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

钒合金（V-4Cr-4Ti）具有较低的中子活化率和较快的衰变速度，在高温区具有良好的热力学性能，已被提出作为核聚变反应堆的结构材料。一个重要的问题是由于氧和氢的吸收导致v合金的脆化。在2001年度，我们主要研究了氢在v合金中的溶解度以及氧化钛涂层对氢吸收的抑制作用。我们发现(1)当操作温度高于400℃，压力低于1 Pa时，高纯V合金不会发生吸氢脆；(2)在较低的操作温度（<300℃）下，在较长的操作时间内可能会发生吸氢脆；(3)在300℃的吸收温度下，t -氧化物涂层V合金的吸氢率随着ti -氧化物厚度的增加而降低。(4)若能在v -合金表面镀上一层致密、厚实、无裂纹的氧化钛膜，则可延长v -合金的使用寿命。2002年，采用热解吸光谱法研究了氘离子辐照后V合金的氘保留性能。氘离子辐照进行了110和500℃(操作温度的毯子由V合金)离子能量为1.7 keV我们发现(1)氘保留样本眠的形式的D_2,高清,HDO和D_2O和(2)保留氘量在110℃增加离子影响,和不饱和影响多达x10 ^ ^ < 19 > D /厘米2,和(3)留存金额在110℃是一个和两个数量级大于石墨、钨、(4)近表面存在分离的氧化钛对氘的截留量影响较大；(5)500℃时的截留量比110℃时的截留量小一个数量级。2003年度在300℃（ITER壁温）和500℃高离子影响区对V合金进行了氘离子辐照。我们发现(1)氘离子辐照在500℃,保留氘的数量减少和增加离子高影响地区的影响,和(2)的保留氘量是0.1氘离子影响总数的-3%,和(3)留存金额几乎是一样的,对于石墨或钨,和(4)V合金毯子的操作温度高(400 - 700℃),保留浴氢气和离子量非常小。2003年度又对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解吸性能。结果表明：(1)氦气在230℃、530℃和930℃左右发生解吸，(2)氦气的截留量随着注注量的增加而增加，(3)氦气的截留量达到饱和，注注量达到5x10^<18> He/cm^2。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