• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05b
• /
• pp.305-311
• /
• 1998

최근들어 고려된 LBB(Leak Before Break) 적용요건중 동적파괴시힘 절차에는 울진 3&4호기 이후 파단전누설개념이 적용되는 배관이 탄소강으로 제작될 경우. 이 배관이 Dynamic Strain Aging (DSA)에 의해 파괴저항치가 감소되지 않는다는 것이 정량적으로 입증되지 않는 한, 동 배관의 파괴 물성치 결정시 DSA의 영향이 고려되어야 하며, DSA 영향을 평가하기 위해서는 동적과괴시험이 수행되어야 함을 요건화 하고 있다. 본 연구에서는 DSA 효과에 의한 파괴저항(J-R) 특성의 저하가차세대원전 원자로냉각재배관 파단전누설개넘(LBB) 적용시 설계 안전여유도에 영향을 미치지 않는 정도임을 평가하는데 있다. 따라서 ASME Section III에서 탄소강으로 분류하고 있는 강종별 파괴인성 변화를 고찰하고, 차세대원전 주냉각재배관 재료인 SA508 Class la의 최대 파괴인성 감소치를 예측하여, 울진 3&4호기에서 측정된 엘보우용 SA516-Gr.70 강의 DSA 영향 평가 결과와 비교 분석하여 차세대원전 주냉각재배관의 DSA영향을 평가하였다. 도출된 결론으로는 DSA 영향을 고려한 SA508 Class la의 J 및 dJ/dA 값은 극히 보수적으로 추정할 때 50% 이상 감소하는 것으로 예측된다. 이러한 DSA 영향을 고려하였을 경우 배관재 모재의 파괴인성치는 Weld-SAW의 J/T 값 수준으로 감소하였다. 그러나 현 LRB 해석이 가장 낮은 J/T값을 갖는 Weld-SAW Auto의 균열길이 2a인 J/T선도에 의거하여 수행되고 있다는 점을 고려한다면 비록 DSA가 배관재에 영향을 주는 가장 보수적인 값(J 및 dJ/dA값을 50% 이상)을 사용한다고 하더라도 차세대원전 LBB 적용에 문제가 되지 않음을 알 수 있다. 즉 차세대원자로 주냉각재배관에 LBB를 적용하는데는 DSA 영향은 상대적으로 중요하지 않다는 결론을 얻었다. 표면에 수소화물이 농축되어 있는 hydride layer가 형성됨을 관찰하였으며 ～5,000ppm 이상의 경우에는 수소화물의 방향성이 random하였으며 특히, ZIRLO$^{TM}$ 시편의 경우에서는 원주방향으로 길게 이어진 수소화물과 기계적 성질에 치명적인 반경방향의 수소화물이 평행하게 배열된 것을 관찰하였다.하였을 때는 Li$_2$O의 첨가에 의해 치밀화가 주로 일어났고, 반면에 $N_2$-7vol.%H$_2$ 분위기에서 소결하면 Li$_2$O의 첨가에 의해 작은 기공은 소멸되고 큰 기공이 생성되었다.지나치게 모국어의 영향만 강조하고 다른 요인들에 대해서는 다분히 추상적인 언급으로 끝났지만 이 분석을 통 해서 배경어, 목표어, 특히 중간규칙의 역할이 괄목할 만한 것임을 가시적으로 관찰할 수 있 다. 이와 같은 오류분석 방법은 학습자의 모국어 및 관련 외국어의 음운규칙만 알면 어느 학습대상 외국어에라도 적용할 수 있는 보편성을 지니는 것으로 사료된다.없다. 그렇다면 겹의문사를 [-wh]의리를 지 닌 의문사의 병렬로 분석할 수 없다. 예를 들어 누구누구를 [주구-이-ν가] [누구누구-이- ν가]로부터 생성되었다고 볼 수 없다. 그러므로 [-wh] 겹의문사는 복수 의미를 지닐 수 없 다. 그러면 단수 의미는 어떻게 생성되는가\ulcorner 본 논문에서는 표면적 형태에도 불구하고 [-wh]의미의 겹의문사는 병렬적 관계의 합성어가 아니라 내부구조를 지니지 않은 단순한 단어(minimal $X^{0}$ elements)로 가정한다. 즉, [+wh] 의미의 겹의문사는 동일한 구성요 소를 지닌 병렬적 합성어([$[W1]_{XO-}$ $[W1]_{XO}$ ]$_{XO}$)로

• Journal of the Korean Society for Nondestructive Testing
• /
• v.33 no.2
• /
• pp.138-146
• /
• 2013

The present paper provides the elastic stress intensity factors(SIFs) of thick-walled cylinder with non-idealized circumferential through-wall cracks. For estimating these elastic SIFs, the systematic three-dimensional(3D) elastic finite element(FE) analyses were performed. In order to consider practical shape of thick-walled cylinder and non-idealized circumferential through-wall crack, the values of thickness of cylinder, reference crack length and crack length ratio were systematically varied. As for loading conditions, axial tension, global bending and internal pressure were considered. In particular, in order to calculate the SIFs of thick-walled cylinder with non-idealized circumferential through-wall crack from those of thick-walled cylinder with idealized circumferential through-wall crack, the correction factor representing the effect of non-idealized crack on the SIFs were proposed in this paper. The present results can be applied to accurately evaluate the rupture probabilities of nuclear piping considering actual crack growth behaviors.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.8 no.3
• /
• pp.13-22
• /
• 2012

According to recent research on leak-rate estimates to assess rupture probabilities of nuclear piping which contains a circumferential surface/through-wall cracks due to PWSCC, i.e., xLPR (Extremely Low Probability of Rupture) program, it has been revealed that the use of crack shape with an idealized circumferential through-wall crack during actual crack growth can lead to overestimate of the leak-rate. Thus, for accurate estimation of the leak-rate during crack growth, the more realistic crack shape that can simulate the crack shape transition from surface crack to through-wall crack should be used. In this context, in the present study, the elastic crack opening displacement of slanted circumferential through-wall crack in thick-walled cylinder was proposed based on 3-dimensional elastic finite element fracture mechanics analyses. To propose the elastic crack opening displacement of slanted circumferential through-wall crack in thick-walled cylinder, the geometric variables affecting crack opening displacement, i.e., thickness of cylinder, reference inner crack length and slant crack ratio were systematically varied. In terms of loading conditions, axial tension, global bending moment and internal pressure were considered. The present results can be applied to calculate the leak-rate considering more realistic crack shape transition from surface crack to idealized through-wall crack, and can be expected to enhance current leak-rate estimation scheme, for instance, in xLPR program etc.

• Journal of Ocean Engineering and Technology
• /
• v.13 no.3 s.33
• /
• pp.36-48
• /
• 1999

Fatigue life and crack retardation behavior after penetration were experimentally examined using surface pre-cracked specimens of aluminium alloy 5083. The Wheeler model retardation parameter was used successfully to predict crack growth behavior after penetration. By using a crack propagation rule, the change in crack shape after penetration can be evaluated quantitatively. Advanced, waveform-based acoustic emission (AE) techniques have been successfully used to evaluate signal characteristics obtained form fatigue crack propagation and penetratin behavior in 6061 aluminum plate with surface crack under fatigue stress. Surface defects in the structural members are apt to be origins of fatigue crack growth, which may cause serious failure of the whole structure. The nondestructive analysis on the crack growth and penetration from these defects may, therefore, be one of the most important subjects on the reliability of the leak before break (LBB) design. The goal of the present study is to determine if different sources of the AE could be identified by characteristics of the waveforms produced from the crack growth and penetration. AE signals detected in four stages were found to have different signal per stage. With analysis of waveform and power spectrum in 6061 aluminum alloys with a surface crack, it is found to be capabilities on real-time monitoring for the crack propagation and penetration behavior of various damages and defects in structural members.

• Transactions of the Korean Society of Mechanical Engineers
• /
• v.19 no.4
• /
• pp.1051-1061
• /
• 1995

In order perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain (.sigma. - .epsilon.) curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to develop two methods for J-R curve prediction. In the first method, elastic-plastic finite element analyses for a series of crack length / specimen width ratio were performed. Accordingly the load versus load line displacement (P .delta.) curve corresponding to the fracture strain is obtained and the J-R curve based on the generalized locus method is obtained. In the second method, the correlation between .sigma.-.epsilon. curves and J-R curves was statistically analyzed and an empirical equation to predict the J-R curve from the .sigma.-.epsilon. test result is proposed. A good correlation between the predicted results based on the proposed methods and the experimental ones is obtained.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.26 no.6
• /
• pp.1099-1106
• /
• 2002

This paper provides engineering J-integral and crack opening displacement (COD) estimation equations for circumferential through-wall cracked pipes under internal pressure and under combined internal pressure and bending. Based on selected 3-D finite element calculations for the circumferential through-wall cracked pipes under internal pressure using the idealized power law materials, the elastic and plastic influence functions for fully plastic J-integral and COD solutions are found as a function of the normalized crack length and the mean radius-to-thickness ratio. These developed GE/EPRI-type solutions are then re-formulated based on the enhanced reference stress method. Such re-formulation not only provides simpler equations for J-integral and COD estimations, but also can be easily extended to combined internal pressure and bending. The proposed equations are compared with elastic-plastic finite element results using actual stress-strain data, which shows overall excellent agreement.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.20 no.6
• /
• pp.1843-1853
• /
• 1996

The objective of this paper is to evaluate the crack stability of the nuclear main stresm pipes, considering the load reduction effect due to the presence of circumferential throuth-wall crack. Also, the optimization techniques are adoped tosimulate the crack effect on the elbow component of the piuping system. By using a general beam elemetn which contains a discontinuous cross-section, the piping analysis is accomplished to acquire the reduced load. Considering this reduced load, it is feasible for the LBB application in nuclear main stresm pipe. Also, by combining an optimization program and a genaral finite element analysis program, the appropriate dimensions of the simplified beam elemtn which represents the effect of crack in elbow could be successfully determined.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.20 no.2
• /
• pp.529-539
• /
• 1996

Fracture resistance(J-R) curves, which are used for elastic-plastic fracture mechanics analyses, are known to be dependent on the specimen geometry. The objective of this paper is to investigate the effect of crack tip constraint an the J-R curves in CT specimens. Fracture toughness tests on CT specimens with varying planform size were performed and test results showed that the J-R curves were increased with an increase in the planform size. Finite element analysis were also performed and the numerical results showed that this experimental phenomenon was probably due to the relaxation of crack tip constraint resulting from the stress triaxiality.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.21 no.11
• /
• pp.1786-1795
• /
• 1997

In order to perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to modify two J-R curve prediction methods previously proposed by the authors and to propose an additional J-R curve prediction method for nuclear piping materials. In the first method which is based on the elastic-plastic finite element analysis, a blunting region handling procedure is added to the existing method. In the second method which is based on the empirical equation, a revised general equation is proposed to apply to both carbon steel and stainless steel. Finally, in the third method, both full stress-strain curve and finite element analysis results are used for J-R curve prediction. A good agreement between the predicted results based on the proposed methods and the experimental ones is obtained.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.62-69
• /
• 2013

A probabilistic assessment code, PRO-LOCA ver. 3.7 which was developed in an international co-operative research program, PARTRIDGE was evaluated by conducting sensitivity analysis. The effect of some variables such as simulation methods (adaptive sampling, iteration numbers, weld residual stress model), crack features(Poisson's arrival rate, maximum numbers of cracks, initial flaw size, fabrication flaws), operating and loading conditions(temperature, primary bending stress, earthquake strength and frequency), and inspection model(inspection intervals, detectable leak rate) on the failure probabilities of a surge line nozzle was investigated. The results of sensitivity analysis shows the remaining problems of the PRO-LOCA code such as the instability of adaptive sampling and unexpected trend of failure probabilities at an early stage.