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The ATWS analysis of one control rod withdraw out of the HTR-10GT core in addition with bypass valve failure

机译:除旁通阀故障外,对一根控制杆从HTR-10GT芯中抽出的ATWS分析

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摘要

The 10 MW high temperature gas cooled test reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700 ℃ To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750 ℃ and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basic accidents of HTR10-GT must be analyzed according to China's nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800℃, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism, in spite of the failure of bypass valve. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1283.3℃. It was a litter than 1230 ℃ - the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation will be raised to 1620℃, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM for they have the same outlet temperature of the core.
机译:10 MW高温气冷试验堆(HTR-10)是在核能和新能源技术研究所(INET)建造的,自2003年初以来已成功运行。HTR-10的核心出口温度为700℃为了验证燃气轮机直接循环技术,INET首先计划将堆芯出口温度提高到750℃,并使用氦燃气轮机代替蒸汽发生器(然后将反应堆称为HTR-10GT)。尽管HTR-10具有良好的本质安全性,但由于运行参数的更改,HTR10-GT的设计基本事故和超设计基本事故必须根据中国的核法规进行分析。 THERMIX代码系统用于研究一个控制杆因错误而从核心抽出的ATWS事故。错误地以1 cm / s的速度将侧反射器中的控制杆抽出后,插入正反应性,反应堆功率增加,堆芯温度升高。当功率测量范围的中子通量超过123%且堆芯出口温度大于800℃时,反应堆应加注。据认为,反射器中的所有控制棒均已被阻塞,反应堆无法放空。因此事故继续进行,堆芯温度和系统压力增加,但尽管旁路阀发生故障,反应堆仍由于其自然的负温度反应性反馈机制而最终关闭。残留的热量将通过型腔冷却系统从铁心中去除。在事故序列中,最高燃油温度为1283.3℃。它比1230℃少-HTR-10的燃料温度限制。现在,HTR-10GT中使用的球形燃料也将用于HTR-PM中,并且温度限制将提高到1620℃,因此在将一根控制棒从堆芯中抽出的ATWS期间,HTR-10GT是安全的。本文还比较了HTR10-GT和HTR-10的分析结果。结果表明,尽管HTR-10GT的工作温度高于HTR-10,但在事故期间仍是安全的。该分析将有助于HTR-PM,因为它们具有与芯相同的出口温度。

著录项

  • 来源
    《Nuclear Engineering and Design》 |2014年第5期|459-464|共6页
  • 作者

    Minggang Lang; Yujie Dong;

  • 作者单位

    Institute of Nuclear and New Energy Technology, Tsinghua University, Room 207, Building B, Nengkelou, Beijing, PR China;

    Institute of Nuclear and New Energy Technology, Tsinghua University, Room 503, Building A, Nengkelou, Beijing, PR China;

  • 收录信息 美国《科学引文索引》(SCI);美国《工程索引》(EI);美国《生物学医学文摘》(MEDLINE);
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类
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