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Journal Articles

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki

Nuclear Engineering and Design, 271, p.309 - 313, 2014/05

 Times Cited Count:9 Percentile:57.39(Nuclear Science & Technology)

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. The air/water ingress accident, one of the DECs for the HTGR, is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO$$_{2}$$) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

Journal Articles

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 6 Pages, 2012/10

A new concept of the High Temperature Gas-cooled Reactor (HTGR), so-called the Naturally Safe HTGR, is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions such as the air/water ingress accidents occur by deterministic approach based on the inherent safety features of the HTGR. For the Naturally Safe HTGR it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

Oral presentation

Characterization of new nuclear-grade graphites for HTGR/VHTR

Nagaishi, Yoshihide*; Fukuda, Toshiaki*; Kondo, Akira*; Sumita, Junya; Sakaba, Nariaki

no journal, , 

Fine-Grained isotropic graphite shows higher strength making it a promising material for the graphite component of High Temperature Gas-cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). There are two kinds of considerations about lifetime of graphite components under neutron irradiation condition. One consideration is determined based on dimensional changes due to neutron irradiation. In this consideration, small dimensional change with much amount of neutron irradiation is preferable. The other consideration is determined based on a margin between the specified minimum ultimate strengths and the residual stresses induced by neutron irradiation. In this consideration, strength of graphite components is important property to decide lifetime. It is expected that development of new fine-grained isotropic nuclear grade graphite possessing higher strength will contribute toward added design margins and an extension of the lifetime of graphite components, which likely improve the reactor economy very significantly. Furthermore, from the viewpoint of graphite waste, it is also expected to decrease the exchange frequency of core components. Tokai Carbon Co., Ltd. has developed new nuclear grade graphite with higher strength and characterization of them is being carried out. This presentation shows the R&D plan and the initial results of mechanical and thermal properties of un-irradiated nuclear grade graphite.

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