氘、氚与RAFM钢相容性研究进展
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摘要
核聚变能是解决人类能源危机和环境问题最有效的途径,其主要是利用氘氚聚合释放的能量。磁约束聚变是目前最可能实现受控热核聚变的方法,但要实现长期且稳态的核聚变反应还面临着诸多挑战,其中材料的研究与开发是聚变堆能否商业化的关键。在服役过程中,包层结构材料不仅受到高热负荷及强腐蚀作用,还受到各种粒子如氘(D)、氚(T)、氦(He)等的轰击和D-T聚变反应产物高能中子的影响。目前,确定的候选结构材料主要有奥氏体不锈钢、低活化铁素体/马氏体(RAFM)钢、钒合金以及碳化硅复合材料四种。而RAFM钢因具有低活性、较低的热膨胀系数、较高的热导率、辐照环境下具有较好的几何稳定性被选为目前最具前景的结构材料。获得氘氚在RAFM钢中的输运参数是未来核数据库建立的基础和前提,近几年关于氘在RAFM钢中输运行为的研究较多,然而不同研究者所得的结果差别很大,且缺乏实际的氚实验的基础数据。因此建立实验测试标准十分必要。RAFM钢主要以板条马氏体结构为主,具有较高的氘、氚渗透率,极易造成氘氚燃料的损失及氚放射性污染。因此,必须减少或避免RAFM钢与氘氚的直接接触。在RAFM钢表面制备一定厚度的阻氚涂层是实现氚自持最有效的途径之一。目前,国内外研究较多的阻氚涂层为Al2O3涂层,其阻氚因子可达103,且已实现工程化应用。此外,RAFM钢在服役过程中产生的辐照损伤及表面状态变化必然会影响氘氚的输运行为,主流观点认为辐照产生的缺陷会增加氘氚在金属材料中的滞留量,当材料中氘原子浓度达到10-6时,塑韧性下降,产生氢脆,尤其对于氚,衰变产生的He-3原子浓度达到10-9时,还会引发更严重的氦脆。除了制备阻氚涂层外,最近的研究多致力于通过成分调控及改善热处理工艺来提高RAFM钢的抗氢性能及抗辐照性能。本文归纳了氘氚在RAFM钢中行为的研究进展,分别对RAFM钢中氘氚渗透和滞留行为及其对力学性能的影响等进行介绍,分析了RAFM钢开发面临的问题并展望其前景,期望为RAFM钢数据库的建立以及服役于聚变堆的工程可行性提供参考。
Nuclear fusion achieved through deuterium and tritium is the most effective way to solve the energy crisis and environmental issues. The magnetic confinement fusion is currently the most promising steady-state research reactor. However,there are many challenges to be solved including materials design and development which is essential to the commercial application of fusion energy. The blanket structural materials are not only subjected to high heat load and severe corrosion in the service process,but also bombarded by various particles such as deuterium( D),tritium( T),helium( He) and high energy neutrons produced by D-T fusion reaction. Up to now there are primarily four kinds of candidate structural materials which are austenitic stainless steel,low activation ferrite/martensitic( RAFM) steel,vanadium alloy and Si C composites,respectively. RAFM steels have been proposed as the most promising candidate structural materials for the experimental fusion reactor ITER owing to their better swelling resistance,improved irradiation resistance and favorable termo-physical properties.It is fundamental to obtain the transport parameters of deuterium and tritium through RAFM steels for constructing the nuclear database. Although the transport behavior of deuterium in RAFM steels has been researched widely in recent years,the results obtained by different researchers are very different,and data on tritium are scared. Therefore,it is very necessary to establish the experimental test standard. RAFM steel has high deuterium and tritium permeability due to its lath martensitic structure,leading to fuel loss and tritium pollution. Thus direct contact between RAFM steel and hydrogen isotopes must be reduced or avoided. The preparation of tritium permeation barrier( TPB) on the surface of RAFM steel is one of the most effective ways to realize tritium self-sustainment. At present,Al2 O3 coating has attracted much attention as typical candidate TPB materials for its self-healing capacity and low permeation reduction factor( PRF),in addition to the realization of engineering application.Moreover,the radiation damage and surface state change of RAFM steel during service will inevitably affect the transport behavior of deuterium and tritium. It is generally believed that the defects produced by irradiation will increase the retention of deuterium and tritium in metal materials. As a result of hydrogen accumulation,a degradation of mechanical performance like the structure strength and ductility would occur for the materials.Especially for tritium,when the He-3 produced by tritium decay reaches ppb level,helium embrittlement will be appeared. In addition to TPB,recent studies have focused on improving the hydrogen resistance and irradiation resistance of RAFM steel by controlling composition and improving heat treatment process.This article reviews the research progress of deuterium and tritium behaviors in RAFM steels. And the permeation and retention behaviors of deuterium and tritium in RAFM steels,as well as their effects on mechanical properties are presented respectively. The issues involved in the development of RAFM steel are analyzed and its prospect is looked forward. It is expected that this will offer reference for the database establishment of RAFM steel and the feasibility of serving in the fusion reactor.
Nuclear fusion achieved through deuterium and tritium is the most effective way to solve the energy crisis and environmental issues. The magnetic confinement fusion is currently the most promising steady-state research reactor. However,there are many challenges to be solved including materials design and development which is essential to the commercial application of fusion energy. The blanket structural materials are not only subjected to high heat load and severe corrosion in the service process,but also bombarded by various particles such as deuterium( D),tritium( T),helium( He) and high energy neutrons produced by D-T fusion reaction. Up to now there are primarily four kinds of candidate structural materials which are austenitic stainless steel,low activation ferrite/martensitic( RAFM) steel,vanadium alloy and Si C composites,respectively. RAFM steels have been proposed as the most promising candidate structural materials for the experimental fusion reactor ITER owing to their better swelling resistance,improved irradiation resistance and favorable termo-physical properties.It is fundamental to obtain the transport parameters of deuterium and tritium through RAFM steels for constructing the nuclear database. Although the transport behavior of deuterium in RAFM steels has been researched widely in recent years,the results obtained by different researchers are very different,and data on tritium are scared. Therefore,it is very necessary to establish the experimental test standard. RAFM steel has high deuterium and tritium permeability due to its lath martensitic structure,leading to fuel loss and tritium pollution. Thus direct contact between RAFM steel and hydrogen isotopes must be reduced or avoided. The preparation of tritium permeation barrier( TPB) on the surface of RAFM steel is one of the most effective ways to realize tritium self-sustainment. At present,Al2 O3 coating has attracted much attention as typical candidate TPB materials for its self-healing capacity and low permeation reduction factor( PRF),in addition to the realization of engineering application.Moreover,the radiation damage and surface state change of RAFM steel during service will inevitably affect the transport behavior of deuterium and tritium. It is generally believed that the defects produced by irradiation will increase the retention of deuterium and tritium in metal materials. As a result of hydrogen accumulation,a degradation of mechanical performance like the structure strength and ductility would occur for the materials.Especially for tritium,when the He-3 produced by tritium decay reaches ppb level,helium embrittlement will be appeared. In addition to TPB,recent studies have focused on improving the hydrogen resistance and irradiation resistance of RAFM steel by controlling composition and improving heat treatment process.This article reviews the research progress of deuterium and tritium behaviors in RAFM steels. And the permeation and retention behaviors of deuterium and tritium in RAFM steels,as well as their effects on mechanical properties are presented respectively. The issues involved in the development of RAFM steel are analyzed and its prospect is looked forward. It is expected that this will offer reference for the database establishment of RAFM steel and the feasibility of serving in the fusion reactor.
引文
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43 Hara S,Abe T,Enoeda M,et al.Journal of Nuclear Materials,1998,S258-263(4),1280.
44 Liu C,Klein H,Jung P.Journal of Nuclear Materials,2004,335(1),77.
45 Zhu S,Zhang C,Yang Z,et al.Nuclear Engineering and Technology,2017,49(8),1748.
2 Kurtz R J,Alamo A,Lucon E,et al.Journal of Nuclear Materials,2009,386(5),411.
3 He P,Yao W Z,Lv J M,et al.Materials Review A:Review Papers,2018,32(1),34(in Chinese).何培,姚伟志,吕建明,等.材料导报:综述篇,2018,32(1),34.
4 Laha K,Saroja S,Moitra A,et al.Journal of Nuclear Materials,2013,439(1-3),41.
5 Jung P.Fusion Technology,1998,33(1),63.
6 Chen C A,Liu L B,Wang B,et al.Fusion Engineering and Design,2016,112,569.
7 Levchuk D,Koch F,Maier H,et al.Journal of Nuclear Materials,2004,328(2),103.
8 Wang B,Liu L B,Xiang X,et al.Journal of Nuclear Materials,2016,470,30.
9 Noh S J,Lee S K,Byeon W J,et al.Fusion Engineering and Design,2014,89(11),2726.
10 Zhou H S,Hirooka Y,Ashikawa N,et al.Journal of Nuclear Materials,2014,455(1-3),470.
11 Aiello A,Ricapito I,Benamati G.Fusion Science and Technology,2002,41,872.
12 Yasuhisa O,Makoto K,Junya O,et al.Fusion Engineering and Design,2012,87(5-6),580.
13 Lee S K,Yun S H,Han G J,et al.Current Applied Physics,2014,14(10),1385.
14 Ueda Y,Lee H T,Peng H Y,et al.Fusion Engineering and Design,2012,87(7-8),1356.
15 Oya Y,Kobayashi M,Osuo J,et al.Fusion Engineering and Design,2012,87(5-6),580.
16 Swansiger W A,Bastasz R.Journal of Nuclear Materials,1979,85(79),335.
17 Ouyang Y J,Yu G,Hu L,et al.Surface Engineering,2013,29(4),312.
18 Dolinsky Y N,Zouev Y N,Lyasota I A,et al.Journal of Nuclear Mate-rials,2002,307(2),1484.
19 Otsuka T,Shimada M,Kolasinski R,et al.Journal of Nuclear Materials,2011,415(1),S769.
20 Fedorov A V,Til S V,Magielsen A J,et al.Fusion Engineering and Design,2014,448(1-3),139.
21 Fan D J,Lu G D,Zhang G K,et al.Acta Metallurgica Sinica,2018,54(4),519(in Chinese).范东军,陆光达,张桂凯,等.金属学报,2018,54(4),519.
22 Schliefer F,Liu C,Jung P.Journal of Nuclear Materials,2000,283-287,540.
23 Fujita H,Chikada T,Mochizuki J,et al.Fusion Engineering and Design,2018,133,95.
24 Terasawa M,Fukushima K,Nakahigashi S,et al.Japanese Journal of Applied Physics,1986,25(7),1106.
25 Xu Y P,Lu T,Li X C,et al.Nuclear Inst & Methods in Physics Research B,2016,388,5.
26 Yao Z,Liu C,Jung P,et al.Fusion Science and Technology,2005,48,1285.
27 Xiang X,Wang X L,Zhang G K,et al.International Journal of Hydrogen Energy,2015,40(9),3697.
28 Janda D,Fietzek H,Galetz M,et al.Intermetallics,2013,41,51.
29 Zhang M,Xu B J,Ling G P,et al.Applied Surface Science,2015,331(1),1.
30 Zhang G K.Theoretical studies on hydrogen behavior in α-Al2O3 tritium permeation barrier material.Ph.D.Thesis,University of Science and Technology of China,China,2014(in Chinese).张桂凯.α-Al2O3阻氚涂层材料中氢行为的理论研究.博士学位论文,中国科学技术大学,2014.
31 Wang P X,Song J S.Helium in materials and the permeation of tritium,National Defense Industry Press,China,2002(in Chinese).王佩璇,宋家树.材料中的氦及氚渗透,国防工业出版社,2002.
32 Wang Z L,Chen C A,Song Y Q,et al.Fusion Engineering and Design,2018,126,139.
33 Noh S J,Kim H S,Byeon W J,et al.Journal of Nuclear Materials,2017,490,1.
34 Yakushiji K,Lee H T,Oya M,et al.Physica Scripta,2016,2016(T167),014067.
35 Ogorodnikova O V,Zhou Z,Sugiyama K,et al.Nuclear Fusion,2017,57,1.
36 Yamauchi Y,Gotoh K,Nobuta Y.Fusion Engineering and Design,2010,85(10),1838.
37 Ito T,Yamauchi Y,Hino T,et al.Journal of Nuclear Materials,2011,417(1-3),1147.
38 Shinoda N,Yamauchi Y,Nobuta Y,et al.Fusion Engineering and Design,2014,89(7-8),921.
39 Shinoda N,Yamauchi Y,Nobuta Y.Journal of Nuclear Materials,2015,463,1001.
40 Beghini M,Benamati G,Bertini L,et al.Journal of Nuclear Materials,2001,288,1.
41 Yagodzinskyy Y,Malitckii E,Ganchenkova M,et al.Journal of Nuclear Materials,2014,444,435.
42 Wang Z L,Zhu K G,Xiang X,et al.Fusion Engineering and Design,2018,137,15.
43 Hara S,Abe T,Enoeda M,et al.Journal of Nuclear Materials,1998,S258-263(4),1280.
44 Liu C,Klein H,Jung P.Journal of Nuclear Materials,2004,335(1),77.
45 Zhu S,Zhang C,Yang Z,et al.Nuclear Engineering and Technology,2017,49(8),1748.