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BackgroundDue to the imaging mechanism of neutron radiography and the limitation of neutron source, neutron radiography images are prone to lack of contrast, and it is a common method to improve image quality by contrast enhancement algorithm.PurposeThis study aims to solve the problems related to low contrast and loss of details in neutron imaging.MethodsAn adaptive contrast enhancement (ACE) algorithm based on the particle swarm optimization (PSO) was proposed. First of all, the gain factor coefficient was treated as a particle and initializes a particle swarm. Then, the optimal gain factor coefficient with the PSO algorithm was calculated, and the contrast of the high frequency part of neutron images was improved through implementing this optimal gain factor coefficient in ACE algorithm. Finally, several neutron images were processed by this PSO algorithm and compared with the results of traditional optimization algorithm.ResultsThe new algorithm has obvious contrast enhancement effect for different neutron images, and the parameters such as entropy value, gray average gradient value and contrast value of the optimized image are improved compared with the traditional optimization algorithm.ConclusionsThis algorithm can enhance the contrast of image stably, which has practical significance for the wide application of neutron radiography.

The effect of 135Xe and 149Sm on the reactivity of a typical megawatt heat pipe reactor had been studied by MCNP and SuperMC-CAS. The results showed that the reactivity introduced by 135Xe was only 8 pcm. The concentration of 149Sm increased with the core running time, highlighting its significant poisoning effect. A simplified calculation method for determining the equilibrium xenon poisoning value of a fast neutron reactor was proposed in this study. These results provided valuable insights for assessing xenon and samarium poisoning in low-power, fast neutron reactors. [ABSTRACT FROM AUTHOR]

BackgroundThe measurement of high-energy neutrons is becoming more and more important in the evaluation of dose caused by space neutrons and the radiation protection of particle accelerators.PurposeThe study aims to investigate the performance of an energy-extended neutron long counter.MethodsThe energy-extended neutron long counter was designed based on De Pangher type. The (n, xn) multiplication reaction between neutrons and metallic materials was utilized to improve the high-energy response of neutron long counter by adding metal chromium and lead materials into the moderator of the neutron long counter. Then, the structure of the neutron long counter was optimized by using Monte Carlo simulation to obtain the maximum value within the energy range of 1 keV~20 MeV and 1 keV~150 MeV, respectively. Finally, the effective center and energy response of the long neutron counter were measured by Am-Be neutron source and D-T neutron source in the standard radiation field.ResultsThe calculated response curve shows that the deviation of the maximum value from the average value is 12.2% in above energy range. The comparison between experimental results and the Monte Carlo simulation results shows that the theoretical calculation is consistent with the experiment within the error range.ConclusionThe upper energy limit of the long neutron counter can be extended by embedding metallic lead and chromium materials in the inner polyethylene layer of the long neutron counter.

BackgroundIn order to accurately predict the heat transfer characteristics in the sub-channel of fuel assembly for the low Prandtl number fluid, it is necessary to select a suitable turbulent-Prandtl-number model.PurposeThis study aims at the applicability of turbulent-Prandtl-number models for liquid lead-bismuth turbulence in triangular fuel rod bundles.MethodsBased on the triangular rod bundle heat transfer correlation, the turbulent models of Shear Stress Transfer (SST) k-ω was adopted to analyze the different rod bundle structures for five different kinds of turbulent-Prandtl-number models. The analysis results were compared with the recommended heat transfer correlation, which had been verified by experiment in the sub-channels for liquid lead-bismuth. The applicability of various turbulent-Prandtl-number models under different rod diameters and pitch diameter ratios was analyzed in details.ResultsThe results show that the global turbulent-Prandtl-number model is not only related to the Reynolds number Re and Peclet number Pe, but also related to the pitch-to-diameter ratio (P/D). In the entire region of (1.3≤P/D≤1.7), simulation results of turbulent-Prandtl-number model proposed by Kays are in good agreement with the values calculated by Mikityuk equation. All sorts of turbulent-Prandtl-number models have the best P/D scope of application.ConclusionsThe related turbulent-Prandtl-number models can be used to predict the heat transfer characteristics of liquid lead-bismuth in the triangular rod bundles under different pitch diameter ratios.

BackgroundWith the development of accelerator technology, the accelerator-based Boron Neutron Capture Therapy (BNCT) has attracted more and more attention, and metallic 9Be is one of the most suitable materials for neutron-producing targets.PurposeThe study aims to investigate the possibility of obtaining BNCT neutron sources based on proton cyclotron with energy of 14 MeV and current intensity of 80 μA.MethodsGeant4 was used to simulate the neutron beam properties of Be target bombarded by 14 MeV proton beams and to optimize the target thickness. In addition, spherical natural uranium was considered as a neutron multiplier, the neutron beams generated by the combination of different materials and thicknesses of Beam Shaping Assembly (BSA) were investigated by simulation.ResultsThe optimized irradiated field at beam outlet of proton cyclotron can reach 95.6% of epithermal neutrons, and 6.26×107 n·cm-2·s-1 of the epithermal flux rate.ConclusionsThis design solution can be initially used for the technical verification of BNCT neutron source experimental terminal.

BackgroundFast neutron detectors based on 4H-SiC semiconductor materials have incomparable advantages over traditional neutron detectors when used in strong radiation and high temperature environments, such as fusion plasma diagnosis, core power monitoring, and neutron oil/gas logging, because they have large threshold displacement energy, wide band gap, and fast saturation speed.PurposeThis study aims to explore the feasibility of fast neutron yield measurement by using 4H-SiC detector.MethodsFirst of all, the main performance of 4H-SiC fast neutron detector was tested by experiments using the high intensity deuterium-tritium fusion neutron generator (HINEG). Then, the reaction characteristic peak of 12C(n,α0)9Be in the high energy section was observed in pulse amplitude spectrum. The full width at half maxima (FWHM) and energy resolution were calculated.ResultsThe experimental results show that the neutron detection efficiency to fast neutrons (εE) is 8.89×10-6 n-1·cm-2 (±7.7%),FWHM (full width at half maxima) of the characteristic peak of 12C(n,α0)9Be is 190 keV, and the energy resolution is 2.26%.ConclusionsThis study verifies that the 4H-SiC fast neutron detector can be used for neutron yield measurement.

In this study, a reduced size neutron long counter with improved response function was designed using SuperMC simulation code. The main parts of the reduced size long counter are 3He proportional counter, cylindrical moderators made of polyethylene, chromium metal converter, neutron absorber material and outer aluminum cover. A 1.5 cm thick layer of chromium metal was embedded into the hydrogen-rich moderator to enhance response of the long counter at high neutron energies. The radius and length of the long counter was reduced to 15 cm and 35 cm, respectively. In this design we managed to obtain a flat response function from a few keV to 20 MeV energy range. The angular response function, determined at 0.5 MeV and 14 MeV neutron energies, confirmed that the reduced size long counter designed in this study is not significantly affected by in-scattered neutrons. The reduced size long counter is suitable for use as a standard transfer instrument for monitoring neutron fluence in mono-energetic neutron fields.

Molecular dynamics methods were utilized to investigate displacement cascades and tritium diffusion in α-MgT2. It was observed from collision cascades results that the stable number of defects weakly depended on temperature, while the peak and stable number of defects linearly increased with increasing the primary knock-on atom energy. The results of the mean square displacement study revealed that defects had a significant impact on tritium diffusion. The clustering of magnesium self-interstitial atoms and diffusing tritium atoms results in an increased diffusion barrier, whereas the formation of clusters between tritium interstitial atoms is relatively difficult and has no significant impact on the diffusion barrier. The presence of magnesium and tritium vacancies has a minimal effect on the diffusion barrier due to the large number of diffusing tritium atoms that offset the adsorption of vacancies on diffusing atoms. Both magnesium and tritium interstitial atoms increase the collision probability of diffusing atoms, leading to an increased diffusion prefactor. Magnesium vacancies cause significant lattice distortion, increasing the diffusion barrier, while the impact of tritium vacancies on the diffusion barrier is small due to their minimal lattice distortion effect. The research uncovered significant disparities in the diffusion properties of hydrogen and tritium, indicating that the results of the study of hydrogen storage could not be applied to tritium. [ABSTRACT FROM AUTHOR]