• Nuclear Engineering and Technology
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• 제40권5호
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• pp.375-386
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• 2008

Conventional static reliability analysis methods are inadequate for modeling dynamic interactions between components of a system. Various techniques such as dynamic fault tree, dynamic Bayesian networks, and dynamic reliability block diagrams have been proposed for modeling dynamic systems based on improvement of the conventional modeling methods. In this paper, we review these methods briefly and introduce dynamic nodes to the existing reliability graph with general gates (RGGG) as an intuitive modeling method to model dynamic systems. For a quantitative analysis, we use a discrete-time method to convert an RGGG to an equivalent Bayesian network and develop a software tool for generation of probability tables.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.167-174
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• 2011

A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.696-706
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• 2023

Consequences of an anticipated Beyond Design Basis Accident (BDBA) Long-Term Station Blackout (LTSBO) event with complete loss of grid power in the VVER-1200 reactor of Rooppur Nuclear Power Plant (NPP) of Unit-1 are assessed using the RASCAL 4.3 code. This study estimated the released radionuclides, received public radiological dose, and ground surface concentration considering 3 accident scenarios of International Nuclear and Radiological Event Scale (INES) level 7 and two meteorological conditions. Atmospheric transport, dispersion, and deposition processes of released radionuclides are simulated using a straight-line trajectory Gaussian plume model for short distances and a Gaussian puff model for long distances. Total Effective Dose Equivalent (TEDE) to the public within 40 km and radionuclides contribution for three-dose pathways of inhalation, cloudshine, and groundshine owing to airborne releases are evaluated considering with and without passive safety Emergency Core Cooling System (ECCS) in dry (winter) and wet (monsoon) seasons. Source term and their release rates are varied with the functional duration of passive safety ECCS. In three accident scenarios, the TEDE of 10 mSv and above are confined to 8 km and 2 km for the wet and dry seasons, respectively in the downwind direction. The groundshine dose is the most dominating in the wet season while the inhalation dose is in the dry season. Total received doses and surface concentration in the wet season near the plant are higher than those in the dry season due to the deposition effect of rain on the radioactive substances.

• 대한기계학회논문집A
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• 제26권3호
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• pp.576-585
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• 2002

A three-dimensional Lagrangian explicit time-integration finite element code for analyzing the dynamic impact phenomena was developed. It uses four node tetrahedral elements. In order to consider the effects of strain rate hardening, strain hardening and thermal softening, which are frequently observed in high-velocity deformation phenomena, Johnson-Cook model is used as constitutive model. For more accurate and robust contact force computation, the defense node contact algorithm was adopted and implemented. In order to evaluate the performance of the newly developed three-dimensional hydrocode NET3D, numerical simulations of the oblique impact of mild steel plate by mild steel sphere were carried out. Ballistic limit about various oblique angle between 0 degree and 80 degree was estimated through a series of simulations with different initial velocities of sphere. Element eroding by equivalent plastic strain was applied to mild steel spheres and targets. Ballistic limits and fracture characteristics obtained from simulation were compared with experimental results conducted by Finnegan et al. From numerical studies, the following conclusions were reached. (1) Simulations could successfully reproduce the key features observed in experiment such as tensile failure termed "disking"at normal impacts and outwards bending of partially formed plus segments termed "hinge-mode"at oblique impacts. (2) Simulation results fur 60 degrees oblique impact at 0.70 km/s and 0.91 km/s were compared with experimental results and Eulerian hydrocode CTH simulation results. The Lagrangian code NET3D is superior to Eulerian code CTH in the computational accuracy. Agreement with the experimentally obtained final deformed cross-sections of the projectile is excellent. (3) Agreement with the experimental ballistic limit data, particularly at the high-obliquity impacts, is reasonably good. (4) The simulation result is not very sensitive to eroding condition but slightly influenced by friction coefficient.

• 수산해양기술연구
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• 제37권1호
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• pp.45-51
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• 2001

그물의 S$_{n}$S값이 서로 다른 2가지의 상자 구조의 우리형 그물을 1/10로 축소한 모형에 대해 실물 그물의 면적 1$m^2$당 각각 0.1, 0.3, 0.5 및 0.7kg에 상당하는 침자를 그물의 각 모서리에 1개씩 총 4개를 부착하여 회류 수조에서 유속의 변화에 대한 그물의 변형을 측정하고, 이를 통해 조류에 의한 우리형 그물의 형상 유지에 효과적인 침자의 적정량에 대하여 검토하였다. 실험에서 얻어진 결과를 요약하면 다음과 같다. 1. 유속이 증가함에 따라 그물이 날림으로써 그것의 용적은 급격히 감소하였으나, 변형 각도 의 변화량은 증가하였다. 2. 그물의 S$_{n}$S값은 유속이 0.4m/s이하일 때는 그물의 변형에 약간의 영향을 끼쳤지만 유속이 0.5m/s 이상일 때는 별다른 영향을 끼치지 않았다. 3. 그물의 변형은 유속, 침자의 무게 및 S$_{n}$S값에 의해 결정되었다. 4. 그물의 용적을 50% 이상으로 유지하기 위한 침자의 총무게는 유속이 0.3~0.6 m/s일 때 31~245kg이고, 침자의 적정량은 그물의 면적 1$m^2$당 0.5kg 정도로 나타났다.

• 정보처리학회논문지:소프트웨어 및 데이터공학
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• 제12권12호
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• pp.519-524
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• 2023

해싱 기반 이미지 검색에서는 조작된 이미지의 해시코드가 원본 이미지와 달라 동일한 이미지 검색이 어렵다. 본 논문은 이미지의 질감, 모양, 색상 등 특징 정보로부터 지각적 해시코드를 생성하는 자기 감독 기반 딥해싱 모델을 제안하고 평가한다. 비교 모델은 오토인코더 기반 변분 추론 모델들이며, 인코더는 완전 연결 계층, 합성곱 신경망과 트랜스포머 모듈 등으로 설계된다. 제안된 모델은 기하학적 패턴을 추출하고 이미지 내 위치 관계를 활용하는 SimAM 모듈을 포함하는 변형 추론 모델이다. SimAM은 뉴런과 주변 뉴런의 활성화 값을 이용한 에너지 함수를 통해 객체 또는 로컬 영역이 강조된 잠재 벡터를 학습할 수 있다. 제안 방법은 표현 학습 모델로 고차원 입력 이미지의 저차원 잠재 벡터를 생성할 수 있으며, 잠재 벡터는 구분 가능한 해시코드로 이진화 된다. CIFAR-10, ImageNet, NUS-WIDE 등 공개 데이터셋의 실험 결과로부터 제안 모델은 비교 모델보다 우수하며, 지도학습 기반 딥해싱 모델과 동등한 성능이 분석되었다.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2616-2628
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• 2021

The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1661-1668
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• 2020

A physical protection system (PPS) is used for the protection of critical facilities. This paper proposes a structure analytic hierarchy approach (SAHA) for the hierarchical evaluation of the PPS effectiveness in critical infrastructure. SAHA is based on the traditional analysis methods "estimate of adversary sequence interruption, EASI". A community algorithm is used in the building of the SAHA model. SAHA is applied to cluster the associated protection elements for the topological design of complicated PPS with graphical vertexes equivalent to protection elements.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.814-826
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• 2015

The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid ($TiO_2$/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1209-1217
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• 2019

Due to the limitation of the computers, the conventional homogenization method is based on many assumptions and approximations, and some tough problems such as energy spectrum and boundary condition are faced. To deal with those problems, the Monte Carlo global homogenization is adopted. The Reactor Monte Carlo code RMC is used to study the global homogenization method, and the one-step global homogenization method is proposed. The superimposed mesh geometry is also used to divide the physical models, leading to better geometric flexibility. A set of multigroup homogenization cross sections is online generated for each mesh under the real neutron energy spectrum and boundary condition, the cross sections are adjusted by the superhomogenization method, and no leakage correction is required. During the process of superhomogenization, the author-developed reactor core program NLSP3 is used for global calculation, so the global flux distribution and equivalent homogenization cross sections could be solved simultaneously. Meanwhile, the calculated homogenization cross section could accurately reconstruct the non-homogenization flux distribution and could also be used for fine calculation. This one-step global homogenization method was tested by a PWR assembly and a small reactor model, and the results show the validity.