• Geophysics and Geophysical Exploration
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• v.15 no.4
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• pp.219-226
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• 2012

This study provides an analysis on results of neutron logging for various borehole environments through numerical simulation based on a Monte Carlo N-Particle (MCNP) code developed and maintained by Los Alamos National Laboratory. MCNP is suitable for the simulation of neutron logging since the algorithm can simulate transport of nuclear particles in three-dimensional geometry. Rather than simulating a specific tool of a particular service company between many commercial neutron tools, we have constructed a generic thermal neutron tool characterizing commercial tools. This study makes calibration chart of the neutron logging tool for materials (e.g., limestone, sandstone and dolomite) with various porosities. Further, we provides correction charts for the generic neutron logging tool to analyze responses of the tool under various borehole conditions by considering brine-filled borehole fluid and void water, and presence of borehole fluid.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1610-1616
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• 2017

To validate the new Evaluated Nuclear Data File $(ENDF)/B-VIII.0{\beta}4$ library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the $ENDF/B-VIII.0{\beta}4$ library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.785-794
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• 2016

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ~0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper.

• Journal of Radiation Protection and Research
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• v.49 no.1
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• pp.50-64
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• 2024

Background: The reference dose coefficients (DCs) of the International Commission on Radiological Protection (ICRP) have been widely used to estimate organ doses of individuals for risk assessments. This approach has been well accepted because individual anatomy data are usually unavailable, although dosimetric uncertainty exists due to the anatomical difference between the reference phantoms and the individuals. We attempted to quantify the individual variation of organ doses for photon external exposures by calculating and comparing organ DCs for 30 individuals against the ICRP reference DCs. Materials and Methods: We acquired computed tomography images from 30 patients in which eight organs (brain, breasts, liver, lungs, skeleton, skin, stomach, and urinary bladder) were segmented using the ImageJ software to create voxel phantoms. The phantoms were implemented into the Monte Carlo N-Particle 6 (MCNP6) code and then irradiated by broad parallel photon beams (10 keV to 10 MeV) at four directions (antero-posterior, postero-anterior, left-lateral, right-lateral) to calculate organ DCs. Results and Discussion: There was significant variation in organ doses due to the difference in anatomy among the individuals, especially in the kilovoltage region (e.g., <100 keV). For example, the red bone marrow doses at 0.01 MeV varied from 3 to 7 orders of the magnitude depending on the irradiation geometry. In contrast, in the megavoltage region (1-10 MeV), the individual variation of the organ doses was found to be negligibly small (differences <10%). It was also interesting to observe that the organ doses of the ICRP reference phantoms showed good agreement with the mean values of the organ doses among the patients in many cases. Conclusion: The results of this study would be informative to improve insights in individual-specific dosimetry. It should be extended to further studies in terms of many different aspects (e.g., other particles such as neutrons, other exposures such as internal exposures, and a larger number of individuals/patients) in the future.

• Progress in Medical Physics
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• v.33 no.4
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• pp.129-135
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• 2022

Purpose: A full-energy-peak (FEP) efficiency correction is required through a Monte Carlo simulation for accurate radioactivity measurement, considering the geometrical characteristics of the detector and the sample. However, a relative deviation (RD) occurs between the measurement and calculation efficiencies when modeling using the data provided by the manufacturers due to the randomly generated dead layer. This study aims to optimize the structure of the detector by determining the dead layer thickness based on Monte Carlo simulation. Methods: The high-purity germanium (HPGe) detector used in this study was a coaxial p-type GC2518 model, and a certified reference material (CRM) was used to measure the FEP efficiency. Using the MC N-Particle Transport Code (MCNP) code, the FEP efficiency was calculated by increasing the thickness of the outer and inner dead layer in proportion to the thickness of the electrode. Results: As the thickness of the outer and inner dead layer increased by 0.1 mm and 0.1 ㎛, the efficiency difference decreased by 2.43% on average up to 1.0 mm and 1.0 ㎛ and increased by 1.86% thereafter. Therefore, the structure of the detector was optimized by determining 1.0 mm and 1.0 ㎛ as thickness of the dead layer. Conclusions: The effect of the dead layer on the FEP efficiency was evaluated, and an excellent agreement between the measured and calculated efficiencies was confirmed with RDs of less than 4%. It suggests that the optimized HPGe detector can be used to measure the accurate radioactivity using in dismantling and disposing medical linear accelerators.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1734-1741
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• 2023

This work presents the synthesis and preparation of a new glass system described by the equation of (70-x) B₂O₃-5TeO₂ -20SrCO₃-5ZnO -xBi₂O₃, x = 0, 1, 5, 10, and 15 mol. %, using the melt quenching technique at a melting temperature of 1100 ℃. The photon-shielding characteristics mainly the linear attenuation coefficient (LAC) of the prepared glass samples were evaluated using Monte Carlo (MC) simulation N-particle transport code (MCNP-5) at gamma-ray energy extended from 59 keV to 1408 keV emitted by the radioisotopes Am-241, Ba-133, Cs-137, Co-60, Na-22, and Eu-152. Furthermore, we observed that the Bi₂O₃ content of the glasses had a significantly stronger impact on the LAC at 59 and 356 keV. The study of the lead equivalent thickness shows that the performance of fabricated glass sample with 15 mol.% of Bi₂O₃ is four times less than the performance of pure lead at low gamma photon energy while it is enhanced and became two times lower the perforce of pure lead at high energy. Therefore, the fabricated glasses special sample with 15 mol.% of Bi₂O₃ has good shielding properties in low, intermediate, and high energy intervals.

• Bulletin of the Korean Space Science Society
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• 2011.04a
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• pp.28.1-28.1
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• 2011

Radiation-induced displacement damage which has caused the increase of the dark current in the focal plane adopted in the Ozone Monitoring Instrument (OMI) was studied in regards of the primary protons and the secondaries generated by the protons in the orbit. By using the Monte Carlo N-Particle Transport Code System (MCNPX) version 2.4.0 along with the Stopping and Range of Ions in Matter version 2010 (SRIM2010), effects of the primary protons as well as secondary particles including neutron, electron, and photon were investigated. After their doses and fluxes that reached onto the charge-coupled device (CCD) were examined, displacement damage induced by major sources was presented.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.751-757
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• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• Journal of Environmental Science International
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• v.13 no.2
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• pp.175-180
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• 2004

A prompt gamma-ray neutron activation (PGNA) system was simulated by the Monte Carlo N-Particle transport code (MCNP-4A) to estimate the level at which the scattered photon fluence rate, the absolute efficiency of the HPGe-detector, the volume of the concrete sample and the $^{35}$ /Cl(n, ${\gamma}$) reaction rate in this sample contribute to the count rate in the NaCl concentration measurement. The n- ${\gamma}$ fluence rates at the ST-2 beam tube exit of the HANARO reactor were used as input data, and the GAMMA-X type HPGe detector was modeled to tally 1.1649 MeV ${\gamma}$ -rays emitted from the $^{35}$ Cl(n, ${\gamma}$) reaction in the concrete sample. For three cylindrical concrete samples of 13.8, 46.8 and 157.1 ㎤ volumes, respectively, the relations between the NaCl weight fractions of 0.1, 1, 2 and 5 % in each of the concrete samples and the 1.1 649 MeV pulses created in the HPGe detector model were studied. As a result, it was found that the count rate at the same NaCl concentration nearly depends on the volume of the samples in a simulated condition of the same NaCl concentration samples, and that the linearities of the NaCl concentration calibration curves were reasonable in the narrow range of the NaCl weight fraction.