• Clinical and Experimental Reproductive Medicine
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• 제36권1호
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• pp.35-43
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• 2009

목 적: 정액 내 tumor necrosis factor-alpha (TNF-${\alpha}$) 농도와 정자 DNA 손상 및 정액 검사 소견과의 관련성을 평가하고자 하였다. 연구방법: 정액 표본은 45명의 건강한 남성에서 자위에 의하여 획득하였다. 정자의 상태는 컴퓨터 정액 분석기를 이용하여 판정하였으며, 두부의 DNA 손상은 TUNEL 분석방법에 의해 측정하였다. TNF-${\alpha}$ 농도는 동결-융해된 정장액에서 ELISA법으로 측정하였다. 결 과: 정자 DNA 손상율은 1.9%에서 53% (mean ${\pm}$ SD, 12.4${\pm}$9.6%)로 매우 광범위하게 나타났다. 단변량분석에 의하면 DNA 손상 정도와 정자의 농도, 운동성과는 관련이 없었으나, 직진운동성 (linearity)과는 음의 상관 관계를 나타내었으며 (r=-0.325, p=0.03) 연구 대상 남성의 연령과는 양의 상관 관계를 나타내었다 (r=0.484, p=0.001). 정액내에 존재하는 TNF-${\alpha}$ (>1 pg/mL)는 연구 대상 남성의 73.3% (33/45)에서 검출되었으며 평균 농도는 4.9 pg/mL, 범위는 1.1에서 22.6 pg/mL이었다. 정액 검사 상의 정자 상태와 정자 DNA 손상과는 유의한 관련성이 나타나지 않았다. 결 론: 본 연구에서는 정자 DNA의 손상이 남성의 연령과 관련성이 있음을 확인하였으나, TNF-${\alpha}$와의 관련성은 확인할 수 없었다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.617-621
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• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전기설비전문위원
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• pp.73-75
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• 2005

A number of the power cables arc installed in the containment of the nuclear power plant. According to the IEEE Standard 835, the calculation of the temperature rise shows the operation possibility of power cables in the containment. In this paper, we expect the integrity of the power cables by using the calculation of the temperature rise and the development of the lifetime extension of the cables.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1495-1506
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• 2023

For the safe disposal of high-level radioactive waste using Engineered Barrier Systems (EBS), bentonite buffer is used by its high swelling capability and low hydraulic conductivity. When the bentonite buffer is contacted to heated pore water containing ions by radioactive decay, chemical alterations of minerals such as illitization reaction occur. Illitization of bentonite indicates the alteration of expandable smectite into non-expandable illite, which threatens the stability and integrity of EBS. This study intends to provide a thorough review on the information underlying in the illitization of bentonite, by covering basic clay mineralogy, smectite expansion, mechanisms and observation of illitization, and illitization in EBS. Since understanding of smectite illitization is crucial for securing the safety and integrity of nuclear waste disposal systems using bentonite buffer, this thorough review study is expected to provide essential and concise information for the preventive EBS design.

• 비파괴검사학회지
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• 제34권5호
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• pp.356-361
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• 2014

회전형 탐촉자(RPC)는 증기발생기 전열관의 결함 탐지 및 크기 측정 목적으로 널리 사용되고 있다. 손상이 탐지된 전열관에 대한 건전성 평가는 비파괴검사에서 얻어진 열화의 크기 정보를 바탕으로 수행되기 때문에 검사기술의 성능은 전열관의 건전성 평가에 직접적으로 영향을 미치게 된다. 동일 전열관의 인접한 거리에 다중균열이 존재할 경우 검사 기술의 결함 분해능에 제약이 따를 수 있으며 그 영향이 클 경우 근접한 다중균열이 상대적으로 큰 단일균열로 평가될 수 있으므로 전열관의 구조건전성 평가에 오류를 유발할 수 있게 된다. 따라서 본 연구에서는 방전가공으로 균열을 모사한 인공결함에 대한 RPC 탐촉자의 결함 분해능을 관찰하고 전열관의 구조건전성 평가에 미치는 영향을 살펴보았다. 동일 직선상에 놓인 다중균열은 매우 근접한 거리까지 개별균열 식별이 가능하여 건전성 평가에 미치는 영향이 없는 반면, 인접한 거리에 평행하게 놓인 균열의 경우는 RPC 탐촉자의 분해능이 낮아서 부정확한 결함 크기 정보가 얻어지므로 결함관의 파열압력 예측에 영향을 미칠 수 있다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2092-2099
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• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.

• 대한기계학회논문집A
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• 제29권5호
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• pp.734-742
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• 2005

The present study proposes the integrity evaluation model for a straight pipe with local wall thinning defect, which reflects the characteristics of training shape and loading condition in the Piping of nuclear power plant. For this purpose, a series of finite element analyses are performed under various defect geometries and loading conditions, and real pipe experiment data performed previously is employed. The model includes the effect of thinning length as well as thinning depth and width, and also it considers the combined loading effect between internal pressure and bending moment. The proposed model has been validated using the results of finite element analysis and pipe experiment data. The results indicate that the proposed model provides more reliable predictions of pipe failure than the current existing model, in terms of accuracy, consistency, and conservativeness of results.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.175-180
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• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.99-107
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• 2000

A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R $T_{NDT}$ , $K_{IC}$ , and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.165-171
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• 2017

To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.