• 환경위생공학
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• 제10권3호
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• pp.67-77
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• 1995

This study was performed for minimization of excessive biofilm effects at the high strength organic wastewater treatment. As a results of biofilm attachment experiment using piggery wastewater, aggravation of water quality due to excessive biofilm showed after 15 days of operating times.4 excessive biofilm phase, the equivalent biofilm thickness and VSS contents per unit aura were observed in the range of 1,100 to $1,200{\mu}m$ and 2.5 to 3.0mg $VSS/cm^{2}$, respectively. In the aerobic fixed biofilm reactor/anoxic fixed biofilm reactor(AFBR/ANFBR) process with endogenous respiration phase, the BOD removal efficiency was obtained more than 90 percentage at the surface loading rate and volumetric loading rate of the AFBR maintained less than 17 g $BOD/m^{2}{\cdot}$day and 1.7kg $BOD/m^{3}{\cdot}$day, respectively. The removal efficiency of TKN and $NH_{3}$-N at the loading rates below 5.60g $NH_{3}-N/m^{2}{\cdot}day$ and 0.56kg $NH_{3}-N/m^{3}{\cdot}$day were above 76 percentage and 82 percentage, respectively. In order to reduced sludge production rate and aggravation of water quality, endogenous respiration phase was accepted at first AFBR reactor. As a results of this operating condition, sludge production was minimized and removal efficiency was maintained stability.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2006년도 추계학술대회
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• pp.477-480
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• 2006

In this study, we investigated operating characteristics of the LNG burner for steam methane reforming. The developed LNG burner and catalytic reactor to supply an efficient heat transfer between the combustion gas and catalyst got a good response of various operating load within 5-7 minute and high efficiency for steam methane reforming as a conversion of methane over 90%. We calculated the volume of catalyst for $1Nm^3/hr$ steam LNG reforming as $211cc/(Nm^3/hr\;H_2)$ and got the operating condition and design data of the burner and steam reforming for LNG.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• 소음진동
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• 제10권2호
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• pp.261-268
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• 2000

The instrumented capsule is subjected to flow-induced vibration(FIV) due to the flow of the primary coolant and then the structural integrity of the capsule during irradiation in the HANARO reactor is an issue of major concern. For this purpose the acceleration was measured by four accelerometers attached to the protection tube of the capsule mainbody and the displacement of test holes was calcultated using commercial finite element program ANSYS to evaluate the structural interference with the neighboring flow tubes under the reactor operating condition. The calculated displacements of test holes in the reactor in-core were found to be lower than the values of allowable design criteria.

• 한국전산유체공학회지
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• 제16권1호
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• pp.42-47
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• 2011

In this paper, various operating parameters of stream reforming process from methane in stream reformer and preconverter for MCFC is studied by numerical method. Commercial code is used to simulated the porous catalyst with user subroutine to model three dominant chemical reactions which are Stream Reforming(SR), Water-Gas Shift(WGS), and Direct Stram Reforming(DSR). The hydrogen production is tested with different wall temperature and different reactor shapes. The calculated results of the concentration of hydrogen in stream reformer are very well consistent with experimental results. This numerical study gives the design reactor wall temperature condition and size of reactor to satisfy the required fuel conversion.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1998년도 춘계학술대회논문집; 용평리조트 타워콘도, 21-22 May 1998
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• pp.582-589
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• 1998

This paper presents the analysis results of ex-core and in-core neutron noise, acceleration signals and pressure fluctuation measured at Ulchin Nuclear Unit 1 & 2 to identify and monitor the reactor internals vibration including fuel motion. A phase separation algorithm developed by authors was applied to the neutron noises to clearly identify the reactor internals vibratory motion. The beam mode frequency of the core support barrel was identified to be 8Hz and the shell mode to be 20Hz. The first frequency of the fuel assembly was also found to be 3Hz, while first two acoustic frequencies of the primary coolant system were 6 and 17.5Hz. By monitoring and analyzing these frequencies periodically, it is possible to diagnose the operating condition of reactor internals and to provide an early detection of faults for the predictive maintenance.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.