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

钒合金（V-4Cr-4 Ti）具有较低的中子活化和较快的衰变速度，以及良好的高温热机械性能，被推荐作为聚变堆的结构材料。其中一个重要的问题是由于氧和氢的吸收，钒合金的脆化。在2001财年，我们主要研究了氢在钒合金中的溶解度和钛氧化物涂层对吸氢的抑制作用。结果表明：（1）高纯V合金在400℃以上、压力低于1 Pa时不会发生吸氢脆化;（2）在较低的工作温度（<300℃）下，长时间工作时可能发生吸氢脆化。在300℃吸氢温度下，Ti氧化物包覆V合金的吸氢速率随Ti氧化物厚度的增加而降低; 关于我们 在V合金表面涂敷抗裂涂层，可提高V合金的使用寿命。2002年FY主要用热脱附谱研究了钒合金在氘离子辐照后的吸氘性能。在110和500 ℃下进行了氘离子辐照结果表明：（1）样品中的氘以D_2、HD、HDO和D_2O的形式脱附;（2）在110 ℃时，样品中的氘量随离子注量的增加而增加;在110℃时氘的保留量分别比石墨和钨大一个和两个数量级;（4）在近表面区域存在的钛氧化物对氘的保留量有很大的影响<19>，500℃时的保留量比110℃时小一个数量级。2003年，在300℃（ITER壁温）和500℃高注量区对钒合金进行了氘离子辐照实验。结果表明：（1）在500℃氘离子辐照下，在高注量区，氘的滞留量随注量的增加而减少;（2）氘的滞留量占总氘注量的0.1-3%;（3）氘的滞留量与石墨或钨的滞留量基本相同。（4）由于钒合金包层的操作温度高（400-700℃），熔池氢气和离子的滞留量很小。在2003年财政年度，还进行了高氦通量氦离子辐照钒合金，以研究氦的保留和热脱附性能。结果表明：（1）氦在230℃、530℃和930℃左右解吸;（2）氦的滞留量随注量的增加而增加;（3）当注量达到5 × 10 ^<18>He/cm ^2时，氦的滞留量达到饱和。少

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