• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.15 no.12 s.105
• /
• pp.1370-1377
• /
• 2005

The bottom structure of an instrumented capsule is a part which is joined at the receptacle of the flow tube in the reactor in-core. A geometrical change of the bottom structure has an effect on the pressure drop and the vibration of the capsule. The out-pile test to evaluate the structural integrity of the material capsule called 04M-17U was performed by using a single channel and a half core test loop. From the pressure drop test, the optimized diameter of the cone shape's bottom structure which satisfies HANARO's flow requirement (19.6 kg/s) is 71 mm. The maximum displacement of the capsule measured at the half core test loop is lower than 1.0 mm. From the analysis results, it is found that the test hole will not be interfered with near the flow tubes because its displacement due to the cooling water is very small at 0.072 mm. The fundamental frequency of the capsule under water is 9.64 Hz. It is expected that the resonance between the capsule and the fluid flow due to the cooling water in HANARO's in-core will not occur. Also, the new bottom structure of a solid cone shape with 71 mm in diameter will be applicable to the material and special capsules in the future.

• Applied Chemistry for Engineering
• /
• v.4 no.1
• /
• pp.171-177
• /
• 1993

In decontamination process to remove radioactive materials of reactor cooling system, the metal ions dissolved by organic acids in decontamination solution are separated by use of ion exchange resin in the column. However, organic acids in decontamination solution decrease the apparent affinity of the resin to metal ions. In light of this, some experiments were carried out on the Amberlite IRN-77 cation resin with cobalt and iron to gain a better understanding of the complexation effects on the ion exchange process. Experimental results showed that EDTA among organic acids used as chemical decontaminants predominantly caused reduction of ion exchange capacity of cobaltous ion to resin since this reagent formed the complex with the cobaltous ion stronger than that with the ferrous ion. In contrast, the effects of oxalic acid and citric acid were found to be negligible. And, single and two-component nonlinear equilibrium relationships of the metal ions were established using experimental data.

• Nuclear Engineering and Technology
• /
• v.27 no.5
• /
• pp.743-752
• /
• 1995

This paper investigates the effects of fuel groupings in the coupled internals and core model on the internals and fuel responses due to pipe breaks. The 177 fuel assemblies for Korean standard nuclear power plant are grouped into several stick models and the responses of internals components are calculated. The analysis results show that the fuel model groupings in the coupled internals and core model have no significant effects on the internals and fuel responses for pipe break excitation. Also, in order to determine the feasibility of constructing a single equivalent stick representation of In or more adjacent fuel bundles, the reduced models, each of which employs a different stiffness lumping rule, are constructed. It is shown that the equivalent stiffness calculated to get the first natural frequency of the original model while preserving net gap between grouping centers gives the minimum modelling error.

• Nuclear Engineering and Technology
• /
• v.26 no.2
• /
• pp.168-178
• /
• 1994

Accident Management involves all measures to prevent core damage and retain the core within the reactor vessel, maintain containment integrity and minimize off-site releases. The accident management approach includes : (1) advanced evaluation of candidate strategies, (2) development of procedures to execute appropriate actions efficiently, and (3) identification and provision for materials, tools, and possible modifications to the plant system that may be needed for such execution. When assessing accident management strategies it effectiveness, adverse effect and its feasibility, including information needs and compatibility with existing procedures, must be considered. The objective of this paper is to introduce analytical tools of decision trees and influence diagrams to develop a framework for modeling and assessing severe accident management strategies. The characteristics associated with these took are presented. Based on decision trees and influence diagrams, the framework is applied to a simple example associated with a single decision.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.58 no.10
• /
• pp.1916-1922
• /
• 2009

Introducing distributed generators(DGs) to utility distribution system can cause malfunction of relay on the grid when ground faults or severe load unbalances are occurred on the system. Because DGs interconnected to the grid can contribute fault currents and make bidirectional power flows on the system, fault currents from DGs can cause an interference of relay operation. A directional over current relay(DOCR) can determine the direction of power flow whether a fault occurs at the source side or load side through detecting the phases of voltage and current simultaneously. However, it is identified in this paper that the contributed fault current(Ifdg) from the ground source when was occurred to contribute single-line-to-ground(SLG) fault current, has various phases according to the distances from the ground source. It means that the directionality of Ifdg may not be determined by simply detecting the phases of voltage and current in some fault conditions. The magnitude of Ifdg can be estimated approximately as high as 3 times of a phase current and its maximum is up to 2,000 A depending on the capacity of generation facilities. In order to prevent malfunction of relay and damage of DG facilities from the contribution of ground fault currents, Ifdg should be limited within a proper range. Installation of neutral ground reactor (NGR) at a primary neutral of interconnection transformer was suggested in the paper. Capacity of the proposed NGR can be adjusted easily by controlling taps of the NGR. An algorithm for unidirectional relay was also proposed to prevent the malfunction of relay due to the fault current, Ifdg. By the algorithm, it is possible to determine the directionality of fault from measuring only the magnitude of fault current. It also implies that the directionality of fault can be detected by unidirectional relay without replacement of relay with the bidirectional relay.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
• /
• v.16 no.6
• /
• pp.40-50
• /
• 2002

In this paper, we present a GA(Genetic Algorithm) approach to select weighting matrices of an optimal LQ(Linear Quadratic) controller for SVC(Static VAR Compensator). A SVC, one of the FACTS(Flexible AC Transmission System), constructed by a FC(Fixed Capacitor) and a TCR(Thyristor Controlled Reactor), was designed and implemented to improve the damping of a synchronous generator, as well as to control the system voltage Also, a design of LQ controller depends on choosing weighting matrices. The selection of weighting matrices which is not a trivial solution is usually carried out by trial and error. We proposed an efficient method using GA of finding weighting matrices for optimal control law. Thus, we proved the usefulness of proposed method to improve the stability of single machine-infinite bus with SVC system by eigenvalues analysis and simulation.

• Proceedings of the Korean Vacuum Society Conference
• /
• 2000.02a
• /
• pp.74-74
• /
• 2000

낮은 유전상수(k$\leq$3)와 높은 열적안정성(>4$25^{\circ}C$)은 초고집적회로(ULSI)기술에서 RC 지연을 해결하기 위한 금속배선의 중간 절연층으로서의 2개의 가장 중요한 특성이다. 본 연구에서는 cyclohezane을 precursor로 사용하여 plasma enhanced chemical vapor deposition(PECVD)방법으로 유기박막을 성장시켰으며 낮은 유전상수와 높은 열적안정성을 동시에 확보하기 위하여 열적안정성은 좋지 않지만 유전상수가 낮은 박막(soft layer)위에 유전상수는 다소 높지만 열적안정성이 좋은 박막(hard layer)을 얇게 증착하여 hard layer/soft layer의 2층 구조를 형성하여서 구조적, 전기적 특성을 조사하였다. 유기박막은 5$0^{\circ}C$로 유지된 reactor 내부에서 argon(Ar) plasma에 의해 증착되었으며 platinum(Pt)기판과 silicon 기판위에 동시에 증착하였다. Pt 기판위에 증착한 시편으로 유전상수, I-V 등 전기적 특성을 측정하였고, silicon 기판위에 증착한 시편으로 열적안정성과 구조적 특성등을 분석하였다. 증착압력 0.2Torr에서 plasma power를 5W에서90W로 증가할 때 유전상수는 2.36에서 3.39로 증가하였으며 열적안정성은 90W에서 180W로 증가하였을 때 유전상수는 2.42에서 2.79로 증가하엿고 열적안정성은 모두30$0^{\circ}C$이하였다. 단일층 구조에서는 유전상수가 낮은 박막은 열적으로 불안정하고 열적 안정성이 좋은 박막은 유전상수가 다소 높은 문제가 나타났다. 이런 문제를 해결하기 위하여 2 Torr, 120W에서 증착한 유전상수가 2.55이고 열적으로 불안정한 박막을 soft layer로 5150 증착하고 그 위에 0.2Torr, 90W에서 증착한 유전상수가 3.39이고 열적으로 45$0^{\circ}C$까지 안정한 박막을 hard layer로 360 , 720 , 1440 증착하였다. 증착된 2층구조 박막의 유전상수는 각각 2.62, 2.68, 2.79이었으며 열적안정성 측정에서는 40$0^{\circ}C$까지 두께 감소가 보이지 않았다. 그러나 SEM 측정에서 열처리 후 표면이 거칠어지는 현상이 발견되었다.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2005.11a
• /
• pp.782-787
• /
• 2005

The bottom structure of an instrumented capsule is a part which is joined at the receptacle of the flow tube in the reactor in-core. A geometrical change or the bottom structure has an effect on the pressure drop and the vibration of the capsule. The out-pile test to evaluate the structural Integrity of the material capsule called 04M-l7U was performed by using a single channel and a half core test loop. From the pressure drop test, the optimized diameter of the cone shape's bottom structure which satisfies HANARO's flow requirement (19 6 kg/s) is 71 mm. The maximum displacement of the capsule measured at the half core test loop is lower than 1.0 mm. From the analysis results, it is found that the test hole will not be interfered with near the flow tubes because its displacement due to the cooling water is very small at 0.072 mm. The fundamental frequency of the capsule under water is 9.64 Hz. It is expected that the resonance between the capsule and the fluid flow due to the cooling water in HANARO's In-core will not occur. Also, the new bottom structure of a solid cone shape with 71 mm in diameter will be applicable to the material and special capsules in the future.

• Journal of the Korean Society of Propulsion Engineers
• /
• v.22 no.2
• /
• pp.74-86
• /
• 2018

This paper presents the one-dimensional model of a hydrogen generator, where the alkali solution was supplied from the top to the dry aluminum powders. Hydrogen was produced as the solution moved downward and reacted with aluminum. The species conservation equations were considered for the hydrogen gas and alkali solution, while the energy conservation equation was applied to the gas-liquid-solid mixture as a single medium. The gas rising velocity and liquid penetration velocity were also included in the theoretical approach. The developed code was validated with the experimental data of the hydrogen production amount and collector pressure. Additionally, the model successfully predicted the various reactor properties, such as the concentrations, volume fractions, and temperatures, and is expected to help significantly in the design of a novel hydrogen generator.

• Nuclear Engineering and Technology
• /
• v.26 no.4
• /
• pp.555-561
• /
• 1994

The program of performing a fast and automatic analysis of gamma ray spectra obtained by a Multi-Channel Analyzer (MCA) is developed for the surveillance of the nuclear fuel integrity. The integrity of the nuclear fuel is confirmed by the measurement of the radiation level of the reactor coolant through the real time monitoring and the periodic sampling analysis. In Yonggwang nuclear power plane 3 and 4, the Process Radiation Monitoring System (PRMS), which is a real time monitoring system, provides a measure of the fuel integrity. Currently, its spectrometer channel can identify only one radionuclide at a time since the signal processing unit of the spectrometer channel is a Single Channel Analyzer (SCA). To improve the PRMS, it is necessary to substitute the MCA for the SCA The program is operated in a real time mode and an on-demand mode, and automatically performed for all procedures. The test results by using the National Bureau of Standards (NBS) mixed standard source are in good agreement with those from Canberra System 100 which is a commercial MCA Consequently, the developed program seems to be employed for automatic monitoring of gamma rays in nuclear power plants.