• Proceedings of the KSME Conference
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• 2007.05a
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• pp.291-295
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• 2007

FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL consists of In-Pile Test Section (IPS) and Out-Pile System (OPS). FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. Task Force Team was organized to do a construction systematically and the communication between members of the task force team was done through the CoP(community of Practice) notice board provided by the Institute. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. Without a sweet of the workers of the participating company of HEC(Hyundae Engineering Co, Ltd), HDEC(HyunDai Engineering & Construction Co. Ltd), equipment manufacturer, and the task force team, it is not possible to install the FTL facility within the planned shutdown period. The Commissioning of the FTL is on due to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

• Nuclear Engineering and Technology
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• v.38 no.3
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• pp.293-300
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• 2006

A liquid metal reactor (LMR) operated at high temperatures is subjected to both cyclic mechanical loading and thermal loading; thus, creep-fatigue is a major concern to be addressed with regard to maintaining structural integrity. The Korea Advanced Liquid Metal Reactor (KALIMER), which has a normal operating temperature of $545^{\circ}C$ and a total service life time of 60 years, is composed of various cylindrical structures, such as the reactor vessel and the reactor baffle. This study focuses on the creepfatigue crack initiation for a cylindrical Y-junction structure made of 316 stainless steel (SS), which is subjected to cyclic axial tensile loading and thermal loading at a high-temperature hold time of $545^{\circ}C$. The evaluation of the considered creep-fatigue crack initiation was carried out utilizing the ${\sigma}_d$ approach of the RCC-MR A16 guide, which is the high-temperature defect assessment procedure. This procedure is based on the total accumulated strain during the service time. To confirm the evaluated result, a high-temperature creep-fatigue structural test was performed. The test model had a circumferential through wall defect at the center of the model. The defect front of the test model was investigated after the $100^{th}$ cycle of the testing by utilizing a metallurgical inspection technique with an optical microscope, after which the test result was compared with the evaluation result. This study shows how creep-fatigue crack initiation for a high-temperature structure can be predicted with conservatism per the RCC-MR A16 guide.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1222-1230
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• 2020

The corrosion rates of the reactor pressure vessel materials of SA508 Grade 3 were measured using a weight loss method in aerated boric acid solutions to simulate the evaporation of leaked PWR primary water in an ambient environment. The corrosion behavior and products were examined using X-ray diffraction and electron microscopy. SA508 showed typical general corrosion characteristics. The corrosion rate increased steadily as the boron concentration was increased. As the immersion time elapsed, the corrosion rate slowly or rapidly decreased according to the oxidation reaction of iron. The corrosion rate showed a complicated pattern depending on the temperature; it increased gradually and then rapidly decreased again when reaching a certain transition temperature. The corrosion products of SA508 were found to be FeO(OH), Fe₂O₃, and Fe₃O₄. As the boron concentration decreased and the temperature was increased, the formation of Fe₃O₄ was more favorable as compared to the formation of FeO(OH) and Fe₂O₃. Consequently, the changes of the corrosion rate and behavior were closely related to the oxidation reaction of iron on the surface. The corrosive damage to SA508 appears to be most severe when the oxidation reaction is such that Fe₂O₃ forms as a corrosion product.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1626-1637
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• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Asian Pacific Journal of Cancer Prevention
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• v.17 no.4
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• pp.2151-2157
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• 2016

Scabraside D, a sulfated triterpene glycoside, was extracted from the sea cucumber Holothuria scabra. It shows anti-proliferation in many of cancer cell lines, but the function and mechanisms of action of scabraside D in human cholangiocarcinoma (HuCCA) have not previously determined. In this study, we investigated the activity of scabraside D on HuCCA cell apoptosis, lymphangiogenesis and metastasis in a nude mouse model. Scabraside D induced signs of apoptosis, such as cell shrinkage, nuclear condensation, nuclear fragmentation and DNA fragmentation on TUNEL assays, while effectively decreasing expression of BCl-2 but increasing caspase-3 gene level expression. Immunohistochemistry revealed that scabraside D significantly reduced lymphatic vessel density (LVD). Moreover, scabraside D treatment significantly decreased VEGF-C, MMP-9 and uPA gene expression, which play important roles in the lymphangiogenesis and invasion of cancer cells in metastasis processes. Quantitative real-time PCR showed that scabraside D significantly decreased iNOS and STAT-3 gene expression. This study demonstrated that scabraside D plays a role in activation of HuCCA tumor apoptosis and inhibition of lymphangiogenesis, invasion and metastasis through decreasing BCl-2, MMP-9, uPA and VEGF-C and increasing caspase-3 expression by suppression of iNOS and STAT-3 expression. Therefore, scabraside D could be a promising candidate for cholangiocarcinoma treatment.

• The Korean Journal of Nuclear Medicine
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• v.6 no.2
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• pp.21-39
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• 1972

The employment of gamma-ray scintillation camera with the use of short-lived radioisotopic pharmaceuticals, sucu as $^{99m}Tc$ sodium pertechnetate, have enabled us to perform RI (radioisotopic) angiocardiography. Although conventional cardiac catheterization or angiocardiography using contrast media have been an important diagnostic tool, they may carry some risks or serious complications. The author investigated on RI angiocardiography in twelve normal and twenty five patients with cardiovascular diseases in an effort to evaluate its diagnostic value. The results obtained with this study are as follows; 1. In normal subjects, the scintillation camera transit time of arm-to-right heart was found to be $2.1{\pm}0.67$ seconds, right heart-to-lung $1.5{\pm}0.40$ seconds and lung-to-left heart $3.5{\pm}0.86$ seconds. 2. Transformation or displacement of the heart and the great vessel are easily discriminated by RI angiocardiography. Both in the cases with tricuspid atresia and tetralogy of Fallot, ventricular septal defects are well recongnized by sequential RI angiocardiography. 3. It is also helpful in determinining the site and extent of obstruction, and estimatiing the postoperative course both in the cases with superior vena cava syndrome and pulmonary stenosis. 4. Pericarditis with effusion is readilly diagnosed by RI angiocardiogram showing characteristic "dead space" between intracardiac and intrapulmonary radioactivity. 5. It was found that the diagnostic accuracy of this study was 78.4%. It is concluded that above results obtained are useful and accurate enough for the diagnostic screening methode for clinical practice.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.403-415
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• 1998

The CANDU 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside the calandria tube and the annulus between these tubes, which forms a closed loop with $CO_2$ gas recirculating, is called the Annulus Gas System(AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tube rupture incident. To judge whether the operator action time is enough or not in the design of Wolsong 2,3 & 4, the Leak Before Break(LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsong Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for the dew point rate of rise of Wolsong Unit 2. It was found that the response of the dew point depends on the moisture injection rate, $CO_2$ gas flow rate and the leak location. The test showed that CANDU 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS $CO_2$ flow rate is approximated.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.281-291
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• 1995

The objective of this work is to obtain the experimental data for the heat transfer processes occurring both on the inside and outside surfaces of containment steel wall with dry and wet outer surface conditions in the passive containment cooling system. The test model represented a 60$^{\circ}$ section of a containment vessel based on the AP 600 geometry. Major linear dimensions of the test model ore reduced tv a factor of ten. To simulate the decay heat a steam generator heated by electricity was placed in the test model. The maximum heat flux was 8.91 kW/$m^2$. Two types of tests were performed. The one was the tort on the natural convection of air without water film flow. The other was the evaporative heat transfer test with the falling water film flow and natural air draft. no test result shooed that the heat transfer capability by the natural convection from the containment to the air without oater film flow was limited at about 1.48 kW/$m^2$ heat flux. It was found that the heat removal capability was remarkably enhanced in the tests with the waster film flow and air draft. The obtained heat transfer data ore compared with the existing correlations.