• Nuclear Engineering and Technology
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• 제45권2호
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• pp.249-256
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• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 추계학술대회 논문집
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• pp.808-813
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• 2005

Structural integrity assessment of defected pipe is important in fitness for service evaluation and proper engineering assessment is needed to determine whether pipelines are still fit for service. This paper present a failure prediction of gas pipes made of APIl X65 steel with gouge using stress-modified true fracture strain, which is regarded as a criterion of ductile fracture. For this purpose, API X65 pipes with gouge are simulated using elastic-plastic FE analyses with the proposed ductile failure criterion and the resulting burst pressures are compared with experimental data. Agreements are quite good, which gives confidence in the use of the proposed criteria to defect assessment fer gas pipelines. Then, further extensive finite element analyses are performed to obtain the burst pressure solution of pipes with gouge as a function of defect depth, length and pipeline geometry.

• 방사성폐기물학회지
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• 제21권4호
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• pp.465-479
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• 2023

To ensure the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. This paper addresses the development of the safety assessment model for the underground silo of Wolseong Low-and Immediate-Level Waste (LILW) disposal facility in Korea. As the simulated result, the nuclides diffused from the waste were kept inside the silo without the leakage of those while the integrity of the concrete is maintained. After the degradation of concrete, radionuclides migrate in the same direction as the groundwater flow by mainly advection mechanism. The release of radionuclides has a positive linear relationship with a half-life in the range of medium half-life. Additionally, the solidified waste form delays and reduces the migration of radionuclides through the interaction between the nuclides and the solidified medium. Herein, the phenomenon of this delay was implemented with the mass transfer coefficient of the flux node at numerical modeling. The solidification effects, which are delaying and reducing the leakage of nuclides, were maintained the integrity of the nuclides. This effect was decreased by increasing the half-life and the mass transfer coefficient of radionuclides.

• 한국재난정보학회 논문집
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• 제12권1호
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• pp.1-9
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• 2016

국부적인 두께감소 문제에 대하여 직관부위를 대상으로 한 건전성 평가는 잘 알려져 있으나 엘보우를 대상으로 한 건전성 평가는 최근에 원자력 분야에서는 많은 연구가 이루어지고 있으나 석유화학 플랜트에서 이용되는 건전성 평가 지침서 중 하나인 API579 코드에는 아직 없는 실정이다. 이에 본 연구에서는 엘보우를 대상으로 엘보우의 외부(extrados)와 내부(intrados)에 두께 감소가 있다고 가정한 후 유한요소해석법을 이용하여 두께감소된 엘부우의 건전성평가 해석을 수행하였다. 본 해석 결과는 석유화학 플랜트에서 이용되는 엘보우의 건전성 평가에 이용될 수 있음을 보여주었다.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• 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.

• Corrosion Science and Technology
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• 제14권1호
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• pp.1-11
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.

• 한국소프트웨어감정평가학회 논문지
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• 제10권1호
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• pp.19-26
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• 2014

본 논문에서는 안드로이드 애플리케이션(앱)을 역공학 공격으로부터 방어하는 기법을 제안한다. 이 기법에서 서버는 안드로이드 패키지 파일인 APK 내에 있는 원본 실행코드(DEX)를 암호화하고, 실행 시 이를 복호화 할 수 있는 스텁(stub) 코드를 APK에 삽입하여 배포한다. 스텁 코드는 자신에 대한 공격을 탐지하기 위해 무결성 검증 코드를 포함한다. 사용자가 해당 APK를 설치·실행할 때, 스텁 코드는 자체의 무결성을 검증한 후, 암호화된 원본 실행코드를 복호화하고, 이를 동적 로딩(dynamic loading)하여 실행한다. 앱의 원본 실행코드는 암호화되어 배포되므로 지적재산권을 효과적으로 보호할 수 있다. 또한, 스텁 코드에 대해 무결성을 검증하므로, 제안 기법의 우회 가능성을 차단한다. 우리는 15개의 안드로이드 앱에 제안 기법을 적용하여 그 유효성을 평가하였다. 실험 결과, 13개의 앱이 정상적으로 동작함을 확인하였다.

• 한국지반공학회논문집
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• 제21권4호
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• pp.51-63
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• 2005

콘크리트 포장에는 재료의 결함이나 구조적 문제점으로 인하여 표면균열, 슬래브의 처짐 등과 같은 문제가 발생할 수 있다. 결함이 발생한 콘크리트 포장이 현재 운용중일 때 심각한 교통문제와 경제적 손실을 초래하지 않기 위해서는 신속하게 유지관리가 이루어져야 한다. 이러한 콘크리트 포장의 신속한 유지관리를 위해서 탄성파 기법이 효율적으로 활용될 수 있는데, 이는 탄성파 기법으로 재료의 공학적 물성과 건전도 상태를 비파괴적, 비관입적으로 신속하게 측정할 수 있고, 또한 콘크리트 하부 기층, 보조기층, 노반의 강성까지 측정할 수 있는 장점을 가지고 있기 때문이다. 본 연구에서는 건전도 평가기법으로서 비파괴 탄성파 기법의 적합성을 평가하기 위하여 탄성파 기법에 관한 수치해석을 수행하였고, 수치해석 결과를 근간으로 하여 콘크리트 활주로의 재료적 결함, 구조적 건전도 상태를 종합적으로 평가하는 비파괴기법 평가체제를 제안하였다. 표면균열이 국부적으로 발달한 $\bigcirc\bigcirc$ 콘크리트 활주로에 대하여 본 연구에서 제안한 평가체제를 적용하여 표면균열의 원인을 규명하였다. 탄성파 건전도 평가로 획득한 활주로의 건전도 상태는 현장에서 채취한 코어와 코어시험 결과를 이용하여 비교함으로써 본 연구에서 제안한 비파괴 탄성파 기법 평가체제의 신뢰성을 확인하였다.

• Asian-Australasian Journal of Animal Sciences
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• 제28권12호
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• pp.1713-1720
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• 2015

The present work describes experiments undertaken to evaluate the usefulness of selected physicochemical indices of semen, cell membrane integrity and sperm chromatin structure for the assessment of boar semen sensitivity to processes connected with pre-insemination procedures. The experiments were carried out on 30 boars: including 15 regarded as providers of sensitive semen and 15 regarded as providers of semen that is little sensitive to laboratory processing. The selection of boars for both groups was based on sperm morphology analyses, assuming secondary morphological change incidence in spermatozoa as the criterion. Two ejaculates were manually collected from each boar at an interval of 3 to 4 months. The following analyses were carried out for each ejaculate: sperm motility assessment, sperm pH measurement, sperm morphology assessment, sperm chromatin structure evaluation and cell membrane integrity assessment. The analyses were performed three times. Semen storage did not cause an increase in the incidence of secondary morphological changes in the group of boars considered to provide sperm of low sensitivity. On the other hand, with continued storage there was a marked increase in the incidence of spermatozoa with secondary morphological changes in the group of boars regarded as producing more sensitive semen. Ejaculates of group I boars evaluated directly after collection had an approximately 6% smaller share of spermatozoa with undamaged cell membranes than the ejaculates of boars in group II ($p{\leq}0.05$). In the process of time the percentage of spermatozoa with undamaged cell membranes decreased. The sperm of group I boars was characterised with a lower sperm motility than the semen of group II boars. After 1 hour of storing diluted semen, the sperm motility of boars producing highly sensitive semen was already 4% lower ($p{\leq}0.05$), and after 24 hours of storage it was 6.33% lower than that of the boars that produced semen with a low sensitivity. Factors that confirm the accuracy of insemination male selection can include a low rate of sperm motility decrease during the storage of diluted semen, low and contained incidence of secondary morphological changes in spermatozoa during semen storage and a high frequency of spermatozoa with undamaged cell membranes.