• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1681-1688
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• 2019

In accordance with ASME Boiler and Pressure Vessel (B&PV) Code Sec.XI Appendix. G, a postulated crack is located at the beltline of a reactor pressure vessel because the neutron flux at the beltline is higher than elsewhere. This means that the distance between the core and the semi-spherical bottom head is longer than the distance between the core and the cylindrical beltline. However, several Small and Medium sized Reactors have bottom heads with diverse shapes, including dished or semi-elliptical shapes, to satisfy the requirement and performance. So, the aim of this paper is to evaluate the effect of crack location on Pressure-Temperature limit curve. To do this, two types of postulated crack location, such as beltline and semi-elliptical bottom head, were adopted to derive the Pressure-Temperature limit curve. Also, parametric studies for neutron flux, crack shape and so on were performed. As a result, core critical temperature of semi-elliptical bottom head is found to higher than that of beltline even when they have same values of thickness and neutron flux. This result will be useful to enhance the understanding of Pressure-Temperature limit curve.

• 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.

• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.647-654
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• 2007

Postulation of flaws, one of the most important areas in RPV integrity assessment, significantly affects the results. In the present work, several parameters, such as orientation, underclad vs. surface cracking, crack depth and shape, etc., are postulated and parametric studies are performed to investigate the influence of the flaw parameters on the structural integrity assessment of the reactor pressure vessel during pressurized thermal shock. The influence of individual parameters describing the crack is evaluated based on sensitivity study results.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.70-76
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• 2013

Several finite element models for the leak-before-break (LBB) assessment of overlay dissimilar metal weldment were constructed and analyzed to develop a simple finite element modeling method. The J-integral, crack opening displacement (COD) and J-integral distribution along the crack front in thickness direction due to the applied moment were obtained from the analysis results of the constructed finite element models, and studied compared to the previous literatures. It is concluded that the modeling with base material only is simple and produces a slightly conservative results compared to the complex modeling composed with weld metal and base metal in the calculation of J-integrals and COD values which are used for the calculation of fracture toughness and postulated leakage crack length respectively.

• 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.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.90-95
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• 2004

The probabilistic fracture analysis is used to determine the effects of uncertainties involved in material properties, location and size of flaws, etc, which can not be addressed using a deterministic approach. In this paper the probabilistic fracture analysis is applied for evaluating the RV(Reactor Vessel) under PTS(Pressurised Thermal Shock). A semi-elliptical axial crack is assumed in the inside surface of RV. The selected random parameters are initial crack depth, neutron fluence, chemical composition of material (copper, nickel and phosphorous) and $RT_{NDT}$. The deterministically calculated $K_I$ and crack tip temperature are used for the probabilistic calculation. Using Monte Carlo simulation, the crack initiation probability for fixed flaw and PNNL(Pacific Northwest National Laboratory) flaw distribution is calculated. As the results show initiation probability of fixed flaw is much higher than that of PNNL distribution, the postulated crack sizes of 1/10t in this paper and 1/4t of ASME are evaluated to be very conservative.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.254-266
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• 2017

Finite-element analysis based on elastic-perfectly plastic material was conducted to examine the influence of structural deformations on collapse loads of circumferential through-wall critically cracked $90^{\circ}$ pipe bends undergoing in-plane closing bending and internal pressure. The critical crack is defined for a through-wall circumferential crack at the extrados with a subtended angle below which there is no weakening effect on collapse moment of elbows subjected to in-plane closing bending. Elliptical and semioval cross sections were postulated at the bend regions and compared. Twice-elastic-slope method was utilized to obtain the collapse loads. Structural deformations, namely, ovality and thinning, were each varied from 0% to 20% in steps of 5% and the normalized internal pressure was varied from 0.2 to 0.6. Results indicate that elliptic cross sections were suitable for pipe ratios 5 and 10, whereas for pipe ratio 20, semioval cross sections gave satisfactory solutions. The effect of ovality on collapse loads is significant, although it cancelled out at a certain value of applied internal pressure. Thinning had a negligible effect on collapse loads of bends with crack geometries considered.

• Structural Engineering and Mechanics
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• v.18 no.1
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• pp.91-112
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• 2004

Concrete is a composite material and at meso-level, may be assumed to be composed of three phases: aggregate, mortar-matrix and aggregate-matrix interface. It is postulated herein that although non-linear material parameters are generally used to model this composite structure by finite element method, linear elastic fracture mechanics principles can be used for modelling at the meso level, if the properties of all three phases are known. For this reason, a novel meso-mechanical approach for concrete fracture which uses the composite material model with distributed-phase for elastic properties of phases and considers the size effect according to linear elastic fracture mechanics for strength properties of phases is presented in this paper. Consequently, the developed model needs two parameters such as compressive strength and maximum grain size of concrete. The model is applied to three most popular fracture mechanics approaches for concrete namely the two-parameter model, the effective crack model and the size effect model. It is concluded that the developed model well agrees with considered approaches.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.4 s.247
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• pp.420-427
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• 2006

The structural and leakage integrity of steam generator tubes should be sustained against all postulated loads even if a crack is present. During the past three decades, most of the efforts with respect to integrity evaluation of steam generator tubes have been focused on limit load solutions but, recently, the applicability of elastic-plastic fracture mechanics was examined cautiously due to its effectiveness. The purpose of this paper is to introduce a testing method to estimate fracture resistance characteristics of steam generator tubes with a through-wall crack. Due to limited thickness and diameter, inevitably, the steam generator tubes themselves were tested instead of standard specimen or alternative ones. Also, a series of three dimensional elastic-plastic finite element analyses were carried out to derive closed-form estimation equations with respect to J-integral and crack extension for direct current potential drop method. Since the effectiveness of $J_{IC}$ as well as J-R curves was proven through comparison with those of standard specimens taken from pipes, it is believed that the proposed scheme can be utilized as an efficient tool for integrity evaluation of cracked steam generator tubes.