• 에너지공학
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• 제8권2호
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• pp.213-223
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• 1999

최근 원자력발전소의 안전계통 소프트웨어는 신뢰성을 향상시키기 위해 ANSI/IEEE-ANS-7-4.3.2-1982 기준에 따라 확인 및 검증이 이루어지고 있다. 이 규정은 안전관련 소프트웨어가 정적 및 동적 환경에서 시험되어야 한다고 요구하고 있다. 부적절노심냉각감시계통의 경우에 정적시험 절차 및 관련기술들은 개발되었으나 동적시험 절차 및 관련기술들은 개발되지 않았다. 따라서, 본 논문에서는 미개발된 기술들을 논의하고, 동적시험 절차와 시험 입력자료 생성 프로그램을 제안한다. 이 프로그램의 성능은 울진 3,4호기 최종 안전성 분석 보고서의 사고해석 결과를 사용하여 확인하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.862-868
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• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.

• Journal of information and communication convergence engineering
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• 제7권2호
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• pp.107-112
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• 2009

Renewable energy resources will be an increasingly important part of power generation in the new millennium. Besides assisting in the reduction of the emission of greenhouse gases, they add the much needed flexibility to the energy resource mix by decreasing the dependence on fossil fuels. Due to their modular characteristics, ease of installation and because they can be located closer to the user, PV system have great potential as distributed power source to the utilities. In this paper, a dc-de power converter scheme with the push-pull based technology is proposed to apply for solar power system which has many features such as high efficiency, stable output, and low acoustic noises, DC-DC converter is used in proposed topology has stable efficiency curve at all load range and very high efficiency characteristics. This paper presents the design of a single-phase photovoltaic inverter model and the simulation of its performance.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.546-551
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• 2010

The objective of this research is to estimate the crack location and size of a carbon steel pipe by using a laser ultrasound guided wave for the wall thinning evaluation of an elbow. The wall thinning of the carbon steel pipe is one of the most serious problems in nuclear power plants, especially the wall thinning of the carbon steel elbow caused by Flow-Accelerated Corrosion (FAC). Therefore, a non-destructive inspection method of elbow is essential for the nuclear power plants to operate safely. The specimens used in this study were carbon steel elbows, which represented the main elements of real nuclear power plants. The shape of the wall thinning was an oval with a width of 120mm, a length of 80mm, and a depth of 5mm. The L(0,1) and L(0,2) modes variation of the ultrasound guided wave signal is obtained from the response of the laser generation/air-coupled detection ultrasonic hybrid system represent the characteristics of the defect. The trends of these characteristics and signal processing were used to estimate the size and location of wall thinning.

• 대한전기학회논문지
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• 제40권6호
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• pp.551-561
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• 1991

The probabilistic generation simulation plays a key role in power system expansion and operational planning especially for the calculation of expected energy, loss of load probaility and unserved energy expected. However, it is crucial to develop a probabilistic generation simulation algorithm which gives sufficiently precise results within a reasonable computation time. In a previous paper, we have proposed an efficent method using Fast Hartley Transform in convolution process for considering the thermal and nuclear units. In this paper, a method considering the scheduling of pumped-storage plants and hydro plants with energy constraint is proposed. The method also adopts FHT techniques. We improve the model to include multi-state and multi-block generation. The method has been applied for a real size model system.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1993년도 하계학술대회 논문집 A
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• pp.159-162
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• 1993

A new approach using fuzzy dynamic programming is proposed for the flexible long-term generation mix under uncertain circumstances. A characteristic feature of the presented approach is that not only fuzziness in fuel and construction cost. load growth and reliability but also many constraints of generation mix can easily be taken into account by using fuzzy dynamic programming. The method can accommodate arbitrary shape of membership function as well as the operation of pump-generator. And so more realistic solution can be obtained. The effectiveness of the proposed approach is demonstrated by the best generation mix problem of KEPCO-system which contains nuclear, coal, LNG, oil and pump-generator hydro plant in multi-years.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.43-47
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• 2003

The performance of the Radwaste System is measured in terms of generation of waste volumes, the release of radioactive materials to the environment and the occupational radiation exposure to workers. Based on our design and operating experience from PWR plants, various design goals for liquid radwaste system were developed to improve system performance. It has been making continuous effort to develop the advanced liquid radwaste processing technology for new PWR plants since 1998. The primary goal of this effort was to obtain better performance and to design a more economical liquid radwaste system. This paper describes lesson learned experience from design of the liquid radwaste system in Korea Nuclear Power Plants.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2011년도 춘계학술대회 초록집
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• pp.136.2-136.2
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• 2011

In this work, we analyzed correlations between life-cycle greenhouse gas (GHG) emissions and life-cycle cost of energy resources. Energy resources studied in this paper include coal, natural gas, nuclear power, hydropower, geothermal energy, wind power, solar thermal energy, and solar photovoltaic energy, and all of them are used to generate electricity. We calculated the mean values, ranges of maximum minus minimum values, and ranges of 90% confidence interval of life-cycle GHG emissions and life-cycle cost of each energy resource. Based on the values, we plotted them in two dimensional graphs to analyze a relationship and characteristics between GHG emissions and cost. Besides, to analyze the technical maturity, the GHG emissions and the range of minimum and maximum values were compared to each other. For the electric generation, energy resources are largely inverse proportional to the GHG emission and the corresponding cost.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.155.2-155
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• 2002

A visual environment for system analysis codes (hereinafter called "ViSA") has been developed to support code users in their input preparations, code executions, and output interpretations. ViSA provides a more convenient way for base input data generation and modification on a user-friendly basis. It also provides on-line graphical displays to give an in-depth understanding of transient thermal-hydraulic behaviors in nuclear power plants. This paper presents the main features of ViSA.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.