Graphite and Carbon Materials in Nuclear Engineering - Xu, Shijiang; Kang, Feiyu; Tsang, Derek
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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…mehr

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Produktbeschreibung
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.
Autorenporträt
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.