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Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
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.
Fukaya, Yuji; Nishihara, Tetsuo
Nuclear Engineering and Design, 307, p.188 - 196, 2016/10
Times Cited Count：4 Percentile：36.53(Nuclear Science & Technology)Reduction of High Level Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and the features are significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing, and effective waste loading method for direct disposal is proposed by applying the feature in this study. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with LWR representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.53 - 62, 2016/09
In order to evaluate adequacy of present safety criteria and safety margins in terms of advanced fuels and provide a database for future regulation on them, JAEA started an extensive research program called ALPS-II program, which has been sponsored by NRA, Japan. This program is primarily composed of tests simulating a RIA and a LOCA on the high-burnup advanced fuels irradiated in commercial PWR or BWR. Recently, the failure limits of the high-burnup advanced fuels under RIA conditions were investigated at NSRR, and post-test examinations on the fuel rods after the pulse irradiation tests are being performed. In terms of the simulated LOCA test, integral thermal shock tests and high temperature oxidation tests were carried out at RFEF, and the fracture limits, high temperature oxidation rate, etc. of the high-burnup advanced fuel cladding were investigated. This paper mainly describes some recent experimental results obtained in this program with respect to RIA and LOCA.
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.
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.
Sekimoto, Hiroshi*
JAERI-Tech 2005-008, 111 Pages, 2005/03
JAERI-Tech-2005-008.pdf:15.9MB
The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. When this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the bunrup of discharged fuel is about 40%. It means that the nuclear energy can be utilized for many hundreds years without new mining, enrichment and reprocessing, and the amount of spent fuel can be reduced considerably. Compared to fast reactors, application of CANDLE burnup to prismatic fuel high-temperature gas cooled reactors is very easy. In this report, the applications of CANDLE burnup to both these types of reactors are studied.
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.
Katanishi, Shoji; Takei, Masanobu; Nakata, Tetsuo*; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.67 - 75, 2004/03
no abstracts in English
Department of Hot Laboratories
JAERI-Review 2002-039, 106 Pages, 2003/01
JAERI-Review-2002-039.pdf:9.46MB
no abstracts in English
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.
Okubo, Tsutomu; Suzuki, Motoe; Iwamura, Takamichi; Takeda, Renzo*; Moriya, Kumiaki*; Kanno, Minoru*
Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10
A small scale around 300 MWe reduced-moderation water reactor (RMWR) concept has been developed. For the core, a BWR type core concept with the tight-lattice fuel rod arrangement and the high void fraction is adopted to attain a high conversion ratio over 1.0. The negative void reactivity coefficients are also required, and the very flat short core concept is adopted to make the natural circulation cooling (NC) possible. The core burn-up of 60 GWd/t and the operation cycle of 24 months are also attained. For the system, simplification of the system with the passive safety features is a basic approach to overcome the scale demerit as well as the NC. For example, the HPCF system is replaced with the passive accumulator system resulting in the expensive emergency DGs reduction. The cost evaluation for concerned NSSS components gives about 20% reduction. Since MOX fuels in the RMWR contains Pu around 30 wt% and is irradiated to a high burn-up, the fuel safety evaluation has been performed and the acceptable results have been obtained from the thermal feasibility point of view.
Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, X.; Kunitomi, Kazuhiko
JAERI-Tech 2002-066, 51 Pages, 2002/09
JAERI-Tech-2002-066.pdf:7.79MB
no abstracts in English
Iwamura, Takamichi; Okubo, Tsutomu; Yonomoto, Taisuke; Takeda, Renzo*; Moriya, Kumiaki*; Kanno, Minoru*
Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 8 Pages, 2002/00
Research and developments of reduced-moderation water reactor (RMWR) have been performed. The RMWR can attain the favorable characteristics such as high burn-up, long operation cycle, multiple recycling of plutonium and effective utilization of uranium resources, based on the matured LWR technologies. MOX fuel assemblies in the tight-lattice fuel rod arrangement are used to reduce the moderation of neutron, and hence, to increase the conversion ratio. The conceptual design has been accomplished for the small 330MWe RMWR core with the discharge burn-up of 60GWd/t and the operation cycle of 24 months, under the natural circulation cooling of the core. A breeding ratio of 1.01 and the negative void reactivity coefficient are simultaneously realized in the design. In the plant system design, the passive safety features are intended to be utilized mainly to improve the economy. At present, a hybrid one under the combination of the passive and the active components, and a fully passive one are proposed. The former has been evaluated to reduce the cost for the reactor components.
Department of Hot Laboratories
JAERI-Review 2001-044, 95 Pages, 2001/12
JAERI-Review-2001-044.pdf:6.55MB
no abstracts in English
Okubo, Tadatsune*; Tsukuda, Yoshiaki*; Kamimura, Katsuichiro*; Murai, Kenji*; Goto, Ken*; Doi, Soichi*; Senda, Yasuhide*; Kosaka, Yuji*; Kido, Toshiya*; Murata, Tamotsu*; et al.
Nihon Genshiryoku Gakkai-Shi, 43(9), p.906 - 915, 2001/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
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
Nagase, Fumihisa
Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.148 - 155, 2001/06
no abstracts in English
