• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.444-452
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• 2019

To assess the availability of a nuclear power plant's dynamic systems, it is necessary to consider the impact of dynamic interactions, such as components, software, and operating processes. However, there is currently no simple, easy-to-use tool for assessing the availability of these dynamic systems. The existing method, such as Markov chains, derives an accurate solution but has difficulty in modeling the system. When using conventional fault trees, the reliability of a system with dynamic characteristics cannot be evaluated accurately because the fault trees consider reliability of a specific operating configuration of the system. The dynamic reliability graph with general gates (DRGGG) allows an intuitive modeling similar to the actual system configuration, which can reduce the human errors that can occur during modeling of the target system. However, because the current DRGGG is able to evaluate the dynamic system in terms of only reliability without repair, a new evaluation method that can calculate the availability of the dynamic system with repair is proposed through this study. The proposed method extends the DRGGG by adding the repair condition to the dynamic gates. As a result of comparing the proposed method with Markov chains regarding a simple verification model, it is confirmed that the quantified value converges to the solution.

• Journal of Environmental Science International
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• v.22 no.1
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• pp.1-6
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• 2013

A fixed biofilm reactor system composed of anaerobic, anoxic(1), anoxic(2), aerobic(1) and aerobic(2) reactor was packed with synthetic activated ceramic (SAC) media and adopted to reduce the inhibition effect of low temperature on nitrification activities. The changes of nitrification activity at different wastewater temperature were investigated through the evaluation of temperature coefficient, volatile attached solid (VAS), specific nitrification rate and alkalinity consumption. Operating temperature was varied from 20 to $5^{\circ}C$. In this biofilm system, the specific nitrification rates of $15^{\circ}C$, $10^{\circ}C$ and $5^{\circ}C$ were 0.972, 0.859 and 0.613 when the specific nitrification rate of $20^{\circ}C$ was assumed to 1.00. Moreover the nitrification activity was also observed at $5^{\circ}C$ which is lower temperature than the critical temperature condition for the microorganism of activated sludge system. The specific amount of volatile attached solid (VAS) on media was maintained the range of 13.6-12.5 mg VAS/g media at $20{\sim}10^{\circ}C$. As the temperature was downed to $5^{\circ}C$, VAS was rapidly decreased to 10.9 mg VAS/g media and effluent suspended solids was increased from 3.2 mg/L to 12.0 mg/L due to the detachment of microorganism from SAC media. And alkalinity consumption was lower than theoretical value with 5.23 mg as $CaCO_3$/mg ${NH_4}^+$-N removal at $20^{\circ}C$. Temperature coefficient (${\Theta}$) of nitrification rate ($20^{\circ}C{\sim}5^{\circ}C$) was 1.033. Therefore, this fixed film nitrogen removal process showed superior stability for low temperature condition than conventional suspended growth process.

• Journal of Energy Engineering
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• v.2 no.2
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• pp.200-207
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• 1993

A methanol reformer was designed and fabricated using a CuO-ZnO low temperature shift catalyst, and its operation characteristics have been studied for the phosphoric acid fuel cell (PAFC) power generation system. The type of reactor was annular Methanol was consumed both for heating and for reforming fuel. Contents of carbon monoxide produced from the reformer increased as the reaction temperatures increased, but decreased as the mole ratios of water to methanol(H$_2$O/CH$_3$OH) increased. At steady state operating conditional, temperature profile of the catalytic reactor of the reformer was well coincide with the model equation, and it took 50 minutes from start to the rated condition of the reformer. When the system was operated at 4/4 and 1/4 of load, thermal efficiencies of the system were 72.3% and 77%, respectively. When the PAFC system was operated with reformed gas in the range of 62 V-37.6 V and 0-147 A, the trend of I-V curve showed a typical fuel tell characteristic. At steady state condition, the flow rates of reforming and combustion methanol were 88.1 mol/h and 50.1 mol/h, respectively.

• The KSFM Journal of Fluid Machinery
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• v.16 no.4
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• pp.35-43
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• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.2
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• Journal of Korean Society of Water and Wastewater
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• v.23 no.5
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• pp.547-555
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• 2009

The objectives of this research were focused on the effects of various operating parameters on nitrous oxide emission such as C/N ratio, ammonia concentration and HRT in the hybrid and suspension reactors. With the decreasing of C/N ratios, $N_2O$ emission rates in the both processes were increased because organic carbon source for denitrification was depleted. In case of biofilm reactor operated using medium, $N_2O$ release from the nitrification was not affected by the variation of ammonia concentration. But in the suspension reactor, $N_2O$ production from the nitrification was rapidly increased with the increase of ammonia. Nitrite accumulation caused by undesirable nitrification conditions could be a important reason for the increase in the $N_2O$ production from the aerobic reactor. And rapid increase in $N_2O$ production was reflected by the decrease of HRT, similar to the results observed in the results of ammonia loading changes. So it could be said that it is very important to put in consideration both its optimum conditions for wastewater treatment efficiency and suitable conditions for $N_2O$ diminish, simultaneously, in order to development an eco-friendly and advanced wastewater treatment, especially in BNR process.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.236-247
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• 1983

Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.

• Journal of the Korean Society of Propulsion Engineers
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• v.11 no.1
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• pp.18-26
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• 2007

A development of monopropellant thruster for microsatellite that uses concentrated hydrogen peroxide is described. Catalyst, the most important component in the thruster, was prepared and coated on a monolith honeycomb. Performance evaluation of thruster was peformed by considering the efficiency of characteristic velocity and ignition delay. As a result, 96.0% of $C^*$ efficiency was obtained at designed propellant flowrate and steady state operating condition.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.167-174
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• 1997

The method that the operating condition of Control Rod Drive Mechanism (CRDM) can be monitored without mounting sensors within CRDM housing was developed, and by using this developed method the closed-loop controller for the CRDM was designed which can optimize the performance and maximize the reliability of CRDM operation. Neural network is utilized as pattern recognition engine in detecting CRDM actuation. In this paper, most problems in previous open loop system are resolved. The control algorithms for closed-loop system ore developed and implemented within the hardware of timing controller based on microprocessor. All functions in the timing controller ore verified by means of real time CRDM simulator. The results show that the timing controller performs its intended functions properly.

• Computational Structural Engineering
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• v.9 no.1
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• pp.47-54
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• 1996

The Control Element Assembly(CEA) shroud is one of the most important components in the reactor vessel internals for the nuclear power plant. Because of the severe modification from its original design the structural integrity of this component has been questioned. In an attempt to resolve this question, the response of the CEA shroud to a random loading in the actual operating condition is calculated analytically and experimentally and compared to the code allowables to show that it is structurally adequate and acceptable for the long term operation.