SUITABILITY OF VANADIUM ALLOY WITH LOW RADIO ACTIVITY USED AS BLANKET MATEIRAL

与其他候选材料相比，钒合金(V-4Cr-4Ti)具有较低的中子活化度和较快的衰变速度，在高温区具有良好的热机械性能，已被推荐作为聚变反应堆的结构材料。值得关注的一个重要问题是V合金由于吸氧和吸氢而发生的脆化。2001财年，我们主要研究了氢在钒合金中的溶解度以及氧化钛涂层对吸氢的抑制作用。结果表明：(1)当操作温度高于400℃，压力低于1pa时，高纯钒合金不会发生吸氢脆化；(2)在较低的操作温度(300℃)下，长时间的操作过程中可能会发生吸氢脆化；(3)在300℃的吸氢温度下，钛氧化物涂层的吸氢率随钛氧化物厚度的增加而降低；(4)如果形成致密而厚的氧化钛薄膜，则不含…在V合金表面可以形成更多的裂纹，从而延长V合金的使用寿命。在2002财年，主要用热脱附谱研究了钒合金在氚离子辐照后的吸氢性能。在离子能量为1.7keV的110和500℃(钒合金包层工作温度)下进行了氚离子辐照。结果表明：(1)样品中残留的氢以D_2、HD、HdO和D_2O的形式解吸；(2)在110℃处的存放量随离子注量的增加而增加，但未饱和到注量为1×10~(-1)；D/cm~2；(3)在110℃时的存留量比石墨和钨分别大一个数量级和两个数量级；(4)近表面区是否存在分凝的氧化钛对氚的存留量有很大影响；(5)在500℃时的存留量比110℃时小一个数量级。2003财年，在300℃(热核实验堆壁温)和500℃的高离子注量区分别对V合金进行了氚离子辐照。我们发现：(1)在5 0 0℃时，在高注量区，随着离子注量的增加，氚的滞留量减少；(2)在高注量区，氚的滞留量只占总的氚离子注量的0.1-3%；(3)与石墨或钨的滞留量几乎相同；(4)由于V合金包层的工作温度很高(4 0 0-70 0℃)，所以用氢气和离子的滞留量很小。2003财年，还对V合金进行了高氦通量的氦离子辐照，研究了氦的保留和热脱附性能。我们发现：(1)氦在230℃、530℃和930℃附近被脱附；(2)氦的滞留量随注量的增加而增加；(3)氦的滞留量已饱和，注量可达5×10^~(-18)He/cm~2。

Vanadium alloy (V-4Cr-4Ti) has been proposed as a structural material for fusion reactors because of its lower neutron induced activation and more rapid decay than other candidate materials and good thermal-mechanical properties in the high temperature region. One important of concern is the embrittlement of V-alloy due to oxygen and hydrogen absorption. In 2001 FY, we mainly investigated the solubility of hydrogen for the V-alloy and the suppression of hydrogen absorption by titanium oxide coatings. We found that (1) the embrittlement of V alloy with high purity due to hydrogen absorption does not occur when the operation temperature is higher than 400℃ and pressure is lower than 1 Pa, and (2) in lower operation temperature (<300℃), the embrittlement might occur during a long operation rime, and (3) the hydrogen absorption rate of T-oxide coated V alloy decreased with increase of Ti-oxide thickness at the absorption temperature of 300℃, and (4) if a dense and thick Ti-oxide film witho … More ut cracking can be coated on the surface of V-alloy, the lifetime of V-alloy might be enlarged. In 2002 FY, the deuterium retention properties of the V alloy, were mainly investigated by thermal desorption spectroscopy after deuterium ion irradiation. The deuterium ion irradiation was carried out at 110 and 500 ℃ (operation temperature of blanket made by V alloy) with ion energy of 1.7 keV We found that (1) deuterium retained in the sample was desorbed in the forms of D_2, HD, HDO and D_2O, and (2) the amount of retained deuterium at 110 ℃ increased with the ion fluence, and did not saturate to a fluence up to 1x10^<19> D/cm^2, and (3) the retained amount at 110℃ was one and two orders of magnitude larger than graphite and tungsten, respectively, and (4) the retained amount of deuterium was strongly affected by the existent of titanium-oxide segregated in the near surface region, and (5) the retained amount at 500℃ was one order of magnitude smaller than that at 110℃. In 2003 FY, deuterium ion irradiation for the V alloy was carried out at 300℃ (wall temperature of ITER) and 500℃ at higher ion fluence region. We found that (1) for the deuterium ion irradiation at 500℃, the amount of retained deuterium decreased with increase in ion fluence in the high fluence region, and (2) the amount of retained deuterium was 0.1-3% of the total deuterium ion fluence, and (3) the retained amount was almost the same as that for graphite or tungsten, and (4) since the operation temperature of V alloy blanket Is high (400-700℃), the retention amount by bath hydrogen gas and ion becomes very small. In 2003 FY, helium ion irradiation with high helium flux was also carried out for the V alloy to Investigate the retention and thermal desorption properties of helium. We found that (1) helium was desorbed around at 230℃, 530℃ and 930℃, and (2) the amount of retained helium increased with the fluence, and (3) the retained amount was saturated to a fluence up to 5x10^<18> He/cm^2. Less