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    Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

    Source: Journal of Nuclear Engineering and Radiation Science:;2018:;volume( 004 ):;issue: 001::page 11010
    Author:
    Yetisir, Metin
    ,
    Hamilton, Holly
    ,
    Xu, Rui
    ,
    Gaudet, Michel
    ,
    Rhodes, David
    ,
    King, Mitch
    ,
    Andrew, Kittmer
    ,
    Benson, Ben
    DOI: 10.1115/1.4037818
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: The Canadian supercritical water-cooled nuclear reactor (SCWR) is a 2540 MWth channel-type SCWR concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. The fuel assembly is designed so that all in-core components exposed to high radiation fields (other than the pressure tube) are part of the fuel assembly, which is removed from the reactor core as part of the assembly after three operating cycles. This design feature significantly reduces the likelihood of component failures due to radiation damage. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 °C at the inlet, 625 °C at the outlet). These conditions lead to fuel cladding temperatures close to 800 °C. Because of the reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that axial ridging, a possible failure mechanism with collapsed fuel cladding, can be avoided if the cladding thickness is larger than 0.4 mm. Detailed numerical analysis showed that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This result was obtained with conservative assumptions, suggesting that the actual margin is greater. Hence, one of the design goals, the exclusion of the possibility of melting of the fuel, which is called the “no-core-melt” concept, seems attainable. However, this needs to be demonstrated more rigorously by removing the conservative assumptions in the analysis and performing supporting experimental work. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.
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      Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

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    http://yetl.yabesh.ir/yetl1/handle/yetl/4252636
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    contributor authorYetisir, Metin
    contributor authorHamilton, Holly
    contributor authorXu, Rui
    contributor authorGaudet, Michel
    contributor authorRhodes, David
    contributor authorKing, Mitch
    contributor authorAndrew, Kittmer
    contributor authorBenson, Ben
    date accessioned2019-02-28T11:05:50Z
    date available2019-02-28T11:05:50Z
    date copyright12/4/2017 12:00:00 AM
    date issued2018
    identifier issn2332-8983
    identifier otherners_004_01_011010.pdf
    identifier urihttp://yetl.yabesh.ir/yetl1/handle/yetl/4252636
    description abstractThe Canadian supercritical water-cooled nuclear reactor (SCWR) is a 2540 MWth channel-type SCWR concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. The fuel assembly is designed so that all in-core components exposed to high radiation fields (other than the pressure tube) are part of the fuel assembly, which is removed from the reactor core as part of the assembly after three operating cycles. This design feature significantly reduces the likelihood of component failures due to radiation damage. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 °C at the inlet, 625 °C at the outlet). These conditions lead to fuel cladding temperatures close to 800 °C. Because of the reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that axial ridging, a possible failure mechanism with collapsed fuel cladding, can be avoided if the cladding thickness is larger than 0.4 mm. Detailed numerical analysis showed that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This result was obtained with conservative assumptions, suggesting that the actual margin is greater. Hence, one of the design goals, the exclusion of the possibility of melting of the fuel, which is called the “no-core-melt” concept, seems attainable. However, this needs to be demonstrated more rigorously by removing the conservative assumptions in the analysis and performing supporting experimental work. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.
    publisherThe American Society of Mechanical Engineers (ASME)
    titleFuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor
    typeJournal Paper
    journal volume4
    journal issue1
    journal titleJournal of Nuclear Engineering and Radiation Science
    identifier doi10.1115/1.4037818
    journal fristpage11010
    journal lastpage011010-7
    treeJournal of Nuclear Engineering and Radiation Science:;2018:;volume( 004 ):;issue: 001
    contenttypeFulltext
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