Graphite and Carbon Materials in Nuclear Engineering
Will be published approx. on 1. January 2029
Book
Paperback/Softback
624 pages
978-0-12-812653-0 (ISBN)
Description
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.
More details
Language
English
Place of publication
San Diego
United States
Publishing group
Elsevier Science Publishing Co Inc
Target group
Professional and scholarly
Product notice
Paperback (trade)
Unsewn / adhesive bound
Dimensions
Height: 235 mm
Width: 191 mm
ISBN-13
978-0-12-812653-0 (9780128126530)
Copyright in bibliographic data is held by Nielsen Book Services Limited or its licensors: all rights reserved.
Schweitzer Classification
Persons
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. 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. 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.
Content
1. Introduction
2. Basic knowledge of nuclear reactor
3. High temperature reactor
4. Molten salt reactor
5. Structure of carbonaceous materials
6. Production of nuclear graphite
7. Properties of nuclear graphite
8. Radiation effect of nuclear graphite
9. Decommissioning of nuclear graphite
10. Carbonaceous materials in fusion reactor
2. Basic knowledge of nuclear reactor
3. High temperature reactor
4. Molten salt reactor
5. Structure of carbonaceous materials
6. Production of nuclear graphite
7. Properties of nuclear graphite
8. Radiation effect of nuclear graphite
9. Decommissioning of nuclear graphite
10. Carbonaceous materials in fusion reactor