• Proceedings of the Korean Society of Precision Engineering Conference
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• 2008.06a
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• pp.719-720
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• 2008

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2010.11a
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• pp.943-944
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• 2010

• Journal of the Korean Society for Precision Engineering
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• v.19 no.11
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• pp.174-182
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• 2002

The integrity evaluation of pressure tube is essential for the safety of CANDU reactor, and integrity must be assured when flaws or contacts between pressure tube and surrounding calandria tube are found. In order to complete the integrity evaluation, not only complicated and iterative calculation procedures but also a lot of data and knowledge are required. For this reason, an integrity evaluation system, which provides an efficient way of the evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec.？ and FFSG issued by the AECL, and applicable for the evaluation of blister, sharp flaw and blunt notch. Delayed hydride cracking and blister evaluation modules are included in the general flaw and notch evaluation module. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Steel and Composite Structures
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• v.3 no.4
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• pp.231-242
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• 2003

This paper deals with the formulation and development of fracture criteria for high-strength structural members containing surface damage in the form of notches (i.e., blunt defects). The important role of the yield strength of the material and its strain hardening capacity (evaluated by means of the constitutive law or stress-strain curve) is analysed in depth by considering the fracture performance of notched samples taken from high-strength steels with different levels of cold drawing (the most heavily drawn steel being commercial prestressing steel used in prestressed concrete). The final aim of the paper is to establish fracture-based design criteria for structural members made of steels with distinct yield strength and containing very different kinds of notch-shape surface damage.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1996.04a
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• pp.704-708
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• 1996

The elastic dynamic crack curving could be obtained by controlling the loading rate, the initial crack-tip blunting to store much energy before crack initiation and the magnitude of reflected wave from finite boundaries. However there is no theoretical and experimental elastic-plastic dynamic curving study. This paper proposes a specimen geometryfor a study of dynamic elastic-plastic crack curving and presents a preliminary result. The specimen has a blunt physical crack tip on a side, and a round notch tip on the other side. From the experiment using this specimen, it is found that the narrow plastic zone ahead of the round notch tip produces the change of load direction and anti-symmetricity of the dynamic isochromatics, and each result causes the crack curving phenomenon. After a certain time, as the elastic-plastic crack gets close to the round notch tip near, the degree of the crack curving get larger. The elastic reack curving propagates more sensitively to the surround of crack tip than the plastic crack curving does. The cynamic elastic-plastic crack curving is found to be proportional to the CTOA(the crack tip opening angle). The dynamic elastic-plastic crack may propagate in the direction perpendicular to the loading. An apparant strip yield zone which is similar to the Dugdale strip yield zone is noted ahead of the physical crack tip.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.1 s.173
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Journal of Mechanical Science and Technology
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• v.17 no.7
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• pp.947-957
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• 2003

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle. In order to complete the integrity evaluation of pressure tube, expert knowledge, iterative calculation procedures and a lot of input data are required. More over, results of integrity assessment may be different according to the evaluation method. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached database, was developed. The present system was built on the basis of 3D FEM results, ASME Sec. XI, and Fitness For Service Guidelines for CANDU pressure tubes issued by the AECL (Atomic Energy Canada Limited). The present system also covers the delayed hydride cracking and the blister evaluation, which are the characteristics of pressure tube integrity evaluation. In order to verify the present system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Transactions of the Korean Society of Mechanical Engineers
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• v.11 no.5
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• pp.740-745
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• 1987

For the prediction of the crack growth initiation from a blunt notch or a precrack in a prestrained material under plane strain tension and small-scale yielding conditions, a microscopic fracture criterion is proposed in terms of the crack tip opening displacement(COD) needed for the attainment of fracture strain at a microstructural distance. Smooth blunting of a crack tip with an initial root radius is assumed, and strain distributions on the crack-line axis are calculated at each deformation stage until the distributions against an original distance normalized to the COD are insensitive to an initial root radius. This case of no initial-root-radius effect is taken as for a sharp crack tip, on which the criterion is applied to determine the characteristic length of material from a critical COD for a fatigue-precracked specimen. The predicted COD at the fracture initiation from a crack with an initial root radius or a prestraining shows reasonable agreement with experimental values.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.639-656
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• 2006

In order to apply the leak before break (LBB) concept to nuclear piping systems, the dynamic strain aging effect of low carbon steel materials has to be taken into account, in compliance with the requirements of the Korean Standard Review Guide (KSRG) 3.6.3-1. For this goal, J-R tests are needed for a range of various temperatures and loading rates, including dynamic loading conditions. In the dynamic loading J-R test, the unloading compliance method can not be applied to measure the crack growth and direct current potential drop (DCPD) method; this method also has a problem defining the crack initiation point. The normalization method is known as a very useful method to determine the J-R curve under dynamic loading because it does not need additional equipment or complicated loading sequences such as electric current or unloading. This method was accepted by the American Society for Testing and Materials (ASTM) as a standard test method E1820 A15 in 2001. However, it has not yet been clearly verified yet if the normalization method is sufficiently reliable to be applied to LBB. In this study, the basic background of the J-integral, LBB and dynamic loading J-R test are explained, and the current status for dynamic loading J-R test methods are reviewed from the view point of LBB for nuclear piping. In particular, the theoretical and historical background of the normalization method which has received attention recently, is summarized. Recent studies for this method are introduced and future works are suggested that may improve the reliability of LBB for nuclear piping.