Material compatibility issues in EU fusion fuel cycle R&D and design
- 作者：D.K. Murdoch ; I. Cristescu ; C. Day ; M. Glugla ; R. Lä ; sser ; A. Mack
- 关键词：Glasses ; Mechanical properties ; Crack growth
- 刊名：Journal of Nuclear Materials
- 出版年：2007
- 期刊代码：51_00223115
- 类别：ms
- 出版时间：1 August 2007
- 卷：367-370
- 期：Complete
- 页码：1366-1370
- 文件大小：127 K
摘要
Many material selections for fusion Fuel Cycle systems are determined by the properties of tritium, including its behaviour as a hydrogen isotope, and its decay product, 3He. Within the EU R&D program, the following issues related to tritium service have been addressed. The mechanical integrity and longevity of the sorbent/bonding agent/substrate system used for cryosorption pumping have been extensively tested under tritium exposure. Extended testing of palladium/silver membranes used for separation of elemental hydrogens from impurities has been carried out to confirm longevity in tritium service. For all high temperature (mg src="http://www.sciencedirect.com/scidirimg/entities/223c.gif" alt="not, vert, similar" border=0>150 °C) components, tritium permeation through primary containments must be confined by outer (low temperature) jackets, and designs have been developed to achieve this. For wetproof catalysts and solid polymer electrolysers used for water detritiation, tests are in progress to determine the operating life. Testing of the ITER reference tritium storage getter material is under way.