• Journal of Environmental Science International
• /
• v.16 no.3
• /
• pp.255-260
• /
• 2007

The photocatalytic decolorization of Rhodamine B (RhB) was studied using packed-bed reactor and immobilized $TiO_2/UV$ System. The 20 W UV-A, UV-B and UV-C lamps were employed as the light source. The effect of shape and surface polishing extent of reflector, distance between the reactor and reflector, reactor material were investigated. The results showed that the order of the initial reaction constant with reflector shape was round > polygon > W > rhombus. The optimum distance between the reactor and reflector was 2 cm. The initial reaction constant of quartz reactor was 1.46 times higher than that of tile PVDF reactor.

• 제어로봇시스템학회:학술대회논문집
• /
• 2001.10a
• /
• pp.100.3-100
• /
• 2001

This article deals with the development of a mathematical model for the finishing polycarbonate polymerization process using a horizontal rotating disk-ring reactor with counter-current gas sweeping and the performance analysis of the reactor system by using the model. Here we intend to propose a model describing the reactor system consisting of two phases, in which by-product phenol is removed from the polymer of high molecular weight compatible with the products of commercial grades. The vapor phase is represented by a tanks-ln-series model while the polymer melt phase is regarded as a plug flow reactor.

• Korean Chemical Engineering Research
• /
• v.47 no.4
• /
• pp.446-452
• /
• 2009

The behavior of a one-step fixed bed reactor which directly synthesizes dimethyl ether(DME) from Natural Gas was simulated. In the reactor, the prevention of the occurrence of hot spots which can cause deactivation of catalysts is pivotal, since methanol synthesis and dehydration reaction involved in the synthesis of DME are highly exothermic. Therefore, we simulated and compared performance of the reactor with counter-current cooling and pool boiling system that can be applied to a commercial plant. As a result, we found that counter-current cooling system is more effective in terms of CO conversion and DME productivity. However, pool boiling system can operate in a small temperature gradient that can decrease problems caused by hot spot. And, the system can operate in a safer range.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2008.11a
• /
• pp.687-692
• /
• 2008

In a nuclear power plant, loose part monitoring and its diagnostic technique is one of the major issues for ensuring the structural integrity of the reactor system. Typically, accelerometers are mounted on the surface of a reactor vessel to localize impact location caused by the impact of metallic substances on the reactor system. However, in some cases, the number of the accelerometers is not enough to estimate the impact location precisely. In such a case, one of alternative plan is to utilize another type sensors that can measure the vibration of the reactor structure even though the measuring frequency ranges are different from each others. The AE sensors installed on the reactor structure can be utilized as additional sensors for loose part monitoring. In this paper, we proposed a new method to estimate impact location by using both accelerometer signal and AE signal, simultaneously. The feasibility of the proposed method is verified by an experiment. The experimental results demonstrate that we can enhance the reliability and precision of the loose part monitoring.

• Proceedings of the KIEE Conference
• /
• 2005.10b
• /
• pp.225-227
• /
• 2005

The in-core neutron Flux Mapping System in a pressurized water reactor yields information on the neutron flux distribution in the reactor core at selected core locations by means of movable detectors. The obtained data are used to verify the reactor core design parameters. The detector cables run through guide tubes(thimbles), and typically thirty-six to fifty-eight thimbles are allocated in the reactor depending on the number of fuel assemblies. These thimbles are inserted into nuclear fuel assemblies through conduits connected from the bottom of the reactor vessel to a seal table. During the plant refueling outage period, the thimbles are withdrawn up to 4m from the seal table, the height of a nuclear fuel. In spite of their importance, however, the thimble handling work has been performed by only human operators. In addition, its efficiency is very low due to narrow working environments on the seal table, thereby resulting in the excessive radiation exposure of maintenance personnel. To solve these problems, a new thimble handling equipment for in-core flux mapping system was developed, and we confirmed its effectiveness through experiments.

• Nuclear Engineering and Technology
• /
• v.52 no.11
• /
• pp.2522-2534
• /
• 2020

It is well known that the model of nuclear reactors features natural nonlinearity, and variable parameters during power tracking operation. In this paper, a disturbance observer-based adaptive sliding mode control (DOB-ASMC) strategy is proposed for power tracking of the pressurized-water reactor (PWR) in the presence of lumped disturbances. The nuclear reactor model is firstly established based on point-reactor kinetics equations with six delayed neutron groups. Then, a new sliding mode disturbance observer is designed to estimate the lumped disturbance, and its stability is discussed. On the basis of the developed DOB, an adaptive sliding mode control scheme is proposed, which is a combination of backstepping technique and integral sliding mode control approach. In addition, an adaptive law is introduced to enhance the robustness of a PWR with disturbances. The asymptotic stability of the overall control system is verified by Lyapunov stability theory. Simulation results are provided to demonstrate that the proposed DOB-ASMC strategy has better power tracking performance than conventional sliding mode controller and PID control method as well as conventional backstepping controller.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.18 no.11
• /
• pp.1143-1149
• /
• 2008

In a nuclear power plant, loose part monitoring and its diagnostic technique is one of the major issues for ensuring the structural integrity of the reactor system. Typically, accelerometers are mounted on the surface of a reactor vessel to localize impact location cavsed by the impact of metallic substances on the reactor system. However, in some cases, the number of the accelerometers is not enough to estimate the impact location precisely. In such a case, one of alternative plan is to utilize another type sensors that can measure the vibration of the reactor structure even though the measuring frequency ranges are different from each others. The AE sensors installed on the reactor structure can be utilized as additional sensors for loose part monitoring. In this paper, we proposed a new method to estimate impact location by using both accelerometer signal and AE signal, simultaneously. The feasibility of the proposed method is verified by an experiment. The experimental results demonstrate that we can enhance the reliability and precision of the loose part monitoring.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.191-196
• /
• 1996

The potential value of the Application Specific Integrated Circuits(ASIC's) in safety systems of Nuclear Power Plants(NPP's) is being increasingly recognized because they are essentially hardwired circuitry on a chip, the reliability of the system can be proved more easily than that of software based systems which is difficult in point of software V&V(Verification and Validation). There are two types of ASIC, one is a full customized type, the other is a half customized type. PLD(Programmable Logic Device) used in this paper is a half customized ASIC which is a device consisting of blocks of logic connected with programmable interconnections that are customized in the package by end users. This paper describes the RPS(Reactor Protection System) composed of ASICs which provides emergency shutdown of the reactor to protect the core and the pressure boundary of RCS(Reactor Coolant System) in NPP's. The RPS is largely composed of five logic blocks, each of them was implemented in one PLD, as the followings. A). Bistable Logic B). Matrix Logic C).Initiation Logic D). MMI(Man Machine Interface) Logic E). Test Logic.

• Nuclear Engineering and Technology
• /
• v.47 no.4
• /
• pp.434-442
• /
• 2015

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.500-505
• /
• 1995

This paper describes the results of simulation of a CANDU operational transient problem (re-startup after short shutdown) using the Coupled Reactor Kinetics(CRKIN) code developed previously with CANDU Reactor Regulating System(RRS) logic. The performance in the simulation is focused on investigating the behaviours of neutron power and regulating devices in accordance with the changes of xenon concentration following the operation of the RRS.