• 지구물리와물리탐사
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• 제18권3호
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• pp.154-161
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• 2015

본 연구에서는 중성자-감마 스펙트럼검층 존데 설계를 목적으로 Monte Carlo 시물레이션을 이용하여 열중성자 반응의 우세한 영역 파악 및 포획감마 스펙트럼의 에너지피크 값에 기초한 지층 구성 원소 구분을 수행하였다. 14 MeV 에너지준위의 중성자를 방출하는 중성자발생장치를 선원으로 이용하여 선원으로부터 10 cm 간격으로 12개의 중성자 검출기들을 배열함으로써 거리에 따른 열중성자 양을 측정하였다. 시추공 영향 저감을 위해 존데모형에 차폐재를 적용하여 보다 정확한 열중성자 측정을 수행하여 열중성자 반응이 우세한 위치를 분석한 뒤, 이 위치에서 검출된 포획감마 에너지 스펙트럼을 분석하여 지층을 구성하는 주요 원소 및 그 양을 확인하였다. 본 연구 결과는 중성자-감마 스펙트럼검층 존데의 신호대잡음 비 향상과 포획감마 검출기 최적 위치 선정에 도움이 될 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.

• Journal of Radiation Protection and Research
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• 제4권1호
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• pp.14-20
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• 1979

고(高)에너지 방사선(放射線) 치료(治療)에 있어서 정상조직(正常組織)의 완전차폐(完全遮蔽)를 위하여 $5{\sim}8cm$ 납두께의 부정형(不定形) 차폐(遮蔽)벽돌을 제작(製作)해야하는 난점(難點)이 있었다. 저자(著者)들은 납 30.0%, 주석 11.5% 비스므스 48.5%, 카드미늄 10.0%를 사중(四重) 공정결합(共晶結合)시켜 밀도(密度)가 $9.8g/cm^3$ 용융온도(熔融溫度)가 $68^{\circ}C$인 저용융(低熔融) 차폐물질(遮蔽物質)을 개발(開發)하여 이를 Lead Y라고 명명(名命)하였다. 제작(製作)된 Lead Y Block을 $68^{\circ}C$에서 용융(熔融)시켜 보호(保護)해야할 중요(重要)한 장기(臟器)의 형태(形態)대로 제작(製作)된 styrofoam 음형(陰形)에 부어서 차폐효과(遮蔽效果)가 큰 차폐(遮蔽)벽돌을 쉽고 안전(安全)하게 제작(製作)할 수 있었고 납보다 더 단단하고 재현성(再現性)이 크며 저렴(低廉)한 가격(價格)으로 구입(購入)이 가능(可能)하므로 방사선(放射線) 치료효과(治療效果)에 큰 도움을 줄 수 있었다.

• Journal of Radiation Protection and Research
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• 제47권1호
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• pp.50-57
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• 2022

Background: There are several proton therapy facilities in operation or planned in Taiwan, and these facilities are anticipated to not only treat cancer but also provide beam services to the industry or academia. The simplified approach based on the Monte Carlo-based data sets (source terms and attenuation lengths) with the point-source line-of-sight approximation is friendly in the design stage of the proton therapy facilities because it is intuitive and easy to use. The purpose of this study is to expand the Monte Carlo-based data sets to allow the simplified approach to cover the application of proton beams more widely. Materials and Methods: In this work, the MCNP6 Monte Carlo code was used in three simulations to achieve the purpose, including the neutron yield calculation, Monte Carlo-based data sets generation, and dose assessment in simple cases to demonstrate the effectiveness of the generated data sets. Results and Discussion: The consistent comparison of the simplified approach and Monte Carlo simulation results show the effectiveness and advantage of applying the data set to a quick shielding design and conservative dose assessment for proton therapy facilities. Conclusion: This study has expanded the existing Monte Carlo-based data set to allow the simplified approach method to be used for dose assessment or shielding design for beam services in proton therapy facilities. It should be noted that the default model of the MCNP6 is no longer the Bertini model but the CEM (cascade-exciton model), therefore, the results of the simplified approach will be more conservative when it was used to do the double confirmation of the final shielding design.

• Journal of Radiation Protection and Research
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• 제30권2호
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• pp.55-60
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• 2005

SMART 연구로의 노외계측기 설계를 위하여 고온 전출력 조건과 중성자 계수율이 최소가 되는 조건에 대해서 중성자속 분포 평가를 수행하였다. 고온 전출력 조건에서 IST 영역의 에너지 구간별 중성자속 분포 계산은 DORT와 MCNP코드를 이용하였으며, 계산 결과 IST 내의 첫 번째 물 영역에서 최대의 열중성자속을 보였고 두 코드 결과는 대략 10% 이내에서 일치하는 것으로 나타났다. 그리고 중성자 계수율이 최소가 되는 조건에서 노외계측기 설치 영역에서의 중성자속을 계산한 결과, 선원의 세기가 $1.0{\times}10^8(n/sec)$이라고 가정한 경우 최대 열중성자속의 크기는 $6.99{\times}10^{-2}(n/cm^2-sec)$로 전체 중성자속의 80% 이상을 차지하는 것으로 나타났는데 이는 IST 철 구조물을 통과한 속중성자가 감속능이 큰 물 영역에서 에너지를 잃고 열중성자로 변하였기 때문이다. 그러므로 노외계측기 설계시 계측기를 둘러싸는 계측기 안내관 충전물질, 설치위치 그리고 각 계측기 Segment들의 길이 등을 최적화하여 중성자 계수율을 증가시키는 방안을 모색할 필요가 있겠으며, 이러한 중성자속 평가 결과는 노외계측기가 IST 영역에 설치될 경우 노외계측기 선속 요건으로 이용될 수 있다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2335-2347
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• 2023

As medical facilities are usually built at urban areas, special concrete aggregates and evaluation methods are needed to optimize the design of concrete walls by balancing density, thickness, material composition, cost, and other factors. Carbon treatment rooms require a high radiation shielding requirement, as the neutron yield from carbon therapy is much higher than the neutron yield of protons. In this case study, the maximum carbon energy is 430 MeV/u and the maximum current is 0.27 nA from a hybrid particle therapy system. Hospital or facility construction should consider this requirement to design a special heavy concrete. In this work, magnetite is adopted as the major aggregate. Density is determined mainly by the major aggregate content of magnetite, and a heavy concrete test block was constructed for structural tests. The compressive strength is 35.7 MPa. The density ranges from 3.65 g/cm³ to 4.14 g/cm³, and the iron mass content ranges from 53.78% to 60.38% from the 12 cored sample measurements. It was found that there is a linear relationship between density and iron content, and mixing impurities should be the major reason leading to the nonuniform element and density distribution. The effect of this nonuniformity on radiation shielding properties for a carbon treatment room is investigated by three groups of Monte Carlo simulations. Higher density dominates to reduce shielding thickness. However, a higher content of high-Z elements will weaken the shielding strength, especially at a lower dose rate threshold and vice versa. The weakened side effect of a high iron content on the shielding property is obvious at 2.5 µSv=h. Therefore, we should not blindly pursue high Z content in engineering. If the thickness is constrained to 2 m, then the density can be reduced to 3.3 g/cm³, which will save cost by reducing the magnetite composition with 50.44% iron content. If a higher density of 3.9 g/cm³ with 57.65% iron content is selected for construction, then the thickness of the wall can be reduced to 174.2 cm, which will save space for equipment installation.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2010년도 제39회 하계학술대회 초록집
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• pp.314-314
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• 2010

For the optimal design of a tokamak-type reactor, self-consistent determination of a radial build of reactor systems is important and the radial build has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor systems. In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the shield should provide sufficient protection for the superconducting TF coil and the shield plays a key role in determining the size of a reactor. To determine the radial build of a reactor, neutronic effects such as tritium breeding in the blanket, nuclear heating, and radiation damage to toroidal field (TF) coil has to be included in the systems analysis. In this work, the outboard blanket only is considered where tritium self-sufficiency is possible by using an inboard neutron reflector instead of breeding blanket. The reflecting shield should provide not only protection for the superconducting TF coil but also improved neutron economy for the tritium breeding in outboard blanket. Tungsten carbide, metal hydride such as titanium hydride and zirconium hydride can be used for improved shielding performance and thus smaller shield thickness. With the use of advanced technology in the shield, conceptual design of a compact superconducting LAR reactor with aspect ratio of less than 2 will be presented as a viable power plant.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2221-2229
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• 2022

In this research, for improving the measurement performance of Prompt Gamma-ray Neutron Activation Analysis (PGNAA) set-up, a new optimization method for set-up design was proposed and investigated. At first, the calculation method for Signal-to-Noise Ratio (SNR) was proposed. Since the SNR could be calculated and quantified accurately, the SNR was chosen as the evaluation parameter in the new optimization method. For discussing the feasibility of the SNR optimization method, two kinds of PGNAA set-ups were designed in the MCNP code, based on the SNR optimization method and the previous signal optimization method, respectively. Meanwhile, the single element spectra analysis method was proposed, and the analysis effect of single element spectra as well as element sensitivity were used for comparing the measurement performance. Since the simulation results showed the better measurement performance of set-up designed by SNR optimization method, the experimental set-ups were built for the further testing, finally demonstrating the feasibility of the SNR optimization method for PGNAA setup design.

• Journal of Radiation Protection and Research
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• 제21권1호
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• pp.17-25
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• 1996

본 연구에서는 사용 후 핵연료의 안전수송을 위한 수송용기의 설계/해석 항목 중 용기 내부에 장전한 핵연료에서 방출되는 중성자의 방사선량률을 효과적으로 평가하는 방법을 구축하기 위하여 수송용기의 방사선차폐해석을 기존의 해석 수행방법인 결정론적인 방법으로 수행하고 확률론적인 방법으로 그 결과를 검증하였다. 결정론적 방법을 이용한 해석코드로 Discrete Ordinate 방법의 DOT4.2 코드를 사용하였으며, 이에 대한 비교와 검증을 위한 확률론적 방법의 차폐해석 코드로는 Monte Carlo 해법의 해석코드인 MCNP4A을 이용하였다. 동일한 대상물에 대한 방사선량율에 대한 평가를 두 방법으로 수행한 결과 두 방법으로부터의 해석결과는 큰 차이를 보이지 않았다. 이 결과비교를 통하여 사용후 핵연료 수송용기에 대한 방사선량율 평가가 올바르게 수행된 것을 확인할 수 있었고 또한 설계 및 해석에 관한 품질보중사항이 규정된 10CFR71 appendixH의 설계해석 및 전산코드 검증에 따한 요구조건을 만족시킬 수 있었다.