• Journal of Information Management
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• v.35 no.1
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• pp.1-27
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• 2004

To share knowledge efficiently among our members on the basis of knowledge management system, first of all, we need to systematically design the knowledge classification scheme that we can classify these knowledge well. The objective of this study is to construct the most suitable knowledge classification scheme that all of us can share them in Korea Atomic Energy Research Institute (KAERI). To construct the knowledge classification scheme all over the our organization, we established a few principles to design it and examined related many classification schemes. And I carried out 3 steps to complete the best desirable KAERI's knowledge classification scheme, that is, (1) the step to design a draft of the knowledge classification scheme, (2) the step to revise a draft of the knowledge classification scheme, (3) the step to verify the revised scheme and to decide its scheme. The scheme completed as a results of this study is consisted of total 218 items : sections of 8 items, classes of 43 items and sub-classes of 167 items. I expect that the knowledge classification scheme designed as the results of this study can be played an important role as the frame to efficiently share knowledge among our members when we introduce knowledge management system in our organization. In addition, I expect that steps to design its scheme as well as this scheme itself can be applied when design a knowledge classification scheme at the other R&D institutes and enterprises.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.529-538
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• 1993

A stochastic way using continuous time Markov process is presented to model the one-dimensional nuclide transport in fractured rock matrix as an extended study for previous work ［1］. A nuclide migration model by the continuous time Markov process for single planar fractured rock matrix, which is considered as a transient system where a process by which the nuclide is diffused into the rock matrix from the fracture may be no more time homogeneous, is compared with a conventional deterministic analytical solution. The primary desired quantities from a stochastic model are the expected values and variance of the state variables as a function of time. The time-dependent probability distributions of nuclides are presented for each discretized compartment of the medium given intensities of transition. Since this model is discrete in medium space, parameters which affect nuclide transport could be easily incorporated for such heterogeneous media as the fractured rock matrix and the layered porous media. Even though the model developed in this study was shown to be sensitive to the number of discretized compartment showing numerical dispersion as the number of compartments are decreased, with small compensating of dispersion coefficient, the model agrees well to analytical solution.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.539-547
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• 1993

A stochastic approach using continuous time Markov process is presented to model the one-dimensional nuclide transport in fractured rock media as a further extension for previous works［1-3］. Nuclide transport of decay chain of arbitrary length in the single planar fractured rock media in the vicinity of the radioactive waste repository is modeled using a continuous time Markov process. While most of analytical solutions for nuclide transport of decay chain deal with the limited length of decay chain, do not consider the case of having rock matrix diffusion, and have very complicated solution form, the present model offers rather a simplified solution in the form of expectance and its variance resulted from a stochastic modeling. As another deterministic way, even numerical models of decay chain transport, in most cases, show very complicated procedure to get the solution and large discrepancy for the exact solution as opposed to the stochastic model developed in this study. To demonstrate the use of the present model and to verify the model by comparing with the deterministic model, a specific illustration was made for the transport of a chain of three member in single fractured rock medium with constant groundwater flow rate in the fracture, which ignores the rock matrix diffusion and shows good capability to model the fractured media around the repository.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.337-344
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• 1994

Enthalpy rise hot channel factor($F_{\Delta{H}}$$^{N}$) is one of the most limiting constraints in determining the fuel loading pattern(LP) for PWR's. In order to enhance the LP design flexibility without any changes of not only basic fuel specifications but also Technical Specifications and Operation Procedures, we apply the enrichment zoning concept to Westinghouse designed PWR's to flatten the rod power distributions within the fuel assembly and thus to reduce $F_{\Delta{H}}$$^{N}$. Enrichment zoning is described that each assembly consists of two different enrichment fuels ; the lower enriched fuels are located in positions which are expected to have the higher rod power and vice versa for the higher enriched fuels. As a result of unit assembly calculations to flatten the rod power distribution within the assembly, the appropriate enrichment difference is found to be 0.3~0.4w/o. Through core depletion calculations for the 18-month cycle of Kori Unit 4, the $F_{\Delta{H}}$$^{N}$ behavior in core with the enrichment zoning concept is investigated. A comparison with the reference case without the enrichment zoning results in a reduction in $F_{\Delta{H}}$$^{N}$ of approximately 1.5%.TEX>H/$^{N}$ of approximately 1.5%.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.401-410
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• 1994

Validation study of two radionuclide migration models for single fracture developed in geologic medium the hydrodynamic dispersion diffusion model(HDDM) and the channeling dispersion diffusion model(CDDM), was studied by migration experiment of tracers through an artificial granite fracture on the labolatory scale. The tracers used were Uranine and Sodium lignosulfonate know as nonsorbing material. The flow rate ranged 0.4 to 1.5 cc/min. Related parameters for the models were estimated by optimization technique. Theoretical breakthrough curves with experimental data were compared. In the experiment, it was deduced that the surface sorption for both tracers did not play an important role while the diffusion of Uranine into the rock matrix turned out to be an important mass transfer mechanism. The parameter characterizing the rock matrix diffusion of each model agreed well The simulated result showed that the amount of flow rate could not tell the CDDM from the HDDM quantitatively. On the other hand, the variation of fracture length gave influence on the two models in a different degree. The dispersivity of breakthrough curve of the CDDM was more amplified than that of the CDDM when the fracture length was increased. A good agreement between the models and experimental data gave a confirmation that both models were very useful in predicting the migration system through a single fracture.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.140-148
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• 1994

A qualification test was performed for the iron removal chemical cleaning of the secondary side of nuclear steam generators at the selected temperature, 1$25^{\circ}C$, higher than the standard application temperature, 93$^{\circ}C$. The field cleaning condition for a nuclear unit was tested in a bench scale test loop including a SUS 316 stainless steel autoclave with one gallon capacity as a test vessel. The kinetics of sludge dissolution, corrosion of the secondary side materials and change of solvent chemistry were monitored. Test results indicated that more thorough cleaning was accomplished in less than half of the cleaning time required at 93$^{\circ}C$. And the total corrosions of the secondary side materials were found to be less than the values at 93$^{\circ}C$. While the solvent is recirculated and heated by an external chemical cleaning equipment for the conventional 93$^{\circ}C$ process, the secondary side is heated by the lateral heat of the primary coolant without the recirculation of the cleaning solution, and the solvent is mixed by vigorous boiling induced by periodic ventilation for the high temperature process. The requirement that the reactor coolant pumps should be running during the cleaning operation is the major disadvantage of the high temperature process which also should be considered when chemical cleaning is planned for steam generators under operation.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.859-869
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• 1995

The present study is to understand the physical phenomena anticipated during the accident with RHR loss under mid-loop operation in a PWR and, at the same time, to examine the prediction capability of RELAP5/MOD3.1 on such an accident, by simulating an integral test relevant to this accident for reliable analysis in an actual PWR. The selected experiment, i.g. BETHSY Test 6.9a, represents the configuration with the pressurizer manway open and steam generators unavailable during the accident. Accordingly, the results of this ok are sure to contribute to understanding both the key events as well as the sensitive parameters, anticipated in the accident, for validity of the actual analysis. In the simulation result overall behavior as well as major phenomena observed in the experiment have been predicted reasonably by RELAP5/MOD3.1, however, the problem associated with enormous computing time .due to small time step size has been encountered. Besides, the code prediction of higher swollen level in the pressure vessel has given rise to overestimation of both pressurizer level and RCS pressure. Subsequently, overprediction of the break flow through the manway has led to earlier core uncovery than that in the experiment by about 400 seconds. As a whole, it is demonstrated from both the experiment and the analysis that gravity feed has not been sufficient to recover the core level and thus additional forced feed has been necessary in this configuration.

• Nuclear industry
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• v.23 no.1 s.239
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• pp.64-78
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• 2003

한국원자력연구소에서는 과학기술부에서 주관하는 국제 협력 기반 조성 사업 과제의 일환으로 한국-핀란드 양국간 원자력 협력 증진을 위한 프로젝트를 수행하고 있다. 특히 이 연구에서는 방사성 폐기물 관리와 관련된 양국간 이해 증진과 향후 협력을 모색하기 위한 방안을 수립하고자 하였다. 본 연구에서는 이와 같은 관점에서 세계 최초로 사용후 핵연료 영구 처분장 부지를 확보하고 우리나라와 지질 조건이 유사한 결정질 암반에 신규로 심지층 처분 연구 실증 시설인 온칼로(Onkalo) 프로젝트를 계획하고 있는 핀란드의 방사성 폐기물 관리기관인 POSIVA 등과 관련 협력 기관, 정부 기관 등과 함께 향후 구체적인 협력 방안을 모색하고, 핀란드의 사용후 핵연료 직접 처분 연구사업 계획을 벤치 마킹하여 2003년도에 시작하는 국내 고준위 방사성 폐기물 처분 연구 과제 계획 수립에 도움을 주고자 하였으며, 이와 병행하여 핀란드 신규 원전 사업과 관련된 국내 산업체의 참여 가능성을 타진해 보고자 하였다.

• Nuclear industry
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• v.28 no.3
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• pp.50-59
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• 2008

미국의 원자력발전소는 당초 인가된 운전 기간은 40년이었으나 지금까지 많은 발전소가 운전 인가 갱신에 의해 운전 기간을 20년 연장하고 있다. 한편 일본에서는 발전소의 장기 운전을 가정한 고경년화 대책을 수립하고 있으며, 전력 회사는 고경년화 기술 평가 등 보고서를 작성하여 국가의 평가를 받고 있다. 이 분석에서는 경년 열화(經年劣化) 대책 상황을 조사하여 미.일 양국을 비교하였다. 그 결과 미국과 일본의 진행 방법, 배경, 노력(努力) 및 심사 기간 등에 다른 면이 있으나 미.일 모두 60년간의 장기 운전을 예정한 기기 등의 건전성 확인을 목적으로 하고 있는 면에서는 다름이 없다. 또, 원자력안전시스템연구소(INSS : Institute of Nuclear Safety System)의 해외 부적절한 데이터베이스를 이용한 경향 분석을 하고, 운전 인가 갱신신청의 유무가 기기 등의 경년 열화에 미치는 영향에 대해서분석 평가를 했다. 그 결과 인가 갱신 미신청 Unit에 경과년수의 증가에 수반하여 경년 열화 과실 발생 건수가 증가되는 향이 있는 것을 알았다. 이에 의해 미국의 인가 갱신 제도가 발전소의 고경년화 대응에 유효하게 기능하고 있다고 생각되어, 동등한 제도를 운용하고 있는 일본의 고경년화 대책의 유효성을 시사하는 것이었다.

• 기계와재료
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• v.21 no.3
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• pp.42-51
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• 2009

원자력/화력 발전소에서 사용중인 터빈출력제어장치(turbine power control device)는 유압 서보액추에이터(hydraulic servo actuator)로 구동하는 특수 스팀 밸브(steam valve)로서 터빈의 속도를 제어하고 스팀을 차단하는 기능이 있다. 대형 발전기(500~1000Mw)를 구동하여 양질의 전기를 생산하기 위해서는 발전기에 연결된 고압 및 저압터빈에 최적량의 스팀을 공급하여야 하고, 고속(화력 3600 rpm, 원자력 1800 rpm)으로 회전하는 터빈이나 스팀계통에 이상이 발생할 경우 터빈의 과속(over speed) 방지를 위하여, 즉시 터빈으로 공급되는 스팀을 차단하여 터빈을 보호해야 한다. 따라서 터빈의 속도제어와 계통의 스팀 량을 감시하여 차단하는 발전소의 특수 밸브의 신뢰성확보기술이 요구된다. 특히 원자력발전소의 경우 핵연료교환주기(약 24개월)에 밸브들을 정비 또는 교체하고 있어 이때마다 시스템과 매칭(튜닝)기술이 요구되었다. 본 연구에서는 전량 수입에 의존했던 원자력/화력 발전소의 특수 밸브인 터빈출력제어장치의 국산화 개발과 신뢰성확보기술 효과에 대하여 논하였다.