• 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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• 제29권5호
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• pp.375-384
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• 1997

Relative power density distributions of the Kori Unit 1 pressurized water reactor are calculated by Monte Carlo modeling with the MCNP code. The Kori Unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starling with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system converges to a k value of 1.00039 $\geq$ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori Unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root moan square error of 2.159%.

• Journal of Radiation Protection and Research
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• 제48권1호
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• pp.44-51
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• 2023

Background: Aptamers are currently being used in various fields including medical treatments due to their characteristics of selectively binding to specific molecules. Due to their special characteristics, the aptamers are expected to be used to remove radionuclides from a large amount of liquid radioactive waste generated during the decommissioning of nuclear power plants. The radiological effects on the aptamers should be evaluated to ensure their integrity for the application of a radionuclide removal technique. Materials and Methods: In this study, Monte Carlo N-Particle transport code version 6 (MCNP6) and Monte Carlo damage simulation (MCDS) codes were employed to evaluate the radiological effects on the aptamers. MCNP6 was used to evaluate the secondary electron spectrum and the absorbed dose in a medium. MCDS was used to calculate the DNA damage by using the secondary electron spectrum and the absorbed dose. Binding experiments were conducted to indirectly verify the results derived by MCNP6 and MCDS calculations. Results and Discussion: Damage yields of about 5.00×10^-4 were calculated for 100 bp aptamer due to the radiation dose of 1 Gy. In experiments with radioactive materials, the results that the removal rate of the radioactive ⁶⁰Co by the aptamer is the same with the non-radioactive ⁵⁹Co prove the accuracy of the previous DNA damage calculation. Conclusion: The evaluation results suggest that only very small fraction of significant number of the aptamers will be damaged by the radioactive materials in the liquid radioactive waste.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3591-3598
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• 2023

Photoneutron background spectroscopy observations at linac are essential for directing accelerator shielding and subtracting background signals. Therefore, we constructed a Bonner Sphere Spectrometer (BSS) system based on an array of BF₃ gas proportional counter tubes. Initially, the response of the BSS system was simulated using the MCNP5 code. Next, the response of the system was calibrated by using neutrons with energies of 2.86 MeV and 14.84 MeV. Then, the system was employed to measure the spectrum of the ²⁴¹Am-Be neutron source, and the results were unfolded by using the Gravel and EM algorithms. Using the validated system, the undesirable neutron spectrum of the 120 MeV electron linac was finally measured and acquired. In addition, it is demonstrated that the equivalent undesirable neutron dose at a distance of 3.2 m from the linac is 19.7 mSv/h. The results measured by the above methods could provide guidance for linac-related research.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.397-402
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• 1996

A remote operational system for the tomographical assay was designed to scan the sample and to assay the inside radioactive materials distribution three dimensionally, composed of 3 axes moving table, collimator, data acquisition system in a PC control. The system design was done by considering that how the accurate assay be affected by the modeling or by the other system components. In the system design, MCNP code simulation was done to find the optimum condition for the spatial assay of the radioactive materials.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.408-422
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• 1999

To reduce the fast neutron fluence which deteriorates the RPV integrity, additional shields were assumed to be installed at the outer core structures of the Kori Unit 1 reactor, and its reduction effects were examined. Full scope Monte Carlo simulation with MCNP4A code was made to estimate the fast neutron fluence at the RPV. An optimized design option was found from various choices in geometry and material for shield structure. It was expected that magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV, resulting in extension of plant lifetime by 4.6 EFPYs based on the criterion of PTS requirement It was investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were only negligibly influenced by the installation of additional shield structure.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3816-3823
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• 2021

The mechanical features evaluated theoretically using Makishima-Mackenzie's model for glasses xBaO-(50-x) PbO-50P₂O₅ where x = 0, 5, 10, 15, 20, 30, 40, and 50 mol%. Wherefore, the elastic characteristics; Young's, bulk, shear, and longitudinal modulus calculated. The obtained result showed an increase in the calculated values of elastic moduli with the replacement of the PbO by BaO contents. Moreover, the Poisson ratio, micro-hardness, and the softening temperature calculated for the investigated glasses. Besides, gamma and neutron shielding ability evaluated for the barium doped lead phosphate glasses. Monte Caro code (MCNP-5) and the Phy-X/PSD program applied to estimate the mass attenuation coefficient of the studied glasses. The decrease in the PbO ratio has a negative effect on the MAC. The highest MAC decreased from 65.896 cm²/g to 32.711 cm²/g at 0.015 MeV for BPP0 and BPP7, respectively. The calculated values of EBF and EABF showed that replacement of PbO with BaO contents in the studied BPP glasses helps to reduce the number of photons accumulated inside the studied BPP glasses.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3692-3699
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• 2023

The Rare isotope Accelerator complex for ON-line experiments in Korea houses several accelerator complexes. Among them, the µSR facility will be initially equipped with a 600 MeV and 100 kW proton beam to generate surface muons, and will be upgraded to 400 kW with the same energy. Accelerated proton beams lose approximately 20% of the power at the target, and the remaining power is concentrated in the beam direction. Therefore, to ensure safe operation of the facility, concentrated protons must be distributed and absorbed at the beam dump. Additionally, effective dose levels must be lower than the legal standard, and the beam dumps used at 100 kW should be reused at 400 kW to minimize the generation of radioactive waste. In this study, we introduce a tailored method for designing beam dumps based on the characteristics of the µSR facility. To optimize the geometry, the absorbed power and effective dose were calculated using the MCNP6 code. The temperature and stress were determined using the ANSYS Mechanical code. Thus, the beam dump design consists of six structures when operated at 100 kW, and a 400 kW beam dump consisting of 24 structures was developed by reusing the 100 kW beam dump.

• Journal of Radiation Protection and Research
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• 제24권4호
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• pp.205-213
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• 1999

지표에 오염된 방사성핵종의 단위방사능당 유효선량환산계수를 남성과 여성 인형모의피폭체와 MCNP4A 코드를 이용하여 계산하였다. 모사실험은 40 keV에서 10 MeV 영역의 19개 단일 에너지에 대한 유효선량 계산을 수행하였다. 에너지에 따른 단위 선원강도에 대한 유효선량 E를 기존 연구자들의 결과물인 유효선량당량 $H_E$와 비교한 결과, 본 연구의 E값이 USEPA의 FGR에 주어진 $H_E$ 값에 비해 30%의 편차를 보였다. 에너지와 유효선량의 관계를 polynomial fitting을 통해 구한 유효선량 감응함수는 다음과 같다. $f({\varepsilon})[fSv\;m^2]=\;0.0634\;+\;0.727{\varepsilon}-0.0520{\varepsilon}^2+0.00247{\varepsilon}^3$ 여기서, ${\varepsilon}$는 감마선의 에너지(MeV)이다. 감응함수와 ICRP 38의 방사성핵종 붕괴 자료를 이용하여 지표면과 공기 오염의 단위 방사능농도에 대한 유효선량환산계수를 계산한 후 DOSEFACTOR코드를 사용하여 계산한 베타선에 의한 피부선량을 합하여 90개의 중요 핵종들에 대한 환산계수를 평가하여 도표로 제시하였다. 기존 자료들과 비교를 통해 기존 환산계수를 사용할 경우 특히 저에너지 감마선이나 고에너지 베타선을 방출하는 핵종에 대해서 상당한 과소평가가 이루어질 수 있음을 확인할 수 있었다.