• 한국결정성장학회지
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• 제15권4호
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• pp.135-140
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• 2005

Thermal-fluid analysis was performed to understand the thermal behavior in the horizontal CVD reactor thereby to design a susceptor which has a uniform deposition rate during silicon EPI growing. Four different types of susceptor designs, standard (no hole susceptor), hole $\sharp$1 (240 mm), hole $\sharp$2 (150 mm) and hole $\sharp$3 (60 mm), were simulated by CFD (Computational Fluid Dynamics) tool. Temperature, gas flow, deposition rate and growth rate were calculated and analyzed. The degree of flatness of EPI wafer loaded on the susceptor was computed in terms of silicon growth rate. The simulation results show that the temperature and thermal distribution in the wafer are greatly dependent on inner diameter of hole susceptor and demonstrate that the introduction of hole in the susceptor can degrade wafer flatness. Maximum temperature difference appeared around holes. As the diameter of the hole decreases, flatness of the wafer becomes poor. Among the threes types of susceptors with the hole, optimal design which resulted a good uniform flatness ($5\%$) was obtained when using hole $\sharp$1.

• 한국기계가공학회지
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• 제11권4호
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• pp.160-167
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• 2012

This paper is concerned with a particular fracture mechanics parameter ${\Delta}K_{th}$, known as the 'threshold stress intensity range', or 'fatigue threshold'. This threshold ${\Delta}K_{th}$ constitutes, as it were, a hinge between the notion of crack initiation and the notion of crack growth. It has often been thought that, like the endurance limit, it could be an intrinsic criterion of the material. The study was conducted on a SA-508 pressure vessel steel used in the nuclear power industry. This material exhibits a typical threshold effect in the range of the crack growth rates which were determined; that is, below approximately $da/dN=10^{-6}mm/cycle$, the slope of the da./dN versus ${\Delta}K$ curve is almost vertical. The value of ${\Delta}K_{th}$ was determined at a growth rate of $10^{-7}$ mm/cycle according to the ASTM Standard for threshold testing. The fatigue threshold values are in the range 21 $kg/mm^{3/2}$ to 12 $kg/mm^{3/2}$ depending on the stress ratio effect.

• Nuclear Engineering and Technology
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• 제29권3호
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• pp.218-228
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• 1997

Correlations between the amount of DAW (Dry Active Waste) generated from present Korean PWRs and their operating parameters were analyzed. As the result of multi-variable linear regressions, a model predicting the volume of DAW using the number of shutdowns ( $f_{FS}$ ) and total personnel exposure ( $P_{\varepsilon}$) was derived. Considering one standard error bound, the model could successfully simulate about 8575 of the real data. In order to predict the amount of DAW to be generated from a KNGR another model was derived by taking into account the additional volume reduction by supercompaction system. In addition, the volume of WAW (Wet Active Waste) to be generated from KNGR (Korean Next Generation Reactor) was calculated by considering conceptual design data and replacement effect of radwaste evaporator with selective ion exchangers. Finally, total volume of SRW (Solid Radioactive Waste) to be generated from KNGR was predicted by inserting design goal values of $f_{FS}$ and $P_{\varepsilon}$ into the model. The result showed that the expected amount of SRW to be generated from KNGR would be in the range of 33~44㎥. $y^{-1}$ . It was proved that the value would meet the operational target of KNGR proposed by KEPCO, that is, 50㎥. $y^{-1}$ .

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(4)
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• pp.182-187
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• 1996

Deformation characteristics of miniature plate tensile specimens have been studied to develop the thickness requirement and a correlation to estimate the mechanical properties of bulk material from miniature specimen data. The material used was a SA 508 C1.3 reactor pressure vessel steel and the thicknesses of miniature tensile specimens varied from ().12 m to 2 mm. The effects of thickness on the tensile deformation properties such as strength, ductility, and necking characteristics were analyzed. The yield and ultimate tensile strengths were independent of specimen thickness when the thickness was larger than about 0.2 mm. The uniform and total elongations decreased as the specimen thickness decreased. It was also observed that the uniform strain component in the width direction decreased with decrease in the specimen thickness, however, that in the thickness direction was rather constant in total thickness range studied. Based on this observation and a relationship between the necking angle and the ratio between strain components, a correlation between the uniform elongations of miniature specimen and standard specimen was derived. The uniform elongations calculated by this new correlation agreed well with the measured values.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1298-1303
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• 2004

The characteristics of soot near the soot inception point for an ethene-air flame was carried out in a WSR (well-stirred reactor). The new sampling tool like the temperature controlled filter system was introduced to minimize the condensation during sampling. The new analysis tools applied include the real time size distribution analysis with the Nano-DMA, particle size by transmission electron microscopy, C/H analysis, g filter analysis, and thermogravimetric analysis using both non-oxidative and oxidative pyrolysis. The WSR can generate young soot particles that can be collected and examined to gain insight into inception. For the current conditions, soot does not form for ${\Phi}=1.9$, inception occurs at or before ${\Phi}=2.0$, and inception combined with soot surface growth and/or coagulation occurs for ${\Phi=2.1}$. The filter samples for ${\Phi}$=1.9 are composed of volatile compounds that evolve at relatively low temperatures when heated in the presence or absence of $O_2$. The samples collected from the WSR at ${\Phi}=2.0$ and ${\Phi}=2.1$ are precursor-like in morphology and size. They have higher C/H ratios and lower organic percentages than precursor particles, but they are clearly not fully carbonized soot. The WSR PAH distribution is similar to that in young soot from inverse flames.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.690-703
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• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.979-988
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• 2017

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2788-2802
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• 2021

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW_th metal-fueled SFR (MET-1000) and a 3600 MW_th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k_eff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k_eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.

• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3202-3205
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• 2023

The k₀-standardization neutron activation analysis method has successfully determined the mass fraction of elements of interest using around a hundred analytical radionuclides. However, several very short-lived nuclides with half-life less than 100 s have not been used at Dalat research reactor. One of the reasons is that the values of k₀-factors of these nuclides are significantly different. Therefore, this work focused on re-determination and evaluation of k₀-factors of very short-lived nuclides ¹¹⁰Ag (T_1/2 = 24 s) and ^46mSc (T_1/2 = 18.75 s). The results of determination of the short-lived nuclides revealed that k₀-factor of ¹¹⁰Ag is significantly difference between the existing data and the obtained results in this work. The evaluation of the k₀-factors was done by using the obtained results for application of k₀-NAA for NIST-1566b and NIST-2711A standard reference materials.