• Journal of Radiation Protection and Research
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• 제48권4호
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• pp.175-183
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• 2023

Background: Recently, biological adsorbents have been developed for removing radionuclides from radioactive liquid waste due to their high selectivity, eco-friendliness, and renewability. However, since they can be damaged by radiation in radioactive waste, a method for estimating the bio-adsorbent performance as a time should consider the radiation damages in terms of their renewability. This paper aims to develop a simulation method that applies a deep learning technique to rapidly and accurately estimate the adsorption performance of bio-adsorbents when inserted into liquid radioactive waste. Materials and Methods: A model that describes various interactions between a bio-adsorbent and liquid has been constructed using numerical methods to estimate the adsorption capacity of the bio-adsorbent. To generate datasets for machine learning, Monte Carlo N-Particle (MCNP) simulations were conducted while considering radioactive concentrations in the adsorbent column. Results and Discussion: Compared with the result of the conventional method, the proposed method indicates that the accuracy is in good agreement, within 0.99% and 0.06% for the R² score and mean absolute percentage error, respectively. Furthermore, the estimation speed is improved by over 30 times. Conclusion: Note that an artificial neural network can rapidly and accurately estimate the survival rate of a bio-adsorbent from radiation ionization compared with the MCNP simulation and can determine if the bio-adsorbents are reusable.

• 지질공학
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• 제27권4호
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• pp.429-438
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• 2017

지질학, 자원공학의 다양한 분야에서 적용되는 중성자검층의 활용성과 측정값의 신뢰성을 높이기 위하여 Monte Carlo 알고리듬에 기초한 MCNP 모델링을 수행하였다. 본 연구에서는 중성자 존데와 지층의 수치모형화를 기본으로 존데 고유의 교정곡선과 MCNP 모델링으로 계산한 교정곡선과의 비교를 통해 모델링의 적정성을 확인하고 암상변화, 공극 유체 특성, 시추공 지름 변화, 케이싱 영향, 공내수 영향을 모델링 결과를 이용하여 고찰하였다. 모델링 결과, 암상 변화에 따른 중성자 계수율 비율의 변화를 정량적으로 파악하였다. 시추공 지름이 존데와 비슷한 3인치의 경우, 지름이 큰 경우보다 계수율의 비가 예상보다 높게 나타났다. 이와 같은 결과는 공내수 영향이 작은 영향으로 해석되었다. 케이싱 내에서의 반응과 나공에서의 반응을 비교할 때 전반적으로 차이가 작았으며 특히 지층의 공극률이 증가하면 케이싱 영향이 감소하여 구분이 어려웠다. 지하수위 상부에 대한 모델링 결과는 지하수위 하부와는 반대의 경향을 나타냈으며 지하수위 파악에도 정성적으로 이용이 가능할 것으로 예상된다. 다양한 시추공 환경에 대한 모델링 결과는 중성자검층 현장자료의 자료처리와 해석이 유용하게 이용될 것으로 예상된다.

• 한국의학물리학회지:의학물리
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• 제33권4호
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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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• 제53권3호
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• pp.942-948
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• 2021

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1885-1891
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• 2023

The present work aims to fabricate Na_1+xZr₂Si_xP_3-xO₁₂ compound at various calcination temperatures based on the zircon mineral. The fabricated compound was calcinated at 250, 500, and 1000℃. The effect of calcination temperature on the structure, crystal phase, and radiation shielding properties was studied for the fabricated compound. The X-ray diffraction diffractometer demonstrates that, the monoclinic crystal phase appeared at a calcination temperature of 250℃ and 500℃ is totally transformed to a high-symmetry hexagonal crystal phase under a calcination temperature of 1000℃. The radiation shielding capacity was also qualified for the fabricated compounds using the Monte Carlo N-Particle transport code in the g-photons energy interval between 15keV and 122keV. The impacts of calcination temperature on the g-ray shielding behavior were clarified in the present study, where the linear attenuation coefficient was enhanced by 218% at energy of 122keV, when the calcination temperature increased from 250 to 1000℃, respectively.

• Nuclear Engineering and Technology
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• 제55권5호
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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.

• Nuclear Engineering and Technology
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• 제48권3호
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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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• 제49권1호
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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.

• 한국의학물리학회지:의학물리
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• 제24권4호
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• pp.253-258
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• 2013

본 연구의 목적은 electronic portal imaging device (EPID)를 통하여 방사선 치료를 받는 환자로부터 투과해 나오는 선량으로 외부적인 선량 재구성과 몬테카를로 전산모사로부터 도출되는 내부 선량 계측과의 관계를 도출하고 이를 분석하기 위한 연구로 진행되었다. 본 연구는 전산모사 연구로써 두 가지의 경우를 비교 분석하고 이와 비슷한 연구에 대한 기본적인 지표를 제공하고자 시행되었다. 실험에 관한 기하학적 정보와 방사선 소스에 대한 정보를 몬테카를로 전산모사 툴인 Monte Carlo n-particle (MCNPX)에 입력하였고 EPID 이미지 도출을 위하여 MCNPX 내에 tally카드를 이용하여 선량정보를 도출하고 이를 영상화 할 수 있도록 하였다. 또한 내부적인 계측을 위하여 물 팬텀을 소스와 표면의 거리(source to surface distance, SSD)가 100 cm이 되도록 설정하였으며, 그보다 10 cm 아래에 EPID를 위치시켰다. 내부 계측은 물팬텀 자체에서 흡수되는 흡수 선량을 mesh tally로 수집하였고, 4문 조사를 통하여 중첩된 선량에 대한 데이터를 획득하였다. 그와 동시에 EPID에서 물을 투과해 나오는 선량을 획득 한 뒤 역 투사 방법을 사용하여 선량 재구성을 하였다. 이둘의 경우를 비교하기 위해 자체적인 교정(calibration)을 통하여 투과해 나온 선량과 흡수된 선량과의 관계를 비교하고 4문 조사를 통하여 물 팬텀 내의 특정 부분에 대한 중첩된 선량 데이터와 EPID를 통해 재구성한 선량 데이터를 분석하였다. 물 팬텀과 EPID에서 획득한 누적 선량의 합은 각각 평균 3.4580 MeV/g과 3.4354 MeV/g이었다. 이는 앞서 계측된 물 팬텀 내부의 누적 선량과 0.6536% 선량 오차를 보였다.

• 방사선산업학회지
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• 제9권1호
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• pp.1-7
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• 2015

The trend of x-ray image sensor has been evolved from an amorphous silicon sensor to a crystal silicon sensor. A crystal silicon X-ray sensor, meaning a X-ray CIS (CMOS image sensor), is consisted of three transistors (Trs), i.e., a Reset Transistor, a Source Follower and a Select Transistor, and a photodiode. They are highly sensitive to radiation exposure. As the frequency of exposure to radiation increases, the quality of the imaging device dramatically decreases. The most well known effects of a X-ray CIS due to the radiation damage are increments in the reset voltage and dark currents. In this study, a pixel array of a X-ray CIS was made of $20{\times}20pixels$ and this pixel array was exposed to a high radiation dose. The radiation source was Co-60 and the total radiation dose was increased from 1 to 9 kGy with a step of 1 kGy. We irradiated the small pixel array to get the increments data of the reset voltage and the dark currents. Also, we simulated the radiation effects of the pixel by MCNP (Monte Carlo N-Particle) simulation. From the comparison of actual data and simulation data, the most affected location could be determined and the cause of the increments of the reset voltage and dark current could be found.