• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.211-218
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• 2013

Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 ($^{14}C$), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the $^{14}C$, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create $CO_x$ gases, i.e. "gasify" graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl-like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide ($CO_2$). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].

• Proceedings of the Korean Vacuum Society Conference
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• 1998.02a
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• pp.77-77
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• 1998

Present silicon dioxide (SiOz) 떠m as intennetal dielectridIMD) layers will result in high parasitic c capacitance and crosstalk interference in 비gh density devices. Low dielectric materials such as f f1uorina뼈 silicon oxide(SiOF) and f1uoropolymer IMD layers have been tried to s이ve this problem. I In the SiOF ftlm, as fluorine concentration increases the dielectric constant of t뼈 film decreases but i it becomes unstable and wa않r absorptivity increases. The dielectric constant above 3.0 is obtain어 i in these ftlms. Fluoropolymers such as polyte$\sigma$따luoroethylene(PTFE) are known as low dielectric c constant (>2.0) materials. However, their $\alpha$)Or thermal stability and low adhesive fa$\pi$e have h hindered 야1리ru뚱 as IMD ma따"ials. 1 The concept of a plasma processing a찌Jaratus with 비gh density plasma at low pressure has r received much attention for deposition because films made in these plasma reactors have many a advantages such as go여 film quality and gap filling profile. High ion flux with low ion energy in m the high density plasma make the low contamination and go어 $\sigma$'Oss피lked ftlm. Especially the h helicon plasma reactor have attractive features for ftlm deposition 야~au똥 of i앙 high density plasma p production compared with other conventional type plasma soun:es. I In this pa야Jr, we present the results on the low dielectric constant fluorocarbonated-SiOF film d밑JOsited on p-Si(loo) 5 inch silicon substrates with 00% of 0dFTES gas mixture and 20% of Ar g gas in a helicon plasma reactor. High density 띠asma is generated in the conventional helicon p plasma soun:e with Nagoya type ill antenna, 5-15 MHz and 1 kW RF power, 700 Gauss of m magnetic field, and 1.5 mTorr of pressure. The electron density and temperature of the 0dFTES d discharge are measUI벼 by Langmuir probe. The relative density of radicals are measured by optic허 e emission spe따'Oscopy(OES). Chemical bonding structure 3I피 atomic concentration 따'C characterized u using fourier transform infrared(FTIR) s야3띠"Oscopy and X -ray photonelectron spl:’따'Oscopy (XPS). D Dielectric constant is measured using a metal insulator semiconductor (MIS;AVO.4 $\mu$ m thick f fIlmlp-SD s$\sigma$ucture. A chemical stoichiome$\sigma$y of 야Ie fluorocarbina$textsc{k}$영-SiOF film 따~si야영 at room temperature, which t the flow rate of Oz and FTES gas is Isccm and 6sccm, res야~tvely, is form려 야Ie SiouFo.36Co.14. A d dielec$\sigma$ic constant of this fIlm is 2.8, but the s$\alpha$'!Cimen at annealed 5OOt: is obtain려 3.24, and the s stepcoverage in the 0.4 $\mu$ m and 0.5 $\mu$ m pattern 킹'C above 92% and 91% without void, res야~tively. res야~tively.

• Journal of the Korean Vacuum Society
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• v.5 no.3
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• pp.239-244
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• 1996

We firstly report that formation of $SiO_2$ layer on Si surface can be effectively prevented by flowing the $Si_2H_6$ gas during the heating-up procedure for amorphous Si depositions. In this way, amorphously deposited Si layer onto crystalline Si substrates can be grown epitaxially during the post-deposition heat treatments. The suppression of surface $SiO_2$ can be explained in terms of adsorption of SiHx adspecies, instead of oxygen from residual gases in the reactors, to Si surfaces after desorption of hydrogen from H-passivated Si surfaces. Employing $Si_2H_6$ flowing and soild phase epitaxial growth, high-quality epitaxial Si layer can be obtained at low temperatures below $600^{\circ}C$ without conventional high temperature cleaning procedures.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.797-801
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• 2020

Most of irradiated graphite that should be disposed comes from moderators and reflectors of nuclear power plants. The quantity of irradiated graphite could be higher in the future if high-temperature reactors (HTRs) will be deployed. In this case noteworthy quantities of fuel pebbles containing semi-graphitic carbonaceous material should be added to the already existing 250,000 tons of irradiated graphite. Industry graphite is largely used in industrial applications for its high thermal and electrical conductivity and thermal and chemical resistance, making it a valuable material. Irradiated graphite constitutes a waste management challenge owing to the presence of long-lived radionuclides, such as ¹⁴C and ³⁶Cl. In the ENEA Nuclear Material Characterization Laboratory it has been successfully designed a procedure based on the exfoliation process organic solvent assisted, with the purpose of investigate the possibility of achieving graphite significantly less toxic that could be recycled for other purpose [1]. The objective of this paper is to evaluate the possibility of the scalability from laboratory to industrial dimensions of the exfoliation process and provide the prototype of a chemical plant for the treatment of irradiated graphite.

• Korean Chemical Engineering Research
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• v.52 no.6
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• pp.826-833
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• 2014

Driven by both environmental and economic reasons, the development of small to medium scale GTL(gas-to-liquid) process for offshore applications and for utilizing other stranded or associated gas has recently been studied increasingly. Microchannel GTL reactors have been prefrered over the conventional GTL reactors for such applications, due to its compactness, and additional advantages of small heat and mass transfer distance desired for high heat transfer performance and reactor conversion. In this work, multi-microchannel reactor was simulated by using commercial CFD code, ANSYS FLUENT, to study the geometric effect of the microchannels on the heat transfer phenomena. A heat generation curve was first calculated by modeling a Fischer-Tropsch reaction in a single-microchannel reactor model using Matlab-ASPEN integration platform. The calculated heat generation curve was implemented to the CFD model. Four design variables based on the microchannel geometry namely coolant channel width, coolant channel height, coolant channel to process channel distance, and coolant channel to coolant channel distance, were selected for calculating three dependent variables namely, heat flux, maximum temperature of coolant channel, and maximum temperature of process channel. The simulation results were visualized to understand the effects of the design variables on the dependent variables. Heat flux and maximum temperature of cooling channel and process channel were found to be increasing when coolant channel width and height were decreased. Coolant channel to process channel distance was found to have no effect on the heat transfer phenomena. Finally, total heat flux was found to be increasing and maximum coolant channel temperature to be decreasing when coolant channel to coolant channel distance was decreased. Using the qualitative trend revealed from the present study, an appropriate process channel and coolant channel geometry along with the distance between the adjacent channels can be recommended for a microchannel reactor that meet a desired reactor performance on heat transfer phenomena and hence reactor conversion of a Fischer-Tropsch microchannel reactor.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• Journal of Environmental Science International
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• v.25 no.11
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• pp.1493-1498
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• 2016

The efficiencies of Gang-Byeon sewage treatment facilities, which are based on GPS-X modelling, were analysed and used to design recycle water treatment processes. The effluent of an aeration tank contained total kjeldahl nitrogen (TKN) of 1.8 mg/L with both C-1 and C-2 conditions, confirming that most ammonia nitrogen ($NH_3{^+}-N$) was converted to nitrate nitrogen ($NO_3{^-}-N$). The concentrations of $NH_3{^+}-N$ and $NO_3{^-}-N$ were found to be 222.5 and 227.2 mg/L, respectively, with C-1 conditions and 212.2 and 80.4 mg/L with C-2 conditions. Although C-2 conditions with higher organic matter yielded a slightly higher nitrogen removal efficiency, sufficient denitrification was not observed to meet the discharge standards. For the total nitrogen (T-N) removal efficiency, the final effluent concentrations of T-N were 293.8 mg/L with biochemical oxygen demand (BOD) of 2,500 mg/L, being about 1.5 times lower than that (445.3 mg/L) with BOD of 2,000 mg/L. Therefore, an external carbon source to increase the C/N ratio was required to get sufficient denitrification. During the winter period with temperature less than $10^{\circ}C$, the denitrification efficiency was dropped rapidly even with a high TKN concentration (1,500 mg/L). This indicates that unit reactors (anoxic/aerobic tanks) for winter need to be installed to increase the hydraulic retention time. Thus, to enhance nitrification and denitrification efficiencies, flexible operations with seasons are recommended for nitrification/anoxic/denitrification tanks.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.75-83
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• 2017

In severe accident conditions of light water reactors, the loss of coolant may cause problems in integrity of zirconium fuel cladding. Under the condition of the loss of coolant, the zirconium fuel cladding can be exposed to high temperature steam and reacted with them by producing of hydrogen, which is caused by the failure in oxidation resistance of zirconium cladding materials during the loss of coolant accident scenarios. In order to avoid these problems, we develop a multi-metallic layered composite (MMLC) fuel cladding which compromises between the neutronic advantages of zirconium-based alloys and the accident-tolerance of non-zirconium-based metallic materials. Cold pilgering process is a common tube manufacturing process, which is complex material forming operation in highly non-steady state, where the materials undergo a long series of deformation resulting in both diameter and thickness reduction. During the cold pilgering process, MMLC claddings need to reduce the outside diameter and wall thickness. However, multi-layers of the tube are expected to occur different deformation processes because each layer has different mechanical properties. To improve the utilization of the pilgering process, 3-dimensional computational analyses have been made using a finite element modeling technique. We also analyze the dimensional change, strain and stress distribution at MMLC tube by considering the behavior of rolls such as stroke rate and feed rate.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1138-1147
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• 2018

Delayed hydride cracking (DHC) is an important failure mechanism for Zircaloy tubes in the demanding environment of nuclear reactors. The threshold stress intensity factor, $K_{IH}$, and critical hydride length, $l_C$, are important parameters to evaluate DHC. Theoretical models of them are developed for Zircaloy tubes undergoing non-homogenous temperature loading, with new stress distributions ahead of the crack tip and thermal stresses involved. A new stress distribution in the plastic zone ahead of the crack tip is proposed according to the fracture mechanics theory of second-order estimate of plastic zone size. The developed models with fewer fitting parameters are validated with the experimental results for $K_{IH}$ and $l_C$. The research results for radial cracking cases indicate that a better agreement for $K_{IH}$ can be achieved; the negative axial thermal stresses can lessen $K_{IH}$ and enlarge the critical hydride length, so its effect should be considered in the safety evaluation and constraint design for fuel rods; the critical hydride length $l_C$ changes slightly in a certain range of stress intensity factors, which interprets the phenomenon that the DHC velocity varies slowly in the steady crack growth stage. Besides, the sensitivity analysis of model parameters demonstrates that an increase in yield strength of zircaloy will result in a decrease in the critical hydride length $l_C$, and $K_{IH}$ will firstly decrease and then have a trend to increase with the yield strength of Zircaloy; higher fracture strength of hydrided zircaloy will lead to very high values of threshold stress intensity factor and critical hydride length at higher temperatures, which might be the main mechanism of crack arrest for some Zircaloy materials.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3236-3246
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• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.