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Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09
Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.912 - 932, 2005/10
High burnup fuel cladding can fail due to mechanical interaction with expanding fuel pellet under reactivity initiated accident (RIA) conditions. In order to evaluate the cladding failure limit, investigations to modify ring tensile test have been performed to measure mechanical properties of Zircaloy cladding properly. JAERI developed the test method and geometry minimizing undesirable effects of friction and bending moment in the specimen. Using the modified test method, mechanical properties of unirradiated Zircaloy-4 cladding were evaluated as functions of hydrogen concentration and temperature. For hydrogen concentrations above 700 ppm, obvious increase of ductility is observed with the temperature increase from 300 to 473 K. For hydrogen concentrations below 500 ppm, on the other hand, temperature dependence of ductility is relatively small in the present temperature range from 300 to 573 K.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10
LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 
81
m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.
Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11
Times Cited Count：10 Percentile：55.72(Nuclear Science & Technology)The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1m and 10m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.
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.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi
HPR-362, Vol.2, 12 Pages, 2004/05
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.
Kitano, Koji*; Fuketa, Toyoshi; Uetsuka, Hiroshi
JAERI-Research 2001-041, 24 Pages, 2001/08
JAERI-Research-2001-041.pdf:2.75MB
no abstracts in English
Kusagaya, Kazuyuki*; Nakamura, Takehiko; Yoshinaga, Makio; Okonogi, Kazunari*; Uetsuka, Hiroshi
JAERI-Research 2001-010, 44 Pages, 2001/03
JAERI-Research-2001-010.pdf:9.91MB
no abstracts in English
Akie, Hiroshi; Nakamura, Takehiko
Progress in Nuclear Energy, 38(3-4), p.363 - 370, 2001/02
Times Cited Count：6 Percentile：44.09(Nuclear Science & Technology)no abstracts in English
Kuroda, Masatoshi*; Yamanaka, Shinsuke*; Nagase, Fumihisa; Uetsuka, Hiroshi
Nuclear Engineering and Design, 203(2-3), p.185 - 194, 2001/01
Times Cited Count：14 Percentile：68.97(Nuclear Science & Technology)no abstracts in English
Yamashita, Toshiyuki
Nihon Genshiryoku Gakkai "Kodo Nenryo Gijutsu" Kenkyu Senmon Iinkai Hokokusho, p.467 - 474, 2001/00
no abstracts in English
JNC TN1400 2000-012, 250 Pages, 2000/11
JNC-TN1400-2000-012.pdf:10.18MB
no abstracts in English
Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Kikuchi, Keiichi*
Journal of Nuclear Science and Technology, 37(5), p.455 - 464, 2000/05
no abstracts in English
Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi; Kikuchi, Keiichi*; Abe, Tomoyuki*
JAERI-Research 99-060, p.62 - 0, 2000/03
JAERI-Research-99-060.pdf:12.05MB
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
Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; ; ; *
JAERI-Tech 98-031, 225 Pages, 1998/08
no abstracts in English
Akie, Hiroshi; Anoda, Yoshinari; Takano, Hideki; *; *
JAERI-Research 98-009, 44 Pages, 1998/03
JAERI-Research-98-009.pdf:2.49MB
no abstracts in English
Ishijima, Kiyomi; Fuketa, Toyoshi
Purutoniumu Nenryo Kogaku, p.421 - 451, 1998/01
no abstracts in English
Nakamura, Takehiko; *; Sasajima, Hideo; Fuketa, Toyoshi; *
JAERI-Research 96-060, 110 Pages, 1996/11
JAERI-Research-96-060.pdf:2.74MB
no abstracts in English
