• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.72-77
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• 2010

During the period of reinforcement work for the licensing renewal of CANDU NPP, the fuel channels, Calandria tubes and feeders of CANDU Reactor are replaced. The remote visual inspection of Calandria internal is also performed during the period of reinforcement work. This period is a unique opportunity to inspect the inside of the Calandria. The visual inspection for the Calandria vessel and its internals of Wolsong NPP Unit 1 was performed by Nuclear Engineering & Technology Institute(NETEC) of KHNP. To perform this inspection, NETEC developed equipment applied new technology such as the synchronization of 3D CAD, automatic alignment and control system. The inspection confirmed that the Calandria integrity of Wolsong NPP Unit 1 is perfect.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.429-436
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• 1995

This paper describes a methodology for determination of representative CANDU feeder geometry and the pressure drops between inlet/outlet header and fuel channel in the primary loop. A code, MEDOC, was developed based on this methodology and helps perform a calculation of equivalent feeder geometry for a selected channel group on the basis of feeder geometry data (fluid volume, mass flow rate, loss factor) and given property data pressure, quality, density) at inlet/outlet header. The equivalent feeder geometry calculated based on this methodology will be useful fur the transient thermohydraulic analysis of the primary heat transport system for the CANDU heavy water-cooled pressure tube reactor.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.152-157
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• 1997

To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled ,and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.286-300
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• 2002

An algorithm for the sodium boiling model has been developed for calculation of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. Modeling of sodium boiling in liquid metal reactors using sodium as a coolant is necessary because of phenomenon difference comparing with that observed generally in light water reactor systems. The applied model to the algorithm is the multiple-bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbies that (ill the whole cross section of the coolant channel except for the liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble The present study is focused on not only demonstration of the vapor bubble behavior predicted by the developed model, but also confirmation of a qualitative acceptance for the model. As a result, the model can represent important phenomena in the sodium boiling, but it is found that further effort is also needed for its completition.

• Journal of Radiation Protection and Research
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• 제49권1호
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• pp.29-39
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• 2024

Background: The gamma emission tomography (GET) device has been reported a reliable technique to inspect partial defects within spent nuclear fuel (SNF) of pin-by-pin level. However, the existing GET devices have low accuracy owing to the high attenuation and scatter probability for SNF inspection condition. The purpose of this study is to design and optimize a Yonsei single-photon emission computed tomography version 2 (YSECT.v.2) for fast inspection of SNF in water storage by acquisition of high-quality tomographic images. Materials and Methods: Using Geant4 (Geant4 Collaboration) and DETECT-2000 (Glenn F. Knoll et al.) Monte Carlo simulation, the geometrical structure of the proposed device was determined and its performance was evaluated for the ¹³⁷Cs source in water. In a Geant4-based assessment, proposed device was compared with the International Atomic Energy Agency (IAEA)-authenticated device for the quality of tomographic images obtained for 12 fuel sources in a 14 × 14 Westinghouse-type fuel assembly. Results and Discussion: According to the results, the length, slit width, and septal width of the collimator were determined to be 65, 2.1, and 1.5 mm, respectively, and the material and length of the trapezoidal-shaped scintillator were determined to be gadolinium aluminum gallium garnet and 45 mm, respectively. Based on the results of performance comparison between the YSECT.v.2 and IAEA's device, the proposed device showed 200 times higher performance in gamma-detection sensitivity and similar source discrimination probability. Conclusion: In this study, we optimally designed the GET device for improving the SNF inspection accuracy and evaluated its performance. Our results show that the YSECT.v.2 device could be employed for SNF inspection.

• Journal of Electrochemical Science and Technology
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• 제10권2호
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• pp.104-114
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• 2019

Energy is an essential driving force for modern society. In particular, electricity has become the standard source of power for almost every aspect of life. Electric power runs lights, televisions, cell phones, laptops, etc. However, it has become apparent that the current methods of producing this most valuable commodity combustion of fossil fuels are of limited supply and has become detrimental for the Earth's environment. It is also self-evident, given the fact that these resources are non-renewable, that these sources of energy will eventually run out. One of the most promising alternatives to the burning of fossil fuel in the production of electric power is the proton exchange membrane (PEM) fuel cell. The PEM fuel cell is environmentally friendly and achieves much higher efficiencies than a combustion engine. Water management is an important issue of PEM fuel cell operation. Water is the product of the electrochemical reactions inside fuel cell. If liquid water accumulation becomes excessive in a fuel cell, water columns will clog the gas flow channel. This condition is referred to as flooding. A number of researchers have examined the water removal methods in order to improve the performance. In this paper, a new water removal method that investigates the use of vibro-acoustic methods is presented. Piezo-actuators are devices to generate the flexural wave and are attached at end of a cathode bipolar plate. The "flexural wave" is used to impart energy to resting droplets and thus cause movement of the droplets in the direction of the traveling wave.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.687-700
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• 1995

제안된 CANDU-9 원자로의 열수력 과도변화상태가 해석되었으며 주요한 몇개의 과도변화가 열수송 계통의 설계요건을 만족시키는지에 대해 평가되었다. 열수송계통의 과도변화시 핵연료의 건전성과 계통압력상승의 제한 측면에서 분석된 본 해석결과에 따라서 제안된 열수송계통형상과 열수송계통기기의 예비 크기가 확정 및 검증되었다. AECB R-77 요구조건에 대한 CANDU-9 원자로의 만족여부를 평가하였다. 해석결과, 각 과도변화시 원자로 모관의 고압첨두치가 ASME코드의 요구조건에 따른 허용범주내에 있었으며 핵연료의 건전성이 확인되었다. 원자로 가동운전시 제안된 CANDU-9 원자로의 고유적인 핵연료채널을 통한 역류현상을 규명하기 위하여 한개의 펌프가 시동될때의 과도변화현상을 해석하였다.

• Nuclear Engineering and Technology
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• 제12권2호
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• pp.88-98
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• 1980

원통형 관속을 흐르는 뜨거운 단상 유체의 응고 과정을 해석적인 방법과 실험적인 방법으로 연구 하였다. 파라핀초와 Wood's Metal을 뜨거운 유체로 사용하여 일련의 실험을 하였다. 이 실험 데이터로 부룩해븐 연구소에서 개발한 응고과정에 대한 기존 준정적 수리해석 모델을 증명하였다. 또한 이 실험결과, 수직관속을 순간적으로 흐르며 응고하는 과정에 미치는 여러가지 매개변수의 영향에 관한 자료를 얻게 되었다. 이 실험에 사용한 기구와 실험 방법 도 아울러 기술하였다. 녹은 유체의 순간적으로 흘러내리는 양에 대한 수학적 모델의 예측 결과를 실험데이터와 비교하기 위해 도표로 제시하였다. 또한, 수학적 모델을 고속증식로(LMFBR)에 사고가 일어 났을 경우에 응용하여 보았다.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.