• Proceedings of the Korean Institute of Building Construction Conference
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• 2017.11a
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• pp.183-184
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• 2017

Some of the information created during the design phase of an New NPP life cycle is useful only for the execution of the construction phase; however, much of the information greatly impacts the longer-term operational phase. To most make use of design and construction information produced by data based design system during the construction and operation phase, This research is identified controlled data and drawn design information of controlled equipment from documents generated during the life-cycle stages. This study aimed to analyze related documents to assure traceability of controlled equipment from design phase through O&M and then suggested DB(Data Base) based control method on technical information of major equipment throughout nuclear power plant lifecycle.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.69-76
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• 2018

The paper presents preliminary investigation results for the effect of the baseline correction in the acceleration excitation method on finite element seismic analysis results (such as accumulated equivalent plastic strain, equivalent plastic strain considering cyclic plasticity, von Mises effective stress, etc) of nuclear safety Class I components. For investigation, finite element elastic-plastic time-history seismic analysis is performed for a surge line including a pressurizer lower head, a pressurizer surge nozzle, a surge piping, and a hot leg surge nozzle using the Chaboche hardening model. Analysis is performed for various seismic loading methods such as acceleration excitation methods with and without the baseline correction, and a displacement excitation method. Comparing finite element analysis results, the effect of the baseline correction is investigated. As a result of the investigation, it is identified that finite element analysis results using the three methods do not show significant difference.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.6
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• pp.683-691
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• 2011

In order to develop technology for an equipment qualification (EQ) test, which is an important process in localizing solenoid-operated valves used in nuclear power plants, we analyzed related regulations, test procedures, conditions, equipment, and acceptance criteria. EQ regulations for the solenoid-operated valve are classified as law, guide, and standard, and are subdivided according to test specimens and contents. The EQ test is composed of functional, normal-, and accident- condition tests. The solenoid-operated valve is aged under normal and accident conditions, which are predicted in the design conditions of a nuclear power plant, and the performance of the valve is measured by a functional test. The test method and procedure analyzed in this paper might be very useful for manufacturers as well as EQ testers.

• Proceedings of the Materials Research Society of Korea Conference
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• 2011.10a
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• pp.39.1-39.1
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• 2011

가압경수로의 압력경계기기는 약 $300^{\circ}C$, 150기압의 고온고압수환경에서 가동되고 있다. 특히 가압기 밀림관은 고온수와, 저온수가 교차하는 부분으로 열성층 형성으로 열적, 기계적 피로 및 수화학환경이 더해진 부식피로 등에 의하여 손상을 받는다. PWR 원전에서 수화학환경은 대표적으로 용존산소(DO) 5ppb, pH 6~8, 용존수소(DH) <30 cc/kg, 온도 $316^{\circ}C$의 환경을 유지하게 된다. 가압기 밀림관에는 오스테나이트계 스테인리스강이 사용되는데, 오스테나이트계 스테인리스강은 고온 수화학환경에 민감한 것으로 알려져 있다. 따라서 오스테나이트계 스테인리강을 공기중에서의 기계적특성 및 피로특성을 향상시키기 위하여 질소를 첨가한 스테인리스강을 제조하여 PWR 원전환경에서의 피로균열성장특성을 평가하였다. 실험에 사용된 재료는 PWR 원전 가압기 밀림관 소재인 Type 347 스테인리스강에 0.0005 wt%가 첨가된 상용재와 0.11 wt% 질소가 첨가된 재료이다. 사용된 시편형상은 두께 5 mm, 폭 25.4 mm의 CT 시편이다. 수화학환경은 150기압, 온도 $316^{\circ}C$, 용존산소(DO) 5ppb, 용존수소(DH) 30 cc/Kg, pH는 약 7로 유지 하였으며, 응력비 0.1, 하중 반복속도 10Hz의 기계적 조건에서 하중제어로 시험을 진행하였다. 균열길이는 직류전위차법(Direct Current Potential Drop: DCPD)을 이용하여 측정하였다. 질소함량이 증가할수록 동일 사이클에서 균열길이가 늦게 성장하였고, 피로균열성장속도도 약간 늦어지는 것으로 나타났다. 각 스테인리스강의 피로파면 관찰결과 상용재는 약 1 ${\mu}m$의 산화물들이 생성되는 반면 질소첨가 스테인리스강은 약 0.1 ${\mu}m$정도 산화물이 생성되었다. 산화막의 두께도 질소가 첨가됨으로써 상용재에 비해 얇게 생성되었다. 따라서 질소가 첨가됨으로써 부식환경에서 내산화성이 향상되었으며, 이는 피로균열성장특성에 영향을 미치는 것으로 판단된다.

• KIPS Transactions on Computer and Communication Systems
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• v.2 no.12
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• pp.561-568
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• 2013

Instrumentation and control system in nuclear power plant performs protecting, controling and monitoring safety operation of Nuclear Power Plant. As cyber attack to the control equipment of instrumentation and control system can cause reactor shutdown and radiation release, it is required to design the instrumentation and control system considering cyber security in accordance with regulatory guides and industrial standards. In this paper, we proposed a design method of uni-directional communication structure which is required in the design of defense-in-depth model according to regulatory guides and industrial standards and we implemented a communication board with the proposed method. This communication board was tested in various test environments and test items and we concluded it can provide uni-directional communication structure required to design of defense-in-depth model against cyber attack by analyzing the results. The proposed method and implemented communication board were applied in the design of SMART (system-integrated modular advanced reactor) I&C (instrumentation and control) systems.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.35 no.2
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• pp.83-91
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• 2022

Seismic design and analysis of nuclear power plant components are performed based on an decoupled model. However, this decoupled analysis has a limitation in that it generates inaccurate results compared to the coupled analysis because it cannot simulate actual phenomena such as the interaction between structures and components. Thus, this study performed seismic coupled and decoupled analysis on an existing nuclear containment structure and related components, considering the mass and natural frequency ratios. And based on these results, comparative analyses of responses of components were conducted. Consequently, the seismic coupled analysis result generally gave a smaller value than the decoupled analysis result. These results were similar to the analysis results for the simple coupled model, which was an existing study, but the difference in component responses was much more pronounced. Also, this was influenced by the installation location of the component rather than the influence of the input frequency of the input seismic motions. Finally, the difference between the decoupled and coupled seismic analysis occurred in the region where the mass ratio of the components was large, and the natural frequencies were almost similar due to the considerable dynamic interaction between the structure and the component in this realm.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3497-3500
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• 2007

The internal structure is subjected to dynamic analysis due to the structural integrity. The internal structure shall be installed in the vertical hole call IR1 of reactor core. In order to verify the deflection of the internal structure, the mode and response spectrum analysis of the internal structure was performed. The natural frequency of the internal structure is 11.6 Hz(mode 1 and 2) and deflections of the internal structure are less than values of allowable design (3.2 mm).

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2000.10a
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• pp.215-223
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• 2000

In this study, the effect of stiffness ratio between base frame and anchorgae is evaluated and the seismic verification of nuclear power plant equipment anchorage is performed for typical equipment. The stiffness ratio between base frame and anchorage is mainly controlled by the effective height of side wall plate. And, the change of that stiffness ratio cause the large shift or ovreturning axis of equipment base. This shift of overturning axis of equipment base is able to reduce the factor of safety about 10%. Therefore, the adequate method for evaluating of effective height of side wall is required as further study.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2002.11a
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• pp.69-74
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• 2002

원자력발전소에서 발생하는 화재사건은 원자로 정지를 유발함과 통시에 안전정지 또는 사고완화 기능을 수행하는 다수의 기기를 동시에 손상시킬 수 있어 원자력발전소의 안전성에 적지 않은 영향을 줄 수 있다. 미국에서는 1975년 Browns Ferry 원전 1호기 케이블 포설실(cable spreading room)의 케이블 관봉투 밀봉재에 대한 건전성을 검사하는 과정 중에 화재가 발생하여 원자로 건물로 화재가 확산되는 심각한 사고가 발생하였다.(중략)

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2008.06a
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• pp.173-174
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• 2008