• Nuclear Engineering and Technology
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• 제41권1호
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• pp.135-142
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• 2009

In this investigation, we prepared a 1 and 3-parallel serpentine single PEMFC, which has an active area of $100\;cm^2$ and a flow channel cross section of $1{\times}1mm$. Distribution and transport of water in a non-operating PEMFC were observed by varying flow types and the flow rates (250, 400, and 850 cc/min). This investigation was performed at the neutron imaging facility at the CO1d Neutron RAdiography facility (CONRAD), HMI, Germany of which the collimation ratio and neutron fluence rate are 250, $1{\times}10^{6}n/s/cm^2$, respectively. The neutron image was continuously recorded by a scintillator and lens-CCD coupled detector system every 10 seconds. It has been observed that although the distilled water was supplied into the cathode channel only, the neutron image showed a water movement from the cathode to the anode channel. The water at the cathode channel was completely discharged as soon as the pressurized air was supplied. But the water at the anode channel was not easily removed by the pressurized air except for the 3-parallel serpentine type with 850cc/min of air flow rate. Moreover, the water at the MEA wasn't removed for any of the cases.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.103-108
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• 1995

By applying the P$_1$ and P$_2$ equations to the operator form of a synthetic acceleration, we derive the p,-acceleration (diffusion synthetic acceleration: DSA) and P$_2$-acceleration (p$_2$SA) schemes in one dimensional slab geometry. We Fourier-analyze the derived acceleration schemes with the discrete-ordinates transport equation and showed that the DSA outperforms the P$_2$SA. These results confirm that one cannot simply assume that replacement of the DSA with a higher order approximation will lead to a better acceleration performance.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1860-1865
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• 2019

The confinement degradation of the energetic particles during RMP would be a key issue in success of realizing the successful energy production using fusion plasma, because a 3.5 MeV energetic alpha particle should be able to sustain the burning plasma after the ignition. As KSTAR recent results indicate the generation of high-performance plasma(${\beta}_p{\sim}3$), the confinement of the energetic particles is also an important key aspect in neutral beam driven plasma. In general, the measured absolute value of the neutron intensity is generally used for to estimating the confinement time of energetic particles by comparing it with the theoretical value based on transport calculations. However, the availability of, but for its calculation process, many accurate diagnostic data of plasma parameters such as thermal and incident fast ion density, are essential to the calculation process. In this paper, the time evolution of the neutron signal from an He3 counter during the beam blank has permitted to facilitate the estimation of the slowing down time of energetic particles and the method is applied to investigate the fast ion effect on ELM suppressed KSTAR plasma which is heated by high energy deuterium neutral beams.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1769-1780
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• 2022

A new method of dose calculation algorithm, called GPU-accelerated Monte Carlo and collapsed cone Convolution (GMCC) was developed to improve the calculation speed of BNCT treatment planning system. The GPU-accelerated Monte Carlo routine in GMCC is used to simulate the neutron transport over whole energy range and the Collapsed Cone Convolution method is to calculate the gamma dose. Other dose components due to alpha particles and protons, are calculated using the calculated neutron flux and reaction data. The mathematical principle and the algorithm architecture are introduced. The accuracy and performance of the GMCC were verified by comparing with the FLUKA results. A water phantom and a head CT voxel model were simulated. The neutron flux and the absorbed dose obtained by the GMCC were consistent well with the FLUKA results. In the case of head CT voxel model, the mean absolute percentage error for the neutron flux and the absorbed dose were 3.98% and 3.91%, respectively. The calculation speed of the absorbed dose by the GMCC was 56 times faster than the FLUKA code. It was verified that the GMCC could be a good candidate tool instead of the Monte Carlo method in the BNCT dose calculations.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1071-1083
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• 2023

The on-the-fly energy release per fission (OTFK) model is implemented in STREAM to continuously update the Kappa values during the depletion calculation. The explicit neutron and photon energy distribution, which has not been considered in previous STREAM versions, is incorporated into the existing on-the-fly model. The impacts of the modified OTFK model with explicit neutron and photon heating in STREAM on the power distribution, fuel temperature, and other core parameters during depletion with feedback calculations are studied using several problems from the VERA benchmark suit. Overall, the explicit heating calculation provides a better power map for the feedback calculations particularly when strong gamma emitters are present. Generally, the fuel temperature decreases when neutron and photon heating is employed because fission neutrons and gamma rays are transported away from their points of generation. This energy release model in STREAM indicates that gamma energy accounts for approximately 9.5%-10% of the total energy released, and approximately 2.4%-2.6% of the total energy released will be deposited in the coolant for the VERA 5, NuScale, and Yonggwang Unit 3 2D cores.

• Journal of Radiation Protection and Research
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• 제38권1호
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• pp.37-43
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• 2013

TLD를 이용하여 중성자 선량을 측정할 경우, TLD는 중성자 에너지에 대한 반응도 차이가 크기 때문에 현장 중성자장의 스펙트럼 특성에 맞는 에너지 반응도 보정이 반드시 필요하다. 본 실험에는 소형으로 가공된 TLD 소자를 사용하여 $^{252}Cf$ 중성자장에 설치된 내부구조가 복잡하고 좁은 Long-Counter (중성자 검출기) 내외부에서의 중성자 주위선량당량(ambient dose equivalent)을 측정하였다. 측정결과는 입자수송해석코드(MCNPX)를 이용한 계산결과와 비교하였다. 기존의 TLD 교정 선원인 $D_2O$ 감속 $^{252}Cf$만으로 교정하여 판독한 결과값은 전산모사 계산값과 많은 차이를 보였다. 그러나 bare 및 $D_2O$ 감속 $^{252}Cf$ 선원을 사용하여 생산한 두 교정인자를 혼용한 판독값은 계산값과 비슷하였다. 결과적으로, TLD 소자는 사용 현장과 비슷한 특성을 가지는 중성자장에서 교정되어야지만 올바른 선량평가가 가능함을 확인하였다.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1593-1603
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• 2023

In deterministic and Monte Carlo transport codes, b-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available b-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code - named Time Dependent OpenMC or OpenMC(TD)- time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.182-183
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• 2004

• 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.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.225-229
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• 2001

In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is $5.89 {\times} 10^{17}\;n/cm^2$ and that on the radial surface of reactor vessel is $4.49 {\times} 10^[16}\;n/cm^2$. These results meet the requirement, $1.0 {\times} 10^{20}\;n/cm^2$, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed.