• 제목/요약/키워드: Nuclear power plant structure concrete

검색결과 58건 처리시간 0.029초

원전 구조물의 건조수축 저감을 위한 실험적 연구 (Experimental study to improve drying shrinkage durability performance of Nuclear Power Plant Structure)

  • 임상준;이병수;방창준
    • 한국건축시공학회:학술대회논문집
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    • 한국건축시공학회 2012년도 추계 학술논문 발표대회
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    • pp.205-206
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    • 2012
  • In general, nuclear power plant concrete structure's performance has been very good with the majority of identified problems initiating during construction and corrected at that time. This study is experiments to improve drying shrinkage using glycol ether-based material for the durability of nuclear power plants. Thus, this study evaluated the obtained data from a mock up test for the practical use of concrete containing glycol ether. According to the results of this study, the concrete showed resistance performance of around 40% to drying shrinkage.

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월성 원자력발전소 격납건물의 극한내압평가 (Evaluation of Ultimate Pressure Capacity of Wolsong Containment Structure)

  • 곽효경;김재홍;김선훈;정연석
    • 한국전산구조공학회:학술대회논문집
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    • 한국전산구조공학회 2005년도 춘계 학술발표회 논문집
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    • pp.183-189
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    • 2005
  • Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

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개구 저감률에 의한 원전 SC벽체의 내력 평가 (Evaluation of Structural Capacity of SC Walls in Nuclear Power Plant accounting for the Area Lost to Openings)

  • 정철헌;정래영;문일환;이정휘
    • 대한토목학회논문집
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    • 제33권6호
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    • pp.2181-2193
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    • 2013
  • 지금까지 수행된 개구부가 존재하는 벽체에 대한 연구는 대부분 RC 벽체에 대해서 수행되었으며, SC(Steel plate Concrete) 벽체에 설치되는 개구부에 대한 연구는 수행된 예가 적다. 최근에 국내에서 개발된 SC 벽체는 원전구조물에 일부 적용되고 있지만, 관련 설계기준인 KEPIC-SNG에서도 개구부를 갖는 SC 벽체에 대한 설계법은 명확하게 정의되지 않았다. 본 연구에서는 원전구조물내 벽체에 설치되는 SC 벽체를 대상으로 개구 저감률이 구조내력에 미치는 영향을 평가하였다. 개구 저감률을 고려한 구조내력 평가 결과는 실험 및 수치해석 결과와 비교분석하였다.

스마트 구조물용 광섬유 격자센서의 원전격납건물 적용 실험 연구 (Study on the Fiber Bragg Grating Smart Sensors for Containment Structure in Nuclear Power Plant)

  • 김기수;송영철;방기성;윤덕중
    • 한국콘크리트학회:학술대회논문집
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    • 한국콘크리트학회 2004년도 춘계 학술발표회 제16권1호
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    • pp.412-415
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    • 2004
  • This study was performed to verify the behaviors of fiber Bragg grating (FBG) sensors attached to the containment structure in the nuclear power plant as a part of structural integrity test which demonstrates that the structural response of the non-prototype primary containment structure is within predicted limits plus tolerances when pressurized to $115\%$ of containment design pressure, and that the containment does not sustain any structural damage.

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Safety Analysis of Concrete Treatment Workers in Decommissioning of Nuclear Power Plant

  • Hwang, Young Hwan;Kim, Si Young;Lee, Mi-Hyun;Hong, Sang Beom;Kim, Cheon-Woo
    • 방사성폐기물학회지
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    • 제20권3호
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    • pp.349-356
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    • 2022
  • Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers' radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers' radiation exposure, and improved the treatment process's efficiency. The shielding package's effect on workers' radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

Development of Micro-Blast Type Scabbling Technology for Contaminated Concrete Structure in Nuclear Power Plant Decommissioning

  • Lee, Kyungho;Chung, Sewon;Park, Kihyun;Park, SeongHee
    • 방사성폐기물학회지
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    • 제20권1호
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    • pp.99-110
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    • 2022
  • In decommissioning a nuclear power plant, numerous concrete structures need to be demolished and decontaminated. Although concrete decontamination technologies have been developed globally, concrete cutting remains problematic due to the secondary waste production and dispersion risk from concrete scabbling. To minimize workers' radiation exposure and secondary waste in dismantling and decontaminating concrete structures, the following conceptual designs were developed. A micro-blast type scabbling technology using explosive materials and a multi-dimensional contamination measurement and artificial intelligence (AI) mapping technology capable of identifying the contamination status of concrete surfaces. Trials revealed that this technology has several merits, including nuclide identification of more than 5 nuclides, radioactivity measurement capability of 0.1-107 Bq·g-1, 1.5 kg robot weight for easy handling, 10 cm robot self-running capability, 100% detonator performance, decontamination factor (DF) of 100 and 8,000 cm2·hr-1 decontamination speed, better than that of TWI (7,500 cm2·hr-1). Hence, the micro-blast type scabbling technology is a suitable method for concrete decontamination. As the Korean explosives industry is well developed and robot and mapping systems are supported by government research and development, this scabbling technology can efficiently aid the Korean decommissioning industry.

Safety assessment of nuclear fuel reprocessing plant under the free drop impact of spent fuel cask and fuel assembly part I: Large-scale model test and finite element model validation

  • Li, Z.C.;Yang, Y.H.;Dong, Z.F.;Huang, T.;Wu, H.
    • Nuclear Engineering and Technology
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    • 제53권8호
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    • pp.2682-2695
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    • 2021
  • This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

전단벽 구조물의 진동대 시험결과와 유한요소 내진해석결과 비교 (Comparison of Shaking Table Test Results and Finite Element Seismic Analysis Results of Shear Wall Structures)

  • 김기현;장영선
    • 한국지진공학회논문집
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    • 제25권3호
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    • pp.137-144
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    • 2021
  • In this study, the seismic safety of nuclear power plant structures is evaluated and verified by performing a vibration test on a relatively simple shear wall structure. The shear walls are the prominent members of nuclear power plants and resist the seismic load. The shear wall structure is designed and manufactured to perform shaking table tests and is used to increase the accuracy of the analytical method by comparing them with the numerical analysis results. Different results will be checked and more efficient application methods will be studied depending on the method of designing reinforced concrete structures.