• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Proceedings of the Korean Vacuum Society Conference
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• 2000.02a
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• pp.162-162
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• 2000

We have studied magnetic properties of Co/Pt multilayered films which have attracted great interest as high-density magneto-optical (MO) recording media due to their good MO properties. For this study, [Pt(45 )/Co(35 )]$\times$8 films were deposited with a Pt buffer layer of 60 on Si(100) substrate by alternating electron-beam evaporation in a high vacuum and were ion beam mixed by using 80keV Ar+ at 25$0^{\circ}C$. Especially, an external magnetic field was added to help changing magnetic property during ion beam mixing (IBM). The intermixing of Co and Pt layers after IBM was confirmed with Rutherford Backscattering Spectroscopy (RBS) and Transmission Electron Microscopy (TEM). The MO property of the film was measured with magneto-optical Kerr spectrometer and the change of magnetic easy axis in the film plane was observed from Ker loop data. This anomalous result might be correlated with the change of atomic structure due to the intermixing effect.

• Journal of Aerospace System Engineering
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• v.14 no.2
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• pp.50-56
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• 2020

In this study, we investigated whether long short-term memory (LSTM) can be used in the future to predict F10.7 index data; the F10.7 index is a space environment factor affecting atomic oxygen erosion. Based on this, we compared the prediction performances of LSTM, the Autoregressive integrated moving average (ARIMA) model (which is a traditional statistical prediction model), and the similar pattern searching method used for long-term prediction. The LSTM model yielded superior results compared to the other techniques in the prediction period starting from the max/min points, but presented inferior results in the prediction period including the inflection points. It was found that efficient learning was not achieved, owing to the lack of currently available learning data in the prediction period including the maximum points. To overcome this, we proposed a method to increase the size of the learning samples using the sunspot data and to upgrade the LSTM model.

• Journal of Mechanical Science and Technology
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• v.20 no.3
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• pp.335-344
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• 2006

A comprehensive review of through thickness transverse residual stress distributions in a range of as-welded and mechanically bent components made up of a range of steels has been carried out, and simplified generic transverse residual stress profiles for a plate and pipe components have been proposed. The geometries consisted of welded pipe butt joints, T-plate joints, tubular T-joints, tubular Y-joints and a pipe on plate joints as well as cold bent tubes and pipes. The collected data covered a range of engineering steels including ferritic, austenitic, C-Mn and Cr-Mo steels. Measured residual stress data, normalised with respect to the parent material yield stress, has shown a good linear correlation versus the normalised depth of the region containing the residual stress resulting from the welding or cold-bending process. The proposed simplified generic residual stress profiles based on the mean statistical linear fit of all the data provides a reasonably conservative prediction of the stress intensity factors. Whereas the profiles for the assessment procedures are fixed and case specific, the simple bilinear profiles for the residual stresses obtained by shifting the mean and bending stress from the mean regression line have been proposed and validated.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.201-210
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• 1997

A thermal conductivity correlation has been proposed which can be applied to high turnup fuel by considering both of thermal conductivity with turnup across fuel pellet and additional degradation at pellet rim due to very high porosity. In addition, a correlation has been developed that can estimate the porosity of rim region as a function of rim burnup under the assumptions that all the produced fission gases are retained in the in porosity and threshold pellet average burnup required for the formation of rim region is 40 MWD/㎏U. Rim width is correlated to rim burnup using measured data. For the RISO experimental data obtained at pellet average turnup of 43.5 MWD/㎏U for three linear heat generation rates of 30, 35 and 40 ㎾/m, radial temperature distributions ore calculated using the present correlation and compared with the measured ones. This comparison shows that the present correlation gives the best agreement with the measured data when it is combined with the HALDEN's correlation for thermal conductivity considering its degradation with burnup. Another comparison with the HALDEN's measured fuel centerline temperature as a function of burnup at 25 ㎾/m up to about 44 MWD/㎾U also suggest that the present correlation yields the best agreement when it is combined with the HALDEN's thermal conductivity.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.191-195
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• 2016

Background: The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Materials and Methods: Validation calculations with recent nuclear data evaluations ENDF/BV-II. 0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. Results and Discussion: It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and ${chi}^2$ values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the $k_{eff}$ values. Conclusion: It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.543-550
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• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• Journal of Radiation Industry
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• v.6 no.3
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• pp.231-237
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• 2012

This paper introduces a gamma-ray measurement system for a transportable tomography which is applicable for an industrial process diagnosis. The gamma-ray measurement system consists of pulse mode operating 72 channel CsI detectors, main AMP-pulse shaper, single channel analyzer, counter and control PC. The CsI crystal is coupled with a PIN diode which is connected to an amplifier and pulse shaper. For a compact design, the amplifier and pulse shaping circuit are included in a single package. 36 sets of CsI detectors are connected to a multi-channel counter through single channel analyzers. A computer controls and collects data from two multi-channel counters. This configuration results in 72 channel counting system in total. The CT rotator and radiation measurement system are controlled by a PC with LabVIEW program. Tomographic data were measured for a phantom by the measurement system and transportable gamma-ray CT. From the experimental data image reconstructions were performed by ML-EM algorithm. The result showed that the CsI detector system can be a suitable component for transportable gamma-ray CT system.

• Korean Journal of Metals and Materials
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• v.47 no.10
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• pp.613-620
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• 2009

The Kachanov-Rabotnov (K-R) creep model was proposed to accurately model the long-term creep curves above $10^5$ hours of Alloy 617. To this end, a series of creep data was obtained from creep tests conducted under different stress levels at $950^{\circ}C$. Using these data, the creep constants used in the K-R model and the modified K-R model were determined by a nonlinear least square fitting (NLSF) method, respectively. The K-R model yielded poor correspondence with the experimental curves, but the modified K-R model provided good agreement with the curves. Log-log plots of ${\varepsilon}^{\ast}$-stress and ${\varepsilon}^{\ast}$-time to rupture showed good linear relationships. Constants in the modified K-R model were obtained as ${\lambda}$=2.78, and $k=1.24$, and they showed behavior close to stress independency. Using these constants, long-term creep curves above $10^5$ hours obtained from short-term creep data can be modeled by implementing the modified K-R model.