退役核石墨中氚的去污

邓珂; 马玉华; 秦来来; 刘佳煜; 张钦; 马兆伟; 杨果; 卫飞; 王广华; 刘卫

doi:10.11889/j.1000-3436.2018.rrj.36.020101

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退役核石墨中氚的去污

综述 | 更新时间：2021-12-10

- 退役核石墨中氚的去污
- Decontamination of tritium from irradiated nuclear graphite
- 辐射研究与辐射工艺学报 2018年36卷第2期
- 作者机构：
  
  1.中国科学院上海应用物理研究所上海 201800
  2.中国科学院大学北京 100049
- 作者简介：
  
  [ "邓珂, 男, 1990年出生, 2015年于英国伦敦帝国理工学院获硕士学位, 现为博士研究生, 研究方向为退役核石墨中氚的去污工艺, E-mail:dengke@sinap.ac.cn", "DENG Ke (male) was born in 1990, and received his master’s degree from Imperial College London in 2015. Now he is a doctoral candidate focusing on the study of the removal of tritium from irradiated nuclear graphite. E-mail: dengke@sinap.ac.cn" ]
  刘卫, 博士, 教授, E-mail:liuwei@sinap.ac.cn Ph.D. LIU Wei, professor, E-mail:liuwei@sinap.ac.cn
- 基金信息：
  
  国家自然科学基金(21601202)
- DOI：10.11889/j.1000-3436.2018.rrj.36.020101
  中图分类号：
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邓珂, 马玉华, 秦来来, 等. 退役核石墨中氚的去污[J]. 辐射研究与辐射工艺学报, 2018,36(2):020101-7.

Ke DENG, Yuhua MA, Lailai QIN, et al. Decontamination of tritium from irradiated nuclear graphite[J]. Journal of Radiation Research and Radiation Processing, 2018,36(2):020101-7.
邓珂, 马玉华, 秦来来, 等. 退役核石墨中氚的去污[J]. 辐射研究与辐射工艺学报, 2018,36(2):020101-7. DOI： 10.11889/j.1000-3436.2018.rrj.36.020101.

Ke DENG, Yuhua MA, Lailai QIN, et al. Decontamination of tritium from irradiated nuclear graphite[J]. Journal of Radiation Research and Radiation Processing, 2018,36(2):020101-7. DOI： 10.11889/j.1000-3436.2018.rrj.36.020101.

摘要

核石墨在反应堆的运行过程中，产生大量的放射性核素。其中，氚由于其较高的含量以及极易与环境中的水发生同位素交换而进入环境的特性，被视作核石墨退役中需要特殊处理的对象。根据国际上该领域的相关研究，对核石墨的热处理被认为是现阶段最佳的去氚方法，但是，最终氚的去除效率还受到热处理工艺中各种参数的影响。核石墨中氚去除的热处理温度为至少600℃，载气流速越高，氚的去除效率也就越高，并且，向载气中加入少量的水蒸气能够有效地提高氚的去除效率，而核石墨由于其本身性质的差异以及工作环境的区别，不同核石墨之间的热处理除氚效果也有很大区别。现阶段该领域都是一些定性的研究，还需要通过更多更细致的研究来优化各参数以达到最佳的氚去除效率。国内对于退役核石墨中氚的去除工艺研究还非常少，本文对国际上核石墨去氚的研究进行了综述，希望能为国内相关研究的开展提供参考。

Abstract

Due to the operation of nuclear reactors, the nuclear graphite used in these reactors contains large amount of radionuclides. Tritium is considered one of the most important radionuclides, which should be specially treated because of its high concentration and the ease of entering the environment via isotopic reaction with water. According to relevant studies worldwide, thermal treatment is the most applicable technology for the removal of tritium; however, the efficiency of the removal is influenced by the process parameters. The removal temperature should be higher than 600℃, and the removal efficiency increases with increasing flow rate of the carrier gas and upon addition of steam to the carrier gas. Most significantly, the tritium removal efficiency is greatly influenced by the nature and working condition of the graphite. Current research in this field is mostly qualitative. To optimize tritium removal rates, the process should be systematically studied with respect to the different process parameters involved. However, there have been few such studies reported in China. This manuscript summarizes the main findings of various international research groups regarding tritium removal from nuclear graphite, in order to provide some guidance to related domestic researches.

关键词

核石墨去污热处理氚

Keywords

Decomtamination of nuclear graphiteThermal treatmentTritium

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