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Graphite and Carbon Materials in Nuclear Engineering
1st Edition - January 1, 2019
Authors: Shijiang Xu, Feiyu Kang, Derek Tsang
Paperback ISBN:
9 7 8 - 0 - 1 2 - 8 1 2 6 5 3 - 0
eBook ISBN:
9 7 8 - 0 - 1 2 - 8 1 2 6 5 4 - 7
Graphite and Carbon Materials in Nuclear Engineering provides basic knowledge of carbonaceous materials used in High Temperature Reactor (HTR) and Molten Salt Reactor (MSR)… Read more
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Graphite and Carbon Materials in Nuclear Engineering provides basic knowledge of carbonaceous materials used in High Temperature Reactor (HTR) and Molten Salt Reactor (MSR) systems. The book covers nuclear engineering, working environment and requirements of nuclear graphite; R&D and production of nuclear graphite; irradiation effect (or irradiation damage) of nuclear graphite; and issues the must be resolved for the healthy development of HTR and MSR. This valuable book will serve as a reference book not only for new researchers entering this field from diversified backgrounds, but also for experts including nuclear materials scientists and engineers, particularly those who work in HTR and MSR material section, reactor designers, project managers and governmental nuclear authorities.
- Provides comprehensive knowledge for people involved in nuclear engineering, especially in R&D of High Temperature Reactor (HTR) and Molten Salt Reactor (MSR)
- Describes the R&D methodology of carbonaceous materials in nuclear engineering
- Features material from authors who are internationally known for their expertise in this field and are still active researchers
Nuclear materials Scientists and engineers particularly those who work in HTR and MSR material section, reactor designers, project managers, and governmental nuclear authorities
1. Introduction2. Basic knowledge of nuclear reactor3. High temperature reactor4. Molten salt reactor5. Structure of carbonaceous materials6. Production of nuclear graphite7. Properties of nuclear graphite8. Radiation effect of nuclear graphite9. Decommissioning of nuclear graphite10. Carbonaceous materials in fusion reactor
- No. of pages: 624
- Language: English
- Edition: 1
- Published: January 1, 2019
- Imprint: Academic Press
- Paperback ISBN: 9780128126530
- eBook ISBN: 9780128126547
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Shijiang Xu
Prof. Xu has worked in nuclear materials field since 1958, took part in building 2MW experimental shield reactor; R&D in nuclear fuel (UO2), graphite for liquid fuel reactors, BeO; in charge of R&D in HTR fuel technology (laboratory scale). Since 1993, Prof. Xu has interested in R&D of nuclear graphite and he has devoted to promote nuclear graphite development in China.
Affiliations and expertise
Retired Professor, Institute of Nuclear Energy Technology, Tsinghua UniversityFK
Feiyu Kang
Feiyu Kang received his PhD from The Hong Kong University of Science and Technology in 1997. He is honorary editorial advisory board of international journal CARBON, Joint Chairmen of international symposiums: CARBON2002 (Beijing), Carbon2011 (Shanghai) and 15th International Symposium on Intercalation Compounds (ISIC15), Coordinators of international research projects: Professor M. Inagaki (NSFC-JSPS) and Professor I. Mochida (JST-MOST).
Prof. Kang has investigated graphite and carbon materials since 1988. His research interest includes nano-carbon materials, graphite producing process, porous carbon and nuclear graphite. Prof. Kang had published more than 200 scientific papers and 3 books.
Affiliations and expertise
Professor, Department of Materials Science and Engineering, Tsinghua University, ChinaDT
Derek Tsang
Prof. Tsang has more than 10 years experiences in nuclear graphite. Have been worked for Magnox nuclear reactor and Advanced Gas-cooled Reactor in United Kingdom and High Temperature Test Reactor (HTTR) in Japan. Now Prof. Tsang is working on the Thorium Molten Salt Reactor development in China. Prof. Tsang is committee members of ASME and ASTM.
Affiliations and expertise
Professor, Center for Excellence TMSR Energy, Shanghai Institute of Applied Physics, Chinese Academy of Sciences