• Nuclear Engineering and Technology
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• 제54권2호
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• pp.617-626
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• 2022

To effectively monitor the variety of distributions of neutron flux, fuel power or temperatures in the reactor core, usually the ex-core and in-core neutron detectors are employed. The thermocouples for temperature measurement are installed in the coolant inlet or outlet of the respective fuel assemblies. It is necessary to reconstruct the measurement information of the whole reactor position. However, the reading of different types of detector in the core reflects different aspects of the 3D power distribution. The feasibility of reconstruction the core three-dimension power distribution by using different combinations of in-core, ex-core and thermocouples detectors is analyzed in this paper to synthesize the useful information of various detectors. A comparison of multilayer perceptron (MLP) network and radial basis function (RBF) network is performed. RBF results are more extreme precision but also more sensitivity to detector failure and uncertainty, compare to MLP networks. This is because that localized neural network could offer conservative regression in RBF. Adding random disturbance in training dataset is helpful to reduce the influence of detector failure and uncertainty. Some convolution neural networks seem to be helpful to get more accurate results by use more spatial layout information, though relative researches are still under way.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3140-3153
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• 2022

A numerical feasibility study was conducted to investigate the thermal-hydraulic characteristics of a steam generator with corrugated plates for a small modular reactor. Accordingly, a one-dimensional thermal-hydraulic analysis code was developed based on the existing state-of-the-art thermal-hydraulic models and correlations for corrugated plate heat exchangers. Subsequently, the pressure loss, heat transfer, and instability characteristics of the steam generator with corrugated plates were investigated according to the chevron angle and mass flux. Additionally, the characteristics of rectangular and disk-type corrugated plate steam generators with equivalent heat transfer areas were analyzed. The steam generator with disk-type corrugated plates exhibited better performance in terms of pressure loss and heat transfer rate than the rectangular type. In addition, when the mass flux decreased from the onset of boiling points, reverse gradients of the total pressure change were observed in both types. Thus, it was confirmed that Ledinegg instability could occur in the steam generator with corrugated plates. However, it was dependent on the chevron angle, and the optimal chevron angle to minimize instability was 45° under the conditions of the present analysis.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3429-3439
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• 2022

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this paper, the radiation shielding capability of the MCS code is validated by simulating some selected SINBAD (Shielding Integral Benchmark Archive and Database) experiments. The whole validation was performed in two ways. Firstly, the functionality and computational rationality of the MCS code was verified by comparing the simulation results with those of MCNP code. Secondly, the validity and computational accuracy of the MCS code was confirmed by comparing the simulation results with the experimental results of SINBAD. The simulation results of the MCS code are highly consistent with the those of the MCNP code, and they are within the 2σ error bound of the experiment results. It shows that the calculation results of the MCS code are reliable when simulating the radiation shielding problems.

• International Journal of Advanced Culture Technology
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• 제10권3호
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• pp.295-306
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• 2022

The NOx removal performance of the SCR process depends on various factors such as catalytic factors (catalyst composition, shape, space velocity, etc.), temperature and flow rate distribution of the exhaust gas. Among them, the uniformity of the flow flowing into the catalyst bed plays the most important role. In this study, the flow characteristics in the SCR reactor in the design stage were simulated using a three-dimensional numerical analysis technique to confirm the uniformity of the airflow. Due to the limitation of the installation space, the shape of the inlet duct was compared with the two types of inlet duct shape because there were many curved sections of the inlet duct and the duct size margin was not large. The effect of inlet duct shape, guide vane or mixer installation, and venturi shape change on SCR reactor internal flow, airflow uniformity, and space utilization rate of ammonia concentration were studied. It was found that the uniformity of the airflow reaching the catalyst layer was greatly improved when an inlet duct with a shape that could suppress drift was applied and guide vanes were installed in the curved part of the inlet duct to properly distribute the process gas. In addition, the space utilization rate was greatly improved when the duct at the rear of the nozzle was applied as a venturi type rather than a mixer for uniform distribution of ammonia gas.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4628-4636
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• 2023

Cables are indispensable in nuclear power plants for transmitting data measured by various types of detectors, such as self-powered neutron detectors (SPNDs). These cables will generate disturbing signals that must be accurately distinguished and eliminated. Given that the cable current is not very significant, previous research has focused on SPND, with little attention paid to cable evaluation and validation. This paper specifically focuses on the quantitative analysis of cables and proposes a theoretical model to predict cable noise. In this model, the reaction characteristics between irradiated neutrons and cables were discussed thoroughly. Based on the Monte Carlo method, a comprehensive simulation approach of neutron sensitivity was introduced and long-term irradiation experiments in a heavy water reactor (HWR) were designed to verify this model. The theoretical results of this method agree quite well with the experimental measurements, proving that the model is reliable and exhibits excellent accuracy. The experimental data also show that the cable current accounts for approximately 0.2% of the total current at the initial moment, but as the detector gradually depletes, it will contribute more than 2%, making it a non-negligible proportion of the total signal current.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1372-1384
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• 2024

The discrimination of the source for xenon gases' release can provide an important clue for detecting the nuclear activities in the neighboring countries. In this paper, three machine learning techniques, which are logistic regression, support vector machine (SVM), and k-nearest neighbors (KNN), were applied to develop the predictive models for discriminating the source for xenon gases' release based on the xenon isotopic activity ratio data which were generated using the depletion codes, i.e., ORIGEN in SCALE 6.2 and Serpent, for the probable sources. The considered sources for the neighboring countries of South Korea include PWRs, CANDUs, IRT-2000, Yongbyun 5 MWe reactor, and nuclear tests with plutonium and uranium. The results of the analysis showed that the overall prediction accuracies of models with SVM and KNN using six inputs, all exceeded 90%. Particularly, the models based on SVM and KNN that used six or three xenon isotope activity ratios with three classification categories, namely reactor, plutonium bomb, and uranium bomb, had accuracy levels greater than 88%. The prediction performances demonstrate the applicability of machine learning algorithms to predict nuclear threat using ratios of xenon isotopic activity.

• 한국유기농업학회지
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• 제20권4호
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• pp.703-713
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• 2012

가축 분뇨 액비화 시설의 폭기효과 개선과 악취발산 감소효과를 동시에 달성하기 위한 목적으로 폭기방법이 돈분뇨 슬러리 액비 주요 부숙요인인 액비조내 슬러리의 혼합효율, 질소와 SS 그리고 냄새 발생정도에 미치는 영향을 분석하였다. 파일롯 시험조를 대상으로 한 폭기방법의 차이에 따른 혼합효과를 분석한 결과, 배관 내 혼합형 시험조의 교반 및 혼합효과가 더 높았다. 폭기방법별 처리효과 시험결과, 질소와 SS 모두 폭기가 진행됨에 따라 바닥층에서의 농도는 감소하고 반응조 상층부에서의 농도는 증가하는 경향을 보였다. 액비제조용 돈분뇨 슬러리 내에 녹아들어 간 용존산소 변화정도를 조사한 결과, 배관 내공기 혼합방식에 의한 용존산소농도가 더 높게 조사됨으로써 더 높은 산소전달 효율을 가지는 것으로 분석되었다. 바닥 고정식 공기공급 방식(A)과 배관 내 산소 혼합방식(B)에 의한 용존산소 농도는 큰 차이를 보였다. 반응조 A에서는 폭기 개시시 4.3mg/L 수준이었던 용존산소 농도가 30 분간 폭기한 후에 분석한 결과 7.0mg/L로 상승하였다. 반면에 반응조 B에서는 폭기 개시시 5.0mg/L 수준이던 용존산소 농도가 반응조 A와 동일한 폭기시간이 지난 후에 14.5mg/L로 상승함으로써 더 높은 산소전달 효율을 가지는 것으로 분석되었다. 현수식 폭기방법에 따른 용존산소 변화 정도는 바닥 고정식 공기공급 방식에서와 유사한 결과를 나타냈다. 슬러리 내부에서의 기포형성 정도를 비교하였을 때 배관 내 혼합형 방식에서 기포의 형성이나 상승 정도가 바닥 고정식 공기공급 방식에서보다 상대적으로 더 적었다.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• 한국연소학회:학술대회논문집
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• 한국연소학회 2014년도 제49회 KOSCO SYMPOSIUM 초록집
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• pp.33-36
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• 2014

Gasification of biomass produces syngas containing CO, $H_2$ and/or $CH_4$, which can then be converted into energy or value-added fuels. One of key issues for efficient gasification is to minimize tar concentration in the syngas for use in a final conversion device such as gas engine. This study investigated the decomposition of primary tar by catalytic cracking using char as catalyst, of which the feature can be integrated into a fixed bed gasifier design. The pyrolysis vapor containing tar from pyrolysis of wood at $500^{\circ}C$ was passed through a reactor filled with or without char at $800^{\circ}C$ for a residence time of 1, 3 or 5 sec. Then, the condensable vapor (water and tar) and gases were analyzed for the yields and elemental composition. Four types of char particles with different microscopic surface area and pore size distribution: wood, paddy straw, palm kernel shell and activated carbon. The results were analyzed for the mass and carbon yields of tar and the composition of product gases to conclude the effects of char types and residence time.

• 대한인간공학회지
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• 제33권1호
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• pp.27-37
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• 2014

Objective: This study aims to improve human performance by analyzing the operators' tasks and providing input data on the composition of future SMART operators. Background: SMART is a nuclear reactor for export which needs operators who can satisfy both safety and economic feasibility. Therefore, this study is fundamental research on the composition of operators and this research analyzed SMART tasks in terms of human safety performance. Method: After analyzing 10 SMART EOG in hierarchical task analysis, this study classified task performance types according to task requirements of NUREG-0711 (Rev.3). Results: This study found the task frequency of SMART EOG and 12 operating task types. Conclusion: This study expects that human performance can be improved by analyzing the personal errors, which have the highest task frequency among 12 operating task types. Application: The results of this study can be applied as base data when licensing needs to be acquired.