• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3080-3087
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• 2023

Thermoelectric (TE) materials working in radioisotope thermoelectric generators are irradiated by neutrons throughout its service; thus, investigating the neutron irradiation stability of TE devices is necessary. Herein, the influence of neutron irradiation with fluences of 4.56 × 10¹⁰ and 1 × 10¹³ n/cm² by pulsed neutron reactor on the electrical and thermal transport properties of n-type Bi₂Te_2.7Se_0.3 and p-type Bi_0.5Sb_1.5Te₃ thermoelectric alloys prepared by cold-pressing and molding is investigated. After neutron irradiation, the properties of thermoelectric materials fluctuate, which is related to the material type and irradiation fluence. Different from p-type thermoelectric materials, neutron irradiation has a positive effect on n-type Bi₂Te_2.7Se_0.3 materials. This result might be due to the increase of carrier mobility and the optimization of electrical conductivity. Afterward, the effects of p-type and n-type TE devices with different treatments on the output performance of TE devices are further discussed. The positive and negative effects caused by irradiation can cancel each other to a certain extent. For TE devices paired with p-type Bi_0.5Sb_1.5Te₃ and n-type Bi₂Te_2.7Se_0.3 thermoelectric legs, the generated power and conversion efficiency are stable after neutron irradiation.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1350-1364
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• 2023

2D/1D coupling method is an important neutron transport calculation method due to its high accuracy and relatively low computation cost. However, 2D/1D coupling method may diverge especially in small axial mesh size. To analyze the convergence behavior of 2D/1D coupling method, a Fourier analysis for k-eigenvalue neutron transport problems is implemented. The analysis results present the divergence problem of 2D/1D coupling method in small axial mesh size. Several common attempts are made to solve the divergence problem, which are to increase the number of inner iterations of the 2D or 1D calculation, and two times 1D calculations per outer iteration. However, these attempts only could improve the convergence rate but cannot deal with the divergence problem of 2D/1D coupling method thoroughly. Moreover, the choice of axial solvers, such as DGFEM SN and traditional SN, and its effect on the convergence behavior are also discussed. The results show that the choice of axial solver is a key point for the convergence of 2D/1D method. The DGFEM SN based 2D/1D method could converge within a wide range of optical thickness region, which is superior to that of traditional SN method.

• Journal of Radiation Protection and Research
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• 제42권4호
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• pp.212-221
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• 2017

Background: This study aims to calculate detector positions as a design of a radioactive source localizing radiation portal monitor (RPM) system using an improved genetic algorithm. Materials and Methods: To calculate of detector positions for a source localizing RPM system optimization problem is defined. To solve the problem, a modified iterative genetic algorithm (MIGA) is developed. In general, a genetic algorithm (GA) finds a globally optimal solution with a high probability, but it is not perfect at all times. To increase the probability to find globally optimal solution rather, a MIGA is designed by supplementing the iteration, competition, and verification with GA. For an optimization problem that is defined to find detector positions that maximizes differences of detector signals, a localization method is derived by modifying the inverse radiation transport model, and realistic parameter information is suggested. Results and Discussion: To compare the MIGA and GA, both algorithms are implemented in a MATLAB environment. The performance of the GA and MIGA and that of the procedures supplemented in the MIGA are analyzed by computer simulations. The results show that the iteration, competition, and verification procedures help to search for globally optimal solutions. Further, the MIGA is more robust against falling into local minima and finds a more reliably optimal result than the GA. Conclusion: The positions of the detectors on an RPM for radioactive source localization are optimized using the MIGA. To increase the contrast of the measurements from each detector, a relationship between the source and the detectors is derived by modifying the inverse transport model. Realistic parameters are utilized for accurate simulations. Furthermore, the MIGA is developed to achieve a reliable solution. By utilizing results of this study, an RPM for radioactive source localization has been designed and will be fabricated soon.

• 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.

• Composites Research
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• 제36권6호
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• pp.435-440
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• 2023

본 연구에서는 기존 사용후핵연료(Spent nuclear fuel) 운반/저장 용기에 사용되는 중성자 흡수용 B₄C/Al 복합소재의 열전도도를 개선하기 위해 탄화붕소(B₄C)와 입방정 질화붕소(cBN)를 동시에 강화재로 사용한 알루미늄(Al) 기지 복합소재를 제조하고 평가를 진행하였다. 이를 위해서 교반주조 공정을 통해 복합재 잉곳을 제조하고 이를 압연하여 중성자 흡수용 소재를 성공적으로 제조하였다. 제조된 소재의 평가를 위해 cBN 첨가에 따른 열전도도와 중성자 흡수능 변화를 관찰하였다. 열전도도 측정 결과, B₄C 단일 입자만을 사용한 복합소재 대비 B₄C, cBN을 함께 사용한 복합소재가 동일 체적률 조건 하에서 약 3%의 열전도도 증가가 발생하는 것을 확인하였으며 중성자 흡수 단면적 계산을 통해 중성자 흡수능이 무시할 수 있는 수준으로 저하가 발생하는 것을 확인하였다. 본 연구의 결과를 바탕으로, 중성자 흡수 소재의 새로운 설계 방안을 제시하고 고성능 운반/저장 용기의 개발에 기여할 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2452-2459
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• 2020

Neutron spectrum is essential to the safe operation of reactors. Traditional online neutron spectrum measurement methods still have room to improve accuracy for the application cases of wide energy range. From the application of artificial neural network (ANN) algorithm in spectrum unfolding, its accuracy is difficult to be improved for lacking of enough effective training data. In this paper, an adaptive deviation-resistant neutron spectrum unfolding method based on transfer learning was developed. The model of ANN was trained with thousands of neutron spectra generated with Monte Carlo transport calculation to construct a coarse-grained unfolded spectrum. In order to improve the accuracy of the unfolded spectrum, results of the previous ANN model combined with some specific eigenvalues of the current system were put into the dataset for training the deeper ANN model, and fine-grained unfolded spectrum could be achieved through the deeper ANN model. The method could realize accurate spectrum unfolding while maintaining universality, combined with detectors covering wide energy range, it could improve the accuracy of spectrum measurement methods for wide energy range. This method was verified with a fast neutron reactor BN-600. The mean square error (MSE), average relative deviation (ARD) and spectrum quality (Qs) were selected to evaluate the final results and they all demonstrated that the developed method was much more precise than traditional spectrum unfolding methods.

• 한국전기전자재료학회논문지
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• 제21권4호
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• pp.381-387
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• 2008

Neutron production rate in spherically convergent beam fusion(SCBF) device as a portable neutron source strongly depends on the ion current and the grid cathode structure. In this paper, as the process of design and analysis, Design of Experiment(DOE) based on the results by Finite Element Method-Flux Corrected Transport(FEM-FCT) method is employed to calculate the ion current. This method is very useful to find optimal design conditions in a short time. Number of rings, radius of rings, and distance between the grid cathode and center are selected as control factors. From the results in the optimized model, the higher ion current is calculated and deeper potential well is also observed.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2013년도 제45회 하계 정기학술대회 초록집
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• pp.311.2-311.2
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• 2013

Transmutation characteristics of minor actinides in a transmutation reactor based on a Low Aspect Ratio (LAR) tokamak are investigated. One-dimensional neutron transport and burn-up calculation coupled with the tokamak systems analysis were performed to find the optimal system parameters. The dependence of the transmutation characteristics such as neutron multiplication factor, produced power and transmutation rate on an aspect ratio A in the range of 1.5 to 2.0 was investigated. By adding Pu239 in the transmutation blanket as a neutron multiplication material, it was shown that the one unit of the transmutation reactor based on the LAR tokamak producing fusion power of 150 MWth can destroy the minor actinides contained in the spent fuels produced from more than 19 units of l GWe PWRs with production of the power being in the range of 0.9 - 3.4 GWth.

• 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.