• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2004년도 가을 학술발표회 논문집
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• pp.245-253
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• 2004

There has been recent demand for extending the life of age-degraded structures and equipment by such techniques as diagnosis, maintenance, safety assessment, and estimating residual life on iron-making plants and hydraulic, thermal, and nuclear power plants. These techniques take into account comprehensive scenarios that may cause malfunction and structural damage and allow an assessment of risk based on the likely scenarios. In particular the safety assessment and residual life estimation of age-degraded ships and equipment facilities require consideration of various factors such as mechanical and thermal stresses, corrosion, hardness, load variation due to changes of operating condition, crack generation and strength reduction of material by fatigue. In this study, a detail thermal stress analysis, one of useful techniques of safety assessment and maintenance, is performed on a blast furnace by using general FEM code (MSC/NASTRAN).

• 한국농촌생활과학회지
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• 제5권1호
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• pp.57-66
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• 1994

The purpose of the study was to analyze the family structure in rural Korea systematically and comprehensively according to the broad concept. The data was collected from 810 rural households by interview method with questionnaire. For the analysis, family structure was divided into aspects of static structure and dynamic structure. The static structure was constructed by two components of demographic structure and typological structure. The dynamic structure was also constructed by three components of decision making structure, role structure, and dynamic relationship structure of family members. In demographic structure, family size was 4.1 persons, families, with more female were 35.2%, and families with elder husband than wife were 82.5%, In the typological structure, nuclear family type with two-generation was predominant. In dynamic structure, role structure was autonomic type while conjugal power structure was compounded type with autonomic, syncratic, and husband-dominant type.

• Journal of Mechanical Science and Technology
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• 제15권9호
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• pp.1274-1280
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• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• 한국막학회:학술대회논문집
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• 한국막학회 1994년도 심포지움시리즈 Jan-94 기체분리막 기술 및 응용
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• pp.23-52
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• 1994

The development if separation technology is an important research subject as is clear from its role in the Japanese government's research abd development program for basic technology for the next generation (1981~1990). Japan is poor not only in mineral resources but also in energy resources and if a sudden change occurs in oil producing facility or an accident occurs in a nuclear power plant, then energy policy must undergo changes and economic foundations may collapse. Japan has already experienced this. Although, oil prices are stable at present and Japan can import oil at low cost due to the yen appreciation, Japan needs to promote development work for any new energy crisis that may come in the future. This has been the motive for gas separation membrane development in Japan.

• 한국안전학회지
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• 제16권3호
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• pp.7-11
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• 2001

In this study, specimens were classified four welded specimens and a base metal to investigate fatigue life and crack growth rate of pure titanium welding materials, and Ti was used in turbine equipment of nuclear power generation, etc. The summarized results are as follows; 1) Specimen-2 was bigger 712% than base metal, when it was compared with other welding materials, 2) As the result of specimens data, specimen-2 crack behavior rate res lower 30 times than base metal, and so total fracture life was very influenced by it, 3) Notch tip of Specimen-2 was offsetted 6.7mm from boundary H.A.Z, and if formed 25% in total fracture length, 4) As the considering of da/dN and $\Delta$K, Paris' law is incongruous in this study, because fro inclines nsf on one date.

• 한국안전학회지
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• 제11권2호
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• pp.52-59
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• 1996

This paper presents a new approach for assessing the dynamic reliability in a complex system such as a nuclear power plant. The method is applied to a dynamic analysis of the potential accident sequences which may occur during mid-loop operation. Mid-loop operation is defined as an operation to make RCS water level below the top of the flow area of the hot legs at the junction with the reactor vessel for repairs and maintenance of steam generators and reactor coolant pumps for a specific time. The Idea behind this approach consists of both the use of the concept of the performance achievement/requirement correlation and of a dynamic event tree generation method. The assessment of the system reliability depends on the determination of both the required performance distribution and the achieved performance distribution. The quantified correlation between requirement and achievement represents a comparison between two competing variables. It is demonstrated that this method is easily applicable and flexible in that it can be applied to any kind of dynamic reliability problem.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 2004년도 추계 학술발표회 논문집
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• pp.277-282
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• 2004

본 연구에서는 환경경영의 중요한 수단으로 대두되고 있는 전과정분석(Life Cycle Analysis; LCA)을 통해 원자력 발전기술을 평가한 주요 사례의 분석과정과 결과들을 비교하였다. 이는 선진국들이나 글로벌 선진기업들에서 채택하기 시작한 글로벌 환경친화성의 입증 요구 추세 속에서, 모든 재화의 생산에 전력은 필수 투입물이라는 점에서 그 중요성을 감안하여 수행되었다. 본 연구에서는 국내 전력의 40% 이상 공급을 담당하고 있는 원자력발전 기술을 중심으로 비교하였다. 우리나라의 대표적인 연구를 간 비교 결과 그 값의 차이는 최대 10＾5 정도에 이르러, 이는 외국의 경우 10＾1 수준의 차이를 보이는 데 비해 다소 많은 것으로 조사되었다. 이러한 차이는 사례에 따라 모형의 단순화 정도에 기인한 것으로 판단되며, 이는 전력의 타산업 기반성이나 국내 전력LCA 분석능력의 함양이 중요하다는 점에 비추어 시급히 개선되어야 할 것이라 제안한다.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2006년도 학술발표회 논문집
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• pp.637-644
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• 2006

Acceleration time history used in the seismic analysis of nuclear porter plant structure should envelop a target power spectral density (PSD) function in addition to design response spectrum. Current regulation guide defines the target PSD function only for the U.S. URC RG 1.60 Design Response Spectrum. This paper proposes a technical scheme to obtain the target PSD function compatible with generally defined design response spectrum. The scheme includes the methodology for design-spectrum compatible motion history in order to minimize the variation of the derived target PSD function. The PSD calculation procedure follows simple and practical methods allowed within regulation. Effectiveness of the proposed scheme is identified through an example problem. The design response spectrum In the example is based on U.S. NRC RG 1.60 but amplifies the spectral acceleration amplitudes above 9Hz. The target PSD function with little variation can be constructed with the reduced time history ensemble.

• 에너지공학
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• 제11권4호
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• pp.370-379
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• 2002

우리나라에서 전력 공급의 40％ 이상을 차지하는 원자력 발전의 중요성은 더 말할 나위가 없으나 방사성 폐기물 관리 시설의 운영 지연이 안정적인 원자력 발전에 저해 요소로 작용할 우려가 있다. 정부 당국의 정책적 배려, 방사성폐기물 관리 사업 주체의 노력, 관련 기술의 축적에도 불구하고 방사성 폐기물관리 사업은 가시적인 진전이 없는 상태이다. 그러므로 우리나라는 과거 어느 때보다도 더욱 효율적인 정책 기획과 집행으로 방사성폐기물을 관리해야 하는 시점에 처해 있다고 할 수가 있다. 본 연구는 이러한 배경에서 과거의 정책을 평가하고 미래를 위한 새로운 방사성폐기물 관리 정책을 모색한 결과 연구 개발. 부지 선정. 시설 건설. 운영, 폐쇄 후 감시에 이르는 전 과정에서 정책의 최우선 기조를 관리시설의 장기적인 안전성 확보에 두고, 안전 규제 기관의 참여 하에 공개적으로 입지를 추진해야 하며, 방사성폐기물 관리 사업과 이해 관계가 상충되지 아니하는 중립적인 위치의 공공법인체가 범국가적 차원에서 연구개발 조정 기능을 수행하여야 한다는 요지의 결론을 도출하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.513-518
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• 1996

This report presents a demonstration of application of realistic evaluation methodology to a posturated cold leg large break LOCA in a Westinghouse three-loop pressurized water reactor with 17$\times$17 fuel. The new method of this analysis can be divided into three distinct step: 1) Best Estimate Code Validation and Uncertainty Quantification 2) Realistic LOCA Calculation 3) Limiting Value LOCA Calculation and Uncertainty Combination RELAP5/MOD3/K ［1］, which was improved from RELAP5/MOD3.1, and CONTEMPT4/MOD5 code were used as a best estimate thermal-hydraulic model for realistic LOCA calculation. The code uncertainties which will be determined in step 1) were quantified already in previous study ［2］, and thus the step 2) and 3) for plant application were presented in this paper. The application uncertainty parameters are divided into two categories, i.e. plant system parameters and fuel statistical parameters. Single parameter sensitivity calculations were performed to select system parameters which would be set at their limiting value in Limiting Value Approach (LVA) calculation. Single run of LVA calculation generated 27 PCT data according to the various combinations of fuel parameters and these data provided input to response surface generation. The probability distribution function was generated from Monte Carlo sampling of a response surface and the upper 95$^{th}$ percentile PCT was determined. Break spectrum analysis was also made to determine the critical break size. The results show that sufficient LOCA margin can be obtained for the demonstration NPP.