• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.10
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• pp.1567-1574
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• 2001

In order to operate an RPV safely it is necessary to keep the pressure-temperature (P-T) limit during the heatup and cooldown process. While the ASME Code provides the P-T limit curve for safe operation, this limit curve has been prepared under conservative assumptions In this paper the effects of conservative assumptions involved in the P-T limit curve specified in the ASME Code Sec. XI were investigated. Three different parameters the crack depth the cladding thickness and the cooling rate, were reviewed based on 3-D finite element analyses. Also the constraint effect on P-T limit curve generation was investigated based on J- T approach. It was shown that the crack depth and the constraint effect change the safe region in P-T limit curve significantly Therefore it is recommended to prepare a more precise P-T limit curve based on finite element analysis to obtain P-T limit for safe operation of an RPV.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.3
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• pp.505-513
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• 2002

During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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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.

• Structural Engineering and Mechanics
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• v.14 no.2
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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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• v.33 no.5
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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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• v.53 no.6
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• pp.1271-1286
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• 2015

Recently, US NRC revised fracture toughness requirements as 10CFR50.61a to reduce the conservatism of 10CFR50.61. However, unlike previous studies relating to the initial regulation, structural integrity evaluations based on the alternative regulation are not sufficient. In the present study, PTS and P-T limit curve evaluations were carried out by using both regulations and resulting data were compared. With regard to the PTS evaluation, the results obtained from the alternative requirements were satisfied with the criterion whereas those obtained from the initial requirements did not meet the criterion. Also, with regard to the P-T limit curve evaluation, operating margin by 10CFR50.61a was greater than that by 10CFR50.61.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.201.1-201
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• 2002

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.2
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• pp.99-107
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• 2009

This study produced a design curve and fatigue limit for a variation in volume ratio and reduction ratio of TiNi/Al composites. In many cases, stress-life curve does not indicate fatigue limit, so it was presented by probabilistic-stress-life curve. Goodman diagram was used to analyze the fatigue strength of materials with a finite life determined by repeated load and the fatigue strength of endurance limit with an infinite life. The fatigue experiment was conducted using the scenk-type plane bending specimen in same shape. The result of the fatigue test, which had been conducted under consistent stress amplitude, was examined. (i) The optimal condition for TiNi/Al in accordance with hot pressing (ii) Impacts of fatigue limit caused by a variation in reduction ratio and volume ratio of TiNi/Al composites (iii) Probability distribution for fatigue limit of TiNi/Al2024 and TiNi/Al6061.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.347.2-347
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• 2001

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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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.