• Journal of the Computational Structural Engineering Institute of Korea
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• v.37 no.4
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• pp.267-274
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• 2024

Hydrogen embrittlement fracture poses a challenge in ensuring the structural integrity of materials exposed to hydrogen-rich environments. This study advances our comprehension of hydrogen-induced fracture through an integrated numerical modeling approach. In addition, it employs a ductile fracture model named the Gurson-cohesive model (GCM) and hydrogen diffusion analysis. GCM is employed as a fracture model that combines the Gurson model to illustrate the continuum damage evolution and the cohesive zone model to describe crack surface discontinuity and softening behavior. Moreover, porosity and stress triaxiality are considered as crack initiation criteria . A hydrogen diffusion analysis is also integrated with the GCM to account for hydrogen enhanced decohesion (HEDE) mechanisms and their subsequent impacts on crack initiation and propagation. This framework considers the influence of hydrogen on the softening behavior of the traction-separation relationship on the discontinuous crack surface. Parametric studies explore the sensitivity to diffusion properties and hydrogen-induced fracture properties. By combining numerical models of hydrogen diffusion and the ductile fracture model, this study provides an understanding of hydrogen-induced fracture and thereby contributes significantly to the ongoing efforts to design materials that are resilient to hydrogen embrittlement in practical engineering applications.

• Transactions of Materials Processing
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• v.26 no.4
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• pp.210-215
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• 2017

In this paper, utilizing the theory of ductile fracture a chevron crack in a 4-stage open cold extrusion process is predicted by the finite element methods and then compared with previous experiments. The normalized Cockcroft-Latham damage model is employed and the material is identified using a tensile test based material identification technique that gives fracture information as well as flow stress at large strain. A large difference between the predicted cracks and actual experiments is observed, specifically narrower width and greater maximum height of the crack. This reveals the limitation of this approach based on the conventional theory of ductile fracture. Based on the observations and the related criticisms, a new approach for predicting the chevron crack is proposed, suggesting that either the critical damage should not be a fixed material constant, or that the conventional fracture theory should be considered with the effects of embrittlement due to accumulated plastic deformation while the duration of crack generation and plastic deformation should be reduced.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.6 s.177
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• pp.1398-1407
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• 2000

The structural steels of power plant show the decrease of mechanical properties due to degradation such as temper embrittlement, creep damage and softening during long-term operation at high temper ature. The typical causes of material degradation damage are the creation and coarsening of carbides(M23C6, M6C) and the segregation of impurities(P, Sb and Sn) to grain boundary. It is also well known that material degradation induces the cleavage fracture and increases the ductile-brittle transition temperature of steels. So, it is very important to evaluate degradation damage to secure the reliable and efficient service condition and to prevent brittle failure in service. However, it would not be appropriate to sample a large test piece from in-service components. Therefore, it is necessary to develop a couple of new approaches to the non-destructive estimation technique which may be applicable to assessing the material degradation of the components with not to influence their essential strength. The purpose of this study is to propose and establish a new electrochemical technique for non-destructive evaluation of material degradation damage for Cr-Mo steels which is widely used in the high temperature structural components. And the electrochemical anodic polarization test results are compared with those of semi-nondestructive SP test.

• Journal of Hydrogen and New Energy
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• v.31 no.3
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• pp.284-292
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• 2020

In this study, hydrogen charging was conducted for API X70 steel by the electro-chemical hydrogen charging method. Right after hydrogen was diffused from the specimen surface to the inside of the X70, the small punch tests and hydrogen concentration analysis was conducted within 5 minutes. Hydrogen was analyzed by melting the whole specimen and detect the gas after melting. Mechanical properties were measured by the small punch (SP) testing. Fracture surface and specimen surface were observed using scanning electron microscope. Three tests were repeated for study sensitivity of the SP test results under a same charging condition. It was observed that the variation of the maximum load, SP displacement at failure, hydrogen concentration as the charging period was not much in the case of X70 as the other steel such as Inconel. It can be argued that X70 base metal may have high hydrogen damage resistance and hydrogen diffusion in the base metal would not cause much embrittlement. Limitations of the SP test with 0.5 mm thickness for hydrogen damage test for X70 were discussed.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.49-54
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• 2000

Cr-Mo-V steel is widely used as a material for the turbine structural component in fossil power plants. It is well known that this material shows the various material degradation phenomenons such as temper embrittlement, carbide coarsening. and softening etc. or ins to the severe operation conditions as high temperature and high pressure. These deteriorative factors cause tile change of mechanical properties as reduction of fracture toughness. Therefor it is necessary to evaluate tile extent of degradation damage for Cr-Mo-V steel in life assessment of turbine structural components. In this paper. the electrochemical potentiokinetic reactivation(EPR) test in $50wt%-Ca(NO_3)_2$ solution is performed to develop the newly technique for degradation damage evaluation of Cr-Mo-V steel. The results obtained from the EPR test are compared with those in small punch(SP) tests recommended by semi-nondestructive testing method using miniaturized specimen. The evaluation parameters used in EPR test are tile reactivation current density$(I_R)$ and charge$(Q_{RC})$ reactivation rate$(I_R/I_{Crit},\;Q_R/Q_{Crit})$. The results suggest that $I_R/I_{Crit}$ in these parameters shows a good correlation with SP test results.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.113-121
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• 2014

In this work, an integrated model including molecular dynamics and chemical rate theory was implemented to calculate the growth of point defect clusters(PDC) and copper-rich precipitates(CRP) which could change the mechanical properties of reactor pressure vessel(RPV) steels in a nuclear power plant. A number of time-dependent differential equations were established and numerically integrated to estimate the evolution of irradiation defects. The calculation showed that the concentration of the vacancies was higher than that of the self-interstitial atoms. The higher concentration of vacancies induced a formation of the CRPs in the later stage. The size of the CRPs was used to estimate the mechanical property changes in RPV steels, as is the same case with the PDCs. The calculation results were compared with the measured values of yield strength change and Charpy V-notch transition temperature shift, which were obtained from the surveillance test data of Korean light water reactors(LWRs). The estimated values were in fair agreement with the experimental results in spite of the uncertainty of the modeling parameters.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.2
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• pp.187-193
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• 2015

The STS310S tube has excellent heat transfer ability and is widely used as the material for heat transfer tubes in heat exchange devices. Mixtures of gas and water flow inside the tube whereas hot flame flows outside it. In this environment, the material of the tube may undergo embrittlement, which can cause leakage. Cracks can propagate from the inside of the tube to its outside and result in brittle fracture. This study identified the cause of brittle fracture in the STS310S tube through experiments and discussion, and proposed solutions to prevent fracture.

• Journal of the Korean Magnetics Society
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• v.8 no.5
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• pp.249-254
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• 1998

In relation to the application of magnetic method to the evaluation of irradiation damage (embrittlement) changes in the magnetic parameters(hysteresis loop and Barkhausen noise) and Vickers microhardness due to neutron irradiation and heat treatment were measured and compared. In the case of irradiation $(2.3{\times}10^{19}\;n/cm^2,\; E{\ge}1\;Mev,\; 288{\circ}C)$ hysteresis loop measurements show that susceptibility decreases as coercivity increase. Saturation magnetization do not show any change. Barkhausen noise amplitude and Barkhausen noise energy have decreased while Vickers microhardness has increased. For isothermally heat treated condition of irradiated specimen at 470 $^{\circ}C$ and 540 $^{\circ}C$, Barkhausen noise energy has increased while Vickers microhardness has decreased. Results of BNE and Vickers microhardness are reversed to the results on irradiated condition. All these consistent changes in magnetic parameter and Vickers microhardness measurement, which are thought to be resulted from the interaction between irradiation-induced defects and dislocation, and magnetic domain, respectively, show a possibility that magnetic measurement may be used to the evaluation of material degradation and recovery due to neutron irradiation and heat treatment, respectively, if a relevant large database in prepared.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1887-1894
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• 2024

Fatigue strength of the structural materials of lead-cooled fast reactors (LFRs) and accelerator-driven systems (ADS) may be degraded in liquid metal (Lead or lead-bismuth eutectic (LBE)) environments. The fatigue crack growth (FCG) data of structural materials in liquid LBE are necessary for damage tolerance design, safety assessment and life management of key equipment. A novel monitoring system for fatigue crack length was designed on the compliance method and the monitor technology of crack opening displacement (COD) of CT specimens by the linear variable differential transformers (LVDT) system. It can be used to predict the crack length by monitoring the COD of CT specimens in harsh high-temperature liquid LBE using a LVDT system. The prediction accuracy of this system was verified by FCG experiments in room temperature air and liquid LBE at 150, 250 and 350 ℃. The first results obtained in the FCG test for T91 steel in liquid LBE at 350 ℃ are presented.