• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1031-1035
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• 2023

Neutron is an indirectly ionizing particle without charge, which is normally measured by detecting reaction products. Neutron detection system based on measuring gadolinium-converted gamma-rays is a good way to monitor the neutron because the representative prompt gamma-rays of gadolinium have low energies (79, 89, 182, and 199 keV). Low energy gamma-rays and their high attenuation coefficient on materials allow the simple design of a detector easier to manufacture. Thus, we designed a cadmium zinc telluride detector to investigate feasibility of simultaneous detection of gamma-rays and neutrons by using the Monte-Carlo simulation, which was divided into two parts; first was gamma-detection part and second was gamma- and neutron-simultaneous detection part. Consequently, we confirmed that simultaneous detection of gamma-rays and neutrons could be feasible and valid, although further research is needed for adoption on real detection.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2572-2577
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• 2023

A field determination method for small D-T neutron source yield based on the oxygen prompt gamma rays was established. A neutron-gamma transport equation of the determination device was developed. Two yield field determination devices with a thickness of 20 mm and 50 mm were made. The count rates of the oxygen prompt gamma rays were calculated using three energy spectra processing approaches, which were the characteristic peak of 6.13 MeV, the overlapping peak of 6.92 MeV and 7.12 MeV, and the total energy area. The R-square of the calibration curve is better than 94% and the maximum error of the yield test is 5.21%, demonstrating that it is feasible to measure the yield of D-T neutron source by oxygen prompt gamma rays. Additionally, the results meet the requirements for field determination of the conventional D-T neutron source yield.

• Journal of Sensor Science and Technology
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• v.22 no.3
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• pp.223-226
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• 2013

We designed a single polyethylene bonner sphere with several thermo-luminescence dosimeters (TLD), for measurement of neutron energy spectrum. For the separation of the neutron dosage in the neutron-gamma mixed field, we used 21 ea TLD-600s and TLD-700s, respectively. Because, TLD-600 is sensitive to neutron and gamma rays, and, TLD-700 is sensitive only to gamma-rays, we could determine the each dose by neutron and gamma rays. The neutron response function of the bonner sphere with TLDs was calculated by MCNPX (ver. 2.5.0) Monte Carlo simulation in the energy range from $10^{-1}$ to 20 MeV. For the Cf-252 standard neutron source in KRISS, we could estimate the neutron energy spectrum by unfolding method using the response function.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.715-727
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• 2024

During fast neutron imaging, besides the dark current noise and readout noise of the CCD camera, the main noise in fast neutron imaging comes from high-energy gamma rays generated by neutron nuclear reactions in and around the experimental setup. These high-energy gamma rays result in the presence of high-density gamma white spots (GWS) in the fast neutron image. Due to the microscopic quantum characteristics of the neutron beam itself and environmental scattering effects, fast neutron images typically exhibit a mixture of Gaussian noise. Existing denoising methods in neutron images are difficult to handle when dealing with a mixture of GWS and Gaussian noise. Herein we put forward a deep learning approach based on the Swin Transformer UNet (SUNet) model to remove high-density GWS-Gaussian mixture noise from fast neutron images. The improved denoising model utilizes a customized loss function for training, which combines perceptual loss and mean squared error loss to avoid grid-like artifacts caused by using a single perceptual loss. To address the high cost of acquiring real fast neutron images, this study introduces Monte Carlo method to simulate noise data with GWS characteristics by computing the interaction between gamma rays and sensors based on the principle of GWS generation. Ultimately, the experimental scenarios involving simulated neutron noise images and real fast neutron images demonstrate that the proposed method not only improves the quality and signal-to-noise ratio of fast neutron images but also preserves the details of the original images during denoising.

• Journal of Radiation Protection and Research
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• v.44 no.2
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• pp.53-63
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• 2019

Background: As neutron fields are always accompanied by gamma rays, it is essential to distinguish neutrons from gamma rays in the detection of neutrons. Neutrons and gamma rays can be separated by pulse shape discrimination (PSD) methods. Recently, we performed characterization of a stilbene scintillator detector and an EJ-301 liquid scintillator detector with a high-speed digitizer DT5730 and investigated optimized PSD variables for both detectors. This study is for providing a basis for developing fast neutron/gamma-ray dual-particle imager. Materials and Methods: We conducted PSD experiments using stilbene scintillator and EJ-301 liquid scintillator and evaluated neutron and gamma ray discriminability of each PSD method with a $^{137}Cs$ gamma source and a $^{252}Cf$ neutron source. We implemented digital signal processing techniques to apply two PSD methods - the charge comparison (CC) method and the constant time discrimination (CTD) method - to distinguish neutrons from gamma rays. We tried to find optimized PSD variables giving the best discriminability in a given experimental condition. Results and Discussion: For the stilbene scintillator detector, the charge comparison method and the constant time discrimination method both delivered the PSD FOM values of 1.7. For the EJ-301 liquid scintillator detector, both PSD methods delivered the PSD FOM values of 1.79. With the same PSD variables, PSD performance was excellent in $300{\pm}100keVee$, $500{\pm}100keVee$, and $700{\pm}100keVee$ energy regions. This result shows that we can achieve an effective discrimination of neutrons from gamma rays using these scintillator detector systems. Conclusion: We applied both PSD methods to a stilbene and a liquid scintillator and optimized the PSD performance represented by FOM values. We observed a good separation performance of both scintillators combined with a high-speed digitizer and digital PSD. These results will provide reference values for the dual-particle imager we are developing, which can image both fast neutrons and gamma rays simultaneously.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.262-268
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• 2022

This paper introduces a novel concept of the pulsed neutron facility (PNF) for maximizing the production of the thermal neutrons and its application to medical use based on prompt gamma neutron activation analysis (PGNAA) using Monte Carlo simulations. The PNF consists of a compact D-T neutron generator, a graphite pile, and a detection system using Cadmium telluride (CdTe) detector arrays. The configuration of fuel pins in the graphite monolith and the design and materials for the moderating layer were studied to optimize the thermal neutron yields. Biological samples - normal and cancerous breast tissues - including chlorine, a trace element, were used to investigate the sensitivity of the characteristic γ-rays by neutron-trace material interactions and the detector responses of multiple particles. Around 90 % of neutrons emitted from a deuterium-tritium (D-T) neutron generator thermalized as they passed through the graphite stockpile. The thermal neutrons captured the chlorines in the samples, then the characteristic γ-rays with specific energy levels of 6.12, 7.80 and 8.58 MeV were emitted. Since the concentration of chlorine in the cancerous tissue is twice that in the normal tissue, the count ratio of the characteristic g-rays of the cancerous tissue over the normal tissue is approximately 2.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3545-3552
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• 2024

Polymers have become widely used substrate materials for shielding neutron rays because of their high hydrogen content and easy processing procedures. Rare-earth materials are also being gradually adopted as neutron absorbers because of their considerable thermal neutron absorption cross-sections. This paper utilizes the FLUKA Monte Carlo simulation program to compare the shielding effects of various polymers and rare-earth oxides on neutron rays across different energy ranges. The study investigates the superior shielding materials for neutron radiation in each energy range. Subsequently, a series of materials are simulated by combining the preferred shielding materials for neutron rays in each energy range, exploring the influence of material composition and composite structure on the effectiveness of neutron ray shielding. It is revealed that the preferred material for shielding neutron rays changes for different energy ranges. For low-energy neutron rays, rare-earth oxides such as Sm₂O₃ and Gd₂O₃ demonstrate the most effective shielding, whereas for high-energy neutron rays, polyethylene (PE) provides the best shielding performance. Materials with different compositions show varying preferred structures when dealing with a ²⁵²Cf neutron source. However, in mitigating the secondary gamma rays generated during the neutron shielding process, stacked-type materials exhibit the most effective shielding performance.

• Progress in Medical Physics
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• v.29 no.4
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• pp.101-105
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• 2018

In this paper, a neutron moderation system for boron neutron capture therapy (BNCT) based on a $^{252}Cf$ neutron source is proposed. Different materials have been studied in order to produce a high percentage of epithermal neutrons. A moderator with a construction mixture of $AlF_3$ and Al, three reflectors of $Al_2O_3$, BeO, graphite, and seven filters (Bi, Cu, Fe, Pb, Ti, a two-layer filter of Ti+Bi, and a two-layer filter of Ti+Pb) is considered. The MCNPX simulation code has been used to calculate the neutron and gamma flux at the output window of the neutronic system. The results show that the epithermal neutron flux is relatively high for four filters: Ti+Pb, Ti+Bi, Bi, and Ti. However, a layer of Ti cannot reduce the contribution of ${\gamma}$-rays at the output window. Although the neutron spectra filtered by the Ti+Bi and Ti+Pb overlap, a large fraction of neutrons (74.95%) has epithermal energy when the Ti+Pb is used as a filter. However, the percentages of the fast and thermal neutrons are 25% and 0.5%, respectively. The Bi layer provides a relatively low epithermal neutron flux. Moreover, an assembly configuration of 30% $AlF_3+70%$ Al moderator/$Al_2O_3$ reflector/a two-layer filter of Ti+Pb reduces the fast neutron flux at the output port much more than other assembly combinations. In comparison with a recent model suggested by Ghassoun et al., the proposed neutron moderation system provides a higher epithermal flux with a relatively low contamination of gamma rays.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.446-452
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• 2018

To improve the accuracy and safety of irradiation tests in High flux Advanced Neutron Application ReactOr (HANARO), the nuclear energy deposition rate, which is called nuclear heating, was estimated for an irradiation capsule with an iridium sample in the irradiation hole in order. The gamma rays emitted from the radioisotopes (RIs) of the structural materials such as flow tubes of fuel assemblies and heavy water reflector tank were considered as radiation source. Using the ORIGEN2.1 code, emission rates of delayed gamma rays were calculated in consideration of the activation procedure for 8 years and 2 months of HANARO operation. Calculated emission rates were used as a source term of delayed gamma rays in the MCNP6 code. By using the MCNP code, the nuclear heating rates of the irradiation capsules in the inner core, outer core, and heavy water reflector tank were estimated. Calculated nuclear heating in the inner core, outer core, and heavy water reflector tank were 200-260 mW, 80-100 mW, and 10 mW, respectively.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1763-1774
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• 2023

In this study, a series of flexible neutron/gamma shielding composites are fabricated through the doping of Gd₂O₃ into the matrix of SEBS with (M_Gd2O3: M_SEBS) % from 5% to 100%. Neutron transmittance test shows an exponential attenuation with the increase of areal density of Gd, in which the transmittance T ranges from 59.1440% to 35.3026%, with standard deviation less than 2.2743%, mass attenuation coefficient 𝜇_m from 0.3194 cm²/g to 0.4999 cm²/g, and half value layer-HVL value from 2.4530 mm to 1.1313 mm. Shielding efficiency of the Gd₂O₃/SEBS composites is basically improved in comparison with that of B₄C/SEBS. The transmittance T, mass/linear attenuation coefficient 𝜇_m and 𝜇, HVL and effective atomic number Z_eff for the shielding of γ rays (39 keV, 59 keV and 122 keV) are measured and calculated with XCOM as well as MCX programs. Finally, plots of the three dimensional relationships between transmittance, doping amount and thickness are provided to the guidance for engineering shielding design. In summary, the Gd₂O₃/SEBS composite is proved to be an effective flexible neutron/low energy γ rays shielding material, which could be of potential applications in the field of nuclear technology and nuclear engineering.