Water is the primary resource required for the exploitation of lunar resources. Investigating the distribution of water on the surface of the moon has become a focal point in the lunar exploration plans of several nations.

This study aims to quantify the presence of hydrogen and analyze its spatial distribution on the lunar surface using data from the Chang'E-2 gamma-ray spectrometer (CE2-GRS).

Firstly, an analytical method combining branch specific stripping with the nonlinear least-square fitting Gaussian function was proposed to subtract the characteristic γ rays of interfering nuclides (²¹⁴Bi@2.204 MeV, ²⁷Al@2.210 MeV, ⁴⁹Ca@2.371 MeV) ranging from 2.1 MeV to 2.5 MeV. Then, a characteristic function between the abundance and counts of aluminum γ rays around the moon was established to subtract the counts of aluminum in the mixed peak. Finally, the spatial distribution of hydrogen γ rays for counts per 3 s on the whole lunar surface was obtained.

The analysis results show that high-value characteristics exhibited in some areas, including the Aiken basin, Mare Ingenii, Mare Imbrium, and Oceanus Procellarum, are approximately 2.6 times the average value of hydrogen counts among the 14 major maria. Comparison between the distribution characteristics of hydrogen elements on the lunar surface and the data of epithermal neutrons from Lunar Prospector (LP) reveals a highly negative correlation between the distribution characteristics of the two in these regions.

The distribution characteristics of hydrogen elements on the lunar surface further predict that there may be a large amount of structural water in the Mare Ingenii, Mare Imbrium, and Oceanus Procellarum, formed by the combination of hydroxyl groups or molecular hydrogen (H₂), achieving a better understanding of the orbit γ deep mining and scientific application of energy spectrum data.

The barrel-sampled electromagnetic calorimeter (Ecal) is an important part of the Multi-Purpose Detector (MPD) in Nuclotron-based Ion Collider fAcility (NICA) that is being built in Russia. It is primarily used to detect energy, time, and position information of electrons and photons in the energy domain from 10 MeV to a few GeV. MPD-Ecal is comprised of 2 400 modules with 16 towers per module. Each tower is made up of alternating layers of 211 scintillator sheets and 210 lead sheets, as well as 16 wave-length shift fibers.

This study aims to evaluate the performance, such as energy resolution, time resolution, and coordinate resolution, etc., of the Ecal by simulation.

The Geant4 software was employed to simulate single-energy electron incident on Ecal to examine the effects of several parameters on the performance of Ecal. Influences of the position of the particle incidence point, the number and thickness of the scintillator and lead layers, the polish of the optical fibre end-face, and the energy and type of incident particles on the energy deposition and resolution, time distribution and time resolution, and coordinate resolution were investigated in details. Finally, the time resolution of a single tower was simulated using the natural cosmic ray package, and the tower's expected time resolution in the cosmic ray test was obtained.

As the electron incidence position moves from the edge to the center of the module, the energy deposition within the scintillator rises from 718 MeV to 758 MeV. With a limited tower length of 415.5 mm, increasing the number of scintillator layers decreases the energy resolution of the module, improves the time resolution of the tower, and worsens the coordinate resolution of the Ecal. Taking into account the performance of the Ecal gauge, the optimal number of scintillator layers in the tower is 211. SiPM detects 42% more photoelectrons at a polish of 0.6 when the fiber end is coated with a reflective material than when it is not. As the polish of the fiber end-face increases, so does the number of photoelectrons detected by the SiPM, and the time resolution of the tower improves. When the fibre end-face polish is 1.0, the time resolution of the tower is less than 103 ps while time resolution of the tower (with 211 layers) in the cosmic ray is 185 ps.

The time and coordinate resolutions of Ecal improve with increasing electron energy under the same circumstances.

Due to the specific imaging structure of neutron imaging system, the geometric unsharpness is unavoidable in neutron photographic images after imaging.

This study aims to design a geometric unsharpness correction method for neutron photographic image based on improved Richardson–Lucy (RL) algorithm.

First, according to the geometric unsharpness of the neutron photographic image, the mathematical model of the point spread function (PSF) was established. Then, the Laplace operator and median filter were used to remove the PSF related gamma spots noise (GSN) in the image, and the neutron photographic image was restored by RL algorithm. Finally, the restoration quality of neutron photographic images of the line-pair sample was evaluated using average gradient (AG) and spatial frequency (SF).

The results of the line-pair sample demonstrate that compared with four existing correction algorithms of geometric unsharpness, the proposed method improves AG and SF by 60.23% and 29.90%, respectively.

The proposed method of this study can effectively correct the geometric unsharpness caused by the amplification of gamma spots noise in the process of neutron photographic image restoration, providing an important technical support for performing high-resolution neutron radiography.

Nickel-, iron- and tungsten-based alloys are commonly used as structural materials of reactors. During their operational life, these alloys undergo intense neutron irradiation.

This study aims to analyze the post-irradiation defect evolution and its mechanisms in these materials for comprehending the effects of irradiation on them.

The displacement cascades in nickel, iron, and tungsten were examined at various temperatures (300?500 K), primary knock-on atom (PKA) energies (<20 keV), and directions (<135>, <122> and <100>) by using molecular dynamics (MD) simulations. Firstly, the model was initially relaxed at each specified temperature under a canonical ensemble for 10 ps, applying periodic boundary conditions in every direction. Then, an atom was randomly chosen as a PKA and assigned kinetic energy to initiate the cascade collision simulation in the micro-canonical ensemble. Finally, the Open Visualization Tool package was employed for visualization and data analysis of the irradiation cascade processes.

The simulation results reveal that nickel and iron exhibit similar steady-state defects. At lower PKA energies (<5 keV), nickel exhibits marginally fewer defects than iron. However, as the PKA energy surpasses 5 keV, the number of defects in nickel becomes slightly more than that in iron. Furthermore, under identical irradiation conditions, tungsten demonstrates superior radiation resistance, with fewer steady-state defects when compared with both nickel and iron.

The defect evolution during various cascade displacement phases in three metals and their defect recombination rates are crucial to understanding the disparities in radiation damage resilience. The derived results help to comprehend the radiation characteristics of these metals. Additionally, the primary radiation damage dataset compiled for these metals lays a foundation for further larger-scale simulations of their radiation attributes using rate theory or cluster dynamics methods.

The targeted acquisition of the radioactive element content or radioactivity in a radioactivity is an important task in geological exploration and radioactive pollution investigation. During the process of targeted gamma radiation sampling, gamma rays from non-target areas significantly interfere with the measurement results.

This study aims to design a dual-energy targeted gamma radiation sampling probe that uses a cerium bromide scintillation detector on the basis of the difference in the linear attenuation coefficients of the high- and low-energy gamma rays from the same radioactive decay series in the lead shielding layer of the detector.

Firstly, Monte Carlo (MC) numerical simulations were employed to determine the optimal lead shielding layer thickness for the dual-energy targeted gamma radiation sampling probe detecting targets of the 0.609 MeV and 1.764 MeV gamma rays emitted by ²¹⁴Bi. Then, the directional proportionality coefficients were calculated and applied to obtaining 0.609 MeV gamma ray counts of dual energy γ radiation probe within lead shielded angle. Finally, MC numerical simulations with four types of interfering radiation sources and physical experiments with two radium sources were conducted to validate the results that calculated using the directional proportionality coefficients.

The simulation result of optimal lead shielding layer thickness for the dual-energy targeted gamma radiation sampling probe is 6 mm, and the calculated directional proportionality coefficients of a and A are 0.268 and 0.451, respectively. Validation results show that the relative error between the counts for the 0.609 MeV gamma rays within the shielded angle and the net peak area counts measured with the dual-energy targeted gamma radiation sampling probe for two radium sources is within ±2.52% with average relative error of 0.63%. The relative errors between the measured uranium content and recommended values in the tested models are all less than 5%.

The dual-energy targeted gamma radiation sampling results for a radioactive mixed standard model and three radioactive models indicated that the designed dual-energy targeted gamma radiation sampling probe is capable of targeted gamma radiation sampling.

Studying the transport behavior of impurities in plasma and developing effective impurity control methods are important for achieving high-performance plasma discharge and ensuring the safe operation of the device.

This study aims to design a control system for the experimental advanced superconducting Tokamak (EAST) laser blow-off (LBO) impurity system.

A new automatic control process was adopted to enable to inject tracer particles of different elements into the EAST plasma repeatedly, quantitatively, and controllably. An accurate control of the focusing lens displacement and laser triggering time were achieved through the STM32 microcontroller system and its output PWM waves for stepper motor driving, hence the diameter of the laser spot was adjustable to change the amount of impurities injected. Finally, the designed control system for LBO was tested in staging environment to verify its practicability and accuracy.

The test results show that the system can rapidly detect the external trigger signal and achieve precise timing, with less than 0.4 mm position error for laser spot focusing.

The design of control system meets the requirements of the laser blowing impurity injection experiment. This study is of considerable significance for research on EAST plasma impurity transport.

Large-area plastic scintillators are widely used in radiation monitoring. They are typically employed as gamma radiation counters through signal over-threshold detection. The value of the threshold voltage affects the detector efficiency and minimum detectable activity. When rays are incident at different positions of the large-area plastic, the photon collection efficiencies generated through collecting end pair differ, leading to differences in the detector efficiency.

This study aims to explore appropriate threshold voltage to reduce the differences and maintain a low minimum detectable activity (MDA), hence achieving high detection efficiency for large-area plastic scintillator.

Firstly, an energy spectra acquisition system was designed using STM32F429 produced by STM semiconductor as the main control chip and a counter for detector efficiency testing of large-area plastic scintillator in size of 400 mm×300 mm×50 mm. Then, the reasonable threshold voltage was determined according to the background energy spectra of the large-area plastic scintillator, the energy spectrum of the ⁶⁰Co and ¹³⁷Cs sources at different positions of the large-area plastic scintillator with acquisition time of 3 min, and signal-amplification relationship.

The determined threshold voltage using above mentioned method for ⁶⁰Co source is 93.7 mV. To achieve high detector efficiency and low MDA for the large-area plastic scintillator, the best determined threshold voltage for ⁶⁰Co source is 95 mV. At this voltage, the detector efficiency for ⁶⁰Co source is 22%, the MDA is 78 Bq.

The proposed method has reference value for large-area plastic scintillation designing and manufacturing detectors.

During the operation of nuclear power plants, a large amount of low and intermediate level waste (LILW) is generated, which is usually prepared into 200-L and 400-L waste drums. To ensure the safe disposal of these waste drums, they must be analyzed to determine the type and activity of the nuclides contained within them. Non-destructive assay (NDA) has been widely used in the detection of waste drums in nuclear power plants, along with segmented gamma scanning (SGS) and tomographic gamma scanning (TGS). However, the low measurement accuracy of SGS and the long measurement time of TGS limit the practical application of these methods.

This sudy aims to shorten the measurement time while maintaining high measurement accuracy by proposing a new neural network-based method for measuring the activity of waste drum.

When the waste drum was filled with a uniform distribution of medium and rotated at a constant speed during measurement, the point source was equivalent to a ring source. The equivalent ring source in the waste drum possessed an activity equal to the total activity of all sources. The neural network model is established, the count rate of the detector at different positions is used as input, and the radius of the equivalent ring source is used as output. Finally, the total activity of the waste drum is calculated. The simulated measurement is carried out in a 400-L waste drum, the medium is concrete, the radioactive source is Co-60, and 50 groups of single-source and 10 groups of multi-source are generated randomly. Different methods are used to reconstruct the activity of the waste drum.

When there is only one radioactive source in the waste drum, the mean relative error (MRE) of activity reconstruction by the new method is 4.26%, which is much lower than that of SGS (68.15%) and close to that of TGS with 60 grids (3.97%). When there are multiple radioactive sources in the waste drum, the MRE of activity reconstruction by the new method is 24.27%, which is lower than that of SGS (48.02%) and close to that of TGS with 60 grids (28.61%). This new method achieves the equal measurement accuracy of TGS but reduce the measurement time to 1/20 of TGS.

Compared to traditional measurement methods, the new method greatly shortens the measurement time while maintaining high precision, thereby providing technical support for the measurement of LILW.

A general-purpose heat source (GPHS) is the most established heat source module for radioisotope thermoelectric generators (RTGs) and a key reference for the equivalence of electrically heated analog heat sources during the development and testing of radioisotope power supplies.

The study aims to develop a highly efficient electric heating simulation heat source to meet the requirements for equivalent testing and verification of non-nuclear units of radioisotope power systems.

Firstly, an electrically heated analog heat source was designed and imitated based on various actual GPHS performance parameters. Based on the simulation calculations, the thermodynamic equivalent substitutability between the simulated heat source and real GPHS in terms of material and dimensional differences was evaluated. Then, the performance of its operation in different application scenarios was analyzed, and an optimized application environment was proposed. Finally, based on the experimental test results, the uniformity of the thermal output characteristics of the imitation GPHS-simulated heat source was compared with that of a real GPHS, and the practical application characteristics of the imitation GPHS in RTGs were also evaluated.

At an input power of 250 W, the average surface temperature of the GPHS-simulated heat source reaches 515 K. The temperature variation trend with power is consistent with that of the simulation results. Experimental test results show that the energy conversion efficiency of the RTG module is increased to ~6% with a 250-W heat source power supply.

The proposed and constructed good equivalent simulated GPHS heat source with reference to the thermal properties of a real GPHS can be applied to RTGs and provides an effective and unified reference standard for the performance evaluation of radioisotope power sources.

Scintillator-based fast-ion loss detector (FILD) can measure the velocity-space distribution of fast-ion loss and are key in studying the control mechanism of fast-ion loss in nuclear fusion devices.

This study aims to obtain the velocity-space distribution of fast-ion loss from corresponding FILD data and evaluate the capacity of current FILD probe for the further improvement of diagnostic design.

The FILDSIM code was employed to establish the linkage between the fast-ion image digitized by FILD and the velocity-space distribution of fast-ion loss. The detection coverage of fast ions on the scintillator was assessed through reverse tracing of the lost fast ions, considering the geometry of the FILD probe as well as the pitch angle and gyroradius of fast ions entering the pinhole of the FILD probe.

The obtained velocity-space distribution of fast-ion loss under ICRH indicates that the energy of lost fast hydrogen minority ions is above 200 keV. Moreover, analysis shows that the geometry of the probe, particularly the shell behind the scintillator, obstructs the diagnostic detection range, creating a null region on the scintillator.

The acquisition of the velocity-space distribution of fast-ion loss lays the foundation for further evaluation and control of fast-ion loss under ion cyclotron resonance heating. In addition, the investigation of the probe detection range provides a basis for upgrading diagnostic systems.

Traditional safety analysis methods rely on expert advice and user self-evaluation, lacking the ability to quantify output uncertainty. In contrast, the best estimation plus uncertainty (BEPU) methodology can quantify the uncertainty of the output, thereby avoiding unnecessary conservative assumptions and improving the economic viability of nuclear power. It is now widely used in the design and safety analysis of nuclear reactors. However, owing to the cognitive limitations of science and numerical approximation in programs, most thermal-hydraulic programs lack sufficient input uncertainty information related to internal models, often relying on expert advice.

This study aims to investigate the uncertainty quantification methodology for model parameters in sub-channel codes using Markov Chain Monte Carlo (MCMC) sampling.

Firstly, the PSBT void fraction distribution experiments were employed to evaluate the prediction ability of the subchannel program COBRA-IV, and a Python-based uncertainty analysis methodology was developed to quantitatively analyze the model parameter uncertainties that affect the void fraction. Then, the model parameters were assumed to be independent, with their uncertainties following a normal distribution. Based on the Bayesian principle, the most likely maximum a posteriori probability function (PDF) of the model parameters were obtained by combining the prior and observed information, despite the limited actual uncertainty information. Finally, an MCMC sampling methodology was adopted to solve the Bayesian relation, and the statistical uncertainty information of the model parameters were obtained using a stable a posteriori Markov chain, which requires at least 104 magnitudes to achieve convergence and the corresponding forward program runs. Therefore, to reduce the calculation cost and improve the calculation efficiency, a high-precision adaptive BPNN surrogate model was constructed to replace the complex and time-consuming forward program code. Furthermore, a set of uncertainty quantification methods with Python was developed to simultaneously quantify the uncertainty of the model parameters using a statistical method. During the selection of a slip model we discovered that both the slip ratio and turbulence mixing coefficient significantly affected the void fraction. Therefore, we developed.

The results indicate that after obtaining the uncertainty of the model parameters, the 95% confidence interval of the results generated by the forward propagation of input uncertainty enveloped the experimental values well. Furthermore, by incorporating the mean value of the model parameter uncertainties, obtained via uncertainty quantification, the modified model output exhibited a closer agreement with the experimental values than with the reference values.

The uncertainty quantification analysis methodology established in this study can be applied to the uncertainty analysis of subchannel program model parameters.

The sodium-cooled fast reactor adopts the three loops design with sodium-sodium-water. When a double-ended guillotine (DEG) break occurs in the steam generator (SG) tube, a large leakage sodium-water reaction (SWR) accident occurs, which threatens the safety and integrity of the secondary loop. A protection system is therefore designed to ensure secondary loop integrity.

This study aims to analyze the influence of protection system critical parameters on the large leakage SWR with paralleling SGs.

First of all, a large leakage SWR model, including the water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and protection system models were established. Then, the large leakage SWR model was verified using the experimental data, and the 3-DEG large leakage SWR was simulated on the basis of the secondary loop structure. The integrity of the secondary loop and the protection system response were analyzed. Finally, a sensitivity analysis was performed for the critical parameters of the protection system, including the bursting pressure of liquid rupture disks, bursting delay time of rupture disks, location of liquid rupture disks, length of the release pipe, and volume of the primary accident discharged tank. Parameters with key influence on the integrity of the secondary loop and the protection system response were determined.

The 3-DEG large leakage sodium water reaction accident results in a peak pressure of 2.003 MPa in the reaction zone and 1.329 MPa in the critical equipment of the secondary circuit except for the reaction zone. The smaller bursting pressure and delay time, and the location of liquid bursting disks at the bottom chamber and the release pipe shorter length are more conducive to the integrity of the secondary loop and the protection system response.

This study provides a reference value for the design requirements of large leakage SWR protection systems with paralleling SGs.

NECP-SARAX is a neutronics analysis code system for advanced reactors developed by the Nuclear Engineering Computational Physics Laboratory team of Xi'an Jiaotong University. In the past few years, a considerable amount of verification and validation work has been done based on CEFR, PHENIX, SUPERPHENIX, JOYO MK-I, ZPR, and ZPPR reactors. The results indicate that NECP-SARAX offers high performance for fast spectrum reactor analysis. Meanwhile, the fuel and control rod assemblies of these reactors are used for verification of the cross-section generation code TULIP. While TULIP has demonstrated promising preliminary results in fast spectrum system analysis, a comprehensive systematic verification and validation process remains essential.

This study aims to validate the applicability of TULIP code for various fast spectrum systems.

Firstly, a total of 147 critical experiment benchmarks were selected from ICSBEP and used for analysis. The initial results demonstrated that the k_eff bias between TULIP and Monte Carlo codes exceeded 10^-2 for an experimental benchmark with a thick reflector. Then, a homogeneous two-nuclide problem simplified from the HMF021-002 benchmark was subsequently used to analyze this phenomenon, and the intermediate-weight nuclides had resonance-like fluctuating scattering cross sections above the resonance energy. Finally, to address this phenomenon, the TULIP code was undergone enhancements, mainly focusing on optimizing the resonance calculation strategy and method using ultra fine group to deal with thee self-shielding effect of resonance-like cross sections in the non resonant region under high loading of intermediate-weight nuclides.

In a fast spectrum system with a large amount of structural material, the self-shielding effect of the resonance-like cross section of the intermediate-weight nuclides above the resonance range becomes non-negligible. The optimized TULIP method reduces the k_eff bias to within 3×10^-3 for these benchmarks with a thick reflector.

New numerical results indicate that the enhanced TULIP code has good performance for various fast spectrum system analyses.