• Journal of the Korean institute of surface engineering
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• v.51 no.4
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• pp.218-223
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• 2018

In this study, we proposed a new ultrasonic-chemical decontamination process for decontaminating radioactive corrosion products during the maintenance of reactor coolant pump (RCP). The actual decontamination process was reproduced in the laboratory. And the corrosion characteristics of stainless steel (STS), constituting the RCP interior parts, were examined. The weight-loss measurment and polarization experiment were carried out in order to determine the corrosion characteristics of STS 304 and STS 415 by repeated decontamination processes. The STS 304 presented a little corrosion damage, which was almost indistinguishable from visual observation. The weight-loss rate of STS 304 was also significantly lower. On the other hand, STS 415 showed severe corrosion damage on its surface, greater weight-loss rate and higher corrosion current density than STS 304.

• Journal of Information Technology Services
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• v.20 no.3
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• pp.1-12
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• 2021

Reactor Coolant Pump (RCP) is core part of nuclear power plant to provide the forced circulation of reactor coolant for the removal of core heat. Properly monitoring vibration of RCP is a key activity of a successful predictive maintenance and can lead to a decrease in failure, optimization of machine performance, and a reduction of repair and maintenance costs. Here, we developed real-time RCP Vibration Analysis System (VAS) that web based platform using NoSQL DB (Mongo DB) to handle vibration data of RCP. In this paper, we explain how to implement digital signal process of vibration data from time domain to frequency domain using Fast Fourier transform and how to design NoSQL DB structure, how to implement web service using Java spring framework, JavaScript, High-Chart. We have implement various plot according to standard of the American Society of Mechanical Engineers (ASME) and it can show on web browser based on HTML 5. This data analysis platform shows a upgraded method to real-time analyze vibration data and easily uses without specialist. Furthermore to get better precision we have plan apply to additional machine learning technology.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.3
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• pp.255-262
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• 2011

In a loss-of-coolantaccident, considerable debris may be generated and transported to the recirculation sump. The accumulation of debris will reduce the netpositivesuctionhead and threaten the safety of thenuclear power plant. Both NEI 04-07 and USNRC SER suggesteda CFD methodology. However, additional investigation is needed to consider the unique characteristics of nuclear power plants. The transport of the generated debris is strongly influenced by the break location and the plant characteristics, including the configuration.In this paper, a CFD methodology for blow-down transport evaluation is proposed and applied to an OPR1000 nuclear power plant. The results show that the percentage of small debris transported to the upper containment is 32%, which is 7% larger than the valuegiven in the NEI 04-07 baseline analysis. This result may be used as a point of reference in future analytical studies.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.325-330
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• 2011

RCP(Reactor coolant pump) must be designed to preserve it's functions on normal or abnormal environments and seismic event same as operating basis earthquake(OBE) and safe shutdown earthquake(SSE). Generally, there are static and dynamic analytical method which can be applied by a floor response spectrum or time history analysis for the seismic qualification. Initially, It was accomplished a detailed structural FE-model for finite element analysis on the bases of 3-dimensional solid model which was made by the RCP drawing. As the result of dynamic characteristic using the detailed FE-model, it's shown about 12Hz natural frequency of 1st bending mode shape and maximum displacement has 11mm with the structural bending by single-point response spectrum(SPRS) method at all elevation. But maximum displacement has 7.6mm by multi-point response spectrum(MPRS) method which was applied to the three floor response spectrum at each elevation. Therefore, On a large heighten structures as RCP, The application by SPRS method causes to be more conservative results. Finally, A simpled equivalent beam model which was developed by use of iteration of detailed FE-model is shown the result more similar with those of natural frequencies and SPRS analysis. And maximum equivalent stress and displacement of the simpled beam has verified with 180MPa and 7.1mm each at 15sec as results by SSE time history method.

• Nuclear industry
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• v.27 no.4 s.290
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• pp.46-50
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• 2007

고리 1호기는 한국에서 최초로 규제 기관에 계속운전을 신청한 원전이다. 2007년 6월에 설계 수명 기간 만료가 되는 고리 1호기는 규제 기관으로부터 계속운전(Continued Operation)에 대한 안전성 심사를 받고 있다. 한수원은 고리 1호기 계속운전 승인을 금년 12월에 받기 위해 최선을 다하고 있으며 지역 주민의 사회적 수용성 확보를 위해 노력중이다. 고리 1호기의 계속운전 기간 동안 안전성을 평가하고 정리한 안전성평가보고서를 한수원은 2006년 6월에 정부에 제출하였다. 고리 1호기는 웨스팅하우스의 2루프 가압경수로이다. 이와 동일한 원전인 일본의 미하마 1,2호기와 겐까이1호기가 계속운전중이며, 미국의 기네이와 포인트 비치 1,2호기가 계속운전 승인을 받았다. 제출한 안전성평가보고서에 대해 한국원자력안전기술원이 심사중이며, 해외 원전과 같이 계속운전을 할 수 있을 것으로 예상하고 있다. 또한 계속운전을 위한 사회적 수용성(Public Acceptance) 확보는 설비의 철저한 안전성 확보 및 지역 주민의 공감대 형성을 통해서 이루어질 것이다. 설계 수명 이후 원자력발전소를 계속 운전하는 것은 이미 선진국에서 시행되고 있다. 2007년 3월 기준으로 미국에서 48기가 운영 허가 갱신 승인을 받았고, 영국은 8기, 일본은 12기가 계속운전중이다. 고리 1호기 성능 지표를 개선시키기 위해서 한수원은 증기발생기, 저압 터빈, 원자로 냉각재 펌프 내장품, 주변압기, 주발전기 등을 교체하였으며, 수명관리 연구, 주기적안전성 평가, 환경 영향 평가를 수행하였다. 2005년 9월에는 미국의 운영 허가 갱신 제도를 참조하여 원자력법이 개정되었다. 이에 한수원은 개정된 원자력법에 맞추어 주기적 안전성평가, 주요 기기에 대한 수명 평가 및 방사능 환경 영향평가를 하였다. 이 세가지 보고서들로 구성된 안전성평가보고서를 2006년 6월에 규제 기관에 제출하였다. 계속운전은 한국을 비롯하여 부존 자원이 부족한 국가들에게는 에너지 자원의 효율적 활용 및 온실 가스 배출을 고려할 때 반드시 필요한 것이다.

• Journal of Radiation Protection and Research
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• v.35 no.2
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• pp.57-62
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• 2010

Maintenance on the water chamber of steam generator, the change of pressurizer heater, the removal of pressure tube feeder, and so on during outage in nuclear power plants (NPPs) has a likelihood of high radiation exposure to whole body of workers even short time period due to the high radiation exposure rates. In particular, it is expected that hands would receive the highest radiation exposure because of its contact with radiation materials. In this study, field tests on extremity dose assessment of radiation workers for contact works with high radiation exposure were conducted during the maintenance periods in Korean pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs). In this field test, radiation workers were required to wear additional TLDs on the back and wrist, and an extremity dosimeter on fingers including a main TLD on the chest, while performing maintenance. As a result, it was found that the equivalent dose for fingers was distributed in the fixed range of deep dose and the equivalent dose for wrists.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.9
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• pp.599-605
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• 2015

The revised U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20, rev.3 requires the evaluation of the potential adverse effects from pressure fluctuations and vibrations on piping and components for the reactor coolant, steam, feedwater, and condensate systems. Detailed vibration analyses for the systems attached to the steam generator are very difficult, because these piping systems are very complicated. This paper suggests a screening method for the flow-induced vibration of acoustic resonances and pump-induced vibration of the piping systems attached to the steam generator in order to conduct the APR1400 comprehensive vibration assessment program. This paper seeks to address the areas such as potential vibration sources, and methods to prevent the occurrence of acoustic resonances and pump-induced vibration of piping systems attached to the steam generator, for conducting the APR1400 comprehensive vibration assessment program. The screening method in this paper will be used to estimate the flow-induced vibration of the piping systems attached to the steam generator for the APR1400.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.97-106
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• 1986

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAPS/MODl/NSC), based upon the sequence of events for the KNU1 (Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximates the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV settings etc. are recognized to be important in the transient analyses on a bestestimate basis.