• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1222-1230
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• 2020

The corrosion rates of the reactor pressure vessel materials of SA508 Grade 3 were measured using a weight loss method in aerated boric acid solutions to simulate the evaporation of leaked PWR primary water in an ambient environment. The corrosion behavior and products were examined using X-ray diffraction and electron microscopy. SA508 showed typical general corrosion characteristics. The corrosion rate increased steadily as the boron concentration was increased. As the immersion time elapsed, the corrosion rate slowly or rapidly decreased according to the oxidation reaction of iron. The corrosion rate showed a complicated pattern depending on the temperature; it increased gradually and then rapidly decreased again when reaching a certain transition temperature. The corrosion products of SA508 were found to be FeO(OH), Fe₂O₃, and Fe₃O₄. As the boron concentration decreased and the temperature was increased, the formation of Fe₃O₄ was more favorable as compared to the formation of FeO(OH) and Fe₂O₃. Consequently, the changes of the corrosion rate and behavior were closely related to the oxidation reaction of iron on the surface. The corrosive damage to SA508 appears to be most severe when the oxidation reaction is such that Fe₂O₃ forms as a corrosion product.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.5
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• pp.451-457
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• 2010

In the present study, the influences of grains and precipitates of microstructural evolution on the ultrasonic nonlinearity have been experimentally investigated. The prior-austenite grain and precipitate size are controlled by the variation in austenitizing and tempering conditions in reactor pressure vessel materials of nuclear power plant, SA508 Gr.3 low alloys. The ultrasonic nonlinearity was found to have strong correlations with grains and precipitates since the ultrasonic nonlinear parameter $\beta$ shows decrease trend with coarsening of grains and precipitates. Although the prior-austenite grain size increased, the $\beta$ changed little due to the effects of subgrains, packets and laths. For the preciptate effects, the $\beta$ decreased sharply due to decrease in $Mo_2C$ causing the coherency stain in addition to the precipitate size. The results in this study may provide a potential for characterizing the microstructural evolution, grains and precipitates, by measuring the ultrasonic nonlinearity.

• Journal of the Korean Society for Nondestructive Testing
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• v.20 no.5
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• pp.390-396
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• 2000

The dynamic elastic constants of the simulated weld HAZ (heat-affected zone) of SA 508 Class 3 reactor pressure vessel (RPV) steel were investigated by resonant ultrasound spectroscopy (RUS). The resonance frequencies of rectangular parallelepiped samples woe calculated from the initial estimates of elastic stiffness $c_{11},\;c_{12}\;and\;c_{44}$ with an assumption of isotropic property, dimension and density. Through the comparison of calculated resonant frequencies with the measured resonant frequencies by RUS, very accurate elastic constants of SA 508 Class 3 steel were determined by iteration and convergence processes. Clear differences of Youngs modulus and shear modulus were shown from samples with different thermal cycles and microstructures. Youngs modulus and shear modulus of samples with fine-grained bainite were higher than those with coarse-grained tempered martensite. This tendency was confirmed from other results such as micro-hardness test.

• Journal of Welding and Joining
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• v.2 no.1
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• pp.4-17
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• 1984

Overlaid weld metals of austenitic stainless steel in a pressure vessel of power reactor are usually post-weld heated for a long period of time after welding. The PWHT is considered as a kind of sensitizing and it is important to check the soundness of the weld metal after PWHT, especially about the precipitation of carbides. The purpose of this report is to obtain information on the relation between the change of microstructure and Post-Weld Heat Treatment in the overlaid weld metals. Metallurgical aspects of the problem on austenitic stainless steel heated at $625^{\circ}C$, $670^{\circ}C$, $720^{\circ}C$ and $760^{\circ}C$ for 3, 10, 30, 100 and 300 hours have been investigated by means of optical-micrography, micro-hardness measurement, scanning electron microscope and electron-probe micro analysis. From the results obtained, the following conclusions are drawn; 1) The PWHT above $625^{\circ}C$ for a long time causes a diffusion of carbon atoms from low alloy steel into stainless steel, and consequently carbon is highly concentrated at the boundary layer of stainless steel. 2) C in ferritic steel migrated to austenitic steel and carbides precipitated in austenitic steel along fusion line. At higher temperatures, the ferrite grains coarsened in the decarburized zone. 3) In the change of microstructure of stainless steel overlaid weld metal, the width of carbides precipitated zone and decarburized zone increased with increase of PWHT temperature and time. 4) At about $625^{\circ}C$ to $760^{\circ}C$, chromium carbides, mainly $M_{23} C_6$, precipitate very closely in the carburized layer with remarkable hardening. 5) Precipitation of delta ferrite from molten weld metal depends on solidification phenomenon. There was a small of ferrite near the bond in which the local solidification time was short, comparing with after parts of weld metal. Shape and amount of ferrite were not changed by Post-Weld Heat Treatment after solidification.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.340-350
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• 2015

This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the overconservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Proceedings of the KIEE Conference
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• 1995.07b
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• pp.761-765
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• 1995

The old Timing Controller for Control Element Drive Mechanism (CEDM) is designed as an open loop control system because it is difficult to mount sensors within the Control Element Drive Mechanism(CEDM) which is operating under the pressure boundary of the reactor vessel. In this work new method which can be used to detect the CEDM operational conditions without mounting sensors within the CEDM housing is developed in order to resolve problems of the old Timing Controller. By using the developed new method, the new Timing Controller for the CEDM is designed as a closed loop controller which has features of the control rod drop prevention, fine position control and the coil life time extension. The algorithm developed under closed loop control concept resolves most problems occurred in the old Timing Controller and improves the performance and reliability of the system. During designing and testing of the Timing Controller algorithm, the real time CEDM simulator developed here was used. And all functions of the developed algorithm were verified using CEDM simulator with the real data collected from the site. The results show that the Timing Controller performs its intended functions properly.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.246-258
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• 2016

The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH) and finite element method (FEM) coupling method is used to numerically simulate the fluid and structure interaction (FSI) between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs) in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.760-768
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• 2017

The transition temperature shift (TTS) of the reactor pressure vessel materials is an important factor that determines the lifetime of a nuclear power plant. The prediction of the TTS at the end of a plant's lifespan is calculated based on the equation of Regulatory Guide 1.99 revision 2 (RG1.99/2) from the US. The fluence factor in the equation was expressed as a power function, and the exponent value was determined by the early surveillance data in the US. Recently, an advanced approach to estimate the TTS was proposed in various countries for nuclear power plants, and Korea is considering the development of a new TTS model. In this study, the TTS trend of the Korean surveillance test results was analyzed using a nonlinear regression model and a mixed-effect model based on the power function. The nonlinear regression model yielded a similar exponent as the power function in the fluence compared with RG1.99/2. The mixed-effect model had a higher value of the exponent and showed superior goodness of fit compared with the nonlinear regression model. Compared with RG1.99/2 and RG1.99/3, the mixed-effect model provided a more accurate prediction of the TTS.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.9
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• pp.1047-1052
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• 2012

Primary water stress corrosion cracking (PWSCC) instances have been reported in the Alloy 600 reactor pressure vessel head penetration nozzle and the Alloy 82/182 dissimilar metal butt weld nozzle in several PWRs. Therefore, in-service inspection programs have been adopted worldwide to prevent failure at the weld region. If a PWSCC is observed at the dissimilar metal weld region during inspection, its structural integrity should be evaluated; however, this requires considerable time and effort, and this might lead to a decrease in the plant utilization coefficient. To prevent this, KHNP-CRI have established integrity assessment criteria and developed a computer program for the fast evaluation and judgment of PWSCC. In this paper, the results and current status of the same are presented. Through this study, criteria for the structural integrity evaluation of PWSCC have been established, and a computer program has been developed to realize technical means for the evaluation of PWSCC structural integrity.