• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.648-653
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• 2003

In order to predict stress intensity limit of high-temperature creep structures, creep work-time equation, defined as $W_ct^P=B$, was used, and the results of the equation were compared with isochronous stress-strain curve (ISSC) ones of ASME BPV NH Code. For this purpose, the creep strain tests with. time variations for commercial type 316 stainless steel were conducted with different stresses; 160 MPa, 150 MPa, 145 MPa, 140 MPa and 135 MPa at $593^{\circ}C$. The results of log $W_c$ and log t plots showed a good linear relation up to $10^5$ hr. The constants p, B and stress intensity limit values showed comparatively good agreement to those of ASME NH ISSC. It is believed that the relation can be simply obtained with only several short-term 1% strain data without ISSC which can be obtained by long-term creep data.

• 대한기계학회논문집A
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• 제27권5호
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• pp.750-757
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• 2003

In order to determine creep stress intensity limit of high-temperature components, the usefulness of the creep work and time equation, defined as W$\_$c/t$\^$p/ = B(where W$\_$c/ = $\sigma$$\varepsilon$ is the total creep work done during creep, and p and B are constants), was investigated using the experimental data. For this Purpose, the creep tests for generating 1.0% strain for commercial type i16 stainless steel were conducted with different stresses; 160 MPa, 150 MPa, 145 MPa, 140 MPa and 135 MPa at 593$^{\circ}C$. The plots of log W$\_$c/ - log t showed a good linear relation up to 10$\^$5/ hr, and the results of the creep work-time relation for p, B and stress intensity values showed good agreement to those of isochronous stress-strain curves (ISSC) presented in ASME BPV NH. The relation can be simply obtained with only several short-term 1% strain data without ISSC which can be obtained by long-term creep data. Particularly, this relation is useful in estimating stress intensity limit for new and emerging class of high-temperature creeping materials.

• 대한기계학회논문집
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• 제8권2호
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• pp.141-144
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• 1984

본 연구에서는 완전 역 피로응력(completely reversed fatgue stress)를 받는 타원 노치에서 균열 발생과 전파에 대한 이론 임계 피로 한도를 응력 세기 계수 개념 을 도입하여 임의의 재질과 임으의 타원 노치 형상에 대해서 적용할 수 있도록 유도하 며 그 결과는 기존 이론보다 Frost의 실험치에 더 잘 일치함을 보이고자 한다.

• Journal of Advanced Marine Engineering and Technology
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• 제31권4호
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• pp.385-393
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• 2007

The strength evaluation of the most weakest junction part is required for the safety design of all structures. Most of all. in order to enhance the reliability and safety of the welding part. whose use is the highest, it is very important to establish the efficient structure manufacturing technology by studying and investigating the evaluation of fatigue strength in various environments. This study analyzed the relations of da/dN, and th according to the welding methods of SMAW, FCAW, and SAW. In the stage II. the value of stress intensity factor range was the highest in SMAW welding method of stress ration R=0.1, and appeared under the sequence of FCAW and SAW and as the completion section of stress intensity factor was low, threshold stress intensity factor was lowly formed in da/dN - The fatigue life of each welding method is sensitively worked in high stress ratio. judging from the fact that the width of life reduction increases in the high stress ratio zone compared to the width of life reduction in the low stress ratio zone. In the fatigue limit of welding methods before corrosion. the welding of SMAW and FCAW shows the same fatigue limit compared to Base metal, and SAW holds the lowest fatigue limit value.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.498-513
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• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

• Structural Engineering and Mechanics
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• 제14권2호
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• pp.191-208
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• 2002

A comparative assessment study for a generation of the pressure-temperature (P-T) limit curve of a reactor vessel is performed in accordance with ASME code. Using cooling or heating rate and vessel material properties, stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during reactor cool-down and heat-up. P-T limit curves are analyzed with respect to defect orientation, clad thickness, toughness curve, cooling or heating rate and neutron fluence. The resulting P-T curves are compared each other.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2188-2197
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• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

• 한국산업융합학회 논문집
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• 제25권2_1호
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• pp.141-148
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• 2022

In this paper, the evaluation equations proposed by Tange et al. and Ando et al. were used to evaluate the threshold stress intensity factor ∆K^R_th(s) and fatigue limit ∆𝜎^R_wc, according to the small crack of offshore structural steel F690. Despite the differences in concept and shape of the two equations, the ∆K^R_th(s) and ∆𝜎^R_wc proved completely consistent. It is possible to use these equations to evaluate the dependence of the crack length on the ∆K^R_th(s) and ∆𝜎^R_wc of structures made of all steel grades. With these equations, the characteristics of microcracks can be quantitatively evaluated, and the safety and reliability of the structure can be secured.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.1-8
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• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

• 열처리공학회지
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• 제16권6호
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• pp.315-319
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• 2003

Effect of nitriding on fatigue crack initiation and growth rate has been studied in Ni-Cr-Mo steel. Specimens were nitrided at $860^{\circ}C$ for 15 hr. The fatigue limit of nitrided specimens were superior to those ofannea1ed($860^{\circ}C$, 15 hr) specimens. Based on detailed observations of slip band and micro crack initiation, it is concluded that the excellent fatigue limit of nitrided specimens is attributed to improved slip initiation resistance by nitriding. The characteristic of fatigue crack growth rate of nitrided specimens was investigated by comparing with those of annealed specimens. It was found that the crack growth rate was markedly decreased and the threshold stress intensity factor range was improved by nitriding. It is concluded that the excellent fatigue limit of nitrided specimens is also attributed to improved fatigue crack growth rate and threshold stress intensity factor range by nitriding.