• 한국세라믹학회지
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• 제43권5호
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• pp.315-319
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• 2006

There have been some studies on the use of coal waste as a raw material for clay bricks due to the lack of naturally producing minerals. It can help resolving the problems of pollution, forest conservation and flood control by utilizing coal waste. However, high content of carbon materials usually leads to the black core in clay bricks after firing process, and diminishes the mechanical and aesthetical properties of clay brick. In this study, the effect of firing process is investigated for the removal of black core in clay bricks with carbon content. The removal kinetics of black core are also compared and investigated with the firing schedule and black core removal.

• 자원리싸이클링
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• 제23권6호
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• pp.47-53
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• 2014

런던협약으로 인하여 하수 슬러지 및 유기성 폐기물의 해양투기가 전면 금지되어 이의 효과적인 처리 및 에너지 전환 기술에 대한 요구가 증대되고 있다. 하수 슬러지의 빠른 감량과 에너지화가 가능한 열적 에너지 전환 기술의 적용을 위해서는 하수 슬러지의 열분해 및 연소 특성에 대한 기본적인 kinetics 자료가 필수적이다. 본 연구에서는 열중량 분석기(thermogravimetric analyzer, thermobalance)를 이용하여 하수 슬러지의 열분해 및 연소 kinetics를 도출하였다. 열분해의 경우 총 세 단계의 반응이 일어나 각각에 대하여 subtraction method에 의하여 activation energy와 빈도 인자를 도출하였다. 촤 연소의 경우 반응 kinetics 해석은 기체-고체 화학반응의 세 가지 모델이 이용되었고 shrinking core model이 연소 특성을 가장 잘 나타내어 이 모델을 기준으로 activation energy와 빈도 인자를 도출하고 산소농도 영향을 살펴보았다.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• 시스템엔지니어링학술지
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• 제19권2호
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Bulletin of the Korean Chemical Society
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• 제34권1호
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• pp.33-34
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• 2013

• 시스템엔지니어링학술지
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• 제18권2호
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• pp.58-74
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• 2022

Deterministic accident analysis plays a central role in the nuclear power plant (NPP) safety evaluation and licensing process. Traditionally the conservative approach opted for the point kinetics model, expressing the reactor core parameters in the form of reactivity and power tables. However, with the current advances in computational power, high fidelity multi-physics simulations using real-time code coupling, can provide more detailed core behavior and hence more realistic plant's response. This is particularly relevant for transients where the core is undergoing reactivity anomalies and uneven power distributions with strong feedback mechanisms, such as reactivity initiated accidents (RIAs). This work addresses a RIA, specifically a control element assembly (CEA) withdrawal at power, using the multi-physics analysis tool RELAP5/MOD 3.4/3DKIN. The thermal-hydraulics (TH) code, RELAP5, is internally coupled with the nodal kinetics (NK) code, 3DKIN, and both codes exchange relevant data to model the nuclear power plant (NPP) response as the CEA is withdrawn from the core. The coupled model is more representative of the complex interactions between the thermal-hydraulics and neutronics; therefore the results obtained using a multi-physics simulation provide a larger safety margin and hence more operational flexibility compared to those of the point kinetics model reported in the safety analysis report for APR1400. The systems engineering approach is used to guide the development of the work ensuring a systematic and more efficient execution.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Journal of Microbiology and Biotechnology
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• 제15권6호
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• pp.1330-1336
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• 2005

Polyhydroxyalkanoic acid (PHA) inclusion bodies were analyzed in situ by $^{13}C$-nuclear magnetic resonance ($^{13}C$-NMR) spectroscopy. The PHA inclusion bodies studied were composed of poly(3-hydroxybutyrate) or poly(3hydroxybutyrate-co-4-hydroxybutyrate), which was accumulated in Hydrogenophaga pseudoflava, and medium-chain-length PHA (MCL-PHA), which was accumulated in Pseudomonas fluorescens BM07 from octanoic acid or 11-phenoxyundecanoic acid (11-POU). The quantification of the $^{13}C$-NMR signals was conducted against a standard compound, sodium 2,2-dimethyl-2-silapentane-5-sulfonate (DSS). The chemical shift values for the in vivo NMR spectral peaks agreed well with those for the corresponding purified PHA polymers. The intracellular degradation of the PHA inclusions by intracellular PHA depolymerase(s) was monitored by in vivo NMR spectroscopy and analyzed in terms of first-order reaction kinetics. The H. pseudoflava cells were washed for the degradation experiment, transferred to a degradation medium without a carbon source, but containing 1.0 g/l ammonium sulfate, and cultivated at $35^{\circ}C$ for 72 h. The in vivo NMR spectra were obtained at $70^{\circ}C$ for the short-chain-length PHA cells whereas the spectra for the aliphatic and aromatic MCL-PHA cells were obtained at $50^{\circ}C\;and\;80^{\circ}C$, respectively. For the H. pseudoflava cells, the in vivo NMR kinetics analysis of the PHA degradation resulted in a first-order degradation rate constant of 0.075/h ($r^{2}$=0.94) for the initial 24 h of degradation, which was close to the 0.050/h determined when using a gas chromatographic analysis of chloroform extracts of sulfuric acid/methanol reaction mixtures of dried whole cells. Accordingly, it is suggested that in vivo $^{13}C$-NMR spectroscopy is an important tool for studying intracellular PHA degradation in terms of kinetics.

• Carbon letters
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• 제9권2호
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• pp.121-126
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• 2008

High modulus pitch based carbon fibers (HM) were exposed to isothermal oxidation using tube furnace in carbon dioxide gas to study the oxidation kinetics under the temperature of $800-1100^{\circ}C$. The kinetic equation $f=1-{\exp}(-at^b)$ was introduced and the constant b was obtained in the range of 1.02~1.42. The oxidation kinetics were evaluated by the reaction-controlling regime (RCR) depending upon the apparent activation energies with the conversion increasing from 0.2 to 0.8. The activation energies decrease from 24.7 to 21.0 kcal/mole with the conversion increasing from 0.2 to 0.8, respectively. According to the RCR, the reaction was limited by more diffusion controlling regime for the HM fibers with the conversion increasing. Therefore, it seems that the oxidation which is under the diffusion controlling regime takes place continuously from the skin to the core of the fiber.