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Journal Articles

Design challenges for sodium cooled fast reactors

Konomura, Mamoru; Ichimiya, Masakazu

Journal of Nuclear Materials, 371(1-3), p.250 - 269, 2007/09

 Times Cited Count:17 Percentile:76.46(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Measurement of the fuel pin deflection in an assembly irradiated in FBR "JOYO"

Katsuyama, Kozo; Nagamine, Tsuyoshi; Nakamura, Yasuo; Matsumoto, Shinichiro; Asaka, Takeo; Furuya, Hirotaka

Transactions of the American Nuclear Society, 94(1), p.771 - 772, 2006/06

no abstracts in English

Journal Articles

Present status of ZrC coated fuel particle development for very high temperature reactors in JAEA

Sawa, Kazuhiro; Ueta, Shohei; Aihara, Jun

Transactions of the American Nuclear Society, 94(1), P. 705, 2006/06

The Very-High-Temperature Reactor (VHTR) is the one of the most promising candidates for the Generation IV Nuclear Energy System. The VHTR fuel should exhibit excellent safety performance up to burn-up of about 15 to 20 % fissions per initial metal atom (FIMA) and fluence of 6$$times$$10$$^{25}$$m$$^{-2}$$ (E$$>$$0.1MeV). There is no experimental data which has proved the intactness of conventional SiC-coated fuel particles under such severe condition. Japan Atomic Energy Agency (JAEA) has developed Zirconium carbide (ZrC)-coated fuel particle, the ZrC coating layer of which is expected to maintain its intactness under higher temperature and burn-up compared with SiC-coating layer. JAEA started (1) ZrC-coating process development by large-scale coater, (2) inspection method development of ZrC coating and (3) irradiation test and post irradiation experiment of ZrC coated particles. This paper presents present status of ZrC-coated fuel particle development in JAEA.

Journal Articles

Pyrometallurgical production of U-Pu alloy and injection casting of U-Pu-Zr

Nakamura, Kinya*; Yokoo, Takeshi*; Arai, Yasuo

Transactions of the American Nuclear Society, 94(1), P. 780, 2006/06

no abstracts in English

Oral presentation

Development of optimized martensitic 9Cr-ODS steel cladding

Ukai, Shigeharu; Kaito, Takeji; Otsuka, Satoshi; Narita, Takeshi; Sakasegawa, Hideo

no journal, , 

The composition of martensitic 9Cr-ODS steels was optimized from the balancing of the residual alpha hard and F/M soft grains. The claddings with the optimized composition of 9Cr-0.14C-2W-0.3Ti-0.35Y$$_{2}$$O$$_{3}$$ as well as standard were manufactured and their mechanical properties satisfied the design criteria for SFR fuel.

Oral presentation

Burn-leach; The Most important test in the manufacture of HTGR fuel

Nabielek, H.*; Verfondern, K.*; Tang, C.*; Ueta, Shohei

no journal, , 

Burn-leach is the most sensitive method for the determination of High-Temperature Gas-Cooled Reactor fuel quality. German fuel manufacture was for the operation of the 46 MWth AVR. Improvements in the fuel quality were due to perfected tabling of kernels, particles and overcoated particles and the introduction of automated overcoating. Chinese HTGR first load fuel manufacture around 2000 was for the operation of the 10 MWth HTR-10. An improvement can be observed after the first few production runs. Japanese HTGR first fuel manufacture was for the operation of the 30 MWth HTTR. The particle volume density of 30 % is much higher than the below 10 % of the spherical fuel elements. Nevertheless, very good results in terms of low defect fractions were also achieved. These results establish the quality standard in modern UO$$_{2}$$ Triso fuel.

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