• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2018년도 추계 학술논문 발표대회
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4447-4464
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• 2023

This paper focuses on the development of a new computational model of the CNESTEN's TRIGA Mark II research reactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was developed to assess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor, control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used to accurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivity and uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were used to build the T4 model. In this regard, the differences between the two models are mainly due to mathematical approaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part deals with the development and the validation of the T4 model. The results obtained with the T4 model were compared to the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report (FSAR). Different core configurations were investigated via simulations to test the computational model reliability in predicting the physical parameters of the reactor. As a fairly good agreement among the results was deduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values. The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out to quantify the nuclear data uncertainty in the multiplication factor k_eff. For that purpose, the T4 model was used to calculate the sensitivity profiles of the k_eff to the nuclear data. The integrated-sensitivities were compared to the results obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools and differences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities. Moreover, the nuclear data uncertainties in the k_eff were derived using the COMAC-V2.1 covariance matrices library and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the k_eff is around 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopes to the nuclear data uncertainty in the k_eff is not negligible and should be taken into account when performing S/U analysis.

• 한국의학물리학회지:의학물리
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• 제19권1호
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• pp.21-34
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• 2008

환자의 CT자료를 기반으로 만들어진 3차원상의 표적물질에 전자 및 광자의 전달 현상을 계산하는 몬테카를로(MC) 도즈계산용 병렬프로그램 (PMCEPT 코드)을 개발하여 베어울프 PC 클러스터에 탑제하였다. 시뮬레이션에서 오차를 최소화하고 코드를 더욱 발전시키기 위해서는 현재의 MC 코드의 한계를 아는 것이 매우 유익하다. 이러한 관점에서 저자는 PMCEPT코드를 이용하여 이질 혹은 동질의 표적물질에서 표준화된 깊이 도즈를 계산하여 잘 알려진 다른 코드들, MCNP5, EGS4, DPM, GEANT4 및 실험결과와 비교를 하였다. PMCEPT결과는 이질 혹은 동질의 표적에서 다른 코드들과 $1{\sim}3%$ 오차 범위 안에서 잘 일치하였다. 계산시간 비교에 있어서도 PMCEPT 코드가 MCNP5 보다는 약 20배, GEANT4코드보다는 약 3배정도 빨랐다. 이러한 결과를 종합하면, PMCEPT코드는 의학물리분야의 시뮬레이션 코드로 사용하기에 매우 좋은 것으로 사료된다.

• 센서학회지
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• 제25권4호
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• pp.264-270
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• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

• Journal of Radiation Protection and Research
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• 제22권4호
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• pp.289-298
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• 1997

중성자 수분함량 측정기의 개발에 있어서 중성자 계측값과 흙속의 수분함량에 대한 관계식을 유도하기 위해서는 공시체 제작 등의 많은 실험을 통해 유도한 교정식이 필요하다. 또한 공시체 제작 및 측정실험의 통계적 오차를 줄이기 위해서는 많은 시간과 노력이 필요하다. 하지만 몬테카를로방법을 사용한 전산코드를 이용하여 수행할 경우 시간과 노력을 줄일 수 있을 뿐만 아니라, 보다 일반적인 흙에 대한 교정식을 얻을 수 있다. 본 연구에서는 중성자의 수송문제를 계산하는데 유용한 MCNP4A 전산코드를 이용하여 실제 실험을 모사하였다. 또한 모사결과를 공시체를 제작하여 실험한 결과와 비교하였다. 비교결과 실제실험의 결과와 모사 범위 내에서 일치함을 알 수 있었다. 중성자 수분함량 측정기의 교정식 도출 및 교정상수를 결정하기 위해 적용할 수 있음을 알 수 있었다. 또한 수분함량 측정기의 계측값에 영향을 미치는 인자중의 하나인 흙의 건조밀도 변화에 대한 영향을 살펴보았다.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• 환경생물
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• 제24권4호
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• pp.387-391
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• 2006

Manchester system 타입의 장착기중 상, 하부에 차폐체가 장착되어 있는 Henschke 장착기를 이용하여 자궁암 근접치료시 자궁 및 주변장기의 선량분포를 평가하기 위하여 치료계획수립에 사용되는 실용프로그램 결과와 몬테칼로 모의계산 결과를 비교하였다. 또한 자궁 및 주변 정상조직이 받은 선량을 계산하기 위해 ORNL(Oak Ridge National Laboratory)에서 수립한 여성의 MIRD (Medical Internal Radiation Dose)형 모의피폭체를 이용 하여 주변장기가 받는 선량을 MCNP로 계산하였다. 몬테칼로 모사에는 MCNP 4B코드를 사용하였으며, 실용계산프로그램에는 GAMMADOT를 이용하였다 MCNP계산에는 $^{192}Ir$ 선원과 장착기의 기하학적 모양을 정밀하게 모사하여 계산 오차를 줄이도록 하였으며, 치료계획용 실용계산프로그램의 계산 조건과 동일하게 치료선원의 강내 체류시간과 체류위치를 적용하여 선량을 계산하였다. 주요 선량 비교 평가점은 Manchester system에서 사용되는 4곳과 ICRU 38에서 Manchester system을 보완하기 위해 제시한 방광표면 및 직장이였다. 실용계산 결과는 MCNP모의계산의 결과와 비교했을 때 대부분 위치에서 상대오차 4% 이내의 결과를 보였고, 난형체의 차폐체 장착효과로 인한 방광과 직장에서의 선량감쇠효과는 각각 19%, 20%였다.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4019-4025
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• 2023

Recently, a new monkey computational phantom, called Visible Monkey, was developed for non-ionizing radiation studies in animal research. In this study, we extended its applications to ionizing radiation studies by implementing the voxel model of the Visible Monkey into three general-purpose Monte Carlo (MC) codes: MCNP6, PHITS, and Geant4. The implementation work for MCNP and PHITS was conducted using the LATTICE, UNIVERSE, and FILL cards. The G4VNestedParameterisation class was used for Geant4. Then, organ dose coefficients (DCs) for idealized photon beams in the antero-posterior direction were calculated using the three codes and compared, showing excellent agreement (differences <3%). Additionally, organ DCs in other directions (postero-anterior, left-lateral, and right-lateral) were calculated and compared with those of the newborn and 1-year-old reference phantoms. Significant differences were observed (e.g., the stomach DC of the monkey was 5-fold greater than that of the 1-year-old phantom at 0.03 MeV) while the differences tended to decrease with increasing energy (mostly <20% at 10 MeV). The results of this study allows conducting MC simulations using the Visible Monkey to estimate organ-level doses, which should be valuable to support/improve monkey experiments involving ionizing radiation exposures.

• Journal of Radiation Protection and Research
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• 제48권4호
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• pp.184-196
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• 2023

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities. Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO₃), and tungsten disulfide (WS₂). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z_eff), and macroscopic effective removal cross-section for fast neutrons (Σ_R) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation. Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF-xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF-0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF-0.20WO₃ and PVDF-0.20WS₂. PVDF-0.20WC also exhibited the highest Z_eff values, while PVDF-0.20WS₂ showed a slightly higher increase in Z_eff at energies of 0.662 and 1.333 MeV. PVDF-0.20WC has demonstrated the highest Σ_R value, indicating effective shielding against fast neutrons, while PVDF-0.20WS₂ had the lowest Σ_R value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF-xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate. Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF-xWC showing the most promising results.