• Journal of Radiation Industry
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• v.5 no.4
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• pp.325-330
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• 2011

To predict the details of flow in industrial process unit, single photon emission computed tomography (SPECT) is a promising technique. Recently, industrial SPECT based on medical system has developed by researchers of the Korea Atomic Energy Research Institute (KAERI) and Hanyang University. In the present study, to confirm the effect of the number of detectors on image quality, and determine the optimal number of detectors in industrial SPECT, industrial SPECT system with various geometries were evaluated by the Monte Carlo simulation. CsI(Tl) detectors ($12mm{\times}12mm{\times}20mm$) with collimators (the geometric resolution of collimator $R_g$ was 4 cm at the center of the 30 cm diameter cylindrical vessel object) were modeled in a hexagonal array, and the point sources of $^{99m}Tc$, $^{68}Ga$, and $^{137}Cs$ were simulated at the center of the cylindrical vessel object using the MCNPX code. Then, the reconstruction images of each geometry were reconstructed using the expectation maximization (EM) algorithm. In this study, the reciprocity theorem was used to improve computation time required for system matrix of the EM algorithm. The result shows that the resolution of the reconstructed image was significantly improved by increasing the number of detectors in industrial SPECT system and more than 60 detectors will be required for the resolution of the reconstructed image.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.875-883
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• 2015

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing $^{99}Mo$. In this method, the medical isotope production system itself is used to extract $^{99}Mo$ or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of $^{99}Mo$ by irradiating targets. In this study, the neutronic performance and $^{99}Mo$, $^{89}Sr$, and $^{131}I$ production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of $^{99}Mo$ at the end of a cycle.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3413-3420
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• 2021

Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D₁) and output (D₂) BT, respectively gives thermal neutron flux almost 25.7, 5.6 and 1.1 times greater than the former design of the TRR (with D₁ = 1.8 cm and D₁ = 9.4 cm), previous design of the TRR with D₁ = 3 cm and D₁ = 9.4 cm, and another design with D₁ = 5 cm and D₁ = 9.4 cm, respectively. Therefore, the design proposed in this paper could be a better alternative to the current BT of the TRR.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• Journal of radiological science and technology
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• v.44 no.3
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• pp.205-210
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• 2021

The purpose of this study was to analyze the spatial dose according to the distance by location of medical workers when using a mobile X-ray fluoroscopy device in the operating room through a simulation experiment. The MCNPX program was used for the simulation, and the location of medical workers was set around the operating table, and the spatial dose distribution according to the distance and changes in imaging conditions was evaluated. As a result, The highest score was 2.74×10^-4 mGy, 2.72×10^-4 mGy, and 1.18×10^-4 mGy based on the 10 cm distance from the operating table. Spatial dose depending on the distance 100cm, A point 5.15×10^-5 mGy is decreased 19% of 10cm, D point 5.12×10^-5 mGy, 19 % of 10cm, and G pint, 1.73×10^-5 mGy is reduced by 15% of 10cm. Based on this study, medical-related workers directly or indirectly participating in surgery carry potential risks of radiation exposure during surgery, but there are difficulties in radiation protection due to the nature of their work. Therefore, efforts to reduce exposure suitable for the operating room environment will be required.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Journal of radiological science and technology
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• v.45 no.4
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• pp.331-339
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• 2022

The purpose of this study was to evaluate maternal organ and fetal doses by week of pregnancy for pregnant women nuclear medicine practitioners in the nuclear medicine field. In addition, we intend to present basic data for the management of exposure doses of female nuclear medicine practitioners. In this study, phantoms of childbearing women, 3, 6, 9 months pregnant women were simulated using MCNPX(Monte Carlo N-Particle Extended) among the Monte Carlo methods. First, volume source was constructed based on 10 cm of the anterior part of the lower abdomen of the phantom, and the organ and fetal doses were evaluated for each week of the pregnant woman according to the type of radioactive isotope. Second, the organ and fetal dose of pregnant women were evaluated by increasing the distance between the source and the abdominal surface by 50 and 100 cm. As a result, ¹⁸F sources showed high organ and fetal doses in pregnant women 0 to 3 months, and the dose distribution gradually decreased in 6 to 9 months pregnant women. The distribution of organ and fetal doses for ⁹⁹mTc and ¹²³I sources showed the same tendency as that of ¹⁸F, and the overall absorbed dose distribution was relatively lower than that of ¹⁸F. Through this study, it is considered that workers in the early stages of pregnancy within 3 months will need appropriate management to minimize occupational exposure dose.

• Journal of Radiation Protection and Research
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• v.46 no.2
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• pp.58-65
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• 2021

Background: The evaluation of skyshine distribution, release of airborne radioactive nuclides, and soil activation and groundwater migration were required for radiological assessment of the impact on the environment surrounding In-Flight (IF)-system facility of the RAON (Rare isotope Accelerator complex for ON-line experiment) accelerator complex. Materials and Methods: Monte Carlo simulation by MCNPX code was used for evaluation of skyshine and activation analysis for air and soil. The concentration model was applied in the estimation of the groundwater migration of radionuclides in soil. Results and Discussion: The skyshine dose rates at 1 km from the facility were evaluated as 1.62 × 10^-3 μSv·hr^-1. The annual releases of ³H and ¹⁴C were calculated as 9.62 × 10^-5 mg and 1.19 × 10^-1 mg, respectively. The concentrations of ³H and ²²Na in drinking water were estimated as 1.22 × 10^-1 Bq·cm^-3 and 8.25 × 10^-3 Bq·cm^-3, respectively. Conclusion: Radiological assessment of environmental impact on the IF-facility of RAON was performed through evaluation of skyshine dose distribution, evaluation of annual emission of long-lived radionuclides in the air and estimation of soil activation and groundwater migration of radionuclides. As a result, much lower exposure than the limit value for the public, 1 mSv·yr^-1, is expected during operation of the IF-facility.

• Journal of radiological science and technology
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• v.47 no.1
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• pp.1-6
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• 2024

In Recent years, there has been a noticeable increase in the global incidence of breast cancer, with approximately 2.3 million cases of female breast cancer reported worldwide in 2020. Numerous studies are currently underway to enhance the accuracy of breast cancer diagnosis through the development of digital mammography detectors. This study aims to create Monte Carlo simulation-based mammographic anti-scatter grids and investigate their utility in evaluating the performance of digital mammography detector. Two types of mammographic anti-scatter grids, MAM-CP and Senographe 600T HF, were created using Monte Carlo simulation software (MCNPX 2.7.0), with grid ratios of 3.7 : 1 and 5 : 1, respectively. The grid physical characteristics (sensitivity, exposure factor, contrast improvement ratio) were calculated based on the KS C IEC60627 in the simulations using two X-ray qualities, RQA-M2 (28 kVp) and MW4 (35 kVp). As the X-ray tube voltage increased from 28 kVp to 35 kVp, sensitivity and exposure factor exhibited a decreasing trend, while contrast improvement ratio demonstrated an increasing trend. With an increase in grid ratio from 3.7 : 1 to 5 : 1, all physical characteristics showed an upward trend. Our results were consistent with a previous study that conducted measurements of physical properties using a real phantom. However, the pattern of change in the contrast improvement ratio with X-ray tube voltage differed from the previous study.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.