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首页> 外文期刊>Journal of Engineering for Gas Turbines and Power >Thermal Aspects of Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors
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Thermal Aspects of Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors

机译：超临界水冷核反应堆中碳​​化铀和二碳化铀燃料的热学方面

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摘要

Supercritical water-cooled nuclear reactors (SCWRs) are a Generation IV reactor concept. SCWRs will use a light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374℃). One reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is to increase the thermal efficiency. The thermal efficiency of existing NPPs is between 30% and 35% compared with 45% and 50% of supercritical water (SCW) NPPs. Another benefit of SCWRs is the use of a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing a pressure tube (PT) type SCWR. This concept refers to a 1200-MW-(el) PT-type reactor. Coolant operating parameters are as follows: a pressure of 25 MPa, a channel inlet temperature of 350℃, and an outlet temperature of 625 ℃. The sheath material and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for the sheath is a zirconium alloy and the fuel is an enriched uranium dioxide (UO2). The sheath-temperature design limit is 850℃, and the industry accepted limit for the fuel centerline temperature is 1850℃. Previous studies have shown that the maximum fuel centerline temperature of a UO_2 pellet might exceed this industry accepted limit at SCWR conditions. Therefore, alternative fuels with higher thermal conductivities need to be investigated for SCWR use. Uranium carbide (UC), uranium nitride (UN), and uranium dicarbide (UC2) are excellent fuel choices as they all have higher thermal conductivities compared with conventional nuclear fuels such as UO_2, mixed oxides (MOX), and thoria (ThO_2)- Inconel-600 has been selected as the sheath material due its high corrosion resistance and high yield strength in aggressive supercritical water (SCW) at high-temperatures. This paper presents the thermalhydraulics calculations of a generic PT-type SCWR fuel channel with a 43-element Inconel-600 bundle with UC and UC_2 fuels. The bulk-fluid, sheath and fuel centerline temperature profiles, together with a heat transfer coefficient profile, were calculated for a generic PT-type SCWR fuel-bundle string. Fuel bundles with UC and UC_2 fuels with various axial heat flux profiles (AHFPs) are acceptable since they do not exceed the sheath-temperature design limit of 850℃, and the industry accepted limit for the fuel centerline temperature of 1850℃. The most desirable case in terms of the lowest fuel centerline temperature is the UC fuel with the upstream-skewed cosine AHFP. In this case, the fuel centerline temperature does not exceed even the sheath-temperature design limit of 850℃.
机译：超临界水冷核反应堆（SCWR）是第四代反应堆的概念。超临界水力发电站将使用轻水冷却剂，其工作参数设定为高于水的临界点（22.1 MPa和374℃）。从当前的核电站（NPP）设计转换为SCW NPP设计的原因之一是提高热效率。现有NPP的热效率介于30％和35％之间，而超临界水（SCW）NPP的热效率则介于45％和50％之间。 SCWR的另一个好处是使用了简化的流路，可以省去蒸汽发生器，蒸汽干燥器，蒸汽分离器等。加拿大正在概念化SCWR型压力管（PT）。这个概念指的是1200-MW-（el）PT型反应堆。冷却液的工作参数如下：压力为25 MPa，通道入口温度为350℃，出口温度为625℃。护套材料和核燃料必须能够承受这些极端条件。通常，护套的主要选择是锆合金，燃料是浓缩的二氧化铀（UO2）。护套温度设计极限为850℃，行业公认的燃料中心线温度极限为1850℃。先前的研究表明，在SCWR条件下，UO_2颗粒的最高燃料中心线温度可能会超过该行业认可的极限。因此，需要研究具有更高热导率的替代燃料以用于SCWR。碳化铀（UC），氮化铀（UN）和二碳化铀（UC2）是极好的燃料选择，因为与传统的核燃料（如UO_2，混合氧化物（MOX）和氧化ria（ThO_2）-）相比，它们都具有更高的导热率Inconel-600已被选作护套材料，因为它在高温下在腐蚀性超临界水（SCW）中具有高耐腐蚀性和高屈服强度。本文介绍了带有UC和UC_2燃料的43元素Inconel-600束的通用PT型SCWR燃料通道的热工水力计算。计算了通用PT型SCWR燃料管束的总流体，护套和燃料中心线温度曲线以及传热系数曲线。可以使用带有各种轴向热通量曲线（AHFP）的UC和UC_2燃料的燃料束，因为它们不超过护套温度设计极限850℃，并且不超过行业公认的燃料中心线温度1850℃的极限。就最低燃料中心线温度而言，最理想的情况是带有上游偏余弦AHFP的UC燃料。在这种情况下，燃料中心线温度甚至不会超过护套温度设计极限850℃。

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  • 来源
    《Journal of Engineering for Gas Turbines and Power》 |2011年第2期|p.022901.1-022901.7|共7页
  • 作者

    Lisa Grande; Bryan Villamere; Leyland Allison; Sally Mikhael; Adrianexy Rodriguez-Prado; Igor Pioro;

  • 作者单位

    Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada;

    Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada;

    Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada;

    Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada;

    Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada;

    Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario, L1H 7K4, Canada;

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  • 原文格式 PDF
  • 正文语种 eng
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  • 关键词

    SCWR; uranium carbide; uranium dicarbide; thermalhydraulics; AHFP;

    机译：SCWR;碳化铀二碳化铀热工液压行动计划;

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