• 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.

• 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.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.441-446
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• 2001

In the integral type reactor, SMART, all the major components such as steam generators, pressurizer and pumps are located inside the single reactor pressure vessel. The objective of this study is to evaluate the structural integrity for RPV of SMART under the postulated pressurized thermal shock by applying the finite element analysis. Input data for the finite element analysis were generated using the commercial code I-DEAS, and the fracture mechanics analysis was performed using the ABAQUS. The crack configurations, the crack aspect ratio and the clad thickness were considered in the parametric study. The effects of these parameters on the reference nil-ductility transition temperature were also investigated.

• Structural Engineering and Mechanics
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• v.8 no.1
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• pp.103-118
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• 1999

The structural integrity of the reactor pressure vessel under pressurized thermal shock (PTS) is evaluated in this study. For given material properties and transient histories such as temperature and pressure, the stress distribution is found and stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. A round robin problem of the PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study, and the evaluation results are discussed. The maximum allowable nil-ductility transition temperatures are determined for various crack sizes.

• Journal of Korea Foundry Society
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• v.2 no.2
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• pp.2-9
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• 1982

This paper is concerned with the improvement of impact and tensile Properties of spheroidal graphite cast iron of the following duplex matricess which were heat treated in the eutectic transformation temperature range (that is, $({\alpha}+{\gamma})$ coexisting range) ; ferrite-martensite, ferrite-bainite and ferrite-pearlite. The absorbed energy and maximum load was measured by recording the load-deflection curve with instrumented Charpy impact testing machine in the temperature range from $+100^{\circ}C$ to $-196^{\circ}C$. It was found the ferrite-bainite duplex matrix showed the highest toughness among the above matrices in the room temperature and the low temperature range. Comparison of this matrix to ferrite-pearlite matrix(that is, as cast) showed a lowering of $27^{\circ}C$ in the nil-ductility transition temperature (NDT) and a lowering of $40^{\circ}C$ in the ductile-brittle transition temperature (TrE), Which seems to result from the finner dimple pattern observed using miorofractography.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.470-484
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• 1998

An analysis program for the evaluation of pressure vessel integrity under pressurized thermal shock (PTS) is developed. For given material properties and transient history such as temperature and pressure, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Using this program a round robin problem of PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study. The allowable maximum reference nil-ductility transition temperatures are determined for various crack sizes.

• 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.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.12 no.3
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• pp.271-279
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• 1999

본 연구에서는 국내에서 가장 취약할 것으로 예상되는 원자력 발전소에 가압열충격 사고를 유발할 수 있는 주증기관 파단사고를 가정하여 열수력 해석과 파괴역학 해석을 수행하였다. 원전수명관리연구의 일환으로 계통열수력 해석 및 혼합열유동 해석에 의하여 구한 냉각제의 온도와 압력의 이력 및 용기의 재질성분으로부터 용기의 응력확대계수와 파괴인성치를 계산하고 이들을 비교하여 균열의 진전여부를 판단하여 형상계수가 1/6인 표면균열이 견딜 수 있는 최대 기준무연성천이온도를 결정하였다.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.7
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.