• 대한기계학회논문집B
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• 제31권8호
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• pp.663-671
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• 2007

In this work, shape optimization of a wire-wrapped fuel assembly in a liquid metal reactor has been carried out by combining a three-dimensional Reynolds-averaged Navier-Stokes analysis with the radial basis neural network method, a well known surrogate modeling technique for optimization. Sequential Quadratic Programming is used to search the optimal point from the constructed surrogate. Two geometric design variables are selected for the optimization and design space is sampled using Latin Hypercube Sampling. The optimization problem has been defined as a maximization of the objective function, which is as a linear combination of heat transfer and friction loss related terms with a weighing factor. The objective function value is more sensitive to the ratio of the wire spacer diameter to the fuel rod diameter than to the ratio of the wire wrap pitch to the fuel rod diameter. The optimal values of the design variables are obtained by varying the weighting factor.

• 한국수소및신에너지학회논문집
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• 제22권2호
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• pp.206-212
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• 2011

The effect of process parameters on $H_2$ production from water vapor excited by HF ICP has been qualitatively examined for the first time. With the increase of ICP power, characteristics of $H_2$ production from $H_2O$ dissociation in plasma was divided into 3 regions according to both reaction mechanism and energy efficiency. At the edge of region (II) in the range of middle ICP power, energy effective hydrogen production from $H_2O$ plasma can be achieved. Furthermore, within the region (II) power condition, heating of substrate up to $500^{\circ}C$ shows additional increase of 70~80% in $H_2$ production compared to $H_2O$ plasma without substrate heating. This study have shown that combination of optimal plasma power (region II) and wall heating (around $500^{\circ}C$) is one of effective ways for $H_2$ production from $H_2O$.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1251-1260
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• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3478-3487
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• 2022

In this study, we incorporate an anisotropic scattering scheme involving spherical harmonics into the method of characteristics (MOC). The neutron transport solution in a light water reactor can be significantly improved because of the impact of an anisotropic scattering source with the MOC flat source approximation. Several problems are selected to verify the proposed scheme and investigate its effects and accuracy. The MOC anisotropic scattering source is based on the expansion of spherical harmonics with Legendre polynomial functions. The angular flux, scattering source, and cross section are expanded in terms of the surface spherical harmonics. Later, the polynomial is expanded to achieve the odd and even parity of the source components. Ultimately, the MOC angular and scalar fluxes are calculated from a combination of two sources. This paper presents various numerical examples that represent the hot and cold conditions of a reactor core with boron concentration, burnable absorbers, and control rod materials, with and without a reflector or baffle. Moreover, a small critical core problem is considered which involves significant neutron leakage at room temperature. We demonstrate that an anisotropic scattering source significantly improves solution accuracy for the small core high-leakage problem, as well as for practical large core analyses.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1845-1853
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• 2024

The setpoint of the reactor trip system shall be set to consider the measurement uncertainty of the instrument channel and provide a reasonable and sufficient margin between the analytical limit and the trip setpoint. A comparative analysis was conducted to find out an appropriate uncertainty combination method through an example problem. The four methods were evaluated; 1) ISA-67.04.01 method, 2) the GUM95 method, 3) the modified GUM method developed by Fotowicz, and 4) the modified IEC61888 method proposed by authors for the pressure instrument channel presented in ISA-RP67.04.02 example. The appropriateness of each method was validated by comparing it with the result of Monte Carlo simulation. As a result of the evaluation, all methods are appropriate when all measurement uncertainty elements are normally distributed as expected. But ISA-67.04 method and GUM95 method overestimated the channel uncertainty if there is a dominant input element with rectangular distribution among the uncertainty input elements. Modified GUM95 methods developed by Fotowicz and modified IEC61888 method by authors are able to produce almost the same level of channel uncertainty as the Monte Carlo method, even when there is a dominant rectangular distribution among the uncertainty components, without computer-assisted simulations.

• 한국미생물·생명공학회지
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• 제25권1호
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• pp.89-95
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• 1997

The current processer of the textile wastewater treatment are mostly consisted of a combination of a physico-chemical and a biological treatment. The overall efficiency of these processes is, however, assessed to be fairly low. It is even worse during the summer season when temperature of the wastewater rises above 40$\circ $C. Therefore, a feasible process of the textile wastewater treatment which can work efficiently at higher temperatures was investigated in this work. We used a bench scale reactor consisted of one 4 liter anaerobic and one 8 liter aerobic tank, and the thermophilic symbiotic PVA degraders, Pasteruella hemolytica KMG1 and Pseudomonas sp. KMG6 that had been isolated in our laboratory. In the preliminary flask experiments, we observed that the thermophilic symbiotic PVA degraders could not grow in the wastewater substrate. Then, we isolated the mutant strains by acclimating the KMG1and KMG6 strains to the wastewater medium. The mutant symbionts (KMG1-1 and KMG6-1) were isolated through 6 times successive transfers into the fresh wastewater medium after 5 days culture for each. The mutant strains obtained grew well in the mixed medium composed of 75% wastewater and 25% synthetic medium, and supplemented with 0.5% PVA as a sole carbon source. During the culture for 14 days at pH 7.0 and 40$\CIRC $C, the bacteria assimilated about 89% of the added PVA. The symbionts degraded equally well all the PVA substrates of different molecular weight (nd=500~30000). In contrast to the flask experiments, in the reactor system the mutant strains showed very low levels of the PVA and COD removal rates. However, the new reactor system with an additional aerobic tank attained 82% removal rate of COD, 94% of PVA degradation and 71% of color index under the conditions of 5% inoculm on the tank 2, incubation temperature of 40$\circ $C, dissolved oxygen level of 2~3 mg/l and retention time of 30 hours. This result ensures that the process described above could be an efficient and feasible treatment for the PVA contained textile wastewater at higher temperatures.

• 대한기계학회논문집A
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• 제25권4호
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• pp.756-763
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• 2001

In the process of integrity evaluation for nuclear power plant components, a series of fracture mechanics evaluation on surface cracks in reactor pressure vessel(RPV) must be conducted. These fracture mechanics evaluation are based on stress intensity factor, K. However, under pressurized thermal shock(PTS) conditions, the combination of thermal and mechanical stress by steep temperature gradient and internal pressure causes considerably high tensile stress at the inside of RPV wall. Besides, the internal pressure during the normal operation produces high tensile stress at the RPV wall. As a result, cracks on inner surface of RPVs may experience elastic-plastic behavior which can be explained with J-integral. In such a case, however, J-integral may possibly lose its validity due to constraint effect. In this paper, in order to verify the suitability of J-integral, tow dimensional finite element analyses were applied for various surface cracks. A total of 18 crack geometries were analyzed, and $\Omega$ stresses were obtained by comparing resulting HRR stress distribution with corresponding actual stress distributions. In conclusion, HRR stress fields were found to overestimate the actual crack-tip stress field due to constraint effect.

• 대한환경공학회지
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• 제28권10호
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• pp.1055-1064
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• 2006

본 연구소에서는 혐기성 입상슬러지를 충진한 UASB 반응기와 탄소섬유를 충진한 배양기를 조합하여 아주 느린 성장특성을 가진 혐기성 암모늄 산화균의 배양을 시도하였다. 연속배양 180일 이후, 유입질소부하가 0.6 kg $N/m^3-d$였을 때, 평균 질소전환율은 0.54 kg $N/m^3-d$로 나타났다. 검은 혐기성 입상슬러지는 연속배양시간이 지남에 따라 갈색과 붉은 색으로 변화되었으며, anammox 반응기는 붉은색 입상슬러지가 많을수록 높은 활성도를 나타내었다. 따라서, 붉은색 입상슬러지를 채취하여 분자생물학적 방법을 이용하여 미생물 군집구조를 분석하였다. 클로닝 및 계통분류학적 분지도 작성 결과, anammox UASB 반응조의 붉은색 입상슬러지에서 발견된 미생물 종류는 anammox 미생물과 더불어 문단위의 4가지 다른 미생물, Proteobacteria, Acidobacteria, Chlorobi와 Chloroflexi로 나타났다. Anammox UASB 반응조내의 붉은색 입상슬러지에서는 clone의 개수를 기준으로 anammox 미생물이 약 25%가 존재하였고 $\beta$-proteobacteria가 우점하고 있는 양상을 보여주었으며, 본 연구의 클로닝 정보와 AMX368 FISH 탐침자를 이용해 in silico 실험을 수행한 결과 AMX368과 정확히 들어맞는 anammox 미생물 clone 하나와 하나의 염기서열이 변이를 일으킨 11개의 anammox 미생물 clone을 확인할 수 있었다. 사상균 형태의 Chloroflexi는 혐기조건 입상 슬러지의 형성과 관련이 있는 것으로 판단되었다. FISH 수행결과, anammox 미생물은 붉은색 입상 슬러지에 우점하고 있는 것으로 나타났다.