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Development of mechanistic cladding rupture model for integrated severe accident code ISAA. Part Ⅱ: DVI line small-break LOCA in CAP1400

机译:综合严重事故代码ISAA机械包层破裂模型的发展。 第二部分:第1400章中的DVI线小型突破LOCA

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摘要

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident (SA). For example, the SA caused by the loss of coolant accident (LOCA), the decrease in primary loop pressure and the increase in core temperature will lead to the cladding ballooning and rupture. The clad-ding deformation affects the flow distribution and causes local flow blockage. Meanwhile, water vapor will enter the fuel gap from the rupture, thereby increasing the surface area of cladding that is oxidized by the steam. At present, the widely used integrated severe accident analysis codes cannot analyze fuel rods thermal-mechanical behavior at the early SA stage, and the judgment of cladding rupture is only based on simple parameter model. This paper integrates the developed FRTMB (core Fuel Rod Thermal-Mechanical Behavior analysis module) module into the integrated severe accident analysis code ISAA, so that the coupled system ISAA-FRTMB can analyze fuel rods thermal-mechanical behavior and judge cladding rupture. Part I introduces the need of developing the FRTMB module, verification of the module, and how the module interactively works in the severe accident analysis code. And evaluates the steady-state fuel rods thermal-mechanical behavior at different enrichment areas of the CAP1400 reactor. Part II focuses on analyzing fuel rods behavior during a hypothetical DVI Line (Direct Vessel Injection line) small break-out accident (break diameter d = 4 in.) of the CAP1400, and predicts the clad-ding rupture time and the corresponding failure temperature. Due to failure of components of the emer-gency core cooling system and assumed unavailability of several preventive and mitigative accident management measures (AMM), the accident develops into a severe accident scenario with core melt and reactor pressure vessel failure. The mechanical analysis results show the top of the fuel rod is the first to rupture, rather than the peak node. The primary factor affecting the fuel pellet strain is the deforma-tion caused by thermal expansion, while the densification and swelling change little. These results demonstrate the applicability and reliability of ISAA-FRTMB in analyzing fuel rods thermal-mechanical behavior and judging cladding rupture during transient accidents. (c) 2021 Elsevier Ltd. All rights reserved.
机译:包层气球和破裂是严重事故早期(SA)的重要现象。例如,由冷却液事故(LOCA)损失引起的SA,初级环压力的降低和核心温度的增加将导致包层膨胀和破裂。包层变形会影响流量分布并导致局部流量阻塞。同时,水蒸气将从破裂进入燃料隙,从而增加通过蒸汽氧化的包层的表面积。目前,广泛使用的集成严重事故分析代码不能分析早期SA阶段的燃料棒热电机械行为,并且覆层破裂的判断仅基于简单的参数模型。本文将开发的FRTMB(核心燃料棒热电机械行为分析模块)模块集成到集成的严重事故分析代码ISAA中,使耦合系统ISAA-FRTMB可以分析燃料棒热电机械行为并判断包层破裂。第一部分介绍了开发FRTMB模块,模块验证的需要,以及模块如何在严重的事故分析代码中交互式工作。并评估CAP1400反应器的不同浓缩区域的稳态燃料棒热电部位。第二部分专注于分析假设的DVI线(直接船舶注射线)的燃料棒行为小爆炸事故(断裂直径d = 4英寸),并预测夹层破裂时间和相应的故障温度。由于EMER-Gency核心冷却系统的组件失败,并且假设几种预防性和减要的事故管理措施(AMM)的不可用,事故发生成严重的事故情景,具有核心熔体和反应堆压力容器故障。机械分析结果显示燃料棒的顶部是第一个破裂,而不是峰节节点。影响燃料颗粒菌株的主要因素是由热膨胀引起的变形,而致密化和肿胀变化很少。这些结果表明了ISAA-FRTMB在分析燃料棒热电机构和瞬态事故中判断包层破裂时的适用性和可靠性。 (c)2021 elestvier有限公司保留所有权利。

著录项

  • 来源
    《Annals of nuclear energy》 |2021年第12期|108613.1-108613.11|共11页
  • 作者单位

    Xi An Jiao Tong Univ Sch Nucl Sci & Technol Xian 710049 Shaanxi Peoples R China;

    Xi An Jiao Tong Univ Sch Nucl Sci & Technol Xian 710049 Shaanxi Peoples R China|Xi An Jiao Tong Univ State Key Lab Multiphase Flow Power Engn Xian 710049 Shaanxi Peoples R China;

    Xi An Jiao Tong Univ Sch Nucl Sci & Technol Xian 710049 Shaanxi Peoples R China;

    Xi An Jiao Tong Univ Sch Nucl Sci & Technol Xian 710049 Shaanxi Peoples R China;

    Xi An Jiao Tong Univ Sch Nucl Sci & Technol Xian 710049 Shaanxi Peoples R China|Xi An Jiao Tong Univ State Key Lab Multiphase Flow Power Engn Xian 710049 Shaanxi Peoples R China;

  • 收录信息
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类
  • 关键词

    ISAA-FRTMB; SBLOCA; Severe accident; Fuel behavior; Cladding rupture; CAP1400 reactor;

    机译:isaa-frtmb;sbloca;严重事故;燃料行为;熔覆破裂;CAP1400反应堆;

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