradiated at different depths of the rGOF material in which the radiation source model was a 2.4 cm×2.4 cm square source vertically incident on a circular film target with a radius of 1 cm.

Experimental results show that the electrical resistance of rGOF is increased and the surface morphology is improved after electron irradiation, hence the irradiated defects are created with the ratio of total carbon atoms to total oxygen atoms (C/O) decreased from 4.7 to 3.0. FLUKA simulation results demonstrate that the 500 keV energy electron can penetrate the entire rGOF with deposited energy per unit depth of ±35 eV.

It can be concluded that the change in resistance is caused by defects, structural disturbances, oxygen-containing functional groups and C/O atomic ratios whilst the thermal stability is improved and the graphene layer spacing is larger after electron irradiation.

Synchrotron radiation X-ray diffraction (XRD) is a powerful tool for material structure analysis, and it is commonly used in structural analysis, phase identification, etc. The X-ray diffraction data are often collected using two-dimensional detectors in Synchrotron Radiation Facility at present, but the experimental system cannot ensure the vertical geometric relationship between the X-ray and the detector area, resulting in systematic errors in the experimental data and impacting on subsequent data processing and analysis seriously. Although existing domestic and foreign software such as FIT2D, pyFAI calib2, Dioptas, SGTools, etc. provide calibration functions, there are still some shortcomings in operability and usability.

This study aims to develop and implement an user-friendly software to calibrate 2D XRD detector experimental parameters through the standard sample independently.

A domestic software, named as Corona, was developed using Python and PyFAI library. The experimental geometric parameters were calibrated in Corona by making use of the Debye Scherrer ring of the standard powder sample. By comparing and testing with existing foreign software FIT2D and Dioptas, the accuracy and usability of Corona calibration parameters and data integration were verified.

Verification results show that Corona can effectively achieve calibration detector parameters and data integration accurately with advantages of friendliness, convenience, and ease of user operation.

Corona proposed in this study provides effective tools to calibration 2D X-ray diffraction detector experimental parameters and data integration, but further improvement need to be implemented for expanding application.

The oxygen octahedral rotation (OOR) in perovskite oxides is closely related to its physical properties. The recent development of synchrotron radiation three-dimensional diffraction provides opportunities for efficient characterization of the superlattice half-order peaks corresponding to OOR, but quantitative analysis is still difficult.

This study aims to provide guidance for the measurement of OOR half-order peaks and lay a foundation for their quantitative analysis through a theoretical simulation of the half-order peak intensity.

Firstly, a universal calculating formula for coordinates of all the oxygen ions in an OOR super cell was provided. Then, starting from the calculation formula of structure factor of a lattice unit cell, a quantitative formula for calculating the half-order peak intensity of lattice with OOR was deduced according to the basic theory of crystal diffraction kinematics. Subsequently, the half-order peak intensity distribution patterns corresponding to the 27 rotation modes were simulated and exhibited by programming, and the appearance rules were summarized. Finally, two typical examples were used to verify the consistency between the simulation results and the measured results.

Two typical examples show that the simulation results are in good agreement with the measured results. Based on these results, the OOR half-order peak pattern can be predicted beforehand and their origins may be verified after hand for experimental measurement of half-order peak. This work may promote the application of synchrotron radiation diffraction in the characterization of perovskite OOR.