• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.9 no.4
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• pp.145-152
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• 2009

Domestic Kori-1 MCR was partially modified in 2007 and will be renovated entirely in 2013. Digital devices partially replacing original analog devices have been introduced and standard alone computer systems such as SPDS have been integrated into the plant computer. Upgrading KSNP's MCR based on the ditalization is planned for 2015. However, the site engineers and operators are reluctant to the advanced systems. Therefore, a prototype for the KSNP's advanced MCR has been developed to increase the acceptance level of the operators and field engineers and also, to evaluate user interfaces and I&C architecture. For enhancing support of the operators' work, a P&ID based display system composed of multi-layers, which are linked through a context sensitive menu each other, has been adopted. The $1^{st}$ layer displays a simplified P&ID, the $2^{nd}$ layer control related diagrams such as controllers and logic diagrams, the $3^{rd}$ layer trends, etc. The end point view of MCR for KSNP is also suggested considering reliability and operability of the digital systems. Additionally, modernization strategies over the overhaul periods, that do not have much impact on operation and configuration efforts are suggested.

• 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%.

• Journal of the Korean Geographical Society
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• v.47 no.5
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• pp.755-774
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• 2012

The main objective of the study is to propose GIS-based methods to assess the population distribution criteria for undesirable facilities such as nuclear power plants. First of all, a review of the relevant criteria was conducted for the official documents compiled by such institutions as IAEA (International Atomic Energy Agency), U.S. NRC (Nuclear Regulatory Commission), and some national institutes including the Korea Institute of Nuclear Safety. It is informed from the review that the fundamental principle underlying the various criteria is to maximize the distance between a plant and the nearest population center. It is realized that two interrelated GIS-based techniques need to be devised to put the principle into practice; sophisticated ways of representing population distribution and identifying population centers. A dasymetric areal interpolation is proposed for the former and cell-based and area-based critical density methods are introduced. Grid-based population distributions at various spatial resolutions are created by means of the dasymetric areal interpolation. By applying the critical density methods to the gridded population distribution, some population centers satisfying the population size and density criteria can be identified. These methods were applied to the case of the Gori-1 nuclear power plant and their strengths and limitations were discussed. It was revealed that the assessment results could vary depending upon which method was employed and what values were chosen for various parameters. This study is expected to contribute to foster the applications of methods and techniques developed in geospatial analysis and modeling to the site selection and evaluation.

• Journal of the Korean Recycled Construction Resources Institute
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• v.6 no.2
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• pp.115-122
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• 2018

In this paper, the feasibility of recycling concrete waste as a method to reduce final disposal amount of wastes generated through decommissioning of nuclear power plant has been analyzed based on experimental results of existing literature. When recycled concrete waste was used as recycled aggregate, it was investigated through literature that the concrete strength decreased by 30~40% depending on the mixing ratio. It was also investigated that concrete with recycled aggregate can be used as a structural material when the quality of recycled aggregate is well managed since no significant problem was found. When recycled cement produced from concrete waste was used, the strength of concrete or mortar decreased considerably as the recycled cement content increased. Therefore, it can be concluded that concrete or mortar with recycled cement can be used as a filling material for final disposal of large radioactive waste rather than for structural use. This paper is expected to be useful for reduction on disposal volume and decommissioning cost for nuclear power plants such as Kori 1.

• Journal of the Korea Institute of Building Construction
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• v.15 no.4
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• pp.425-431
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• 2015

Since the construction of Kori 1 was completed in 1978, the construction projects for nuclear power plant are increasingly expanded into domestic and foreign sites. However, some of construction sites of nuclear power plant have the problems of process delay and cost loss due to lack of ability of risk management. The construction of reactor containment building in nuclear power plant is especially dotted with many risk factors because it needs professional skills and large-scale resources due to long duration compared with different construction phase. Therefore, it needs the study that analyzes risk factors expected in construction of reactor containment building and suggests way of stable performance of projects. So, this study assesses risk factors of construction of reactor containment building. For the objectives, this study uses survey for group of minority specialists of 36 experts. The risks of 24 factors is classified by criterions of process, cost, safety, and quality and the results of assessment is analyzed by analytic hierarchy process. As the results, the importance and priority of risk factors classified by each criterion were calculated and the applicability of analytic hierarchy process was identified to analyze risk factors of nuclear power plant construction. These will be baseline data for risk management in construction phase of reactor containment building.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.267-277
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• 2016

For the decommissioning of a nuclear power plant, a site characterization and final status survey require a site-specific suite of radionuclides that could potentially still be present in the site during or after the decontamination processes. The United States Nuclear Regulatory Commission (U.S. NRC) requires a Decommissioning Technical Base Document (DTBD) along with a Site Characterization and Historical Site Assessment (HSA) from the utility for decommissioning to proceed. Both the DTBD and HSA are preliminary components of the Radiological Site Survey investigation process and should be included in the final License Termination Plan (LTP) for site release and reuse consideration from the U.S. NRC and the utility company. This study reviews the United States Nuclear Power Plants (U.S. NPPs) decommissioning cases and is especially focused on the methodologies used for determining a site-specific suite of radionuclides before and during the site characterization and final status surveys. In 2017, Kori-1 will be ready for decommissioning and related preparations are ongoing, this review will help Korea to prepare regulatory guidelines and give technical background for the safe and successful decommissioning of NPPs.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.335-341
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• 1997

운전중 노심의 출력변화를 감시하는 노외계측기(Excore Detector)는 노내계측기(Incore Detector)를 통하여 측정되어진 축방향 출력편차(Axial Offset)를 이용하여 교정되고 있다. 노외 계측기의 전류와 축방향출력편차의 선형적인 관계를 가정하여 노내계측기로 최소한 4회 노심출력을 측정한후 최소자승법(Least Square Method)으로 비례상수들을 구하는 기존의 방법을 대신하여, 단순 노외계측기 교정법은 노내계측기로 1회 측정되어진 자료들을 이용하여 계측기 반응상수(Detector Response Factor)를 계산한 후 비례상수를 계산한다. 계측기반응상수는 2차원 중성자수송모델로부터 계산된 weighting factor와 3차원 확산이론으로부터 구한 노심출력을 이용하여 계산된다. 중성자수송계산은 (R-Z)와 (R-$ heta$)모델을 합성하여 3차원 weighting factor를 계산하므로 축방향 영향뿐만 아니라 집합체별 영향을 고려하였다. 또한 노심의 복잡한 구조로 인하여 근사적인 weighting (actor와 노심출력분포의 사용은 노외계측기의 전류와 계측기반응율의 불일치를 초래하며, 이를 해결하는 상수를 소개하여 보다 정확한 교정결과를 얻도록하였다. 이와 같은 방법을 고리 3호기 9, 10주기 전주기와 11주기초에 적용하여 노심의 연소분포, 냉각수의 온도분포, 노심의 연소도, 노심출력준위등에 대한 단순 노외계측기 교정법의 오차를 분석하여 최적의 노외계측기 교정모델을 제시하였다. 2차원 중성자수송모델 합성법에 의한 단순노외계측기 교정법은 2차원 (R-Z) 중성자수송모델보다 개선된 결과와 평균오차 0.179% 최대 오차 0.624%를 보여주고 있다.하면 조사 후의 조직안정성에도 크게 기여할 것으로 기대된다.EX>O가 각각 첨가된 경우, Ar-4vol.%H$_2$ 분위기보다 H$_2$분위기에서 소결했을 때 밀도가 더 높았다. 그러나, 결정립은 $UO_2$와 $UO_2$-Li$_2$O의 경우, 수소분위기에서 소결했을 때, (U,Ce)O$_2$와 (U,Ce)O$_2$-Li$_2$O에서는 Ar-4vol.%H$_2$분위기에서 소결했을 때 더욱 성장하였다.설명해 줄 수 있다. 넷째, 불규칙적이며 종잡기 힘들고 단편적인 것으로만 보이던 중간언어도 일정한 체계 속에서 변화한다는 사실을 알 수 있다. 다섯째, 종전의 오류 분석에서는 지나치게 모국어의 영향만 강조하고 다른 요인들에 대해서는 다분히 추상적인 언급으로 끝났지만 이 분석을 통 해서 배경어, 목표어, 특히 중간규칙의 역할이 괄목할 만한 것임을 가시적으로 관찰할 수 있 다. 이와 같은 오류분석 방법은 학습자의 모국어 및 관련 외국어의 음운규칙만 알면 어느 학습대상 외국어에라도 적용할 수 있는 보편성을 지니는 것으로 사료된다.없다. 그렇다면 겹의문사를 [-wh]의리를 지 닌 의문사의 병렬로 분석할 수 없다. 예를 들어 누구누구를 [주구-이-ν가] [누구누구-이- ν가]로부터 생성되었다고 볼 수 없다. 그러므로 [-wh] 겹의문사는 복수 의미를 지닐 수 없 다. 그러면 단수 의미는 어떻게 생성되는가\ulcorner 본 논문에서는 표면적 형태에도 불구하고 [-wh]의미의 겹의문사는 병렬적 관계의 합성어가 아니라 내부구조를 지

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.237-247
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• 1993

Activity ratio of two radioactive primary fission products which had sufficiently different half-lives was expressed as functions of cooling time and irradiation histories in which average burnup, irradiation time, cycle interval time and the dominant fissile material of the spent fuel were included. The gamma-ray spectra of 36 samples from 6 spent PWR fuel assemblies irradiated in Kori unit-1 reactor were obtained by a spectrometric system equipped with a high purity germanium gamma-ray detector. Activity ratio $^{l44}$Ce $^{l37}$Cs, analyzed from each spectrum, was used for the calculation of cooling time. The results show that the radioactive fission products $^{l44}$Ce and $^{l37}$Cs are considered as useful monitors for cooling time determination because the estimated cooling time by detection of activity ratio $^{l44}$Ce $^{l37}$Cs agreed well with the operator declared cooling time within relative difference of $\pm$5 % despite the low counting rate of the gamma-ray of $^{l44}$Ce (about 10$^{-3}$ count per second). For the samples with several different irradiation histories, the determined cooling time by modeled irradiation history showed good agreement with that by known irradiation history within time difference of $\pm$0.5 year. From this result, it would be expected to be possible to estimate reliably the cooling time of spent nuclear fuel without the exact information about irradiation history. The feasibility study on identification of and/or sorting out spent nuclear fuel by applying the technique for cooling time determination was also performed and the result shows that the detection of activity ratio $^{l44}$Ce $^{l37}$Cs by gamma-ray spectrometry would be usefully applicable to certify spent nuclear fuel for the purpose of safeguards and management in a facility in which the samples dismantled or cut from spent fuel assemblies are treated, such as the post irradiation examination facility.mination facility.