• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.5
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• pp.329-337
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• 2016

As a turbulence-enhancing device, a mixing vane, which is installed at a spacer grid of the fuel assembly, plays an important role in improving convective heat transfer by generating either swirl flow in the subchannels or cross flow between the fuel rod gaps. Therefore, both the geometric configuration and the arrangement pattern of a mixing vane are important factors in determining the performance of a mixing vane. In this study, in order to examine the flow-distribution features inside a $5{\times}5$ fuel assembly with split-type mixing vanes, which was used in the benchmark calculation of the OECD/NEA, we conduct simulations using the commercial computational fluid dynamics software, ANSYS CFX R.14. We compare the predicted results with measured data obtained from the MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, we discuss the effect of the split-type mixing vanes on the flow pattern inside the fuel assembly.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.149-156
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• 2016

The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the $3{\times}3$ short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

• Journal of Fisheries and Marine Sciences Education
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• v.26 no.2
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• pp.245-256
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• 2014

잉어, Cyprinus carpio와 동자개, Pseudobagrus fulvidraco의 상악 하악 수염을 조직학적으로 조사하였다. 동자개의 수염은 연골성 증축(axial rod of cartilage), 신경섬유다발(bundle of nerve fiber), 표피(epidermis), 평활근 층(smooth muscle layer) 및 미뢰(taste bud)로 구성되었으며, 잉어의 수염은 표피, 신경섬유다발, 혈관(blood vessel) 및 미뢰로 구성되었다. 수염 길이에서 잉어는 상악 바깥쪽 수염(second maxillary barbel)이 상악 안쪽 수염(first maxillary barbel) 보다 길게 나타났으며, 동자개는 하악 안쪽(inner mandibular barbel), 상악 위쪽(upper maxillary barbel), 하악 바깥쪽(outer mandibular barbel), 상악 아래쪽(lower maxillary barbel) 순으로 길게 나타났다(P<0.05). 미뢰의 수를 고려하였을 때, 동자개와 잉어간의 미각에 대한 유의적 차이가 없었다(P>0.05). 아울러, 두 어종의 모든 수염에서 수염 상부의 미뢰 수가 하부의 미뢰 수 보다 높게 나타났다(P<0.05). 본 연구 결과, 동자개의 수염은 딱딱하며 굴절성인 수염(flexible and stiff type)이었으며 잉어의 수염은 연하고 유연한 수염(tender and yielding type)으로 파악되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.284-289
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• 1996

The predictions of the COBRA-IV-I code with the modified turbulent mixing and void drift models have been compared with the diabatic two-phase flow data on equilibrium quality. The turbulent mixing model based on an equal mass exchange of the existing COBRA-IV-I code has been modified to that based on an equal volume exchange between adjacent subchannels, and a void drift model has been newly incorporated in the code. To evaluate the performance of the equal volume exchange turbulent mixing model and the effects of the void drift model, the diabatic steam-water two-phase flow data obtained for the 9-rod bundle test under the typical operating conditions of the boiling water reactor(BWR) conducted by the General Electric (GE) were analyzed by the modified COBRA-IV-I code. The analysis indicates that the equal volume exchange turbulent mixing model with void drift predicts the observed two-phase flow data trends better than the equal mass exchange model, and to predict the correct data trends a more physically based void drift model need to be developed.

• Proceedings of the KIEE Conference
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• 2000.07a
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• pp.131-133
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• 2000

Single phase simulations were carried out in order to determine a solutive conductor to the aeolian noise which will be locally applied to 765 kV transmission lines Basic solutive conductors have already been proposed including conductors equipped with spiral rod. low noise conductor of a special shape and others. A low aeolian noise conductor, however, should have excellent corona characteristics in addition to aeolian noise reduction function. In this paper, we compared the performances of the audible noises and radio interferences of 6 candidate conductor bundles by using corona cage. We also developed two programs to need for evaluating environmental effects of each conductor bundle. Those are a program to calculate the conductor surface gradient of various special bundles and a conversion program of single phase data to the model of transmission line. The future determination on the final low aeolian noise conductor will be made through a long-term test to verify environmental impacts at the full-scale Kochang 765 kV test line.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.389-400
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• 1994

The predictability of KAERI version of RELAP5/MOD3 on reflood peak cladding temperature during large break loss-of-coolant accident is assessed against 18 test runs in FLECHT SEASET test data. The associated uncertainty is statistically quantified. The selected test runs include a gravity feed test and several forced feed tests with wide range of the parameters such as flooding rate, system pressure, initial clad temperature, rod bundle power. The results show that the code under-predicts the peak cladding temperature by 7.56 K on average. The upper limit of the associated uncertainty at 95% confidence level is evaluated to be about 99 K, It including the bias due to the under-prediction.

• Journal of the Korean Society of Tobacco Science
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• v.36 no.1
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• pp.1-11
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• 2014

Cigarette filters containing Brassica Rapa L. seeds of the dark brown and round shaped were evaluated to determine the effect of seed addition on filter degradability. The seeds with germination capability under the tar/nicotine condition in the preliminary test, were put into the active carbon part of the filter(12mm) during filter rod making by the kit. The $4{\pm}2$ pieces of the seeds were put into the opened fiber bundle of the filter tow. In order to test the germination rate of the seeds, seed filters were placed either in a petri dish or test-pot in a conditioned area ($25^{\circ}C$, 70 % RH). The seed filters were placed outdoors exposed to natural conditions with the periodic water supply. The seeds in the smoked filters showed 90 % germination rate after a month under the open air condition. No significant differences in the sensory evaluation and analysis were obtained when the control sample was compared to the same cigarettes with the seeds.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.23 no.2
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• pp.132-138
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• 2011

In the present article, we investigate numerically turbulent flow of air through compound rectangular channels. Large eddy simulation(LES) is employed for unsteady turbulence modeling. LES gives better predictions for the axial mean velocity distribution than those of other turbulent models. Strong large-scale quasi-periodic flow oscillations are observed in most of the geometries investigated. Such large-scale flow oscillations in compound rectangular channels are similar to the quasi-periodic flow pulsation through the gaps between fuel rod bundle in nuclear reactor. It exists in any longitudinal connecting gap between two flow channels. The frequency of this flow oscillation is determined by the geometry of the gap. The large scale cross motions through the rectangular compound channels induce significant heat transfer enhancement of the compound channel flow.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.79-90
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• 2005

In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.