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Kato, Masato; Oki, Takumi; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.; Ozawa, Takayuki; Uwaba, Tomoyuki; Ikusawa, Yoshihisa; Nakamura, Hiroki; Machida, Masahiko
Journal of the American Ceramic Society, 107(5), p.2998 - 3011, 2024/05
Times Cited Count：0 Percentile：0.01(Materials Science, Ceramics)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.
Kato, Masato; Machida, Masahiko; Hirooka, Shun; Nakamichi, Shinya; Ikusawa, Yoshihisa; Nakamura, Hiroki; Kobayashi, Keita; Ozawa, Takayuki; Maeda, Koji; Sasaki, Shinji; et al.
Materials Science and Fuel Technologies of Uranium and Plutonium mixed Oxide, 171 Pages, 2022/10
Innovative and advanced nuclear reactors using plutonium fuel has been developed in each country. In order to develop a new nuclear fuel, irradiation tests are indispensable, and it is necessary to demonstrate the performance and safety of nuclear fuels. If we can develop a technology that accurately simulates irradiation behavior as a technology that complements the irradiation test, the cost, time, and labor involved in nuclear fuel research and development will be greatly reduced. And safety and reliability can be significantly improved through simulation of nuclear fuel irradiation behavior. In order to evaluate the performance of nuclear fuel, it is necessary to know the physical and chemical properties of the fuel at high temperatures. And it is indispensable to develop a behavior model that describes various phenomena that occur during irradiation. In previous research and development, empirical methods with fitting parameters have been used in many parts of model development. However, empirical techniques can give very different results in areas where there is no data. Therefore, the purpose of this study is to construct a scientific descriptive model that can extrapolate the basic characteristics of fuel to the composition and temperature, and to develop an irradiation behavior analysis code to which the model is applied.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Ozawa, Takayuki; Hirooka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*
Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09
Times Cited Count：4 Percentile：56.94(Materials Science, Multidisciplinary)To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.
Shobu, Takahisa; Shiro, Ayumi*; Kono, Fumiaki*; Muramatsu, Toshiharu; Yamada, Tomonori; Naganuma, Masayuki; Ozawa, Takayuki
Quantum Beam Science (Internet), 5(2), p.17_1 - 17_9, 2021/06
The automotive industries employ laser beam welding because it realizes a high energy density without generating irradiation marks on the opposite side of the irradiated surface. Typical measurement techniques such as strain gauges and tube X-rays cannot assess the localized strain at a joint weld. Herein high-energy synchrotron radiation X-ray diffraction was used to study the internal strain distribution of laser lap joint PNC-FMS steels (2- and 5-mm thick) under loading at a high temperature. As the tensile load increased, the local tensile and compressive strains increased near the interface. These changes agreed well with the finite element analysis results. However, it is essential to complementarily utilize internal defect observations by X-ray transmission imaging because the results depend on the defects generated by laser processing.
Ozawa, Takayuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
In a recycle system for minor actinides (MAs) currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. MA content is expected to be 
5 wt.% in the future recycle system, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important behaviors would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. The MA-containing influences were evaluated with CEPTAR.V2, including fuel properties and analysis models to evaluate the MA-MOX fuel irradiation behavior, by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo. The irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed the influences of Am-containing.
Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.
Ozawa, Takayuki; Ikusawa, Yoshihisa; Kato, Masato
Transactions of the American Nuclear Society, 113(1), p.622 - 624, 2015/10
A recycle system for minor actinides (MAs), in which MAs are recycled by reprocessing and irradiating them in a fast reactor, is studied to reduce the degree of hazard and the amount of high-level radioactive wastes. MAs would be used as mixed oxide (MOX) fuels with plutonium and uranium in fast reactors. Since MA content of MA-bearing MOX (MA-MOX) to be used in fast reactors is assumed to reach 5 wt%HM, the effects on not only fuel properties but also fuel behaviors have to be estimated to use MA-MOX as fast reactor fuels. As the MOX fuels to be used will be irradiated at a comparably high linear power and the fuel center temperature would be assumed to be over 2,273 K during irradiation in the fast reactors, fuel restructuring would take place due to void migration towards the fuel center under the radial temperature gradient, and a central void would be formed. Since the fuel center temperature would be decreased by the effect of formation of the central void, the fuel restructuring is one of the most important behaviors for fast reactor fuels. In this study, the effect of MA content on fuel restructuring behavior was estimated from the results of irradiation experiments such as B11 and B14 performed in Joyo to study the irradiation behaviors of MA-MOX and the calculation results using a fuel restructuring model which can take into account MA-MOX dependence on vapor pressure.
Kono, Fumiaki; Sogame, Motomu; Yamada, Tomonori; Shobu, Takahisa; Naganuma, Masayuki; Ozawa, Takayuki; Muramatsu, Toshiharu
JAEA-Technology 2015-004, 57 Pages, 2015/03
JAEA-Technology-2015-004.pdf:20.87MB
Laser welding of ferritic/martensitic steel (PNC-FMS) sheets with different thicknesses (2 mm and 5 mm) was examined to investigate the weldability between the inner and outer duct in fast reactor fuel assemblies with inner duct structure (FAIDUS); the objective of the inner duct is to avoid the re-criticality in case of the core melting accident. Laser-spot and melt-run welding was performed at various laser powers, welding times and velocities to find out the appropriate welding conditions with few defects and enough penetration depth. As for the spot welding, furthermore, slow cooling rate or pulsed laser irradiation could reduce the crack and porosity in the welded zone. The strain of the welded zone almost disappeared and the hardness was comparable with that of the base metal by applying post welding heat treatment at 690 
C for 103 min. In addition, the shear strength of welded joints was confirmed to be sufficiently higher than the provisional allowance shear stress. These results indicate that laser welding would be probably applied to the PNC-FMS inner and outer ducts.
Kamei, Miho; Ozawa, Takayuki
JAEA-Technology 2014-033, 36 Pages, 2014/11
JAEA-Technology-2014-033.pdf:3.93MB
Annular fuel pellet would be available to improve fast reactor fuel performance, and we have developed the "CEPTAR" to apply the annular fuel design taking into account the irradiation behaviors. CEPTAR computes the stress and strain in fuel pellet and cladding by using the generalized plane strain analysis method and the void migration model is applied to compute the fuel restructuring. On the other hand, taking into account the licensability, the fuel restructuring three-region model is applied to the fast reactor fuel design code. In this study, we developed "CEPTAR-D", in which fuel restructuring model of CEPTAR was exchanged into the "fuel restructuring three-region" model, to apply to the fuel design, and verified thermal and mechanical computations by using the results of short-term and long-term irradiation tests. Consequently, the computation accuracy of CEPTAR-D was as well as that of CEPTAR, and it was confirmed that CEPTAR-D could reasonably evaluate the stress due to PCMI.
Ikusawa, Yoshihisa; Ozawa, Takayuki; Hirooka, Shun; Maeda, Koji; Kato, Masato; Maeda, Seiichiro
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07
In order to develop MA contained MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of irradiation behavior analysis code should be evaluated with the result of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module "TRANSIT" to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code "DIRAD" and developed DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor "JOYO". As the result of the verification, it can be mentioned that the DIRAD-TRANSIT system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.
Kato, Masato; Maeda, Koji; Ozawa, Takayuki; Kashimura, Motoaki; Kihara, Yoshiyuki
Journal of Nuclear Science and Technology, 48(4), p.646 - 653, 2011/04
Physical properties and irradiation behavior of minor actinide-bearing MOX were evaluated for the development of advanced fast reactor fuels. The physical properties of the fuels were described as functions of minor actinide content and oxygen-to-metal (
/
) ratio, and the effect of minor actinide addition into MOX on those properties was found to be negligibly small. The irradiation tests of fuel pellets having / ratios of 1.98 or 1.96 were carried out at high linear heat rate of about 430W/cm. The redistribution of actinide element and oxygen were analyzed by using the evaluated properties, and maximum temperatures of the pellets were estimated. The maximum temperature of the pellets of /=1.96 was estimated to reach 2680K which was 130K higher than that of /=1.98 pellets. The maximum temperature of the pellet was lower as compared with its melting temperature of higher than 3000K. In the results of post-irradiation examination, no trace of melting was observed.
Ozawa, Takayuki; Ikusawa, Yoshihisa
Proceedings of 2010 LWR Fuel Performance Meeting/TopFuel/WRFPM (CD-ROM), p.72 - 81, 2010/09
For the effective utilization of the energy resources, preparations are underway to recycle plutonium separated by reprocessing the spent fuels from nuclear power plants into nuclear fuels in Light Water Reactors (LWRs). In this nuclear fuel cycle, plutonium is reused as uranium-plutonium mixed dioxide (MOX). In Japan, a total of 772 MOX fuel assemblies were used in FUGEN without any failure until the end of its operation in March, 2003, the most MOX fuel usage by a thermal reactor in the world. Several post-irradiation examinations necessary to evaluate the MOX fuel performance were carried out for the MOX fuel assembly irradiated in FUGEN, and consequently we could obtain the usable data to evaluate the irradiation behavior of MOX fuels. Furthermore, several MOX fuel assemblies, which were equipped in-pile instruments, used in the irradiation tests, i.e. the regular operation irradiation tests, the ramp tests, and the load-follow tests, in Norway's "Halden" reactor (HBWR). We developed a MOX fuel database to make the most of our experiences with FUGEN and HBWR in helping improve the reliability of future MOX fuel use in LWRs.
Ikusawa, Yoshihisa; Morimoto, Kyoichi; Ozawa, Takayuki; Kato, Masato
Proceedings of Plutonium Futures; The Science 2010 (CD-ROM), p.341 - 342, 2010/09
Thermal conductivity of oxide fuel is important for fuel design and performance analyses. Uranium dioxide and uranium-plutonium mixed oxide (MOX) are used as fuels in light water reactors (LWRs), and the thermal conductivities of these oxide fuels have been measured in various laboratories. In a review of oxide fuel properties, it was reported that the thermal conductivity of oxide fuel would decrease with burn-up increase. In this study, burn-up effect on MOX fuel thermal conductivity was discussed.
Ozawa, Takayuki; Kato, Masato
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.2036 - 2044, 2009/09
Americium, one of MAs, is contained in MOX fuels due to decay of plutonium-241. The radial redistribution of americium has been observed with that of plutonium in the irradiated MOX fuel. The development of a migration model for plutonium and americium redistribution would be important for fuel design because of their influence on thermal properties, i.e. thermal conductivity and melting temperature. In this study, the migration model for plutonium and americium redistribution was developed by taking into account thermal diffusion concurrently with vapor phase transport via pores in the fuel. The computed radial redistribution of plutonium and americium was found to be in good agreement with the results of post-irradiation examinations after the irradiation test for 2% neptunium and 2% americium doped uranium plutonium mixed oxide (U, Np, Pu, Am)O
fuel in JOYO.
Ikusawa, Yoshihisa; Ozawa, Takayuki
JAEA-Technology 2007-070, 27 Pages, 2008/03
JAEA-Technology-2007-070.pdf:46.26MB
IFA-554/555 load-follow tests were performed in HALDEN reactor to study the MOX fuel behavior under the daily-load-follow operation condition in the framework of ATR-MOX fuel development in JAEA. In this report, FP gas release behavior of MOX fuel rod was evaluated under the daily-load-follow operation condition with the examination data from IFA-554/555 by using the computation code FASTGRASS. As the computation results of FASTGRASS code, which could compute the FP gas release behavior under the transient condition, it could be concluded that FP gas would release due to the relaxation of fuel pellet inner stress, which were caused by the cyclic power change during the daily-load-follow operation. In addition, since the amount of released FP gas would decrease during the steady operation after the daily-load-follow, it could be mentioned that the total of FP gas release at the end of life with the daily-load-follow would not be so much different from that without the daily-load-follow.
Ozawa, Takayuki
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.404 - 408, 2007/09
The probabilistic annular fuel design code "BORNFREE-CEPTAR" was developed for the reasonable design of annular fuels to be applied for fast reactors in future. In the probabilistic design method, the performance parameters, i.e. fuel center temperature, cladding temperature, cladding stress, etc., used to be evaluated with the Monte Carlo method under the irradiation behavior, and the quantitative design margin could be obtained. As the result of probabilistic evaluation with this code, the possibility of the improvement of reactor performance of the advanced fast reactor was quantitatively indicated.
Ikusawa, Yoshihisa; Ozawa, Takayuki
JAEA-Technology 2007-013, 38 Pages, 2007/03
JAEA-Technology-2007-013.pdf:4.14MB
The following generation MONJU core fuel is considered using the high density solid pellet. Although the fuel design code "CEPTAR" was developed for annular fuel pellet, CEPTAR code was not verified with the data of high density solid pellet. In this study, CEPTAR code was verified with irradiation data of JOYO Mk-II driver fuel that used high density solid pellet. To estimate irradiation behavior of JOYO Mk-II driver fuel, the following new equations were added to CEPTAR code; The swelling equation and irradiation creep equation of PNC316. The pellet swelling equation evaluated with the PIE data of JOYO Mk-II driver fuel. As a result of verification by using the irradiation data of JOYO Mk-II driver fuel, the calculated values with CEPTAR code were in agreement with the observed values from the result of PIEs up to pellet peak burn-up 76,000MW d/t.
