• 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.

• Korean Journal of Food Science and Technology
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• v.1 no.1
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• pp.85-89
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• 1969

1. According to the Bergey's manual of determinative bacteriology, the high amylase producing strain A-162 isolated from corn was similar to Bacillus subtilis in the characteristics. 2. The addition of corn powder 30%, milk casein 5% and $CaCO_3$. 5% to wheat bran was excellent as amylase producing media. 3. According to vessel content and quantities of the media, the optimum steaming condition of media was different. Excessive steaming (pressure and time) suppressed the growth of Bacillus subtilis var. A-162. 4. The optimum temperature of amylase produced was about $50^{\circ}C$ and its optimum pH 6.0.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.69-75
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• 2004

DORT and ORIGEN2 code were used for calculation of neutron flux and inventory in reactor pressure vessel(RPV) of Kori unit-1, To calculate neutron flux using DORT code, the reactor was divided into 94 mesh from the center of core to RPV and from 0 to 45 degree along the azimuth. The cross-sections of main nuclides were recalculated using neutron flux in the RPV region. The results showed that 95% of the total activity in RPV came from the nuclides of $^{55}$ Fe, $^{60}$ Co, $^{59}$ Ni and $^{63}$ Ni. And the total activity with cooling of more than 50 years after decommissioning was no more than 0.2% of at the time of shutdown. Considering the weight of RPV is 210 tons, the initial total activity of RPV reached 5.25${\times}$10$^{6}$ GBq. To verify results of ORIGEN2 calculation, comparison between calculated and measured value at RPV of Kori unit-1 was peformed. The comparison results showed a good agreement.

• 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.

• Journal of Hydrogen and New Energy
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• v.13 no.2
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• pp.151-158
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• 2002

The hydorgen storage alloys were produced by melting in arc melting furnace and then solution heat treated at $1,100^{\circ}C$ followed by pulverization. The chemical analysis on the samples showed that the major elements of misch metal(Mm) were La, Ce, Pr and Nd with impurity less than 1wt.%. X-ray diffraction indicated that the structure for these samples were a single phase of hexagonal with $CaCu_5$ type. Compared to the initial particle size $100{\sim}110{\mu}m$, the many fine cracks were found and particle size decreased to $14{\mu}m$ for $MmNi_{4.5}Mn_{0.5}$ after hydriding/dehydring test run. To activate the sample the vessel filled with hydrogen storage alloys was first evacuated for for at $70^{\circ}C$ and then treated for 10.5hr under hydrogen pressure of 20atm for $MmNi_{4.5}Mn_{0.5}$ alloy. The experimental data showed that the hydrogen storage alloy of $MmNi_{4.5}Mn_{0.5}$ had superior adsorption and description properties within a temperature rang of $40^{\circ}C{\sim}80^{\circ}C$ and also they had a good P-C-T curve.

• 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.

• International Journal of Naval Architecture and Ocean Engineering
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• v.6 no.4
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• pp.1130-1147
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• 2014

Large size ships have a very flexible construction resulting in low resonance frequencies of the structural eigen-modes. This feature increases the dynamic response of the structure on short period waves (springing) and on impulsive wave loads (whipping). This dynamic response in its turn increases both the fatigue damage and the ultimate load on the structure; these aspects illustrate the importance of including the dynamic response into the design loads for these ship types. Experiments have been carried out using a segmented scaled model of a container ship in a Seakeeping Basin. This paper describes the development of the model for these experiments; the choice was made to divide the hull into six rigid segments connected with a flexible beam. In order to model the typical feature of the open structure of the containership that the shear center is well below the keel line of the vessel, the beam was built into the model as low as possible. The model was instrumented with accelerometers and rotation rate gyroscopes on each segment, relative wave height meters and pressure gauges in the bow area. The beam was instrumented with strain gauges to measure the internal loads at the position of each of the cuts. Experiments have been carried out in regular waves at different amplitudes for the same wave period and in long crested irregular waves for a matrix of wave heights and periods. The results of the experiments are compared to results of calculations with a linear model based on potential flow theory that includes the effects of the flexural modes. Some of the tests were repeated with additional links between the segments to increase the model rigidity by several orders of magnitude, in order to compare the loads between a rigid and a flexible model.

• 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.

• Nuclear Engineering and Technology
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• v.20 no.4
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• pp.253-264
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• 1988

Our experience of neutron noise analysis in French-type 900 MWe pressurized water reactor (PWR) is presented. Neutron noise analysis is based on the technique of interpreting the signal fluctuations of ex-core detectors caused by core reactivity changes and neutron attenuation due to lateral core motion. It also provides advantages over deterministic dynamic-testing techniques because existing plant instrumentation can be utilized and normal operation of the plant is not disturbed. The data of this paper were obtained in the ULJIN unit 1 reactor during the start-up test period and the statistical descriptors, useful for our purpose, are power spectral density (PSD), coherence function (CF), and phase difference between detectors. It is found that core support barrel (CSB) motions induced by coolant flow forces and pressure pulsations in a reactor vessel were indentified around 8 Hz of frequency.

• Journal of the Institute of Convergence Signal Processing
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• v.9 no.1
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• pp.10-17
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• 2008

The purpose of this thesis is to reduce of error for source location through acoustic emission(AE) signal, generated elastic wave from crack growth to leak for facility diagnosis. Especially, in order to overcome noise from original signal, this paper proposed enhancement of source location by using noise reduction based on wavelet transform. To evaluate actual performance in experiments, Pencil Lead Break is used crack signal source on the aluminum plate and drain valve of air compressor is used as substitute pressure vessel to generate leak signal. In signal processing, wavelet shrinkage and soft threshold are used to discriminate signal source and then source location techniques have been effectively used with group velocity using material property and time difference between sensor using cross correlation. Source location for crack and leak test have some difference, but the result show that improved 30% with a average length within 10.46mm in crack test and improved 2% compare with average filter in leak test when we applied wavelet transform.