DETERMINATION OF CORROSIVE CONDITION IN COOLING SYSTEM OF HIGH IRRADIATION DOSE RATE AND HIGH HEAT LOADING COMPONENT AND ESTABLISHMENT OF ITS MODIFICATION TECHNIQUES

高功率加速器靶、核裂变反应堆和聚变反应堆冷却系统中结构材料的应力腐蚀开裂(SCC)主要由炉尖周围分解物的局部浓度决定。研究的主要目的是确定腐蚀控制物种，确定局部水化学对SCC裂纹扩展速率的影响，从而提出合适的水化学控制，如弱碱控制以减缓SCC。1.确定辐照下裂纹尖端的局部水化学。用缝隙辐解模型可以从理论上估算裂隙尖端的水化学。2.高温高压双氧水回路的设计和施工可以将双氧水的浓度控制在任何所需的浓度。可以测量裂纹扩展速率。3.辐照下裂纹扩展速率的测定CGR…在模拟辐照条件下(由H_20O_2浓度控制)测得更多。讨论了H_2O_2和O_2条件下CGR的差异。确定了确定CGR的主要参数是腐蚀电流，而不是电化学腐蚀电位。4.测定了辐照和弱碱性水化学条件下的CGR。(1)在模拟辐照条件下，在pH小于8.5的条件下测量了CGR。(2)评价了裂尖pH对CGR的影响。5.将测量数据反馈到缝隙辐解模型和操作工厂的腐蚀状况评价程序中。(1)提出了合适的G值和反应速率常数集。(2)提出了包含合适常数集的理论模型。(3)由缝隙辐解模型和裂纹扩展模型组成的SCC CGR评价模型。建议不仅适用于裂变反应堆冷却系统，也适用于加速器和聚变堆的冷却系统。(4)提出了防止应力腐蚀开裂的弱碱性水化学，不仅适用于裂变反应堆冷却系统，也适用于加速器和聚变堆的冷却系统。较少

Stress corrosion cracking (SCC) of structural materials in cooling systems of high power accelerator targets, nuclear fission reactors and fusion reactors reactor are determined mainly by local concentrations of rodiolytic species around crock tips. Major purposes of the studies are to confirm the corrosion determining species, to determine the effects of local water chemistry on SCC crack growth rate and then so propose suitable water chemistry control, e.g.,weak alkali control for mitigation of SCC.1.Determination of local water chemistry of crack tips under irradiation.Energy deposition rate was measured in a simulated crack tip. Water chemistry in a crack tip could be theoretically estimated with the crevice radiolysis model.2.Design and construction of high temperature high pressure hydrogen peroxide (H_2O_2) water loopThe concentration of H_2O_2 could be controlled at any desired concentration. Crack growth rate (CGR) could be measured.3.Determination of CGR under irradiationCGR … More was measured under simulated irradiation conditions (controlled by H_20O_2 concentration). Difference in CGR under H_2O_2 and O_2conditions was discussed. It was confirmed that the major parameter to determine CGR was corrosion current rather than electrochemical corrosion potential.4.Determination of CGR under irradiation and weak alkali water chemistry(1)CGR was measured under simulated irradiation condition and under pH less than 8.5.(2)The effects of crack tip pH on CGR was evaluated.5.Feed back of measured date to crevice radiolysis model and corrosion condition evaluation procedures for operation plants(1)Suitable sets far G values and reaction rate constants was proposed.(2)Theoretical model containing the suitable constant sets was proposed.(3)SCC CGR evaluation model which consisted of the crevice radiolysis model and the crack growth model, was proposed not only for fission reactor cooling systems but also cooling systems of accelerators and fusion reactors.(4)Weak alkali water chemistry was proposed to prevent SCC not only for fission reactor cooling systems but also cooling systems of accelerators and fusion reactors. Less