• 전자공학회논문지 IE
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• 제45권2호
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• pp.6-12
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• 2008

본 연구에서는 아스팔트 함량의 중요성을 인식하여 현재 법적 규제 면제치인 100[${\mu}Ci$]이하의 방사성 동위원소를 이용한 아스팔트 함량측정기의 개발을 목표로 하였다. 이를 위하여 중성자선원의 위치에 따라 아스팔트 함량에 변화를 주는 3가지의 분야로 나누었다. 먼저, 아스팔트 혼합물과 중성자 선원과의 간격을 줄일 경우, 반사체 설치의 경우, 이력수를 변화시켰을 경우로 나누어서 컴퓨터 시뮬레이션을 통하여 살펴보았으며, 만족스러운 오차범위 결과를 얻어 아스팔트 함량측정기기의 개발을 위한 설계 자료로 활용하고자하였다.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3044-3050
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• 2021

This study offers a Monte Carlo alternative for computing mass energy absorption coefficients of any material through calculation of photon energy deposited per mass of the sample and the energy flux obtained inside a sample volume. This approach is applied in this study to evaluate mass energy absorption coefficients of some amino acids found in human body at twenty-eight different photon energies between 10 keV and 20 MeV. The simulations involved a pencil beam source modeled to emit a parallel beam of mono-energetic photons toward a 1 mean free path thick sample of rectangular parallelepiped geometry. All the components in the problem geometry were surrounded by a 100 cm vacuum sphere to avoid any interactions in materials other than the absorber itself. The results computed using the Monte Carlo radiation transport packages MCNP6.2 and GAMOS5.1 were checked against the theoretical values available from the tables of XMUDAT database. These comparisons indicate very good agreement and support the conclusion that Monte Carlo technique utilized in this fashion may be used as a computational tool for determining the mass energy absorption coefficients of any material whose data are not available in the literature.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2614-2624
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• 2022

Monte Carlo simulations were used to model a portable Neutron backscattering (NBT) sensor suitable for detecting plastic anti-personnel mines (APMs) buried in dry and moist soils. The model consists of a 100 MBq ²⁵²Cf source encapsulated in a neutron reflector/shield assembly and centered between two ³He detectors. Multi-parameter optimization was performed to investigate the efficiency of Be/Zr(BH₄)₄ and Be/Be(BH₄)₂ assemblies in terms of increasing the signal-to-background (S/B) ratio and reducing the total dose equivalent rate. The MCNP results showed that 2 cm Be/3 cm Zr(BH₄)₄ and 2 cm Be/3 cm Be(BH₄)₂ are the optimal configurations. However, due to portability requirements and abundance of Be, the ²⁵²Cf-2 cm Be/3 cm Be(BH₄)₂ NBT model was selected to scan the center of APM buried 3 cm deep in dry and moist soils. The selected NBT model has positively identified the APM with a S/B ratio of 886 for dry soils of 1 wt% hydrogen content and with S/B ratios of 615, 398, 86, and 12 for the moist soils containing 4, 6, 10, and 14 wt% hydrogen, respectively. The total dose equivalent rate reached 0.0031 mSv/h, suggesting a work load of 8 h/day for 806 days within the permissible annual dose limit of 20 mSv.

• 방사성폐기물학회지
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• 제17권4호
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• pp.375-387
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• 2019

본 연구에서는 최근 개발중인 360 다발 장전용량의 중수로 사용후핵연료 운반용기에 대한 설계기준연료의 방사선원항 평가와 용기외부에서의 방사선량률 계산을 수행하였다. 그리고 국·내외 방사선적 안전성평가와 관련한 기술기준 부합여부를 판단하고 결과의 적합성을 제시하였다. 방사선원항으로 작용하는 설계기준연료 선정을 위해 월성원전에서 운영중인 운반 용기 및 두 가지 방식의 건식저장시설에 적용된 설계기준연료의 사양 및 특성을 조사하였다. 각 운반·저장 시스템 별 설계 기준연료의 연소도, 최소 냉각기간 및 중간저장시설로의 운반시점 등을 바탕으로 연소도 7,800 MWD/MTU와 최소 냉각기간 6년을 설계기준연료로 설정하였다. 설계기준연료의 방사선원항은 SCALE 전산코드의 ORIGEN-ARP모듈을 이용하여 평가하였다. 운반용기의 방사선차폐평가는 MCNP6 전산코드를 이용하였으며, 기술기준에서 요구하는 운반용기 외부에서의 방사선량률 평가를 정상 및 사고조건으로 구분하여 수행하였다. 방사선량률 평가결과, 정상운반조건의 운반용기 표면 및 운반용기 표면 2 m 이격지점에서 계산된 최대 방사선량률은 각각 0.330 mSv·h^-1와 0.065 mSv·h^-1로 도출되어 선량률 제한치인 2.0 mSv·h^-1와 0.1 mSv·h^-1를 모두 만족하는 결과를 도출하였다. 또한 운반사고조건하 운반용기 표면 1 m 지점에서의 최대 방사선량률은 0.321 mSv·h^-1로서 기술기준인 10.0 mSv·h^-1 미만으로 평가되어, 대용량 중수로 사용후핵연료 운반용기는 방사선적 안전성을 확보하는 것으로 나타났다.

• 한국산업융합학회 논문집
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• 제21권6호
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• pp.337-344
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• 2018

Food irradiation is important not only in ensuring safety but also improving antioxidant activity of peaches. Our objective was to establish the best irradiation treatment for peaches by calculating dose distribution using Monte Carlo simulation. 3-D geometry and component densities of peaches, extracted from CT scan, were entered into MCNP to obtain simulated dose distribution. Radiation energies for electron beam were 1.35 MeV (low energy) and 10 MeV (high energy). Co (1.25 MeV) and the Husman irradiator, containing three sealed Cs source rods in an annular array, were used for gamma irradiation. At 1.35 MeV electron beam simulation, electrons penetrated well beyond the peach skin, enough for surface treatment for microorganisms and allergens. At 10 MeV electron beam simulation, for top-beam only treatment, doses at the core were the highest and for double beam treatment, the electron energy was absorbed by the entire sample. At Co source, the radiation doses were presented on the whole area. At Cs source, the dose uniformity ratios were 2.78 for one source and 1.48 for three ones at 120 degrees interval. Proper control of irradiation treatment is critical to establish confidence in the irradiation process.

• Nuclear Engineering and Technology
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• 제54권3호
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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.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.176-189
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• 2019

This paper presents a feasibility study for producing the medical isotope $^{99m}Tc$ using the hazardous and currently wasted radioisotope $^{99}Tc$. This can be achieved with the nuclear resonance fluorescence (NRF) phenomenon, which has recently been made applicable due to high-intensity laser Compton scattering (LCS) photons. In this work, 21 NRF energy states of $^{99}Tc$ have been identified as potential contributors to the photo-production of $^{99m}Tc$ and their NRF cross-sections are evaluated by using the single particle estimate model and the ENSDF data library. The evaluated cross sections are scaled using known measurement data for improved accuracy. The maximum LCS photon energy is adjusted in a way to cover all the significant excited states that may contribute to $^{99m}Tc$ generation. An energy recovery LINAC system is considered as the LCS photon source and the LCS gamma spectrum is optimized by adjusting the electron energy to maximize $^{99m}Tc$ photo-production. The NRF reaction rate for $^{99m}Tc$ is first optimized without considering the photon attenuations such as photo-atomic interactions and self-shielding due to the NRF resonance itself. The change in energy spectrum and intensity due to the photo-atomic reactions has been quantified using the MCNP6 code and then the NRF self-shielding effect was considered to obtain the spectrums that include all the attenuation factors. Simulations show that when a $^{99}Tc$ target is irradiated at an intensity of the order $10^{17}{\gamma}/s$ for 30 h, 2.01 Ci of $^{99m}Tc$ can be produced.

• Nuclear Engineering and Technology
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• 제53권11호
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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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• 제54권12호
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• pp.4671-4678
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• 2022

The purpose of this comprehensive research is to observe the impact of scintillator crystal type on entire detection process. For this aim, MCNPX (version 2.6.0) is used for designing of a physical plastic scintillation detector available in our laboratory. The modelled detector structure is validated using previous studies in the literature. Next, different types of plastic scintillation crystals were assessed in the same geometry. Several fundamental detector properties are determined for six different plastic scintillation crystals. Additionally, the deposited energy quantities were computed using the MCNPX code. Although six scintillation crystals have comparable compositions, the findings clearly indicate that the crystal composed of PVT 80% + PPO 20% has superior counting and detecting characteristics when compared to the other crystals investigated. Moreover, it is observed that the highest deposited energy amount, which is a result of the highest collision number in the crystal volume, corresponds to a PVT 80% + PPO 20% crystal. Despite the fact that plastic detector crystals have similar chemical structures, this study found that performing advanced Monte Carlo simulations on the detection discrepancies within the structures can aid in the development of the most effective spectroscopy procedures by ensuring maximum efficiency prior to and during use.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.