• 제목/요약/키워드: Nuclear Engineering

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Innovative Nuclear Power Plant Building Arrangement in Consideration of Decommissioning

  • Choi, Won-Jun;Roh, Myung-Sub;Kim, Chang-Lak
    • Nuclear Engineering and Technology
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    • 제49권3호
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    • pp.525-533
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    • 2017
  • A new concept termed the Innovative Nuclear Power Plant Building Arrangement (INBA) strategy is a new nuclear power plant building arrangement method which encompasses upfront consideration of more efficient decommissioning. Although existing decommissioning strategies such as immediate dismantling and differed dismantling has the advantage of either early site restoration or radioactive decommissioning waste reduction, the INBA strategy has the advantages of both strategies. In this research paper, the concept and the implementation method of the INBA strategy will be described. Two primary benefits will be further described: (1) early site restoration; and (2) radioactive waste reduction. Several other potential benefits will also be identified. For the estimation of economic benefit, the INBA strategy, with two primary benefits, will be compared with the immediate dismantling strategy. The effect of a short life cycle nuclear power plant in combination with the INBA strategy will be reviewed. Finally, some of the major impediments to the realization of this strategy will be discussed.

Solidification of Spent Resin for Safe Storage

  • Kim, Won-Seok;Kang, Jaehyuk;Jung, Taeho;Jang, Mi-suk;Jeon, Jong-seon;Um, Wooyong
    • 한국방사성폐기물학회:학술대회논문집
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    • 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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    • pp.204-205
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    • 2018
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The development of high fidelity Steam Generator three dimensional thermal hydraulic coupling code: STAF-CT

  • Zhao, Xiaohan;Wang, Mingjun;Wu, Ge;Zhang, Jing;Tian, Wenxi;Qiu, Suizheng;Su, G.H.
    • Nuclear Engineering and Technology
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    • 제53권3호
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    • pp.763-775
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    • 2021
  • The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

Hybrid medium model for conjugate heat transfer modeling in the core of sodium-cooled fast reactor

  • Wang, X.A.;Zhang, Dalin;Wang, Mingjun;Song, Ping;Wang, Shibao;Liang, Yu;Zhang, Yapei;Tian, Wenxi;Qiu, Suizheng;Su, G.H.
    • Nuclear Engineering and Technology
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    • 제52권4호
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    • pp.708-720
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    • 2020
  • Core-wide temperature distribution in sodium-cooled fast reactor plays a key role in its decay heat removal process, however the prediction for temperature distribution is quite complex due to the conjugate heat transfer between the assembly flow and the inter-wrapper flow. Hybrid medium model has been proposed for conjugate heat transfer modeling in the core. The core is modeled with a Realistic modeled inter-wrapper flow and hybrid medium modeled assembly flow. To validate present model, simulations for a three-assembly model were performed with Realistic modeling, traditional porous medium model and hybrid medium model, respectively. The influences of Uniform/Non-Uniform power distribution among assemblies and the Peclet number within the assembly flow have been considered. Compared to traditional porous medium model, present model shows a better agreement with in Realistic modeling prediction of the temperature distribution and the radial heat transfer between the inter-wrapper flow and the assembly flow.

Uncertainty quantification of PWR spent fuel due to nuclear data and modeling parameters

  • Ebiwonjumi, Bamidele;Kong, Chidong;Zhang, Peng;Cherezov, Alexey;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • 제53권3호
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    • pp.715-731
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    • 2021
  • Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.