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    Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

    Source: Journal of Nuclear Engineering and Radiation Science:;2022:;volume( 008 ):;issue: 003::page 31403-1
    Author:
    Saraswat
    ,
    S. P.;Ray
    ,
    D.;Mishra
    ,
    G.;Yadav
    ,
    D.;Bhadouria
    ,
    V. S.;Munshi
    ,
    P.;Allison
    ,
    C.
    DOI: 10.1115/1.4052014
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: The economic simplified boiling water reactor (ESBWR) is a boiling water nuclear reactor of generation III+. The U.S. Nuclear Regulatory Commission (NRC) approved the ESBWR design as the world's best light-water nuclear reactor in 2014. It has the lowest core damage frequency (industry standard indicator of safety) of any Generation III or III+ reactor. It can cool automatically for more than seven days without using electricity or human intervention. During the operation, the ESBWR is designed to produce electricity while emitting almost no greenhouse gases. The energy generated by an ESBWR will prevent the emission of approximately 7.5 million metric tons of CO2 per year compared to standard electricity production on the U.S. grid. The analysis present in this paper aimed to characterize the thermal-hydraulic simulations of full-scale ESBWR design. The analysis presented will help in recognizing the improvement needed in the reactor design and its passive safety systems. The analysis is performed for normal steady-state and postulated design basis accident scenarios (break in one of the reactor main-steam lines inside the contentment). The simulation results obtained by the code REALP/SCDAPSIM/MOD3.4 are compared with the TRACG04 and MELCOR2.1 code results to determine the code predictability and accuracy under accident conditions of the newly proposed design of the ESBWR nuclear reactor. A parametric study is also carried out with different in-containment break sizes to determine code results' sensitive nature due to natural circulation flow phenomena, which are highly unstable. The analysis shows that the code has produced reasonably accurate results under the tested system thermal-hydraulic conditions compared with the TRACG04 and MELCOR2.1 code simulation results. This agreement between the three codes simulation results demonstrated that the publicly available RELAP5 code model has adequate features to simulate the ESBWR thermal-hydraulics. It has been also demonstrated that for the postulated accident conditions, the design of passive safety systems is capable to capture the accident progression without any active power.
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      Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

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    http://yetl.yabesh.ir/yetl1/handle/yetl/4287298
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    • Journal of Nuclear Engineering and Radiation Science

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    contributor authorSaraswat
    contributor authorS. P.;Ray
    contributor authorD.;Mishra
    contributor authorG.;Yadav
    contributor authorD.;Bhadouria
    contributor authorV. S.;Munshi
    contributor authorP.;Allison
    contributor authorC.
    date accessioned2022-08-18T13:01:48Z
    date available2022-08-18T13:01:48Z
    date copyright5/26/2022 12:00:00 AM
    date issued2022
    identifier issn2332-8983
    identifier otherners_008_03_031403.pdf
    identifier urihttp://yetl.yabesh.ir/yetl1/handle/yetl/4287298
    description abstractThe economic simplified boiling water reactor (ESBWR) is a boiling water nuclear reactor of generation III+. The U.S. Nuclear Regulatory Commission (NRC) approved the ESBWR design as the world's best light-water nuclear reactor in 2014. It has the lowest core damage frequency (industry standard indicator of safety) of any Generation III or III+ reactor. It can cool automatically for more than seven days without using electricity or human intervention. During the operation, the ESBWR is designed to produce electricity while emitting almost no greenhouse gases. The energy generated by an ESBWR will prevent the emission of approximately 7.5 million metric tons of CO2 per year compared to standard electricity production on the U.S. grid. The analysis present in this paper aimed to characterize the thermal-hydraulic simulations of full-scale ESBWR design. The analysis presented will help in recognizing the improvement needed in the reactor design and its passive safety systems. The analysis is performed for normal steady-state and postulated design basis accident scenarios (break in one of the reactor main-steam lines inside the contentment). The simulation results obtained by the code REALP/SCDAPSIM/MOD3.4 are compared with the TRACG04 and MELCOR2.1 code results to determine the code predictability and accuracy under accident conditions of the newly proposed design of the ESBWR nuclear reactor. A parametric study is also carried out with different in-containment break sizes to determine code results' sensitive nature due to natural circulation flow phenomena, which are highly unstable. The analysis shows that the code has produced reasonably accurate results under the tested system thermal-hydraulic conditions compared with the TRACG04 and MELCOR2.1 code simulation results. This agreement between the three codes simulation results demonstrated that the publicly available RELAP5 code model has adequate features to simulate the ESBWR thermal-hydraulics. It has been also demonstrated that for the postulated accident conditions, the design of passive safety systems is capable to capture the accident progression without any active power.
    publisherThe American Society of Mechanical Engineers (ASME)
    titleThermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design
    typeJournal Paper
    journal volume8
    journal issue3
    journal titleJournal of Nuclear Engineering and Radiation Science
    identifier doi10.1115/1.4052014
    journal fristpage31403-1
    journal lastpage31403-13
    page13
    treeJournal of Nuclear Engineering and Radiation Science:;2022:;volume( 008 ):;issue: 003
    contenttypeFulltext
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