• Nuclear Engineering and Technology
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• 제43권6호
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4049-4061
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• 2022

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

• ETRI Journal
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• 제29권6호
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• pp.835-837
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• 2007

A multi-bit sigma-delta modulator architecture is described for low-distortion performance and a high-speed operation. The proposed architecture uses both a delayed code and a delayed differential code of analog-to-digital converter in the feedback path, thereby suppressing signal components in the integrators and relaxing the timing requirement of the analog-to-digital converter and the scrambler logic. Implemented by a 0.13 ${\mu}m$ CMOS process, the sigma-delta modulator achieves high linearity. The measured spurious-free dynamic range is 89.1 dB for -6 dBFS input signal.

• ETRI Journal
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• 제31권2호
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• pp.240-242
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• 2009

A computationally efficient implementation of the progressive edge-growth algorithm is presented. This implementation uses an array of red-black (RB) trees to manage the layered structure of check nodes and adopts a new strategy to expand the Tanner graph. The complexity analysis and the simulation results show that the proposed approach reduces the computational effort effectively. In constructing a low-density parity check code with a length of $10^4$, the RB-tree-array-based implementation takes no more 10% of the time required by the original method.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2001년도 학술대회 논문집
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• pp.53-56
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• 2001

The stability of PF 6-7 has been studied according to the transient analysis code TOKSCPF and quench analysis code QSAIT. We compare the stability and temperature rise with various Cu/SC ratios of 2.8 and 3.5 under the KSTAR normal operating conditions. It shows that the Cu/SC ratio has an influence on the quench propagation and stability margin. In transient operating condition, the Cu/SC ratio weakly influences on the temperature rise in PF magnet.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.1966-1973
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• 2023

Virtual reality technology has been widely used in the field of nuclear and radiation safety, dose rate computing in virtual environment is essential for optimizing radiation protection and planning the work in radioactive-controlled area. Because the CPU-based gamma dose rate computing takes up a large amount of time and computing power for voxelization of volumetric radioactive source, it is inefficient and limited in its applied scope. This study is to develop an efficient gamma dose rate computing code and apply into fast virtual simulation. To improve the computing efficiency of the point kernel algorithm in the reference (Li et al., 2020), we design a GPU-based computing framework for taking full advantage of computing power of virtual engine, propose a novel voxelization algorithm of volumetric radioactive source. According to the framework, we develop the GPPK(GPU-based point kernel gamma dose rate computing) code using GPU programming, to realize the fast dose rate computing in virtual world. The test results show that the GPPK code is play and plug for different scenarios of virtual simulation, has a better performance than CPU-based gamma dose rate computing code, especially on the voxelization of three-dimensional (3D) model. The accuracy of dose rates from the proposed method is in the acceptable range.

• 한국멀티미디어학회논문지
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• 제12권8호
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• pp.1172-1181
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• 2009

전자의무기록(Electronic Medical Record, EMR)은 모든 검사 과정이 텍스트 기반의 데이터 형태로 저장되는 의료 분야의 의무기록 시스템을 의미한다. 그러나 국내의 전자의무기록 시스템은 각 의료기관마다 고유한 의료정보검사코드 형태를 이용하여 기록하는 방식으로 정보를 저장하기 때문에 병원 간의 의료검사 기록 형태들의 공유, 해석, 분석에 많은 문제점들을 가진다. 위의 문제들을 해결하기 위하여 표준화 되어 있지 않은 병원들의 검사코드들을 LOINC (Logical Observation Identifiers Names and Code)로 표준화하려는 연구들이 많다. 현재까지의 연구들은 로컬 의료정보검사코드를 수동으로 LOINC로 변환하는 방법이 연구되었다. 또한 대용량 의학 정보들을 다루기에 적절하지 않은 파일 기반에서 코드들을 관리하는 연구들이 이루어져왔다. 기존의 문제점을 해결하기 위하여 본 논문에서는 의료 용어 표준화 알고리즘을 제안하고, 구현하여 해결하였다. 또한, 대표적인 상용시스템이 가졌던 문제점인 검색어를 의사가 직접 생성해야 했던 부분을 LOINC 의 여섯 가지 자동 속성 추출 및 검색어 자동 생성 기능을 구현하여 해결하였다. 또한, 기존의 시스템들이 고려하지 않았던 대용량 데이터의 매핑 부분을 파일 시스템 기반이 아닌 데이터베이스 기반 검색 프레임웍을 구축하였다.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.19-29
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• 2022

The 2D/1D method has become the mainstream of the direct transport calculation considering the balance of accuracy and efficiency. However, the 2D/1D method still suffers from stability issues. Recently, a quasi-3D method has been proposed with axial Legendre expansion. Analysis and comparison of the 2D/1D and quasi-3D method is conducted in theory from the equation derivation. Besides, the C5G7 benchmark, the KUCA benchmark and the macro BEAVRS benchmark are calculated to verify the theory comparisons of these two methods with the direct transport code SHARK. All results show that the quasi-3D method has better stability and accuracy than the 2D/1D method with worse efficiency and memory cost. It provides a new option for direct transport calculation with the quasi-3D method.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.