• Title/Summary/Keyword: Nil-Ductility Transition Reference Temperature: $RT_{NDT}$

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Evaluation of the Preirradiation Baseline Material Characteristics for Yonggwang Nuclear Reactor Pressure Vessel (영광 원자력 발전소 원자로 소재의 가동전 재료 물성 특성)

  • Kim, K.C.;Kim, J.T.;Suk, J.I.;Kwon, H.K.;Sung, U.H.
    • Proceedings of the KSME Conference
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    • 2000.11a
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    • pp.153-158
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    • 2000
  • Nuclear reactor pressure vessel should be safety even in the case that hypothetical defects with allowable size are in vessel. Therefore, the materials should have excellent fracture resistance characteristics. The purpose of this study is to analyze the results of preirradiation baseline test of nuclear pressure vessel for Yonggwang Unit 5/6. In experiments, drop weight tests and impact tests are carried out to obtain nil-ductility transition reference temperature, $RT_{NDT}$ and static and dynamic fracture toughness tests are performed to compare with $K_{IR}$ curve in accordance with ASME Sec.III. The test results show that the materials had sufficiently fracture resistance characteristics for 40 years of design life.

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Maximum Allowable $RT_{NDT}$ of Nuclear Reactor Vessel for Pressurized Thermal Shock Accident (가압열충격 사고에 대한 원자로 용기의 최대 허용 기준무연성천이온도)

  • 정명조;박윤원;송선호
    • Computational Structural Engineering
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    • v.11 no.1
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    • pp.153-160
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    • 1998
  • A small break loss of coolant accident is postulated as a pressurized thermal shock accident in this study. From the temperature and pressure histories of coolant, distributions of the temperature and stress in a vessel wall are analytically calculated. The stress intensity factor and fracture toughness of the vessel wall are determined at the crack tip using the ASME code method and they are compared to check if cracking is expected to occur during the transient postulated. The maximum allowable reference nil-ductility transition temperatures are determined for various crack sizes and the results are discussed.

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Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

  • Park, Jeong Soon;Choi, Young Hwan;Jhung, Myung Jo
    • Nuclear Engineering and Technology
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    • v.48 no.2
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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.

High Strength SA508 Gr.4N Ni-Cr-Mo Low Alloy Steels for Larger Pressure Vessels of the Advanced Nuclear Power Plant (차세대 원전 대형 압력용기용 고강도 SA508 Gr.4N Ni-Cr-Mo계 저합금강 개발)

  • Kim, Min-Chul;Park, Sang-Gyu;Lee, Ki-Hyoung;Lee, Bong-Sang
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.10 no.1
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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.

Deterministic Fracture Mechanics Analysis of Nuclear Reactor Pressure Vessel Under Rot Leg Leak Accident (고온관 누설에 의한 가압열충격 사고시 원자로 용기의 건전성 평가를 위한 결정론적 파괴역학 해석)

  • Lee, Sang-Min;Choi, Jae-Boong;Kim, Young-Jin;Park, Youn-Won;Jhung, Myung-Jo
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.26 no.11
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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.

Structural Integrity Evaluation for Interference-fit Flywheels in Reactor Coolant Pumps of Nuclear Power Plants

  • Park June-soo;Song Ha-cheol;Yoon Ki-seok;Choi Taek-sang;Park Jai-hak
    • Journal of Mechanical Science and Technology
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    • v.19 no.11
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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.