• Corrosion Science and Technology
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• v.22 no.3
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• pp.214-219
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• 2023

Considering the case in the United States where most nuclear power plants with an initial design life of 40 years continue to operate until 60 or 80 years after undergoing material soundness evaluation, it is time to plan a more robust long-term operation strategy for nuclear power plants in Korea. There are some reports that SRO/LRO might be formed when Alloy 690 is heat treated for 10,000 hours to 100,000 hours at 360 to 450 ℃. The possibility of LRO formation in Alloy 690 steam generator tubings of Kori nuclear power plant unit 1 (Kori-1) was investigated using existing research papers. The mechanism in which SRO/LRO occurred was also surveyed. Alloy 690 was found to be more likely to cause ordering than Alloy 600 in terms of alloy composition. The ordering could be evaluated through changes in material properties. However, it is difficult to evaluate it from a microstructural point of view. The likelihood of LRO in Alloy 690 of the Kori-1 plant operated at 320 ℃ for 19 years seemed to be low in terms of time and exposure temperature.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.137-149
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• 1996

The present study is to analyze an integral test, BETHSY test 6.9c, which represent loss of RURS accident under mid-loop operation. Both the pressurizer manway and the steam generator outlet plenum manway are opened as vent paths in order to prevent the system from pressurization by removing the steam generated in the core. The main purposes are to gain insights into the physical phenomena and identify sensitive parameters. Assessment of capability of CATHARE2 prediction can be established the effective recovery procedures using the code in an actual plant. Most of important physical phenomena in the experiment could be predicted by the CATHARE2 code. The peak pressure in the upper plenum is predicted higher than experimental value by 7 kPa since the differential pressure between the pressurizer and the surge line is overestimated. The timing of core uncovery is delayed by 500 seconds mainly due to discrepancy in the core void distribution. It is demonstrated that openings of the pressurizer manwey and the steam generator manway can prevent the core uncovery using only gravity feed injection. Although some disagreements are found in the detailed phenomena, the code prediction is considered reasonable for the overall system behaviors.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.4
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• pp.339-345
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• 2008

The heat exchanger tube in nuclear power plants is mainly fabricated from nonferromagnetic material such as a copper, titanium, and inconel alloy, but the moisture separator & reheater tube in the turbine system is fabricated from ferromagnetic material such as a carbon steel or ferrite stainless steel which has a good mechanical properties in harsh environments of high pressure and temperature. Especially, the moisture separator & reheater tubes, which use steam as a heat transfer media, typically employ a tubing with integral fins to furnish higher heat transfer rates. The ferromagnetic tube typically shows superior properties in high pressure and temperature environments than a nonferromagnetic material, but can make a trouble during the normal operation of power plants because the ferrous tube has service-induced damage forms including a steam cutting, erosion, mechanical wear, stress corrosion cracking, etc. Therefore, nondestructive examination is periodically performed to evaluate the tube integrity. Now, the remote field testing(RFT) technique is one of the solution for examination of ferromagnetic tube because the conventional eddy current technique typically can not be applied to ferromagnetic tube such as a ferrite stainless steel due to the high electrical permeability of ferrous tube. In this study, we have designed RFT probes, calibration standards, artificial flaw specimen, and probe pusher-puller necessary for field application, and have successfully carry out RFT examination of the moisture separator & reheater tube of nuclear power plants.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.17 no.11
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• pp.981-989
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• 2005

In this study, external condensation heat transfer coefficients (HTCs) are measured on a low fin tube and Turbo-C tubes at the saturated vapor temperature of $30^{\circ}C$, $39^{\circ}C$, and $50^{\circ}C$ for R22, R410A, R407C and R134a with the wall subcooled at $3{\~}8^{\circ}C$. The HTCs of all refrigerants decreased as increasing the saturation temperature from $30^{\circ}C$ to $50^{\circ}C$. This trend is due to better thermodynamic properties of the liquid phase at low temperature Beatty and Katz's prediction yielded a $20.0\%$ deviation for the low fin tube data. The heat transfer enhancement factors for the 26 fpi low fin tube and Turbo-C tubes are 4.0${\~}$5.5 and 3.0${\~}$8.1 respectively for the refrigerants tested. Finally the performance of Turbo-C tube is better than that of the low fin tube.

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.11 no.2
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• pp.420-425
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• 2010

In the process of warm forging, billet is heated up to $800^{\circ}C$ and located in the upper part of die block impression. The scattered oxidized scale may cause workers burn and shortening of die life sticking to the die block impression. The separating materials sprayed in die block cause harmful dust, harmful mist, fume, and bad odor which contaminate workshop environment. The process is classified as one of the avoided jobs and make the planned output achievement difficult. Development of an elimination device to clear out the contaminating materials in the workshop and improvement of the unsatisfactory maintenance method to fix the abrasion of die block impression which delays the dead line, cost increases needs to be developed. In this research, I tried to solve the problems caused in warm forging of bronze pipe joint such as the billet heating process, die maintenance, and manufacturing cost through improvement of warming forging manufacturing method and die maintenance method and eliminating harmful gas which will make the workshop more environment friendly.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.1-6
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• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).

• Journal of the Korean Society of Safety
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• v.35 no.4
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1990.10a
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• pp.173-178
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• 1990

로터-베어링축계는 증기및 가스터빈, 터보 발전기, 압축기등 거의 모든 산업 기계류에서 동력 전달의 기본 도구로써 사용되고 있다. 즉 회전에 의한 동력 의전달은 비교적 간단히 대용량의 동력을 효율적으로 전달할 수 있다. 이에 따라 회전기계류에 대한 연구는 산업 혁명 이후 꾸준히 발전되어 온바, 특히 근래에 들어와 산업기계류의 경쟁이 치열하여짐에 따라 산업기계류의 고정 밀화, 고속화, 고신뢰화 요구가 증대하고 있는 현실을 비추어 볼때, 산업 기 계류의 근간을 이루고 있는 로터-베어링 축계의 안정성을 포함한 진동에 관 한 문제는 회전기계류 설계의 주요 기술로써 연구.개발의 필요성이 매우 높 다 하겠다. 회전축계 진동 관련 연구는 두 분야로 대별될 수 있는데 언밸런 스(Unbalance)에 의한 Synchronous진동과 여러가지 원인에 의해 계의 불안 정성을 유발시키는 Nonsynchronous진동으로 나눌 수 있다. 본 연구에서는 이들 연구의 기본이 되는 회전축-베어링계 동특성 해석 프로그램을 개발하 였다. 여러가지 방법이 있으나 여기서는 Holzer가 비틀림 진동에 적용하고, Mykiestad(2)와 Prohl(3)에 의하여 회전축의 횡 진동에 적용된 이후 Lund(4) 등에 의하여 베어링의 영향등이 첨가된 전달 매트릭스 (Transfer Matrix) 방 법을 이용하여 임계속도(Critical Speed), 모우드 형태(Mode shapes)를 예측 하고 불안정 판정(Instability Criteria)등을 할 수 있는 프로그램을 개발하였 다. 특히 Murphy(1)의 다항식 방법(Polynomial Method)에 기본을 두어 기존 의 전달 매트릭스가 가지고 있던 반복, 수렴 시간 문제와 빠뜨리는 임계속도 예측에 대한 개선을 이루었으며 기존 논문과 실험 결과와의 비교 검토를 통 하여 개발된 프로그램의 신뢰성을 검토하였다. 특히, 각종 회전 기계의 소형 화, 경량화 추세에 따라 지반이나 케이싱이 경량이거나 유연하여 회전축과 동적으로 연성된 경우 회전축-베어링-지반으로 이루어진 2중구조의 회전축 계 동특성을 해석할 수 있는 프로그램을 개발하므로서 회전 기계류의 진동 전반에 걸친 문제점에 대한 그 원인과 현상을 명확히 분석하여 국내의 전기 계류의 보다 신뢰성있는 설계 및 제작자료를 확보하는데 기여할 수 있게 하 였다.