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Ozawa, Masaaki*; Amaya, Masaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.185 - 200, 2020/12
no abstracts in English
Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi
Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08
Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Sugiyama, Tomoyuki; Fuketa, Toyoshi; Ozawa, Masaaki*; Nagase, Fumihisa
Proceedings of 2004 International Meeting on LWR Fuel Performance, p.544 - 550, 2004/09
Two pulse irradiation experiments simulating reactivity initiated accidents were performed on high burnup (
60 GWd/t) PWR UO
rods with advanced cladding alloys. Test OI-10 was performed on an MDA cladded rod with large-grain (25 
m) fuel pellets with a peak fuel enthalpy condition of 435 J/g, and resulted in a peak residual hoop strain of 0.7%. On the other hand, Test OI-11 on a ZIRLO cladded rod with conventional pellets resulted in a fuel failure at a fuel enthalpy of 500 J/g due to the pellet-cladding mechanical interaction (PCMI). A long axial split was generated on the cladding over the active length. The fuel pellets were fragmented and dispersed into the coolant water. The fuel enthalpy at failure is higher than the PCMI failure criterion of 209 J/g at the corresponding burnup. The experimental results suggest that the rods with improved corrosion resistance have much safety margin against the PCMI failure compared to the conventional Zircaloy-4 rod.
Nagase, Fumihisa; Tanimoto, Masataka*; Uetsuka, Hiroshi
IAEA-TECDOC-1320, p.270 - 278, 2002/11
With a view to obtaining basic data for evaluating high burnup fuel behavior under LOCA conditions, a systematic research program is being conducted at JAERI. High-temperature oxidation tests with non-irradiated cladding have been performed to investigate separate effects of pre-oxidation and pre-hydriding on the oxidation kinetics. "Integral thermal shock tests" have been conducted simulating a LOCA condition to examine the influence of pre-hydriding on failure-bearing capability of oxidized cladding upon quenching. Test results showed almost no influence of absorbed hydrogen on the threshold value for oxidation amount under no axial restraint condition. On the other hand, it was shown that the threshold value is reduced by absorbed hydrogen for the restraint condition.
Department of Hot Laboratories
JAERI-Review 2001-044, 95 Pages, 2001/12
JAERI-Review-2001-044.pdf:6.55MB
no abstracts in English
Department of Hot Laboratories
JAERI-Review 2000-015, 113 Pages, 2000/09
JAERI-Review-2000-015.pdf:6.09MB
no abstracts in English
Department of Hot Laboratories
JAERI-Review 99-026, p.118 - 0, 1999/11
JAERI-Review-99-026.pdf:9.93MB
no abstracts in English
Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kikuchi, Keiichi*; Abe, Tomoyuki*
Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00
no abstracts in English
Department of Hot Laboratories
JAERI-Review 98-023, 97 Pages, 1998/12
JAERI-Review-98-023.pdf:5.35MB
no abstracts in English
Department of Hot Laboratories
JAERI-Review 98-001, 92 Pages, 1998/02
JAERI-Review-98-001.pdf:4.14MB
no abstracts in English
Department of Hot Laboratories
JAERI-Review 97-001, 118 Pages, 1997/02
JAERI-Review-97-001.pdf:4.36MB
no abstracts in English
Ishijima, Kiyomi; Fuketa, Toyoshi
NUREG/CP-0157, 1, p.93 - 105, 1996/10
no abstracts in English
Saito, Junichi; ; Oyamada, Rokuro
KAERI-NEMAC/TR-32/95, 0, p.125 - 136, 1995/00
no abstracts in English
Kano, Shigeru; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo
no journal, ,
no abstracts in English
Ueta, Shohei; Aihara, Jun; Sumita, Junya; Shaimerdenov, A.*; Dyussambayev, D.*; Gizatulin, S.*; Chakrov, P.*; Sakaba, Nariaki
no journal, ,
In order to investigate irradiation performance of the newly-designed high temperature gas-cooled reactor (HTGR) fuel for high burnup around 100 GWd/t, a capsule irradiation test has been carried out by WWR-K research reactor in the Institute of Nuclear Physics (INP) of Kazakhstan. A result on evaluation of the fuel integrity based on the fractional release of fission product (FP) released from the fuel during the irradiation and a plan of post-irradiation examination are reported.
Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki
no journal, ,
no abstracts in English
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
no journal, ,
no abstracts in English
