• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3808-3815
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• 2021

In this paper, the EPICS2017 photoatomic database was used to evaluate the photon mass attenuation coefficients and buildup factors of soils collected at different depths in the Philippine islands. The extraction and interpolation of the library was accomplished at the recommended linear-linear scales to obtain the incoherent and total cross section and mass attenuation coefficient. The buildup factors were evaluated using the G-P fitting method in ANSI/ANS-6.4.3. An agreement was achieved between XCOM, MCNP5, and EPICS2017 for the calculated mass attenuation coefficient values. The buildup factors were reported at several penetration depths within the standard energy grid. The highest values of both buildup factor classifications were found in the energy range between 100 and 400 keV where incoherent scattering interaction probabilities are predominant, and least at the region of predominant photoionization events. The buildup factors were examined as a function of different soil silica contents. The soil samples with larger silica concentrations were found to have higher buildup factor values and hence lower shielding characteristics, while conversely, those with the least silica contents have increased shielding characteristics brought by the increased proportions of the abundant heavier oxides.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3051-3057
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• 2021

The concrete is considered as an important radiation shielding material employed widely in nuclear reactors, particle accelerators, laboratory hot cells and other different radiation sources. The present research is dedicated to the shielding properties study of the ordinary concrete reinforced with different weight fractions of lead oxide micro/nano particles. Lead oxide particles were fabricated by chemical synthesis method and their properties including the average size, morphological structure, functional groups and thermal properties were characterized by XRD, FESEM-EDS, FTIR and TGA analysis. The gamma ray mass attenuation coefficient of concrete composites has been calculated and measured by means of the Monte Carlo simulation and experimental methods. The simulation process was based on the use of MCNP Monte Carlo code where the mass attenuation coefficient (μ/ρ) has been calculated as a function of different particle sizes and filler weight fractions. The simulation results showed that the employment of the lead oxide filler particles enhances the mass attenuation coefficient of the ordinary concrete, drastically. On the other hand, there are approximately no differences between micro and nano sized particles. The mass attenuation coefficient was increased by increasing the weight fraction of nanoparticles. However, a semi-saturation effect was observed at concentrations more than 10 wt%. The experimental process was based on the fabrication of concrete slabs filled by different weight fractions of nano lead oxide particles. The mass attenuation coefficients of these slabs were determined at different gamma ray energies using ²²Na, ¹³⁷Cs and ⁶⁰Co sources and NaI (Tl) scintillation detector. The experimental results showed that the HVL parameter of the ordinary concrete reinforced with 5 wt% of nano PbO particles was reduced by 64% at 511 keV and 48% at 1332 keV. Reasonable agreement was obtained between simulation and experimental results and showed that the employment of nano PbO particles is more efficient at low gamma energies up to 1Mev. The proposed concrete is less toxic and could be prepared in block form instead of toxic lead blocks.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.959-964
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• 2022

Lead is an important material, both for fusion or fission reactors. The cross sections of natural lead should be validated because lead is a main component of lithium-lead modules suggested for fusion power plants and it directly affects the crucial variable, tritium breeding ratio. The presented study discusses a validation of the lead transport libraries by dint of the activation of carefully selected activation samples. The high emission standard ²⁵²Cf neutron source was used as a neutron source for the presented validation experiment. In the irradiation setup, the samples were placed behind 5 and 10 cm of the lead material. Samples were measured using a gamma spectrometry to infer the reaction rate and compared with MCNP6 calculations using ENDF/B-VIII.0 lead cross sections. The experiment used validated IRDFF-II dosimetric reactions to validate lead cross sections, namely ¹⁹⁷Au(n, 2n)¹⁹⁶Au, ⁵⁸Ni(n,p)⁵⁸Co, ⁹³Nb(n, 2n)^92mNb, ¹¹⁵In(n,n')^115mIn, ¹¹⁵In(n,γ)^116mIn, ¹⁹⁷Au(n,γ)¹⁹⁸Au and ⁶³Cu(n,γ)⁶⁴Cu reactions. The threshold reactions agree reasonably with calculations; however, the experimental data suggests a higher thermal neutron flux behind lead bricks. The paper also suggests ²⁵²Cf isotropic source as a valuable tool for validation of some cross-sections important for fusion applications, i.e. reactions on structural materials, e.g. Cu, Pb, etc.

• Journal of the Korean Society of Radiology
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• v.15 no.6
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• pp.803-811
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• 2021

In this study, the exposure amount of IASCC test worker was evaluated by applying the process simulation technology. Using DELMIA Version 5, a commercial process simulation code, IASCC test facility, hot cells, and workers were prepared, and IASCC test activities were implemented, and the cumulative exposure of workers passing through the dose-distributed space could be evaluated through user coding. In order to simulate behavior of workers, human manikins with a degree of freedom of 200 or more imitating the human musculoskeletal system were applied. In order to calculate the worker's exposure, the coordinates, start time, and retention period for each posture were extracted by accessing the sub-information of the human manikin task, and the cumulative exposure was calculated by multiplying the spatial dose value by the posture retention time. The spatial dose for the exposure evaluation was calculated using MCNP6 Version 1.0, and the calculated spatial dose was embedded into the process simulation domain. As a result of comparing and analyzing the results of exposure evaluation by process simulation and typical exposure evaluation, the annual exposure to daily test work in the regular entrance was predicted at similar levels, 0.388 mSv/year and 1.334 mSv/year, respectively. Exposure assessment was also performed on special tasks performed in areas with high spatial doses, and tasks with high exposure could be easily identified, and work improvement plans could be derived intuitively through human manikin posture and spatial dose visualization of the tasks.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.38 no.10
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• pp.1038-1048
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• 2010

Radiation exposures of an astronaut during the space travels to the International Space Station(ISS) of the Soyuz and the Moon of the Apollo, were calculated considering the altitude, boarding time, period of stay, kinds of spaceships and space suits. The calculated radiation exposures decrease dramatically according to the thickness of the shielding by the wall of the spaceships and by the space suits. For the space travel to the ISS of Soyuz at Low Earth orbit, the thickness of the spaceship required to optimally reduce the radiation exposure is 3 cm. For the Extravehicle Mobility Unit(EMU) the exposures are minimized at 4 cm of the aluminized Mylar and 5 cm of the Demron, respectively. The aluminized Mylar showed better radiation shielding than the Demron which contains the high Z materials. The radiation exposures of an astronaut were $4.2\times10^{-6}$ Sv for the ISS travel and $4.3\times10^{-5}$ Sv for the Moon explore. The high concentration of the high energy proton flux at the surface of the Moon results in high radiation exposure. The calculation scheme and results of this study can be used in the design of the shielding performance of a spaceship and space suits.

• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.200-208
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• 2011

The neutron fluence measurement and evaluation technology is very important for material irradiation test. The most essential technology in this study is the neutron irradiation evaluation method using a fluence monitor. The fluence monitors were fabricated with metal wires of the purity ${\geq}$ 99.9%, whose dimensions were 0.1mm diameter, about 3 mm length, and around 150-200 ${\mu}g$ mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. After irradiation tests, radiation activities were measured with the high purity germanium (HPGe) detector. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position.

• Journal of Radiation Protection and Research
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• v.29 no.4
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• pp.263-268
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• 2004

Using the Monte Carlo simulation, a study on the lion-proportionality of the prototype phoswich detector with $2'{\times}2'$ CSI(Tl) and plastic scintillator, which was made by KAERI, has been carried. The defector response functions (DRFs) calculated by simulations were compared with the experimental measurement on the $^{137}Cs\;and\;^{60}Co$. To precisely simulate the DRF for the phoswich, the CSI(Tl) non-proportionality was calculated using the electron response and the simplified electron cascade sequence for treating the photoelectric absorption event. The resulting DRFs of $^{137}Cs\;and\;^{60}Co$ sources obtained by simulations were compared with experiments for verification. For $^{137}Cs$, gamma-ray responses simulated by MCNP5 are generally good agreement with the measured ones. But the DRF of $^{60}Co$ does not match well with the results of experiment in the energy region below second peak due to the coincidence effect of two gamma-rays (1.17 MeV and 1.33 MeV). Through the analysis of the non-proportionality of CsI(Tl) in the prototype phoswich, the improved DRFs considering non-proportionality were produced and the simulation results were verified using the experimental measurements. However, to more precisely reproduce the DRF for the phoswich, further studies in relation to the electron channeling effect and the Doppler broadening effect of a scintillator are still needed as well as considering that effect of the transfer contribution.

• 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.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.2
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• pp.145-151
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• 2013

Purpose: In this study, we analyzed how the dose change by field size effects on atomic number of shielding materials while using 6 MeV election beam. Materials and Methods: The parallel plate chamber is mounted in $25{\times}25cm^2$ the phantom such that the entrance window of the detector is flush with the phantom surface. phantom was covered laterally with aluminum, copper and lead which thickness have 5% of allowable transmission and then the doses were measured in field size $6{\times}6$, $10{\times}10$ and $20{\times}20cm^2$ respectively. 100 cGy was irradiated using 6 MeV electron beam and SSD (Source Surface Distance) was 100 cm with $10{\times}10cm^2$ field size. To calculate the photon flux, electron flux and Energy deposition produced after pass materals respectively, MCNPX code was used. Results: The results according to the various shielding materials which have 5% of allowable transmission are as in the following. Thickness change rate with field size of $6{\times}6cm^2$ and $20{\times}20cm^2$ that compared to the field size of $10{\times}10cm^2$ found to be +0.06% and -0.06% with aluminum, +0.13% and -0.1% with copper, -1.53% and +1.92% with lead respectively. Compare to the field size $10{\times}10cm^2$, energy deposition for $6{\times}6cm^2$ and $20{\times}20cm^2$ had -4.3% and +4.85% respectively without shielding material. With aluminum it had -0.87% and +6.93% respectively and with lead it had -4.16% and +5.57% respectively. When it comes to photon flux with $6{\times}6cm^2$ and $20{\times}20cm^2$ of field sizes the chance -8.95% and +15.92% without shielding material respectively, with aluminum the number -15.56% and +16.06% respectively and with copper the chance -12.27% and +15.53% respectively, with lead the number +12.36% and -19.81% respectively. In case of electron flux in the same condition, the number -3.92% and +4.55% respectively without shielding material respectively, with aluminum the number +0.59% and +6.87% respectively, with copper the number -1.59% and +3.86% respectively, with lead the chance -5.15% and +4.00% respectively. Conclusion: In this study, we found that the required thickness of the shielding materials got thinner with low atomic number substance as the irradiation field is increasing. On the other hand, with high atomic number substance the required thickness had increased. In addition, bremsstrahlung radiation have an influence on low atomic number materials and high atomic number materials are effected by scattered electrons.