• Proceedings of the KAIS Fall Conference
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• 2012.05a
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• pp.452-454
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• 2012

본 논문에서는 열분해 및 촉매반응의 의한 Hydrazine의 Ammonia 전환율을 연구하였다. 원자력발전소 2차 계통은 물/증기 순환계통으로 기기 및 배관의 부식을 억제하고, 증기발생기(Steam Generator, SG)의 부식생성물 유입을 최소하기 위해 전휘발성처리법(All Volatile Treatment, AVT)을 적용하여 계통수의 pH를 약염기성으로 유지하고 있다. 또한 Hydrazine을 이용하여 계통수의 용존산소제거 및 환원성 분위기를 유지하고 있다. 현재 사용되는 AVT는 대부분 단일 아민(Ammine)으로 계통 전 영역에서 pHt를 약염기성으로 유지하기 어렵다. 따라서 복합 아민을 이용하여 단일 아민의 상호단점을 보완한 수처리법을 적용해야한다. 하지만 복합 아민을 적용할 경우 추가 아민 주입설비, 설치부지, 시설유지보수 및 관리가 요구되므로 기존 주입약품을 이용하여 아민을 공급할 수 있는 연구가 필요하다. 따라서 본 연구에서는 Hydrazine의 열분해 및 촉매반응을 이용한 Ammonia 전환율을 조사하였다.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.431-438
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• 2014

The effect of various pH agents on the corrosion behavior of carbon steel was investigated under a simulated secondary water chemistry condition of a pressurized water reactor (PWR) in a laboratory, and the steel's corrosion performance was compared with the field data obtained from Uljin NPP unit 2 reactor. All tests were carried out at temperatures of $50^{\circ}C-250^{\circ}C$and pH of 8.5 - 10. The pH at a given temperature was controlled by adding different agents. Laboratory data indicate that the corrosion rate of carbon steel decreased as the pH increased under the test conditions and the highest corrosion rate was measured at $150^{\circ}C$. This high corrosion rate may be related to high dissolution and instability of Fe oxide ($Fe_3O_4$) at $150^{\circ}C$. It was also found that an addition of ethanolamine (ETA) to ammonia was more effectivefor anticorrosion than ammonia alone, and that mixed treatment reduced 50% of iron or more at pHs of 9.5 or higher, especially in the steam generator (SG) and the moisture separator & re-heater (MSR).

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.3
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• pp.174-180
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• 2006

Steam generator (SG) tubes are expanded inside tubesheet holes by using explosive or hydraulic methods to be fixed in a tubesheet. In the tube expansion process, it is important to minimize the crevice gap between expanded tube and tube sheet. In this paper, absolute and differential signals are computed by a numerical method for several different locations of tube expansion inside and outside a tubesheet and signal variations due to tubesheet, tube expansion and operating frequencies are observed. Results show that low frequency is good for detecting tubesheet location in both types of signals and high frequency is suitable for sizing of tube diameter as well as the detection of transition region. Also learned is that the absolute signal is good for measuring tube diameter, while the differential signal is good for locating the top of tubesheet and both ends of the transition region. In the case of mingled anomaly with tube expansion and tubesheet, low frequency inspection is found to be useful to analyze the mixed signal.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Corrosion Science and Technology
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• v.10 no.5
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• pp.175-179
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• 2011

The transportation and deposition of iron corrosion products are important elements that affect both the steam generator (SG) integrity and secondary system in pressurized water reactor (PWR) nuclear power plants. Most of iron corrosion products are generated on carbon steel materials due to flow accelerated corrosion (FAC). The several parameters like water chemistry, temperature, hydrodynamic, and steel composition affect FAC. It is well established that the at-temperature pH of the deaerated water system has a first order effect on the FAC rate of carbon steels through nuclear industry researches. In order to reduce transportation and deposition of iron corrosion products, increasing pH(t) tests were applied on secondary system of A, B units. Increasing pH(t) successfully reduced flow accelerated corrosion. The effect of increasing pH(t) to inhibit FAC was identified through the experiment and pH(t) evaluation in this paper.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2718-2731
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• 2024

To ensure the operational safety of nuclear power plants, we present a Quadruped Inspection Robot that can be used for many types of steam generators. Since the Inspection Robot relies on the Holding Modules to grip the tube-sheet, it can be regarded as a hybrid robot with variable configurations, switching between 4-RRR-RR, 3-RRR-RR, and two types of 2-RRR-RR, and the variable configurations bring a great challenge to dimensional synthesis. In this paper, the kinematic model of the Inspection Robot in multiple configurations is established, and the analytical solution is given. The workspace mapping is analyzed by the solution-space, and the workspace of multiple configurations is decomposed into the workspace of 2-RRR to reduce the analysis complexity, and the workspace calculation is simplified by using the envelope rings. The optimization problem of the manipulator is transformed into the calculation of the shortest contraction length of the swing leg. The switching performance of the Inspection Robot is evaluated by stride-length, turning-angle, and workspace overlap-ratio. The performance indexes are classified and transformed based on the proportions and variation trends of dimensional parameters to reduce the number of optimization objective functions, and Pareto optimal solutions are obtained using an intelligent optimization algorithm.

• Corrosion Science and Technology
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• v.11 no.3
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• pp.96-102
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• 2012

Outer diameter stress corrosion cracking (ODSCC) has occurred for Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) during long term operation. Among many causes for SCC, lead (Pb) is known to be one of the most deleterious species in the secondary system. In the present work, the oxide formed on Alloy 600 was characterized as a function of the PbO content in 0.1 M NaOH at $315^{\circ}C$ by using an electrochemical impedance spectroscopy (EIS), a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDS). The oxide property was analyzed in view of SCC susceptibility.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.2
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• pp.90-98
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• 2006

Signals captured from a Combo calibration standard tube paly a crucial role in the evaluation of motorized rotating pancake coil (MRPC) probe signals from steam generator (SG) tubes in nuclear power plants (NPPs). Therefore, the Combo tube signals should be consistent and accurate. However, MRPC probe signals are very easily affected by various factors around the tubes so that they can be distorted in their amplitudes and phase angles which are the values specifically used in the evaluation. To overcome this problem, in this study, we explored possibility of simulation to be used as a practical calibration tool far the evaluation of real field signals. For this purpose, we investigated the characteristics of a MRPC probe and a Combo tube. And then using commercial software (VIC-3D) we simulated a set of calibration signals and compared to the experimental signals. From this comparison, we verified the accuracy of the simulated signals. Finally, we evaluated two defects using the simulated Combo tube signals, and the results were compared with those obtained using the actual field calibration signals.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.513-522
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• 2013

The ODSCC detected in the TSP position of Ulchin 3&4 SGs are typical ODSCC of Alloy 600MA tubes. The causative chemical environment is formed by concentration of impurities inside the occluded region formed by the tube surface, egg crate strips, and sludge deposit there. Most cracks are detected at or near the line contacts between the tube surface and the egg crate strips. The region of dense crack population, as defined as between $4^{th}$ and $9^{th}$ TSPs, and near the center of hot leg hemisphere plane, coincided well with the region of preferential sludge deposition as defined by thermal hydraulics calculation using SGAP computer code. The cracks developed homogeneously in a wide range of SGs, so that the number of cracks detected each outage increased very rapidly since the first detection in the $8^{th}$ refueling outage. The root cause assessment focused on investigation of the difference in microstructure and manufacturing residual stress in order to reveal the cause of different susceptibilities to ODSCC among identical six units. The manufacturing residual stress as measured by XRD on OD surface and by split tube method indicated that the high residual stress of Alloy 600MA tube played a critical role in developing ODSCC. The level of residual stress showed substantial variations among the six units depending on details of straightening and OD grinding processes. Youngwang 3&4 tubes are less susceptible to ODSCC than U3 and U4 tubes because semi-continuous coarse chromium carbides are formed along the grain boundary of Y3&4 tubes, while there are finer less continuous chromium carbides in U3 and U4. The different carbide morphology is caused by the difference in cooling rate after mill anneal. There is a possibility that high chromium content in the Y3&4 tubes, still within the allowable range of Alloy 600, has made some contribution to the improved resistance to ODSCC. It is anticipated that ODSCC in Y5&6 SGs will be retarded more considerably than U3 SGs since the manufacturing residual stress in Y5&6 tubes is substantially lower than in U3 tubes, while the microstructure is similar with each other.