• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.792-799
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• 2023

Non-reactor nuclear facilities are increasing remarkably in Korea combined with advanced technologies such as life and space engineering, and the diversification of the nuclear industry. However, the absence of a basic classification guideline related to the design of non-reactor nuclear facilities has created confusion whenever related projects are carried out. In this paper, related domestic and international technical guidelines are reviewed to present the classification criteria of non-reactor nuclear facilities in Korea. Based on these criteria, the classification of structures, systems and components (SSCs) for safety controls is presented. Using the presented classification criteria, classification of a hot cell facility, a representative non-reactor nuclear facility, was performed. As a result of the classification, the hot cell facility is classified as the hazard category 3, accordingly, the safety class was classified as non-nuclear safety, the seismic category as non-seismic (RW-IIb), and the quality class as manufacturers' standards (S).

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.1004-1012
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• 2021

High density polyethylene (HDPE) pipe has many advantages over metallic pipe, and has been used in non-safety related application for years in some nuclear power plants (NPPs). Recently, HDPE pipe was introduced into safety related applications. The main difference between safety-related and non-safety-related pipes in NPPs is the design method of extra loadings such as gravity, temperature, and earthquake. In this paper, the mechanical behavior of HDPE pipe under various loads in pipe gallery was studied by finite element analysis (FEA). Stress concentrations were found at the fusion regions on inner surface of mitered elbows of HDPE pipe system. The effects of various factors were analyzed, and the influence of various loads on the damage of HDPE pipe system were evaluated. The results of this paper provide a reference for the design of nuclear safety-related Class 3 HDPE pipe. In addition, as the HDPE pipes analyzed in this paper were suspended in pipe gallery, it can also serve as a supplementary reference for current ASME standard on Class 3 HDPE pipe, which only covers the application for buried pipe application.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.63 no.12
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• pp.1637-1639
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• 2014

In the past, the analog protection relays have been widely used for the safety-related systems in the nuclear power plants due to their stability and reliability. Meanwhile, as the high performance digital system has been developed, the digital systems have been adopted in the non-safety systems. However, since the digital systems currently used in the non-safety systems were not developed according to Q-class standard, Commercial Grade Item Dedication (CGID) procedure should be performed in order to apply them to the safety-related system. The purpose of this paper is to describe the CGID procedure including the analysis of the hardware architecture as well as the software embedded in protective relay to apply to the emergency diesel generator in the nuclear power plant. The entire CGID procedure was performed strictly according to the international standard and regulations.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1451-1469
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• 2019

This paper proposes a residual stress mitigation of a nuclear safety-related austenitic stainless steel TP304 pipe bended by local induction heating process via performing elastic-plastic finite element analysis. Residual stress distributions of the pipe bend were calculated by performing finite element analysis. Validity of the finite element analysis procedure was verified via comparing with temperature histories measured by using thermocouples, ultrasonic thickness measurement results, and residual stress measurement results by a hole-drilling method. Parametric finite element stress analysis was performed to investigate effects of the process and geometric shape variables on the residual stresses on inner surfaces of the pipe by applying the verified procedure. As a result of the parametric analysis, it was found that it is difficult to considerably reduce the inner surface residual stresses by changing the existing process and geometric shape variables. So, in order to mitigate the residual stresses, effect of an additional process such as cooling after the bending on the residual stresses was investigated. Finally, it was identified that the additional heating after the bending can significantly reduce the residual stresses while other variables have insignificant effect.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.2
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• pp.397-406
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• 2010

Because diesel generator excitation systems in the nuclear power generating station are included among safety related c1asses(and Class 1E), they have been supposed to apply in the nuclear power generating stations through equipment qualification by nuclear law and so on. So, they has been controlled and assured completely by quality assurance throughout the total development journey. This paper looks into the journey of development of diesel generator excitation systems in the nuclear power generating station.

• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.3
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• pp.136-142
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• 2019

Nuclear power plant concrete structures are in contact with the coast, and durability due to chloride attack is very important because it is used as cooling water by taking seawater. For this purpose, a 3-year long-term saltwater immersion test was carried out to evaluate chloride ion diffusion coefficient and age apponent (m) The m values of the foundation with 4,000 class was 0.35 ~ 0.39, similar to KCI or ACI suggested values. essential service water constructions and tunnels of 5,000 class were 0.44 ~ 0.53 and 6,000 class, and 0.62 of reactor containment buildings were similar to the proposed values of FIB. As a result of the prediction of the service life with the measured age coefficient, all the safety related concrete structures of the nuclear power plants satisfied the service life of more than 60 years.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.4
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• pp.405-413
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• 2015

In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.

• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.187-198
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• 2014

I developed a curriculum reflecting the perspectives of students, science teachers, and professionals in order to carry out standardized, fundamental nuclear power and radiation education in schools. Among elementary, middle, and high schools, 78.4%, 78.6%, and 93.1% respectively exhibited (with high frequency) a need for nuclear power and radiation education. The proposed elementary and middle/high school course titles are "Radiation and Life" and "Nuclear Power and Radiation" respectively. The courses are offered at every grade level and span one semester each year. The duration of each weekly class varies; at the elementary, middle, and high school levels classes meet for 40, 45, and 50 minutes respectively. Thin textbooks containing an abundance of cartoons and photos were requested. The starting points for education were fixed at the sixth grade, second year of middle school, and the first year of high school. It was stipulated that the education be separate from the regular curriculum, and encompass a creative and experimental field study based on the principal and science teachers' needs. Similar trends were observable according to grade levels regarding class hours, textbook format, form of education, and educational necessity. A simulation of the devised curriculum revealed an overall goodness of fit totaling $3.88{\pm}0.60$, $3.89{\pm}0.60$, and $3.66{\pm}0.63$ out of five for elementary, middle school, and high school students respectively, which are scores equivalent to 70 and above (out of 100). The significance of this study is that it is the first to propose a curriculum designed to cultivate value judgment based on understanding nuclear power and radiation. However, the realization of nuclear power and radiation education requires that follow-up measures be taken regarding textbook development, amendments to related laws, and the providing of teaching plans.

• Proceedings of the KIEE Conference
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• 2005.10c
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• pp.13-16
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• 2005

원자력발전소의 안전과 관련된 기기는 원전의 정상상태 운전조건뿐만 아니라 원전의 설계기준사고 조건에서도 기기의 안전관련 기능을 충분하게 수행할 수 있음이 입증되어야만 한다. 아울러 기기의 설치 환경은 원전의 설계기준사고조건(DBE))으로서 지진만이 고려되는 온화한 환경(mild zone)과 냉각재상실사고(LOCA) 주증기관파단사고(MSLB) 등과 같이 고온, 고압 등의 환경요건이 급격히 변화하는 가혹한 환경(harsh zone)으로 구별되므로 안전관련 기기의 검증 또한 이러한 환경요건에 따라 수행되어져야 한다. 본 연구에서는 당사가 개발한 가혹환경용 안전관련 고압전동기의 개발사례를 중심으로 가혹환경요건에 대한 기기의 검증절차와 방법을 제시하였다.

• Computational Structural Engineering
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• v.7 no.3
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• pp.95-103
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• 1994

The safety related components in the nuclear power plant should be designed to withstand the seismic load. Among these components the integrity of reactor internals under earthquake load is important in stand points of safety and economics, because these are classified to Seismic Class I components. So far the modelling methods of reactor internals have been investigated by many authors. In this paper, the dynamic behaviour of reactor internals of Yong Gwang 1&2 nuclear power plants under SSE(Safe Shutdown Earthquake) load is analyzed by using of the simpled Global Beam Model. For this, as a first step, the characteristic analysis of reactor internal components are performed by using of the finite element code ANSYS. And the Global Beam Model for reactor internals which includes beam elements, nonlinear impact springs which have gaps in upper and lower positions, and hydrodynamical couplings which simulate the fluid-filled cylinders of reactor vessel and core barrel structures is established. And for the exciting external force the response spectrum which is applied to reactor support is converted to the time history input. With this excitation and the model the dynamic behaviour of reactor internals is obtained. As the results, the structural integrity of reactor internal components under seismic excitation is verified and the input for the detailed duel assembly series model could be obtained. And the simplicity and effectiveness of Global Beam Model and the economics of the explicit Runge-Kutta-Gills algorithm in impact problem of high frequency interface components are confirmed.