• Journal of Radiation Industry
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• v.17 no.4
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• pp.519-532
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• 2023

Spent nuclear fuel(SNF) is stored in nuclear power plants for a certain period of time and then transported to an interim storage facility. After that, SNF is finally repackaged in a disposal canister at an encapsulation plant for final disposal. Finland and Sweden, leading countries in SNF disposal technology, have already completed designing of spent fuel encapsulation plant. In particular, the encapsulation plant construction in Finland is near completion. When it comes to South Korea, as the amount of SNF production and disposal plan is different from those in Finland and Sweden, it is difficult to apply the concepts of these contries as is. Therefore, it is necessary to establish the spent fuel repackaging concept and to derive each operating and repackaging procedures by considering annual disposal plan of South Korea. The results of this study is expected to be used to establish the concept of optimized encapsulation plant through further research.

• Proceedings of the IEEK Conference
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• 2003.07d
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• pp.1295-1298
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• 2003

The steam generator level in a PWR is very difficult to control particularly at low power. And the constant control gain and time value are not adaptive in steam generator level controller. In normal operation constant control gain and time value have no problem. But there is problem at low power. So variable control gains based on the temperature are required. The best control gain is decided by the experienced knowledge. A fuzzy gain tuner is used for the gain tuning. In the design of fuzzy gain-tuner processing, the experienced knowledge is employed for making fuzzy rules.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.680-686
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• 1995

It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors ill the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.107-114
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• 2013

It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, $HNO_3$, and Cs are very low.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1171-1180
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• 2009

The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The $Electricit\acute{e}$ De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code_Saturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2878-2887
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• 2021

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1-4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Corrosion Science and Technology
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• v.18 no.4
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• pp.138-147
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• 2019

Pipes and components of the secondary system in the pressurized water reactor (PWR) are mainly comprised of manufactured carbon steel. Thus, the generated carbon steel corrosion products are transported into the steam generator and deposited, thereby deteriorating the integrity of the steam generator. Environmental condition in the secondary system of the PWRs differs across different locations. So, the corrosion rate and types of corrosion products depend on specific locations in the secondary system. In this study, the quantity and chemical compositions of corrosion products generated in various locations that vary in different temperatures and chemistry conditions were investigated. As a result of evaluating the PWR "Unit A" that is in current operation, the amount of corrosion products generated in the section of high temperature feedwater system was identified as the largest source in the secondary system. Major components of corrosion products were iron oxides such as magnetite, hematite, and lepidocrocite.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.