• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.67-71
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• 2001

In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.2005-2012
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• 2019

This work investigated the radiation attenuation characteristics of three series of tellurite glass systems with the following compositions: 30PbO-10ZnO-xTeO₂-(60-x)B₂O₃ where x = 10, 30, 40, 50 and 60 mol%, xBaO-xB₂O₃-(100-2x)TeO₂ with x = 15-40 mol% and 50ZnO-(50-x)P₂O₅-xTeO₂, where x = 0, 10, .40 mol%. The results revealed that the attenuation parameters in all the samples decrease with increase in the energy, which implied that all the samples have better interaction with gamma photons at low energies and thus higher photon attenuating efficiency. From the three systems, the samples coded as PbZnBTe60, BaBTe70 and ZnPTe40 have the lowest half value layer values and accordingly have superior photon attenuation efficacy. The maximum effective atomic number values were found for energy less than 0.1 MeV particularly near the K-edges absorption of the heavy atomic number elements such as Te, Ba and Pb. At the lowest energy, the Z_eff values are found in the range of 62.33-66.25, 49.43-50.81 and 24.99-35.83 for series 1-3 respectively. Also, we found that the density of the glass remarkably affects the photon attenuation ability of the selected glasses. The mean free path results showed that the PbO-ZnO-TeO₂-B₂O₃ glass system has better radiation shielding efficiency than the glass samples in series 2 and 3.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.

• Structural Engineering and Mechanics
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• v.87 no.5
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• pp.499-516
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• 2023

To defend against harmful gamma radiation, new types of materials for use in the construction of heavyweight concrete (HWC) are still needed to be developed. This research introduces new materials to be employed as a partial replacement for fine aggregate (FA) to manufacture high-performance heavyweight concrete (HPHWC). These materials include hematite, black sand, ilmenite, and magnetite, with substitution ratios of 50% and 100% of FA. In this research, the hardening and fresh characteristics of HPHWC were obtained. Concrete samples' Gamma-ray linear attenuation coefficient was evaluated utilizing a gamma source of Co-60 through the thicknesses of 2.5, 5, 7.5, 10, 12.5, and 15 cm. High temperatures were studied for HPHWC samples, which were exposed to up to 700℃ for two hours. Energy-dispersive x-rays and a scanning electron microscope carried out microstructure analyses. Magnetite as an FA attained the lowest compressive strength of 87.1 MPa, but the best radiation protection characteristics and the highest density of 3100 kg/m³ were achieved. After 28 days, the attenuation efficiency of concrete mixtures was increased by 6.5% when fine sand was replaced with black sand at a ratio of 50%. HPHWC, which contains hematite, black sand, ilmenite, and magnetite, is designed to reduce environmental and health dangers and be used in medicinal, military, and civil applications.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.215-224
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• 2001

In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShieid utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Journal of radiological science and technology
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• v.45 no.1
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• pp.57-67
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• 2022

CZT detectors, which are compound semiconductors that have been widely used recently for gamma-ray detection purposes, are difficult to detect neutrons because direct interaction with them does not occur unlike gamma-rays. In this paper, a method of detecting and determining energy levels (fast neutrons and thermal neutrons) of neutrons, in addition of identifying energy and nuclide of gamma-rays, and evaluating gamma dose rates using a CZT semiconductor detector is described. Neutrons may be detected by a secondary photoelectric effect or compton scattering process with a characteristic gamma-ray of 558.6 keV generated by a capture reaction (¹¹³Cd + ¹n → ¹¹⁴Cd + 𝛾) with cadmium (Cd) in the CZT detector. However, in the case of fast neutrons, the probability of capture reaction with cadmium (Cd) is very low, so it must be moderated to thermal neutrons using a moderator and the material and thickness of moderator should be determined in consideration of the portability and detection efficiency of the equipment. Conversely, in the case of thermal neutrons, the detection efficiency decreases due to shielding effect of moderator itself, so additional CZT detector that do not contain moderator must be configured. The CZT detector that does not contain moderator can be used to evaluate energy, nuclide, and gamma dose-rate for gamma-rays. The technology proposed in this paper provides a method for detecting both neutrons and gamma-rays using a CZT detector.

• Journal of Radiation Protection and Research
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• v.47 no.4
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• pp.214-225
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• 2022

Background: The conventional cerium-doped Gd₂Al₂Ga₃O₁₂ (GAGG(Ce)) scintillator-based gamma-ray imager has a bulky detector, which can lead to incorrect positioning of the gammaray source if the shielding against background radiation is not appropriately designed. In addition, portability is important in complex environments such as inside nuclear power plants, yet existing gamma-ray imager based on a tungsten mask tends to be weighty and therefore difficult to handle. Motivated by the need to develop a system that is not sensitive to background radiation and is portable, we changed the material of the scintillator and the coded aperture. Materials and Methods: The existing GAGG(Ce) was replaced with Bi₄Ge₃O₁₂ (BGO), a scintillator with high gamma-ray detection efficiency but low energy resolution, and replaced the tungsten (W) used in the existing coded aperture with lead (Pb). Each BGO scintillator is pixelated with 144 elements (12 × 12), and each pixel has an area of 4 mm × 4 mm and the scintillator thickness ranges from 5 to 20 mm (5, 10, and 20 mm). A coded aperture consisting of Pb with a thickness of 20 mm was applied to the BGO scintillators of all thicknesses. Results and Discussion: Spectroscopic characterization, imaging performance, and image quality evaluation revealed the 10 mm-thick BGO scintillators enabled the portable gamma-ray imager to deliver optimal performance. Although its performance is slightly inferior to that of existing GAGG(Ce)-based gamma-ray imager, the results confirmed that the manufacturing cost and the system's overall weight can be reduced. Conclusion: Despite the spectral characteristics, imaging system performance, and image quality is slightly lower than that of GAGG(Ce), the results show that BGO scintillators are preferable for gamma-ray imaging systems in terms of cost and ease of deployment, and the proposed design is well worth applying to systems intended for use in areas that do not require high precision.

• Journal of Radiation Protection and Research
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• v.6 no.1
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.