基于氢研究核石墨中氚的去除
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摘要
世界现阶段有大量的退役核石墨需要处理,~3H和~(14)C为其中含量最多、需重点去污核素。对于~(14)C来说,低温(不高于700°C)低氧环境下的热处理能比较有选择性地去除核石墨中的~(14)C。基于氚是氢的同位素、与氢具有相同的物理化学特性,本研究通过对三种不同产地的核石墨中氢在350oC的吸附以及400~700oC的解吸行为,探究核石墨中氚的去污工艺。实验发现:三种核石墨的氢吸附量不同,解吸规律大致相同,解吸量随时间的变化上有差异。国产核石墨NG-CT-10、日本核石墨IG-110以及德国核石墨NBG-18的氢总吸附量分别为6.7×10-3 mL·g-1、9.30×10~(-3) mL·g~(-1)以及9.12×10~(-30 mL·g~(-1),其中化学吸附量分别为3.2×10~(-3) mL·g~(-1)、3.0×10~(-3) mL·g~(-1)和0.92×10~(-3) mL·g~(-1)。石墨对氢吸附量上的差异可能来源于三种核石墨的不同制备工艺和物理性质上的一些差异,这些差异主要来自于平均孔径、比表面积、成型工艺以及焦粒粒径上的区别;NG-CT-10有效吸附量所占比最高,表明NG-CT-10有较大量的氚吸附量。400~700oC的核石墨氢解吸实验表明:三种石墨中的氢主要是从700oC开始有效解吸,但各自相对于总吸附量的解吸量有明显区别,NG-CT-10、IG-110和NBG-18在700°C时的解吸量分别为7%、13.5%和70%。由此可得,NBG-18中的氚最易被解吸出来。根据氢在石墨中的吸附模型,700oC解吸出来的氢应该位于石墨晶粒边缘。为了解吸剩余氚,同时不影响~(14)C的有效去除,不提高热处理温度,可能需要改变解吸时的载气组分。
[Background] Nowadays, there are big amount of irradiated nuclear graphite waiting to be decomissioned worldwide. Due to their high content in nuclear graphite, tritium and ~(14)C are the two main radionuclides that should be specially treated before final disposal. The optimal desorption temperature for ~(14)C was reported to be 700 °C. [Purpose] As tritium is the isotope of hydrogen with similar physical and chemical properties, in order to investigate the desorption process for tritium in nuclear graphite, this study investigates the absorption of hydrogen at 350 °C and desorption of hydrogen from 400 °C to 700 °C in three types of nuclear graphite. [Methods] Samples of three types of nuclear graphite were subjected to flowing hydrogen at 350 °C for hydrogen absorption, then the absorbed hydrogen was desorbed by thermal treatment from 400 °C to 700 °C. The hydrogen concentration of the outlet gas was measured by gas chromotagraph. [Results] According to our experiments, three types of nuclear graphite have different amount of absorption, but the following desorption experiments showed similar trend with some differences on the amount of desorption varied with time. The amount of absorption of the domestic nuclear graphite NG-CT-10, Japanese nuclear graphite IG-110 and Germen nuclear graphite NBG-18 were measured to be 6.7×10~(-3) mL·g~(-1), 9.3×10~(-3) mL·g~(-1) and 9.12×10~(-3) mL·g~(-1), respectively, and the amount of hydrogen that is chemically absorbed in graphite were 3.2×10~(-3) mL·g~(-1), 3.0×10~(-3) mL·g~(-1) and 0.92×10~(-3) mL·g~(-1) respectively. The difference in the amount of absorption could be due to the difference in the physical properties and synthetic process. Moreover, the absorbed hydrogen start to desorb effectively after the temperature was raised to 700 °C. For the three types of nuclear graphite investigated in our research, the amount of desorption at 700 °C was different: 7% of hydrogen was desorbed in NG-CT-10 graphite, 13.5% for IG-110 and 70% for NBG-18 nuclear graphite. [Conclusion] Based on our findings, the amount of stability absorbed hydrogen in domestic nuclear graphite NG-CT-10 was the highest, hence amount of tritium absorbed could also be the highest in NG-CT-10 nuclear graphite. According to the model of hydrogen absorption in nuclear graphite proposed by Atsumi, the hydrogen desorbed at 700 °C were the hydrogen absorbed at the edge surface of graphite crystallites. To fully desorb tritium and ~(14)C effectively at 700 °C, the carrier gas during desorption should be varied.
[Background] Nowadays, there are big amount of irradiated nuclear graphite waiting to be decomissioned worldwide. Due to their high content in nuclear graphite, tritium and ~(14)C are the two main radionuclides that should be specially treated before final disposal. The optimal desorption temperature for ~(14)C was reported to be 700 °C. [Purpose] As tritium is the isotope of hydrogen with similar physical and chemical properties, in order to investigate the desorption process for tritium in nuclear graphite, this study investigates the absorption of hydrogen at 350 °C and desorption of hydrogen from 400 °C to 700 °C in three types of nuclear graphite. [Methods] Samples of three types of nuclear graphite were subjected to flowing hydrogen at 350 °C for hydrogen absorption, then the absorbed hydrogen was desorbed by thermal treatment from 400 °C to 700 °C. The hydrogen concentration of the outlet gas was measured by gas chromotagraph. [Results] According to our experiments, three types of nuclear graphite have different amount of absorption, but the following desorption experiments showed similar trend with some differences on the amount of desorption varied with time. The amount of absorption of the domestic nuclear graphite NG-CT-10, Japanese nuclear graphite IG-110 and Germen nuclear graphite NBG-18 were measured to be 6.7×10~(-3) mL·g~(-1), 9.3×10~(-3) mL·g~(-1) and 9.12×10~(-3) mL·g~(-1), respectively, and the amount of hydrogen that is chemically absorbed in graphite were 3.2×10~(-3) mL·g~(-1), 3.0×10~(-3) mL·g~(-1) and 0.92×10~(-3) mL·g~(-1) respectively. The difference in the amount of absorption could be due to the difference in the physical properties and synthetic process. Moreover, the absorbed hydrogen start to desorb effectively after the temperature was raised to 700 °C. For the three types of nuclear graphite investigated in our research, the amount of desorption at 700 °C was different: 7% of hydrogen was desorbed in NG-CT-10 graphite, 13.5% for IG-110 and 70% for NBG-18 nuclear graphite. [Conclusion] Based on our findings, the amount of stability absorbed hydrogen in domestic nuclear graphite NG-CT-10 was the highest, hence amount of tritium absorbed could also be the highest in NG-CT-10 nuclear graphite. According to the model of hydrogen absorption in nuclear graphite proposed by Atsumi, the hydrogen desorbed at 700 °C were the hydrogen absorbed at the edge surface of graphite crystallites. To fully desorb tritium and ~(14)C effectively at 700 °C, the carrier gas during desorption should be varied.
引文
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11 Vulpius D,Baginski K,Kraus B,et al.Thermal treatment of neutron-irradiated nuclear graphite[J].Nuclear Engineering and Design,2013,265:294-309.DOI:10.1016/j.nucengdes.2013.09.007.
12 Dunzik-Gougar M L,Smith T E.Removal of carbon-14from irradiated graphite[J].Journal of Nuclear Materials,2014,451(1-3):328-335.DOI:10.1016/j.jnucmat.2014.03.018.
13 Li H,Yang C,Fang C,et al.Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM[J].Progress in Nuclear Energy,2015,85:676-681.DOI:10.1016/j.pnucene.2015.09.003.
14 Le Guillou M,Toulhoat N,Pipon Y,et al.Deuterium migration in nuclear graphite:consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste[J].Journal of Nuclear Materials,2015,461:72-77.DOI:10.1016/j.jnucmat.2015.03.005.
15 Kowalczyk P,Tanaka H,Holyst R,et al.Storage of hydrogen at 303 K in graphite slitlike pores from grand canonical Monte Carlo simulation[J].Journal of Physical Chemistry B,2005,109(36):17174-17183.DOI:10.1021/jp0529063.
16 Rzepka M,Lamp P,de la Casa-Lillo M A.Physisorption of hydrogen on microporous carbon and carbon nanotubes[J].Journal of Physical Chemistry B,1998,102(52):10894-10898.DOI:10.1021/jp9829602.
17 Atsumi H,Iseki M,Shikama T.Trapping and detrapping of hydrogen in carbon-based materials exposed to hydrogen gas[J].Journal of Nuclear Materials,1994,212:1478-1482.DOI:10.1016/0022-3115(94)91073-1.
18 Franklin R E.The structure of graphitic carbons[J].Acta Crystallographica,1951,4(3):253-261.DOI:10.1107/S0365110X51000842.
19 Atsumi H,Tauchi K.Hydrogen absorption and transport in graphite materials[J].Journal of Alloys and Compounds,2003,356:705-709.DOI:10.1016/S0925-8388(03)00290-1.
20 Wilson K L,Hsu W L.Hydrogen recycling preperties of graphite[J].Journal of Nuclear Materials,1987,145:121-130.DOI:10.1016/0022-3115(87)90317-5.
21 Atsumi H,Tanabe T,Shikama T.Bulk hydrogen retention in neutron-irradiated graphite at elevated temperatures[J].Journal of Nuclear Materials,2009,390-391:581-584.DOI:10.1016/j.jnucmat.2009.01.112.
22 Causey R A,Elleman T S,Verghese K.Hydrogen diffusion and solubility in pyrolytic carbon[J].Carbon,1979,17(4):323-328.DOI:10.1016/0008-6223(79)90003-4.
23 Atsumi H,Tokura S,Miyake M.Absorption and desorption of deuterium on graphite at elevated temperatures[J].Journal of Nuclear Materials,1988,155(88):241-245.DOI:10.1016/0022-3115(88)90247-4.
2 Bushuev A V,Verzilov Y M,Zubarev V N,et al.Quantitative-determination of the amount of H-3 and C-14 in reactor graphite[J].Atomic Energy,1992,73(6):959-962.DOI:10.1007/BF00761431.
3 Bushuev A V,Verzilov Y M,Zubarev V N,et al.Experimental determination of the spent graphite radioactive contamination at plutonium-production reactors of the Siberian Group of Chemical Enterprises(Tomsk-7)[J].Nuclear Technology,2002,140(1):51-62.DOI:10.13182/NT02-A3323.
4 Vulpius D,Baginski K,Fischer C,et al.Location and chemical bond of radionuclides in neutron-irradiated nuclear graphite[J].Journal of Nuclear Materials,2013,438(1-3):163-177.DOI:10.1016/j.jnucmat.2013.02.027.
5 Li J F,Dunzik-Gouga M L,Wang J L.Recent advances in the treatment of irradiated graphite:a review[J].Annals of Nuclear Energy,2017,110:140-147.DOI:10.1016/j.anucene.2017.06.040.
6 Atsumi H,Tanabe T,Shikama T.Hydrogen behavior in carbon and graphite before and after neutron irradiationtrapping,diffusion and the simulation of bulk retention[J].Journal of Nuclear Materials,2011,417(1-3):633-636.DOI:10.1016/j.jnucmat.2010.12.100.
7 Kanashenko S L,Gorodetsky A E,Chernikov V N,et al.Hydrogen adsorption on and solubility in graphites[J].Journal of Nuclear Materials,1996,233:1207-1212.DOI:10.1016/S0022-3115(96)00067-0.
8 Kashcheev V A,Ustinov O A,Yakunin S A,et al.Technology and facility for incinerating irradiated reactor graphite[J].Atomic Energy,2017,122(4):252-256.DOI:10.1007/s10512-017-0263-7.
9 Costes J R,Bournot P C P,Guiberteau P.CO2-laser-aided waste incineration[C].Bohn W L,Hugel H,ed.10th International Symposium on Gas Flow and Chemical Lasers,Freidrichshafen 1994,SPIE-The International Society for Optical Engineering,1994,2502:590-596.DOI:10.1117/12.204975.
10 Fachinger J,von Lensa W,Podruhzina T.Decontamination of nuclear graphite[J].Nuclear Engineering and Design,2008,238(11):3086-3091.DOI:10.1016/j.nucengdes.2008.02.010.
11 Vulpius D,Baginski K,Kraus B,et al.Thermal treatment of neutron-irradiated nuclear graphite[J].Nuclear Engineering and Design,2013,265:294-309.DOI:10.1016/j.nucengdes.2013.09.007.
12 Dunzik-Gougar M L,Smith T E.Removal of carbon-14from irradiated graphite[J].Journal of Nuclear Materials,2014,451(1-3):328-335.DOI:10.1016/j.jnucmat.2014.03.018.
13 Li H,Yang C,Fang C,et al.Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM[J].Progress in Nuclear Energy,2015,85:676-681.DOI:10.1016/j.pnucene.2015.09.003.
14 Le Guillou M,Toulhoat N,Pipon Y,et al.Deuterium migration in nuclear graphite:consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste[J].Journal of Nuclear Materials,2015,461:72-77.DOI:10.1016/j.jnucmat.2015.03.005.
15 Kowalczyk P,Tanaka H,Holyst R,et al.Storage of hydrogen at 303 K in graphite slitlike pores from grand canonical Monte Carlo simulation[J].Journal of Physical Chemistry B,2005,109(36):17174-17183.DOI:10.1021/jp0529063.
16 Rzepka M,Lamp P,de la Casa-Lillo M A.Physisorption of hydrogen on microporous carbon and carbon nanotubes[J].Journal of Physical Chemistry B,1998,102(52):10894-10898.DOI:10.1021/jp9829602.
17 Atsumi H,Iseki M,Shikama T.Trapping and detrapping of hydrogen in carbon-based materials exposed to hydrogen gas[J].Journal of Nuclear Materials,1994,212:1478-1482.DOI:10.1016/0022-3115(94)91073-1.
18 Franklin R E.The structure of graphitic carbons[J].Acta Crystallographica,1951,4(3):253-261.DOI:10.1107/S0365110X51000842.
19 Atsumi H,Tauchi K.Hydrogen absorption and transport in graphite materials[J].Journal of Alloys and Compounds,2003,356:705-709.DOI:10.1016/S0925-8388(03)00290-1.
20 Wilson K L,Hsu W L.Hydrogen recycling preperties of graphite[J].Journal of Nuclear Materials,1987,145:121-130.DOI:10.1016/0022-3115(87)90317-5.
21 Atsumi H,Tanabe T,Shikama T.Bulk hydrogen retention in neutron-irradiated graphite at elevated temperatures[J].Journal of Nuclear Materials,2009,390-391:581-584.DOI:10.1016/j.jnucmat.2009.01.112.
22 Causey R A,Elleman T S,Verghese K.Hydrogen diffusion and solubility in pyrolytic carbon[J].Carbon,1979,17(4):323-328.DOI:10.1016/0008-6223(79)90003-4.
23 Atsumi H,Tokura S,Miyake M.Absorption and desorption of deuterium on graphite at elevated temperatures[J].Journal of Nuclear Materials,1988,155(88):241-245.DOI:10.1016/0022-3115(88)90247-4.