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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.

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.

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.

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.

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.

A FDS-LC long counter was developed to monitor the neutron fluence at the HINEG facility. The response function and the effective centre of long counter were calculated by using SuperMC code. The calibration of this long counter was performed in Research Facility, China institute of atomic energy. The results of experimental and simulative characterization of the FDS-LC long counter are persented in this paper. The comparison of the measured results with the calculated results shown a good agreement within 5%.

A high intensity D-T fusion neutron generator (HINEG) is keenly needed for the research and development (R&D) of nuclear technology and safety of the advanced nuclear energy system, especially for the radiation protection and shielding. The R&D of HINEG includes two phases: HINEG-I and HINEG-II. HINEG-I is designed to have both the steady beam and pulsed beam. The neutron yield of the steady beam is up to 1012 n/s. The width of pulse neutron beam is less than 1.5 ns. HINEG-I is used for the basic neutronics study, such as measurement of nuclear data, validation of neutronics methods and software, validation of radiation protection and so on. HINEG-II aims to generate a high neutron yield of 1013 n/s neutrons by adopting high speed rotating tritium target system integrated with jet/spray array enhanced cooling techniques, and can further upgrade to obtain neutron yield of 1014~1015n/s by using of accelerators-array in a later stage. HINEG-II can be used for fundamentals research of nuclear technology including mechanism of materials radiation damage and neutronics performance of components, radiation shielding as well as other nuclear technology applications.

For validating the current evaluated neutron data libraries, neutron leakage spectra from lead and lead bismuth eutectic (LBE) cylinders have been measured using an intense D-T pulsed neutron source with time-of-flight (TOF) method by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS). The measured leakage spectra have been compared with the calculated ones using Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) with the evaluated pointwise data of lead and bismuth processed from ENDF/B-VII.1, JEFF-3.1 and JENDL-4.0 libraries. This work shows that calculations of the three libraries are all generally consistent with the lead experimental result. For LBE experiment, the JEFF-3.1 and JENDL-4.0 calculations both agree well with the measurement. However, the result of ENDF/B-VII.1 fails to fit with the measured data, especially in the energy range of 5.5 and 7 MeV with difference more than 80%. Through sensitivity analysis with partial cross sections of 209Bi in ENDF/B-VII.1 and JEFF, the difference between the measurement and the ENDF/B-VII.1 calculation in LBE experiment is found due to the neutron data of 209Bi.

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]