• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.788-793
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• 2018

High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3418-3422
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• 2007

The neutron imaging technique was used to investigate the water discharge characteristics at PEMFC. Prior to investigation of water discharge characteristics, the linear attenuation coefficient for water at Neutron Radiography Facility (NRF) was calibrated. The feasibility test apparatus was consisted of pressurized air and water in order to simulate the actual operating PEMFC. The feasibility tests have been performed at 1-parallel serpentine type with 100 $cm^2$ active area and different air flow rate (1, 2, and 4 lpm). The total water volume variations at each condition were calculated from the neutron images. The water at channel is well discharged as soon as supplying the pressurized air into the PEMFC. However, because the water at MEA isn't removed the total water volume is constant after 150. Therefore more effective method is needed in order to discharge water at MEA, and the neutron imaging technique is helpful for it.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3506-3513
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• 2023

In this paper, we demonstrate the neutron attenuation of dry cask shielding materials using the S670e helium-4 detector manufactured by Arktis Radiation Ltd. In particular, two materials expected to be applied to the TN-32 dry cask manufactured by ORANO Korea and KORAD-21 by the Korea Radioactive Waste Agency (KORAD) were utilized. The measured neutron attenuation was compared with our Monte Carlo N-Particle Transport simulation results, and the difference is given as the root mean square (RMS). For the fast neutron case, a rapid decline in neutron counts was observed as a function of increasing material thickness, exhibiting an exponential relationship. The discrepancy between the experimentally acquired data and simulation results for the fast neutron was maintained within a 2.3% RMS. In contrast, the observed thermal neutron count demonstrated an initial rise, attained a maximum value, and exhibited an exponential decline as a function of increasing thickness. In particular, the discrepancy between the measured and simulated peak locations for thermal neutrons displayed an RMS deviation of approximately 17.3-22.4%. Finally, the results suggest that a minimum thickness of 5 cm for Li-6 is necessary to achieve a sufficiently significant cross-section, effectively capturing incoming thermal neutrons within the dry cask.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2607-2612
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• 2020

The present work was aimed to show the possibility of using hydrogel (acrylic/boric acid) for evaluation of the neutron radiation shielding. The influence of acrylic acid concentration, different gamma doses and relative contents of boric acid were studied. The physical properties and the thermomechanical stability of the studied samples were investigated. The shielding property of the composite for neutron was tested by Pu-Be neutron source (5 Ci) under room temperature. The neutron fluence rates and gamma fluxes were measured using a stilbene organic scintillator. The macroscopic effective removal cross-section Σ_R (cm^-1) of fast neutrons and total attenuation coefficient μ (cm^-1) of gamma rays has been studied experimentally. The transmission parameters, the relaxation length (??) and the half-value layer (HVL) were obtained. The obtained results indicated that the addition of boric acid to acrylic acid tends to increase the macroscopic effective removal cross-section Σ_R (cm^-1) to 0.141 compared to 0.094 of ordinary concrete.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.818-824
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• 2019

In the present work, we determined the gamma-ray attenuation characteristics of eight different polymers(Polyamide (Nylon 6) (PA-6), polyacrylonitrile (PAN), polyvinylidenechloride (PVDC), polyaniline (PANI), polyethyleneterephthalate (PET), polyphenylenesulfide (PPS), polypyrrole (PPy) and polytetrafluoroethylene (PTFE)) using transmission geometry utilizing the high resolution HPGe detector and different radioactive sources in the energy range 81-1333 keV. The experimental linear attenuation coefficient values are compared with theoretical data (WinXCOM data). The linear attenuation coefficient of all polymers reduced quickly with the increase in energy, at the beginning, while decrease more slowly in the region from 267 keV to 835 keV. The effective atomic number of PVDC and PTFE are comparatively higher than the $Z_{eff}$ of the remaining polymers, while PA-6 possesses the lowest effective atomic number. The half value layer results showed that PTFE ($C_2F_4$, highest density) is more effective to attenuate the gamma photons. Also, the theoretical results of macroscopic effective removal cross section for fast neutrons ($\sum_{R}$) were computed to investigate the neutron attenuation characteristics. It is found that the $\sum_{R}$ values of the eight investigated polymers are close and ranged from $0.07058cm^{-1}$ for PVDC to $0.11510cm^{-1}$ for PA-6.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2419-2431
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• 2023

Experiments related to Boron Neutron Capture Therapy (BNCT) accomplished at the Institute of Nuclear Techniques (INT), Budapest University of Technology and Economics (TUB) are presented. Relevant investigations are required before designing BNCT for vivo applications. Samples of relevant boron compounds (H₃BO₃, BDTPA) usually employed in BNCT were investigated with neutron beam. Channel #5 in the research reactor (100 kW) of INT-TUB provides the neutron beam. Boron samples are mounted on a carrier for neutron irradiation. The particle attenuation of several carrier materials was investigated, and the one with the lowest attenuation was selected. The effects of boron compound type, mass, and compound phase state were also investigated. To detect the emitted charged particles, a traditional ZnS(Ag) detector was employed. The neutron beam's interaction with the detector-detecting layer is investigated. Graphite (as a moderator) was employed to change the neutron beam's characteristics. The fast neutron beam was also thermalized by placing a portable fast neutron source in a paraffin container and irradiating the H₃BO₃. The obtained results suggest that the direct measurement approach appears to be insufficiently sensitive for determining the radiation dose committed by the Alpha particles from the ¹⁰B (n,α) reaction. As a result, a new approach must be used.

• Journal of Radiation Protection and Research
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• v.43 no.1
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• pp.20-28
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• 2018

Background: Heat energy produced in nuclear reactors and nuclear fuel cycle facilities interactions modifies the physical properties of the shielding materials containing water content. Therefore, in the present paper, effect of the heat on shielding effectiveness of the concretes is investigated for gamma and neutron. The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors. Materials and Methods: The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors of ordinary and heavy concretes were investigated using NIST data of XCOM program and Geometric Progression method. Results and Discussion: The improvement in shielding effectiveness for photon and reduction in fast neutron for ordinary concrete was observed. The change in the neutron shielding effectiveness was insignificant. Conclusion: The present investigation on interaction of gamma and neutron radiation would be very useful for assessment of shielding efficiency of the concrete used in high temperature applications such as reactors.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3012-3017
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• 2021

The spatial distribution and the energy spectrum of the neutron yield were investigated with the neutron activation analysis and MCNP simulation was carried out to verify the analysis results and to extend the results to a 3D mapping of the neutron yield distribution in the KSTAR. High purity Ni specimen was selected in the neutron activation analysis. Total neutron yields turned out to be 3.76 × 10¹² n/s - 7.56 × 10¹² n/s at the outer vessel of the KSTAR, two orders of magnitude lower than those at the inner vessel of the KSTAR, which demonstrates the attenuation of neutron yield while passing through the different structural materials of the reactor. Based on the fully expanded 3D simulation results, 2D cross-sectional distributions of the neutron yield on XY and ZX planes of KSTAR were examined. The results reveal that the neutron yield has its maximum concentration near the center of blanket and decreases with increasing proximity to the vacuum vessel wall.