• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.276-282
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• 1996

일체형원자로는 노심, 증기발생기, 가압기, 펌프 등 1차측 주기기들을 하나의 압력용기안에 모두 포함하고 있고, 또 1차측 냉각재가 원자로 안에서만 순환하므로 기존의 분리형원자로에 비해 구조특성상 대용량 원자로 냉각재 상실사고(LBLOCA)의 발생 가능성을 원천적으로 제거할 수 있다. 반면 원자로 냉각재의 보충 등을 위한 소형 배관의 파단 가능성은 역시 존재하므로 소용량 원자로 냉각재 상실 사고(SBLOCA)는 여전히 존재한다. 따라서 현재 한국원자력연구소에서 연구 개발중인 중소규모 전력생산 및 열 활용 목적의 일체형 원자로에는, 원자로 압력용기 외부에 별도의 압력용기(안전용기)를 설치하여 SBLOCA시 원자로 압력용기로부터 방출되는 냉각수를 안전 용기내에 보관하도록 함으로써 사고시 외부로의 방사성 물질 유출 가능성을 획기적으로 줄 일수 있는 설계 개념을 도입하고 있다. 본 논문에서는 안전용기의 설계시 효율적인 냉각방식에 대한 열유체 해석적 접근을 시도하였고, 예비개념설계된 일체형 열병합원자로의 설계상의 특징들 및 안전용기 설계시 앞으로의 연구방향 등도 간략히 소개하였다.

• Journal of the Korean Institute of Gas
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• v.26 no.6
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• pp.16-23
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• 2022

In this study, characteristics of effect on hydrogen gas was investigated to hydrogen embrittlement by disk and tensile tests. The developed and commercial alloy was fabricated to a plate material made from an alloy ingot. The prepared materials were processed in the form of a disk to measure rupture pressure by hydrogen and helium gas at a rate of 0.1 to 1,000 bar/min. In the hydrogen pre-charged tensile test, a specimen was hydrogenated using an anode charging method, and the yield strength, ultimate tensile strength, elongation, and reduction in area rate were carried by a strain rate test. Also, the microstructure was observed to the fracture surface of the tensile test specimen. As a result, the developed materials satisfied endurable hydrogen embrittlement, and the fractured surface showed a brittleness fracture surface with a depth of several ㎛, but dimple due to ductile fracture could be observed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.10
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• pp.1612-1620
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• 2004

This study presents methods to estimate elastic-plastic crack opening displacement (COD) fur circumferential through-wall cracked pipes for the Leak-Before-Break (LBB) analysis of pressurized piping. Proposed methods are based not only on the GE/EPRI approach but also on the reference stress approach. For each approach, two different estimation schemes are given, one for the case when full stress-strain data are available and the other fur the case when only yield and ultimate tensile strengths are available. For the GE/EPRI approach a robust way of determining the Ramberg-Osgood (R-O) parameters is proposed, not only fur the case when detailed information on full stress-strain data is available but also for the case when only yield and ultimate tensile strengths are available. The COD estimates according to the GE/EPRI approach, using the R-O parameters determined from the proposed R-O fitting procedures, generally compare well with the published pipe test data. For the reference stress approach, the COD estimates according to the method based on both full stress-strain data and limited tensile properties are in good agreement with pipe test data. In conclusion, experimental validation given in the present study provides sufficient confidence in the use of the proposed method to practical LBB analyses even though when information on material's tensile properties is limited.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.1
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• pp.32-38
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• 2001

Although the ultrasonic method has been developed and used widely in the fields, it has been used only for measuring the defect size and thickness loss. In this study, the relationship between surface wave attenuation through micro-crack growth and variation of velocity under repeated cyclic loading has been investigated. The specimens are adopted from 2.25Cr-1Mo steel, which is used for power plant and pipeline system, and have dimensions of $200{\times}40{\times}4mm$. The results of ultrasonic test with a 5MHz transducer show that surface wave velocity gradually decreases from the point of 60% of fatigue life and the crack length of 2mm with the increasing fatigue cycles. From the results of this study, it is found that the technique using the ultrasonic velocity change is one of very useful methods to evaluate the fatigue life nondestructively.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.742-749
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• 2003

Failure of a pipeline due to local wall thinning is getting more attention in the nuclear power plant industry. Although guidelines such as ANSI/ASME B31G and ASME Code Case N597 are still useful fer assessing the integrity of a wall thinned pipeline, there are some limitations in these guidelines. For instance, these guidelines consider only pressure loading and thus neglect bending loading. However, most Pipelines in nuclear power plants are subjected to internal pressure and bending moment due to dead-weight loads and seismic loads. Therefore, an assessment procedure for locally wall thinned pipeline subjected to combined loading is needed. In this paper, three-dimensional finite element(FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Maximum moments based on true ultimate stress(${\alpha}$$\sub$u,t/) were obtained from FE results to predict the failure of the pipe. These results were compared with test results, which showed good agreement. Additional finite element analyses were performed to investigate the effect of key parameters, such as wall thinned depth, wall thinned angle and wall thinned length, on maximum moment. Also, the effect of internal pressure on maximum moment was investigated. Change of internal pressure did not show significant effect on the maximum moment.

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

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.4
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• pp.1051-1061
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• 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
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• v.26 no.6
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• pp.1099-1106
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• 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
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• v.20 no.6
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• pp.1843-1853
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• 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
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• v.21 no.11
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• pp.1786-1795
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• 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.