• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2977-2981
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• 2021

Corrosion fatigue crack growth (FCG) behavior of 316LN stainless steel was investigated in high-temperature pressurized water at different temperatures, load ratios (R = K_max/K_min) and rise times (t_R). The environmental assisted effect on FCG rate was observed when both the R and t_R exceeded their critical values. The FCG rate showed a linear relation with stress intensity factor range (ΔK) in double logarithmic coordinate. The environmental assisted effect on FCG rate depended on the ΔK and quantitative relations were proposed. Possible mechanisms of environmental assisted FCG rate under different testing conditions are also discussed.

• Journal of the Korean Ceramic Society
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• v.30 no.3
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• pp.189-198
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• 1993

A new pressurized sol-gel coating technique forming membrane layers inside pores of the porous support by the simple operation has been developed. Crack-free and reproducible nanoparticulate silica membranes supported on the porous $\alpha$-alumina tube are synthesized by pressurized coating at 600kPa for 2hr. The pore radius and N2 gas permiability at the room temperature of silica membrane layers are 8$\AA$ and 7.0$\times$10-7mol/$m^2$.s.Pa, respectively. The mechanism of N2 gas transfer through synthesized membrane layers is the perfect Knudeen flow, and the thermal stability of the silica composite membranes is excellent upto 40$0^{\circ}C$.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.727-728
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• 2008

The objective of this study is to evaluate the integrity for a reactor pressure vessel under the pressurized thermal shock by applying the probability fracture mechanics. A semi-elliptical axial crack is assumed to be in the beltline region of the reactor pressure vessel. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT. The probabilistic integrity analysis was performed using the Monte Carlo simulation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.39-47
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• 2002

Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock(PTS). The PTS event means an event or transient in pressurized water reactors(PWRs) causing severe overcooling(thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored.

• Journal of the Korean Society of Safety
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• v.16 no.4
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• pp.29-38
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• 2001

The purpose of this paper is to evaluate the structural integrity of a reactor pressure vessel subjected to the pressurized thermal shock(PTS) during the transient events, such as main steam line break(MSLB) and small break loss of coolant accident(SBLOCA). For postulated surface or subsurface cracks, variation curves of stress intensity factor are obtained by using the three different methods, including ASME section XI code anlysis, the finite element alternating method and the finite element method. From the stress intensity factor curves, the maximum allowable nil-ductility transition temperatures(RT/NDT/) are determined by the tangent criterion and the maximum criterion for various crack configurations and two initial transient events. As a result of the analysis, it is noted that axial cracks have smaller maximum allowable RT$_{NDT}$ values than same-sized circumferential cracks for both the transient events in the case of the tangent criterion. Axial cracks have smaller RT$_{NDT}$ values than same-sized circumferential cracks for MSLB and circumferential cracks have smaller values than axial cracks for SBLOCA in the case of the maximum criterion.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.10
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• pp.1612-1620
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• 2004

This study presents methods to estimate elastic-plastic crack opening displacement (COD) fur circumferential through-wall cracked pipes for the Leak-Before-Break (LBB) analysis of pressurized piping. Proposed methods are based not only on the GE/EPRI approach but also on the reference stress approach. For each approach, two different estimation schemes are given, one for the case when full stress-strain data are available and the other fur the case when only yield and ultimate tensile strengths are available. For the GE/EPRI approach a robust way of determining the Ramberg-Osgood (R-O) parameters is proposed, not only fur the case when detailed information on full stress-strain data is available but also for the case when only yield and ultimate tensile strengths are available. The COD estimates according to the GE/EPRI approach, using the R-O parameters determined from the proposed R-O fitting procedures, generally compare well with the published pipe test data. For the reference stress approach, the COD estimates according to the method based on both full stress-strain data and limited tensile properties are in good agreement with pipe test data. In conclusion, experimental validation given in the present study provides sufficient confidence in the use of the proposed method to practical LBB analyses even though when information on material's tensile properties is limited.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.11
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• pp.2219-2227
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• 2002

In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature ($(RT_{NDT})$) was investigated on the basis of finite element analyses.

• Transactions of the Korean Society of Mechanical Engineers
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• v.17 no.10
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• pp.2398-2406
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• 1993

Abdi's numerical method(ref.13) for representing a stress singularity by shifting the mid-side nodes of isoparametric elements is reviewed. A simple technique to obtain the optimal position of the mid-side nodes in quadratic isoparametric finite element is presented. From this technique we can directly obtain the position of the side-nodes adjacent to the crack tip. It is also observed that the present technique provides good accuracy for the expression of the opening displacement and the determination of the mid-side nodes for more wide range of material properties than that obtained by Abdicant the finite element method is applied to determine stress intensity factors for pressurized crack perpendicular to and terminating at the interface of two bonded dissimilar materials. A proper definition for stress intensity factors of a crack perpendicular to bimaterial interface is provided. It is based upon a near-tip displacement solutions on the crack surface for interface crack between two dissimilar materials. Numerical testing is carried out with the eight-node and six-node elements. The results obtained are compared with the previous solutions.

• Corrosion Science and Technology
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• v.11 no.4
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• pp.141-150
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• 2012

Stress corrosion cracks in Alloy 600 compact tension specimens tested at $325^{\circ}C$ in a simulated primary water environment of pressurized water reactor were analyzed by analytical transmission electron microscopy and secondary ion mass spectroscopy (SIMS). From a fine-probe chemical analysis, oxygen was found on the grain boundary just ahead of the crack tip, and chromium oxides were precipitated on the crack tip and the grain boundary attacked by the oxygen diffusion, leaving a Cr/Fe depletion (or Ni enrichment) zone. The oxide layer inside the crack was revealed to consist of a double (inner and outer) layer. Chromium oxides existed in the inner layer, with NiO and (Ni,Cr) spinels in the outer layer. From the nano-SIMS analysis, oxygen was detected at the locations of intergranular chromium carbides ahead of the crack tip, which means that oxygen diffused into the grain boundary and oxidized the surfaces of the chromium carbides. The intergranular chromium carbide blunted the crack tip, thereby suppressing the crack propagation.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1060-1068
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• 2019

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.