• 대한기계학회논문집A
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• 제26권11호
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• pp.2219-2227
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• 2002

In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature ($(RT_{NDT})$) was investigated on the basis of finite element analyses.

• Journal of Mechanical Science and Technology
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• 제19권11호
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• pp.1988-1997
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• 2005

This study is concerned with structural integrity evaluations for the interference-fit flywheels in reactor coolant pumps (RCPs) of nuclear power plants. Stresses in the flywheel due to the shrinkage loads and centrifugal loads at the RCP normal operation speed, design overspeed and joint-release speed are obtained using the finite element method (FEM), where release of the deformation-controlled stresses as a result of structural interactions during rotation is considered. Fracture mechanics evaluations for a series of cracks assumed to exist in the flywheel are conducted, considering ductile (fatigue) and non-ductile fracture, and stress intensity factors are obtained for the cracks using the finite element alternating method (FEAM). From analysis results, it is found that fatigue crack growth rates calculated are negligible for smaller cracks. Meanwhile, the material resistance to non-ductile fracture in terms of the critical stress intensity factor (K$_{IC}$) and the nil-ductility transition reference temperature (RT$_{NDT}$) are governing factors for larger cracks.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.545-553
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• 2016

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.100-106
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• 2014

There is a growing need to introduce advanced pressure vessel steels with higher strength and toughness for the optimizatiooCn of the design and construction of longer life and larger capacity nuclear power plants. SA508 Gr.4N Ni-Cr-Mo low alloy steels have superior strength and fracture toughness, compared to SA508 Gr.3 Mn-Mo-Ni low alloy steel. Therefore, the application of SA508 Gr.4N low alloy steel could be considered to satisfy the strength and toughness required in advanced nuclear power plants. The purpose of this study is to characterize the microstructure and mechanical properties of SA508 Gr.4N low alloy steels. 1 ton ingot of SA508 Gr.4N model alloy was fabricated by vacuum induction melting followed by forging, quenching, and tempering. The predominant microstructure of the SA508 Gr.4N model alloy is tempered martensite having small packet and fine Cr-rich carbides. The yield strength at room temperature was 540MPa, and it was decreased with an increase of test temperature while DSA phenomenon occurred at around $288^{\circ}C$. Overall transition property of SA508 Gr.4N model alloy was much better than SA508 Gr.3 low alloy steel. The index temperature, $T_{41J}$, of SA508 Gr.4N model alloy was $-132^{\circ}C$ in Charpy impact tests, and reference nil-ductility transition temperature, $RT_{NDT}$ of $-105^{\circ}C$ was obtained from drop weight tests. From the fracture toughness tests performed in accordance with the ASTM standard E1921 Master curve method, the reference temperature, $T_0$ was $-147^{\circ}C$, which was improved more than $60^{\circ}C$ compared to SA508 Gr.3 low alloy steels.

• 대한기계학회논문집A
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• 제25권7호
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Journal of Welding and Joining
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• 제31권6호
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• pp.90-95
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• 2013

Pipe inside clad welding is mainly used to the flow pipe of sub-sea or chemical plant. For the inside clad welding to the medium pipe with the diameter of about 12", TIG welding is frequently applied with filler metal. In this case, the clad welding has the very broad weld area over $10m^2$. And, the non-destructive test (NDT) such as ultrasonic test (UT) or radiographic testing (RT) should be conducted on the broad weld area, and it costs very high due to the time-consuming work. Therefore, the present study investigated the variation of arc voltage to develop the in-line quality monitoring system for the pipe inside TIG cladding. The 4 experimental parameters (current, arc length, wire feed position, and shield gas flow rate) varied to observe the change of arc voltage and to establish the model for the monitoring. The arc voltage was decreased when the wire was fed to the backward eccentric position(over 2mm), and the shield gas flow rate was insufficient under 10L/min. In the case of the backward eccentric position over 2mm, the bead appearance was not good and the dilution ratio was increased due to deep penetration. When the shield gas flow rate was lower than 10L/min, the bead surface was oxidized.

• 한국재료학회지
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• 제8권11호
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• pp.1020-1025
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• 1998

Mn-Mo-Ni 저합금강의 중성자 조사에 따른 기계적(미세경도, 인장, 샤피충격시험) 및 자기적(포화자화, 보자력, 잔류자화, Barkhausen Noise(BN)진폭, BN에너지) 성질 변화를 측정하여 이들의 상관관계를 고찰하였다. 기계적 성질시험 결과, 중성자 조사로 인하여 항복강도, 인장강도, 미세경도 및 천이온도($T_{41J}$)는 증가하였고 최대흡수에너지(USE)는 감소하였으며, 인장 시험의 경우 용접금속에서는 모재와 비교했을 때 큰 변화가 없었다. 자기적 성질을 측정한 결과, 잔류자화, BN진폭, BN에너지는 감소하였고 보자력은 급격히 증가하는 것으로 나타났다. 기계적.자기적 성질변화의 상관관계에서 자기적성질인 보자력 증가에 따라 천이온도, 항복강도, 경도는 증가하고 USE는 감소하였고, BN진폭의 경우는 보자력과 반대의 경향을 보였다. 본 실험에서 중성자조사로 인한 기계적.자기적 성질변화가 일관성 있는 상관관계가 있음을 확인하였고, 이들의 변화를 통해 조사손상을 평가하는 데 이용 가능하다.