• 한국안전학회지
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• 제34권3호
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4743-4750
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• 2022

Heat pipe cooled nuclear reactor has been a very attractive technical solution to provide the power for deep space applications. In this paper, a 200 kWe space nuclear reactor power design has been proposed based on the combination of an integrated UN ceramic fuel, a heat pipe cooling system and the Stirling power generators. Neutronics and thermal analysis have been performed on the space nuclear reactor. It was found that the entire reactor core has at least 3.9 $ subcritical even under the worst-case submersion accident superimposed a single safety drum failure, and results from fuel temperature coefficient, neutron spectrum and power distribution analysis also showed that this reactor design satisfies the neutronics requirements. Thermal analysis showed that the power in the core can be successfully removed both in normal operation or under one or more heat pipes failure scenarios.

• 한국물환경학회지
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• 제22권4호
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• pp.599-604
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• 2006

Nitrogen removal with the combined SHARON (Single reactor system for high ammonium removal over nitrite)ANAMMOX (Anaerobic ammonium oxidation) process using the effluent of ADEPT (Anaerobic digestion elutriated phased treatment) slurry reactor with very low C/N ratio for piggery waste treatment was investigated. For the preceding SHARON reactor, ammonium nitrogen loading and removal rate were $0.97kg\;NH_4-N/m^3_{reactor}/day$ and $0.68kg\;NH_4-N/m^3_{reactor}/day$ respectively. In steady state, bicarbonate alkalinity consumption for ammonium nitrogen converted to $NO_2-N$ or $NO_3-N$ was 8.4 gram per gram ammonium nitrogen. The successive ANAMMOX reactor was fed with the effluent from SHARON reactor. The loading and removal rate of the soluble nitrogen defined as the sum total of $NH_4-N$, $NO_2-N$ and $NO_3-N$ in ANAMMOX reactor were $1.36kg\;soluble\;N/m^3_{reactor}/day$ and $0.7kg\;soluble\;N/m^3_{reactor}/day$, respectively. The average $NO_2-N/NH_4-N$ removal ratio by ANAMMOX was 2.41. Fluorescence in situ hybridization (FISH) analysis verified that Candidatus Kuenenia stuttgartiensis were dominate, which means that they played an important role of nitrogen removal in ANAMMOX reactor.

• 전력전자학회논문지
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• 제21권6호
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• pp.514-524
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• 2016

This work suggests a study on 5 kW plasma power supply design and reactor capacitance estimation algorithm for a wide range of linear output control to operate a plasma reactor. The suggested study is designed to use a two-stage circuit and control the full-bridge circuit of the two-stage circuit using the buck converter output voltage of the single-stage circuit. The switching frequency of the full-bridge circuit is designed to operate through high-frequency switching and obtain maximum output using LC parallel resonance. Soft switching technique(ZVS) is used to reduce the loss caused by high-frequency switching, and duty control of the buck converter is applied to control a wide range of linear output. The internal capacitance of the reactor cannot easily be extracted, and thus, the reactor cannot be operated in an optimized resonant state. To address this issue, this work designs the internal capacitance of the reactor such that estimations can be performed with the developed reactor capacitance estimation algorithm applied to the internal capacitance of the reactor. A 5 kW plasma power supply is designed for a wide range of linear output control, and the validity of the study on the reactor capacitance estimation algorithm is verified.

• 한국물환경학회지
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• 제34권1호
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• pp.109-120
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• 2018

Sidestream in domestic wastewater treatment plants contains high concentration of ammonium, which increases nitrogen loading rate in the mainstream. The process for deammonification consisting of partial nitritation-anaerobic ammonium oxidation (ANAMMOX) and heterotrophic denitrification is an economical method of solving this problem. Currently, about 130 full-scale deammonification plants are fully operating around the world, but none is in Korea. In order to transfer the principal information about sidestream deammonification processes to researchers and operators, we summarized basic concepts, processes type, and key influence factors (e.g., concentration of nitrogen compounds, dissolved oxygen (DO), temperature, and pH). This review emphasis on the processes of single-stage sequencing batch reactor (SBR) deammonification, which are widely used as full-scale plants. Since simultaneous processes of partial nitritation, ANAMMOX and heterotrophic denitrification occur in a single reactor, the single-stage SBR deammonification requires appropriate control/monitoring strategies for several operating factors (DO and pH mostly) to achieve efficient and stable operation. In future, AB-process consisting of A-stage (energy harvesting from organics) and B-stage (ammonium removal without organics) will be applied to the wastewater treatment process. Thus, we suggest mainstream deammonification for B-stage connected with the sidestream deammonification as seeding source of ANAMMOX. We expect that many researchers will become more interested in the sidestream deammonification.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.18-24
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• 2018

This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor ($k_{eff}$) versus burnup (B) are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B)-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE)14 $10{\times}10$ boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC)-68 storage cask. The results revealed that the curves of $k_{eff}$ versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of $k_{eff}$, ${\Delta}k$) in some compound effects was not a summation of the all ${\Delta}k$ resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of $k_{eff}$ versus B for both single and compound effects.

• 한국환경과학회지
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• 제20권10호
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• pp.1337-1345
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• 2011

The removal performances of divalent heavy metal ions ($Pb^{2+}$, $Cu^{2+}$, $Cd^{2+}$, $Sr^{2+}$ and $Mn^{2+}$) were studied using the Na-P1 zeolite synthesized from Jeju scoria in the batch and continuous fixed column reactor. The uptakes of heavy metal ions by synthetic Na-P1 zeolite decreased in the order of $Pb^{2+}$ > $Cu2^{2+}$ > $Cd^{2+}$ > $Sr^{2+}$ > $Mn^{2+}$ based on the selectivity of each ion to ionic exchange site of Na-P1 zeolite for single and mixed solutions in batch or continuous fixed column reactor. For mixed solution, each heavy metal ion uptake was lower than that in single solution, and especially the uptake for $Mn^{2+}$ decreased greatly. In batch reactor, the uptakes of heavy metal ions by synthetic Na-P1 zeolite were described by Freundlich or Langmuir equation, but they followed the former better than the latter. In continuous fixed column reactor, the maximum ion exchange capacity obtained for each of heavy metal ions, was about 90----- of that in batch reactor. The uptakes of heavy metal ions by synthetic Na-P1 zeolite increased with the increase of initial heavy metal concentration and solution pH, and the decrease of the amount and particle size of synthetic zeolite.

• 상하수도학회지
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• 제26권3호
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• pp.409-421
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• 2012

Ethanolamine (ETA) is mainly used to prevent corrosion of pipe in secondary cooling system of nuclear power plant. Condensed ETA in wastewater could increase COD and T-N when it was emitted to natural water system. Compared to conventional treatments, electrochemical oxidation process using packed bed bipolar electrodes was adopted to treat COD and T-N. According to arrangement of feeder electrode, single packed bed bipolar electrode reactor and multi-paired packed bed bipolar reactor were developed and conventional zero-valent iron (ZVI) was selected as conducting bipolar electrode. Bipolar electrodes were coordinated three-dimensionally in the reactor. The experimental results showed that COD and T-N was little removed in unit system at different pH condition (pH 8 and 11) on 100V. However, in multi-paired system that applied 600V, COD was eliminated 80.85% (anode-cathode-anode, A-C-A) and 85.11% (cathode-anode-cathode, C-A-C), respectively. T-N was also removed 96.88% (A-C-A) and 90.63% (C-A-C), simultaneously. Current efficiency was estimated both single and multi-paired system. At unit bipolar packed bed reactor, current efficiency was almost zero, however in multi-paired system, current efficiency was 300~500% at A-C-A and 250~350% at C-A-C. Current efficiency was over 100% hence it was confirmed that this system is more effective than conventional electrochemical oxidation system.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.441-446
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• 2001

In the integral type reactor, SMART, all the major components such as steam generators, pressurizer and pumps are located inside the single reactor pressure vessel. The objective of this study is to evaluate the structural integrity for RPV of SMART under the postulated pressurized thermal shock by applying the finite element analysis. Input data for the finite element analysis were generated using the commercial code I-DEAS, and the fracture mechanics analysis was performed using the ABAQUS. The crack configurations, the crack aspect ratio and the clad thickness were considered in the parametric study. The effects of these parameters on the reference nil-ductility transition temperature were also investigated.