• Journal of Advanced Marine Engineering and Technology
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• v.31 no.3
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• pp.235-242
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• 2007

This paper discusses dynamic characteristics of a urea-SCR (Selective Catalytic Reduction) system. The urea flow rate to improve NOx conversion efficiency is generally determined by parameters such as catalyst temperature and space velocity. The urea-SCR system was tested in the various engine operating conditions governing the raw NOx emission levels, space velocity. and SCR catalyst temperature. These experiments include cold-transients to determine catalyst light-off temperature and urea flow rate transients. Likewise. ammonia storage dynamics was also investigated. The cold-transient results indicate the light-off temperature of the catalysts used in these experiments was $200-220^{\circ}C$. The ammonia storage and urea flow rate transients all indicate very slow dynamics (on the order of seconds) which presents control challenges for mobile applications. The results presented in this paper should provide an excellent starting point in developing a functional in-vehicle urea-SCR system.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.113-116
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• 1995

Simplified approach has been adopted for the prediction of the thermal behavior of CANDU reactor core during power transients. Based on the assumption that the ratio of mass flow rate for each core channel does not vary during the transient, quasy-steady state analysis technique is applied with predicted core inlet boundary conditions(total mass flow rate and specific enthalpy). For restricted transient case, the presented method shows functionally reasonable estimation of core thermal behavior which could be implemented in the fast running reactor simulation program.

• Journal of the Korean Institute of Gas
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• v.12 no.2
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• pp.85-91
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• 2008

The deviation of measured pressures in pipeline networks from normal or reference pressures is useful information for judging the operation of the pipeline networks. A cost-effective monitoring of pipeline networks including a leak detection capability can be realized when transient pressure variation is accurately predicted using measured conditions at supply- and demand-sides of the networks. In this study, a pipeline network analysis program was developed based on one-dimensional flow equations for compressible fluids. The validity of the present analysis was demonstrated by simulating the flow in a straight pipeline and comparing the results with the previously reported ones. Pressure and flow rate transients in several simple city-gas pipeline networks were also analyzed to show the usefulness of the developed program.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.4
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• pp.753-764
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• 1992

The steam generator water level is difficult to control at low power due to its reversed responses to the feedwater flow, which are well known as the shrink and swell phenomena. With regard to this problem a new control scheme has been studied by which the level transients could be kept within permissible ranges at low power. The relations between the various input conditions to steam generator and the level transients have been examined to be expressed in the form of process transfer functions. Analog filters have been incorporated to be expressed in the process with proper control constants. This control scheme allows the prediction of level variation together with the corresponding feedwater rate and results in mider transients with good stabilites.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.26 no.11
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• pp.1488-1495
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• 2002

The response time represents how fast a system responds to a given disturbance at the system boundary. Flow restricting devices for controlling transients can result in a decrease in the peakm pressure, but may change response time. Response lag in a hydraulic system leads to inefficient working cycle and operator discomfort. The experiments were conducted in order to get information on the parameters which exert appreciable influence on the response time. The experimental apparatu including a hydraulic actuator, orifice and a hydraulic pump was an idealization of a bucket hydraulic shifting system. Experimental results show that the response time depends on operating pressure and flow rate. The effects of orifice type and size on the response time are quantified.

• 한국전산유체공학회:학술대회논문집
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• 2005.04a
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• pp.173-181
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• 2005

A fluid transient analysis for the propellant flow in a monopropellant propulsion system is conducted using the method of characteristics (MOC). Algebraic simultaneous equations method and Clamor's rule method utilized to drive the compatible and characteristic equations are reviewed to understand MOC more extensively. The identification of fluid transient phenomena of propulsion system of Koreasat 1 is carried out through parametric studies. Also this work describes the reason that the propulsion system of Koreasat 1 has no orifice to control flow transients or to limit the initial hydrazine flow rate for the first-pulse firing.

• Journal of Energy Engineering
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• v.7 no.2
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• pp.202-208
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• 1998

As in the other industrial processes, a nuclear power plant involves a steam relieving process through which condensable steam is discharged and condensed in a subcooled pool. An analysis of steam discharge transients was carried out using the method of characteristics to determine the flow characteristics and dynamic loads of piping that are used for structural design of the piping and its supports. The analysis included not only the steam flow rate but also the flow rates of the air and water which originally exist in the pipe. The analytical model was developed for a uniform pipe with friction through which the flow was discharged into a suppression pool. Including the combinations of system elements such as reservoir, valve and branching pipe lines. The piping flow characteristics and dynamic loads were calculated by varying system pressure, pipe length, and submergence depth. It was found that the dynamic load, water clearing time and water clearing velocity at the water/air interface were dependent not only on the system pressure and temperature but also on the pipe length and submergence depth.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.193-198
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• 1996

In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.597-603
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• 2001

Control of pressure transients in a hydraulic system may be important and necessary to avoid failures and to improve the efficiency of operation. Flow restricting devices can result in a decrease in the peak pressure, but may change the response time. The response time has an important effect on both operator and operator perceived smoothness. The response time should correspond to how fast a system responds to a given disturbance at the system boundary. Occasionally the appropriate response time is not easily determined. This study is on the response time characteristics in the hydraulic system studied for the control of response time.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.